Landolt-Börnstein Numerical Data and Functional Relationships in Science and Technology New Series / Editor in Chief: W. Martienssen
Group VIII: Advanced Materials and Technologies Volume 4
Radiological Protection
Editors: A. Kaul, D. Becker Authors: D. Becker, G. Brix, A. Dalheimer, G. Dietze, H.R. Doerfel, K.F. Eckerman, H. Graffunder, Y. Harima, K. Hayashi, N. Ishigure, A. Kaul, H. Klewe-Nebenius, M. Lasch, H. Paretzke, N. Petoussi-Henss, A. Phipps, H. Smith, J.W. Stather, G.N. Stradling, D.M. Taylor, H.-G. Vogt, W. Weiss
ISSN 1619-4802 (Advanced Materials and Technologies) ISBN 3-540-20207-2 Springer Berlin Heidelberg New York Library of Congress Cataloging in Publication Data Zahlenwerte und Funktionen aus Naturwissenschaften und Technik, Neue Serie Editor in Chief: W. Martienssen Vol. VIII/4: Editor: A. Kaul, D. Becker At head of title: Landolt-Börnstein. Added t.p.: Numerical data and functional relationships in science and technology. Tables chiefly in English. Intended to supersede the Physikalisch-chemische Tabellen by H. Landolt and R. Börnstein of which the 6th ed. began publication in 1950 under title: Zahlenwerte und Funktionen aus Physik, Chemie, Astronomie, Geophysik und Technik. Vols. published after v. 1 of group I have imprint: Berlin, New York, Springer-Verlag Includes bibliographies. 1. Physics--Tables. 2. Chemistry--Tables. 3. Engineering--Tables. I. Börnstein, R. (Richard), 1852-1913. II. Landolt, H. (Hans), 1831-1910. III. Physikalisch-chemische Tabellen. IV. Title: Numerical data and functional relationships in science and technology. QC61.23 502'.12 62-53136 This work is subject to copyright. All rights are reserved, whether the whole or part of the material is concerned, specifically the rights of translation, reprinting, reuse of illustrations, recitation, broadcasting, reproduction on microfilm or in other ways, and storage in data banks. Duplication of this publication or parts thereof is permitted only under the provisions of the German Copyright Law of September 9, 1965, in its current version, and permission for use must always be obtained from Springer-Verlag. Violations are liable for prosecution act under German Copyright Law. Springer-Verlag Berlin Heidelberg New York a member of BertelsmannSpringer Science+Business Media GmbH © Springer-Verlag Berlin Heidelberg 2005 Printed in Germany The use of general descriptive names, registered names, trademarks, etc. in this publication does not imply, even in the absence of a specific statement, that such names are exempt from the relevant protective laws and regulations and therefore free for general use. Product Liability: The data and other information in this handbook have been carefully extracted and evaluated by experts from the original literature. Furthermore, they have been checked for correctness by authors and the editorial staff before printing. Nevertheless, the publisher can give no guarantee for the correctness of the data and information provided. In any individual case of application, the respective user must check the correctness by consulting other relevant sources of information. Cover layout: Erich Kirchner, Heidelberg Typesetting: Authors and Redaktion Landolt-Börnstein, Darmstadt Printing and Binding: AZ Druck, Kempten/Allgäu SPIN: 10723325
63/3020 - 5 4 3 2 1 0 – Printed on acid-free paper
Preface About 10 years before discovery of X-rays and natural radioactivity by W. C. Röntgen and H. Becquerel, more precisely in 1883, Hans Landolt, Richard Börnstein and Julius Springer have started a series of selected and easily retrievable physical data, which became a successful tool for natural scientists working or practising a profession in fields of chemistry, physics or technology. Now, i.e. about 120 years after start of this unique data collection and consequently about 100 years after introduction of ionizing radiations and radionuclides in natural sciences, medicine and technology, the Landolt-Börnstein New Series is submitting to the reader a full volume on protection of man against ionizing radiations and radionuclides, i.e. “Radiological Protection”. A comparison with the 6th edition of Landolt-Börnstein containing merely a six pages chapter on “Strahlenschutz” shows the rapid development of the field within the last five decades. Compared to many of the volumes in the Landolt-Börnstein Series published in the past, the present publication in the group Advanced Materials and Technologies is not only a compilation of numerical data and functional relationships for practical purposes. Rather a comprehensive accompanying text is intended to impart to the scientific or professional user of “Radiological Protection” both data, the concepts and scientific bases of the discipline devoted to prevention of health risks to man from exposure to ionizing radiations and radionuclides. Conceptually, radiological protection is based on the principles of justification of any use of ionizing radiation, of optimization of the application of radiation, and limitation of the radiation risk to man and his environment by acceptable doses, so that use of radiation and radionuclides in scientific research, medicine, technique and daily life is always of net benefit to man. Since findings of various scientific disciplines such as medicine, biology, biophysics, nuclear physics and techniques are the basis for radiological protection, multidisciplinary knowledge of fundamentals of these disciplines is necessary for an effective protection of man against health effects of ionizing radiations. Consequently, the present volume contains contributions of experts internationally qualified in scientific disciplines or subjects such as radiation physics, biology and medicine, external and internal dosimetry of ionizing radiation and radionuclides, decontamination and decorporation of radionuclides, or physical and biological measuring techniques. Although a previous volume in the Landolt-Börnstein Series has already considered shielding against high energy radiation such as of accelerators or of cosmic origin the specific item of assessment of radiation shielding was treated, too, however restricted to an extent being necessary for completion of tasks of practical radiological protection, specifically in the field of lower energies. The present volume addresses to • those already working in radiological protection, under the aspect of making available to them numerical data and functional relationships e.g. on assessment of radiation doses from external and internal sources, or with the aim of further education and impartation of most recent knowledge in radiological protection and scientific disciplines behind; • those participating in post-graduate education programmes in radiological protection with the aim to get a qualified expert e.g. in medical radiation physics, or as an employee in a competent national authority for health protection; • newcomers in the field of radiological protection to submit necessary knowledge on bases and practices of this discipline; • advanced students of physics, techniques or medicine with special interest in a later professional occupation as health physicists, engineers or technicians; • physicians practising in X-ray diagnostics, radiation oncology and nuclear medicine with special interest in medical radiological protection.
In the hardcopy of the present volume a CD-ROM is included containing: • the full text in the multi-platform Adobe-Acrobat(pdf)-format with searchable fulltext index and • additional information and data, which would be beyond the scope of the printed version, within the interactive programme SISy (for MS-Windows only). These refer e.g. to decay data of radionuclides or normalized excretion functions for monitoring workers by quantitative assessment of intakes of radionuclides and calculation of resulting doses. For further numerical data such as dose coefficients for intake of radionuclides by workers or members of the public that are available from publications e.g. of the International Commission on Radiological Protection ICRP or of the International Commission on Radiation Units and Measurements ICRU either as hardcopies or in the Internet are not contained on the CD-ROM. The reader is referred to the relevant original sources. The editors of the present volume want to thank the authors of the contributions for their careful work, the Editor in Chief of the Landolt-Börnstein Series, Prof. Dr. W. Martienssen, for having put "Radiological Protection" on the list of volumes to be prepared for the New Series, and the Publisher, especially Drs. Ch. Meier and R. Poerschke from the editorial office for their permanent and very active engagement in realizing the present opus.
The Editors
Braunschweig/Salzgitter, 2004
Contributors
Editors D. Becker Bundesamt für Strahlenschutz Fachbereich KT 2 Willy-Brandt-Straße 5 38226 Salzgitter-Lebenstedt GERMANY
A. Kaul Physikalisch-Technische Bundesanstalt Bundesallee 100 38116 Braunschweig GERMANY
Authors D. Becker Bundesamt für Strahlenschutz Fachbereich KT 2 Willy-Brandt-Straße 5 38226 Salzgitter-Lebenstedt GERMANY 1 Intoduction
G. Brix Bundesamt für Strahlenschutz Institut für Strahlenhygiene Ingolstädter Landstraße 1 Neuherberg 85764 Oberschleißheim GERMANY 10 Measuring techniques
A. Dalheimer Bundesamt für Strahlenschutz Fachbereich Strahlenschutz und Gesundheit Köpenicker Allee 120-130 10318 Berlin GERMANY 10 Measuring techniques
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Contributors
G. Dietze Physikalisch-Technische Bundesanstalt Abt. 6 Bundesallee 100 38116 Braunschweig GERMANY 4 Radiological quantities and units 10 Measuring techniques
H.R. Doerfel Forschungszentrum Karlsruhe Hauptabteilung Sicherheit Postfach 3640 76021 Karlsruhe GERMANY 10 Measuring techniques
K.F. Eckerman Health Sciences Research Division Oak Ridge National Laboratory 1060 Commerce Park Oak Ridge Tennessee 37831-6480 USA 7 Internal dosimetry of radionuclides
H. Graffunder Ingenieurbüro Graffunder Friedrichstraße 28 76297 Stutensee GERMANY Radiation Protection Information System (SISy)
Y. Harima Tokyo Institute of Tecnology Research Laboratory for Nuclear Reactors 7-3-4-307 Hikarigaoka Nerima-ku Tokyo 179-0072 JAPAN 5 Shielding against ionizing radiation
K. Hayashi Hitachi, Ltd. Nuclear Plant Engineering Department Saiwai-cho, 3-1-1, Hitachi, Ibaraki, 317-8511 JAPAN 5 Shielding against ionizing radiation
Contributors
N. Ishigure Research Center for Radiation Safety National Institute of Radiological Science 4-9-1, Anagawa, Inage, Chiba 263-8555 JAPAN 7 Internal dosimetry of radionuclides
A. Kaul Physikalisch-Technische Bundesanstalt Bundesallee 100 38116 Braunschweig GERMANY 1 Introduction 8 Decontamination 11 Exposures from natural and man-made radiation sources
H. Klewe-Nebenius Forschungszentrum Karlsruhe Institut für Instrumentelle Analytik Postfach 3640 76021 Karlsruhe GERMANY 3 Physical fundamentals
M. Lasch Kernkraftwerke Grundremmingen Betriebsgesellschaft mbH Postfach 89355 Grundremmingen GERMANY 8 Decontamination
H.G. Paretzke Institut für Strahlenschutz GSF-Forschungszentrum für Umwelt und Gesundheit, GmbH Neuherberg, Postfach 11 29 85758 Oberschleißheim GERMANY 6 External dosimetry
N. Petoussi-Henss Institut für Strahlenschutz GSF-Forschungszentrum für Umwelt und Gesundheit, GmbH Neuherberg, Postfach 11 29 85758 Oberschleißheim GERMANY 6 External dosimetry
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Contributors
A. Phipps National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 7 Internal dosimetry of radionuclides
H. Smith † National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 2 Biological effects of ionising radiation
J.W. Stather National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 2 Biological effects of ionising radiation 7 Internal dosimetry of radionuclides
G.N. Stradling National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 9 Decorporation of radionuclides
D.M. Taylor 5, Pan Poeth Pen-y-bont, CF31 5BD Wales UNITED KINGDOM 9 Decorporation of radionuclides
H.-G. Vogt Zentrum für Strahlenschutz und Radioökologie Universität Hannover Am Kleinen Felde 30 30167 Hannover GERMANY 5 Shielding against ionizing radiation
Contributors
W. Weiss Bundesamt für Strahlenschutz Fachbereich Strahlenhygiene Institut für Strahlenhygiene Ingolstädter Landstr. 1 85764 Oberschleißheim GERMANY 10 Measuring techniques
Landolt-Börnstein Editorial Office Gagernstr. 8, D-64283 Darmstadt, Germany fax: +49 (6151) 171760 e-mail:
[email protected] Internet http://www.landolt-boernstein.com
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VIII/4 Radiological Protection Contents 1
The development of the organizational and the conceptual basis of radiological protection ........................................................................................................................................................................................................................ 1-1
2 2.1 2.2 2.2.1 2.2.2 2.2.3 2.2.4 2.3 2.3.1 2.3.2 2.3.3 2.3.4 2.4 2.4.1 2.4.2 2.4.3 2.4.4 2.4.5 2.4.6 2.4.7 2.5 2.5.1 2.5.2 2.6 2.6.1 2.6.2 2.6.3 2.6.4 2.7 2.8 2.9
Biological effects of ionising radiation .......................................................................................................................................... 2-1 Introduction .................................................................................................................................................................................................................... 2-1 Cellular effects ............................................................................................................................................................................................................ 2-2 Primary events following exposure to ionising radiation ................................................................................................ 2-2 Cellular damage and repair following the primary radiation events .................................................................... 2-3 Classification of radiation-induced damage .................................................................................................................................. 2-5 Implications of cellular damage for whole or partial body exposure .................................................................. 2-9 Deterministic effects.............................................................................................................................................................................................. 2-9 Tissue and organ development.................................................................................................................................................................... 2-9 Dose-response relationships for radiation damage ............................................................................................................. 2-10 Deterministic effects in humans following acute whole-body irradiation.................................................. 2-11 Deterministic effects following partial body irradiation............................................................................................... 2-12 Radiation-induced cancer ............................................................................................................................................................................. 2-15 Cancer development .......................................................................................................................................................................................... 2-15 Dose-response relationships ...................................................................................................................................................................... 2-16 Exposures to external radiation .............................................................................................................................................................. 2-18 Exposure to internally incorporated radionuclides ............................................................................................................. 2-20 Dose and dose rate effectiveness factors (DDREFs)........................................................................................................ 2-23 Risk coefficients for protection .............................................................................................................................................................. 2-23 Low dose studies ................................................................................................................................................................................................... 2-25 Hereditary disease ................................................................................................................................................................................................ 2-27 Categories of genetic damage .................................................................................................................................................................. 2-27 Risk coefficients for hereditary disease ......................................................................................................................................... 2-28 Irradiation in utero ............................................................................................................................................................................................... 2-29 Deterministic effects.......................................................................................................................................................................................... 2-30 Brain function........................................................................................................................................................................................................... 2-30 Risk coefficients for cancer........................................................................................................................................................................ 2-31 Hereditary disease ................................................................................................................................................................................................ 2-31 Summary of risk factors for cancer and hereditary disease ....................................................................................... 2-32 Conclusions ................................................................................................................................................................................................................ 2-32 References .................................................................................................................................................................................................................... 2-34
3 3.1 3.2 3.2.1 3.2.2 3.3 3.3.1 3.3.2 3.4 3.4.1 3.4.2 3.4.3 3.5
Physical fundamentals ..................................................................................................................................................................................... 3-1 Natural radioactivity .............................................................................................................................................................................................. 3-1 Elements, isotopes and radionuclides .................................................................................................................................................. 3-2 Atoms, electrons and the Periodic Table of Elements ........................................................................................................ 3-2 Atomic nuclei, nuclides and the Chart of Nuclides ............................................................................................................... 3-3 The structure of the atomic nucleus ....................................................................................................................................................... 3-4 Elementary particles .............................................................................................................................................................................................. 3-4 Nuclear transformations..................................................................................................................................................................................... 3-6 Radioactive decay ................................................................................................................................................................................................ 3-10 Basic properties ...................................................................................................................................................................................................... 3-10 Decay modes ............................................................................................................................................................................................................. 3-15 The natural radioactive decay families ........................................................................................................................................... 3-18 Radioactive radiation ........................................................................................................................................................................................ 3-24
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3.5.1 3.5.2 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 3.7
Types of radiation ................................................................................................................................................................................................ 3-24 Physical properties of radiation.............................................................................................................................................................. 3-26 Nuclear fission and fission products ................................................................................................................................................. 3-30 Particle induced nuclear fission ............................................................................................................................................................. 3-30 Fission products ..................................................................................................................................................................................................... 3-30 Nuclear reactors ..................................................................................................................................................................................................... 3-32 Nuclear explosives .............................................................................................................................................................................................. 3-32 Radioactive inventory and nuclear waste..................................................................................................................................... 3-33 Release of radionuclides from the radioactive inventory of a nuclear reactor ....................................... 3-38 References .................................................................................................................................................................................................................... 3-39
4 4.1 4.2 4.2.1 4.2.2 4.3 4.3.1 4.3.2 4.3.3 4.3.4 4.4 4.4.1 4.4.2 4.5 4.5.1 4.5.2 4.5.3 4.6 4.6.1 4.6.2 4.7 4.8 4.9
Radiological quantities and units........................................................................................................................................................ 4-1 Introduction .................................................................................................................................................................................................................... 4-1 Radiation field quantities .................................................................................................................................................................................. 4-2 Scalar radiation field quantities .................................................................................................................................................................. 4-2 Vectorial radiation field quantities.......................................................................................................................................................... 4-4 Interaction coefficients and quantities................................................................................................................................................. 4-5 Cross section.................................................................................................................................................................................................................. 4-5 Mass attenuation coefficient and mass energy transfer coefficient ....................................................................... 4-5 Mass stopping power and linear energy transfer (LET).................................................................................................... 4-6 Mean energy expended in a gas per ion pair formed ........................................................................................................... 4-7 Quantities related to energy transfer ..................................................................................................................................................... 4-7 Stochastic quantities .............................................................................................................................................................................................. 4-7 Non-stochastic quantities.................................................................................................................................................................................. 4-8 Dose quantities in radiation protection ........................................................................................................................................... 4-10 Concept of radiation protection quantities .................................................................................................................................. 4-10 Protection quantities .......................................................................................................................................................................................... 4-11 Operational quantities ...................................................................................................................................................................................... 4-14 Radioactivity quantities .................................................................................................................................................................................. 4-18 Activity, specific activity, activity concentration, activity per area .................................................................. 4-19 Specific quantities for radon, thoron and their progeny................................................................................................ 4-19 Quantities for internal dosimetry .......................................................................................................................................................... 4-22 Limits, constraints, action levels ........................................................................................................................................................... 4-23 References .................................................................................................................................................................................................................... 4-27
5 5.1 5.2 5.3 5.3.1 5.3.2 5.4 5.4.1 5.4.2 5.4.3 5.5 5.5.1 5.5.2 5.5.3 5.5.4 5.6 5.7
Shielding against ionizing radiation ................................................................................................................................................ 5-1 Introduction .................................................................................................................................................................................................................... 5-1 Stopping power and range ............................................................................................................................................................................... 5-1 Penetration depths of charged particles ............................................................................................................................................. 5-2 Heavy charged particles ..................................................................................................................................................................................... 5-2 Electrons and positrons....................................................................................................................................................................................... 5-4 Photons ................................................................................................................................................................................................................................ 5-5 Basic shielding concept ...................................................................................................................................................................................... 5-5 Attenuation data of radioactive sources in shielding materials.................................................................................. 5-6 An example of the calculation of an ambient dose equivalent rate ....................................................................... 5-9 Neutrons ......................................................................................................................................................................................................................... 5-16 Basic shielding concepts ............................................................................................................................................................................... 5-16 Attenuation data of various neutron sources in shielding materials ................................................................. 5-16 Sample shield calculation ............................................................................................................................................................................. 5-18 Induced activity ...................................................................................................................................................................................................... 5-19 Computer codes and online nuclear data services............................................................................................................... 5-28 References .................................................................................................................................................................................................................... 5-32
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6 6.1 6.1.1 6.1.2 6.2 6.2.1 6.2.2 6.2.3 6.2.4 6.3 6.3.1 6.3.2 6.4 6.5 6.5.1 6.5.2 6.6 6.7
External dosimetry .............................................................................................................................................................................................. 6-1 Protection and operational quantities ................................................................................................................................................... 6-1 Protection quantities .............................................................................................................................................................................................. 6-1 Operational Quantities ......................................................................................................................................................................................... 6-1 Dosimetric models ................................................................................................................................................................................................... 6-2 Models and phantoms of the human body ...................................................................................................................................... 6-2 Idealized geometries representing occupational exposures ........................................................................................... 6-4 Environmental source geometries............................................................................................................................................................ 6-4 Methods of calculating protection quantities in computational models............................................................ 6-5 Conversion coefficients for photons ..................................................................................................................................................... 6-6 Occupational .................................................................................................................................................................................................................. 6-6 Conversion coefficients for environmental gamma ray fields ................................................................................ 6-13 Conversion coefficients for neutrons ............................................................................................................................................... 6-20 Conversion coefficients for electrons .............................................................................................................................................. 6-21 Occupational exposure .................................................................................................................................................................................... 6-21 Environmental exposure ................................................................................................................................................................................ 6-23 Doses from external exposure of radionuclides in the environment ................................................................ 6-23 References .................................................................................................................................................................................................................... 6-42
7 7.1 7.2 7.2.1 7.2.2 7.2.3 7.2.4 7.2.5 7.2.6 7.2.7 7.2.8 7.3 7.3.1 7.3.2 7.3.3 7.3.4 7.3.5 7.3.6 7.4 7.4.1 7.4.2 7.4.3 7.5 7.5.1 7.6 7.6.1 7.6.2 7.6.3 7.6.4 7.6.5 7.7 7.7.1 7.7.2 7.8
Internal dosimetry of radionuclides ................................................................................................................................................ 7-1 Introduction .................................................................................................................................................................................................................... 7-1 Biokinetics of radionuclides in the body .......................................................................................................................................... 7-2 Inhalation .......................................................................................................................................................................................................................... 7-3 Ingestion......................................................................................................................................................................................................................... 7-10 Cuts and wounds ................................................................................................................................................................................................... 7-14 Absorption through intact skin ............................................................................................................................................................... 7-15 Systemic behaviour of radionuclides................................................................................................................................................ 7-15 Excretion ....................................................................................................................................................................................................................... 7-21 Embryo and foetus............................................................................................................................................................................................... 7-21 Transfer in maternal milk ............................................................................................................................................................................. 7-24 Dosimetric models ............................................................................................................................................................................................... 7-25 Introduction ................................................................................................................................................................................................................ 7-25 Absorbed fraction and specific absorbed fraction ............................................................................................................... 7-26 Computational models of the human anatomy ....................................................................................................................... 7-28 Dose rate per unit activity, S-factor ................................................................................................................................................... 7-31 Specific absorbed fractions for various radiations.............................................................................................................. 7-32 Calculation of doses to soft tissues and the skeleton........................................................................................................ 7-36 Dose coefficients ................................................................................................................................................................................................... 7-37 Method of calculation ...................................................................................................................................................................................... 7-38 Sources of dose coefficients ...................................................................................................................................................................... 7-41 Dose coefficients for selected radionuclides ............................................................................................................................ 7-43 Internal monitoring ............................................................................................................................................................................................. 7-46 Methods of individual monitoring....................................................................................................................................................... 7-47 Monitoring programme................................................................................................................................................................................... 7-51 Need for a monitoring programme ..................................................................................................................................................... 7-51 Routine monitoring ............................................................................................................................................................................................. 7-51 Special or task-related monitoring ...................................................................................................................................................... 7-52 Confirmatory monitoring .............................................................................................................................................................................. 7-52 Wound monitoring .............................................................................................................................................................................................. 7-52 Dose Assessment................................................................................................................................................................................................... 7-53 Estimation of intake and dose .................................................................................................................................................................. 7-53 Control of worker doses ................................................................................................................................................................................ 7-54 Monitoring data for radionuclides (H-3, Co-60, Sr-90, Ru-106, I-131, Cs-134, Cs-137, Ce-144 U-234, Pu-239, Am-241).............................................................................................................................................................................. 7-55 References .................................................................................................................................................................................................................... 7-68
7.9
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8 8.1 8.1.1 8.1.1.1 8.1.1.2 8.1.1.3 8.1.1.4 8.1.1.5 8.1.1.6 8.1.1.7 8.1.1.8 8.1.1.9 8.1.2 8.1.2.1 8.1.2.2 8.1.2.3 8.1.2.3 8.1.3 8.1.3.1 8.1.3.2 8.1.3.3 8.1.3.4 8.1.3.5 8.1.3.6 8.1.3.7 8.1.3.8 8.1.3.9 8.1.3.10 8.1.4 8.2 8.2.1 8.2.2 8.2.2.1 8.2.2.2 8.2.3 8.2.3.1 8.2.3.2 8.2.4 8.2.4.1 8.2.4.2 8.2.4.3 8.2.4.4 8.2.5 8.2.5.1 8.2.5.2 8.3
Decontamination .................................................................................................................................................................................................... 8-1 Decontamination of materials ...................................................................................................................................................................... 8-1 General approaches to decontamination ........................................................................................................................................... 8-2 Contamination .............................................................................................................................................................................................................. 8-2 Characteristics of oxide layer in BWRs and PWRs .............................................................................................................. 8-2 Other types of contamination ....................................................................................................................................................................... 8-3 Decontamination ....................................................................................................................................................................................................... 8-4 The use of decontamination in decommissioning ................................................................................................................... 8-4 Identification of decontaminable components ............................................................................................................................ 8-5 Effectiveness of decontamination, decontamination factor ........................................................................................... 8-6 Decontamination techniques (processes) ......................................................................................................................................... 8-7 Decontamination and secondary waste generation................................................................................................................. 8-8 Decontamination techniques for large volume closed systems ................................................................................. 8-8 Reactor decontamination in BWRs and PWRs .......................................................................................................................... 8-8 Fuel assemblies and decontamination.............................................................................................................................................. 8-12 Decontamination of sodium cooled systems............................................................................................................................. 8-12 Gas cooled reactors (WAGR) .................................................................................................................................................................. 8-13 Decontamination techniques for segmented parts ............................................................................................................... 8-13 Chemical decontamination .......................................................................................................................................................................... 8-13 Electrochemical decontamination ........................................................................................................................................................ 8-15 Jetting decontamination techniques ................................................................................................................................................... 8-18 Ultrasonic decontamination ....................................................................................................................................................................... 8-19 Decontamination by foams ......................................................................................................................................................................... 8-21 Decontamination by gels............................................................................................................................................................................... 8-21 Decontamination by pastes ......................................................................................................................................................................... 8-21 Mechanical decontamination techniques ...................................................................................................................................... 8-21 Decontamination by strippable coatings ....................................................................................................................................... 8-22 Melting ............................................................................................................................................................................................................................ 8-22 Decontamination techniques for building surfaces ............................................................................................................ 8-22 Decontamination of skin ............................................................................................................................................................................... 8-24 Introduction ................................................................................................................................................................................................................ 8-24 Transport of radioactive substances via the skin .................................................................................................................. 8-24 Anatomy of the skin........................................................................................................................................................................................... 8-24 Transport procedure ........................................................................................................................................................................................... 8-25 Skin dose at contamination ......................................................................................................................................................................... 8-25 Calculation of the equivalent dose to the skin......................................................................................................................... 8-25 Equivalent dose rate conversion coefficients ........................................................................................................................... 8-26 Decontamination measures ......................................................................................................................................................................... 8-30 Organisational and preliminary measures ................................................................................................................................... 8-30 First aid measures of skin decontamination............................................................................................................................... 8-30 Specific decontamination procedures .............................................................................................................................................. 8-31 Decontamination of specific body regions and organs .................................................................................................. 8-32 Procedure at residual contamination and fixing a reference value..................................................................... 8-32 Frequency of decontamination steps................................................................................................................................................. 8-32 Derivation of the reference value for residual contamination ................................................................................. 8-32 References .................................................................................................................................................................................................................... 8-34
9 9.1 9.2 9.2.1 9.2.2 9.2.3
Decorporation of radionuclides ............................................................................................................................................................ 9-1 Introduction .................................................................................................................................................................................................................... 9-2 General considerations ........................................................................................................................................................................................ 9-2 Factors affecting the efficacy of treatment ..................................................................................................................................... 9-2 Factors influencing treatment decisions ............................................................................................................................................ 9-3 Decision levels ............................................................................................................................................................................................................ 9-3
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9.4.4 9.4.5 9.4.6 9.4.7 9.4.8 9.4.9 9.4.10 9.4.11 9.4.12 9.4.13 9.4.14 9.5 9.5.1 9.5.2 9.5.3 9.6 9.6.1 9.6.2 9.6.3 9.6.4 9.6.5 9.6.6 9.6.7 9.7 9.8
Perception of risk and its implications ................................................................................................................................................ 9-4 Approaches to treatment.................................................................................................................................................................................... 9-5 Methods of treatment ............................................................................................................................................................................................ 9-6 Non-specific procedures .................................................................................................................................................................................... 9-6 Procedures to enhance systemic radionuclide excretion................................................................................................... 9-7 General comments on the efficiacy of chelating agents for the actinides ....................................................... 9-8 What are the factors that govern the efficacy of chelating agents ? ..................................................................... 9-9 Can the efficacy of treatment be predicted from animal studies ? .......................................................................... 9-9 Are chelating agents always most effective when the radionuclides are present in circulating blood ? ................................................................................................................................................................................................................................. 9-9 Is DTPA effective for all actinides ? ................................................................................................................................................ 9-10 Will the administration of chelating agents result in enhanced tissue deposition ? .......................... 9-10 Is the administration of sodium carbonate effective for uranium ? ................................................................... 9-10 Must chelating agents be administered promptly to be effective ? .................................................................... 9-10 Is intravenous injection the best mode of administration ?........................................................................................ 9-11 How can judgements on efficacy be made ?............................................................................................................................. 9-11 When should treatment start ? ................................................................................................................................................................. 9-11 When should treatment stop ? ................................................................................................................................................................. 9-11 For which materials are chelating agents likely to be effective ? ........................................................................ 9-11 For which materials are chelating agents unlikely to be effective ? ................................................................. 9-12 Is lung lavage more effective than chelation treatment for inhaled materials ?.................................... 9-12 Recent developments ........................................................................................................................................................................................ 9-13 Plutonium and americium ............................................................................................................................................................................ 9-13 Thorium .......................................................................................................................................................................................................................... 9-13 Uranium.......................................................................................................................................................................................................................... 9-13 Optimum treatment protocols................................................................................................................................................................... 9-14 Tritium ............................................................................................................................................................................................................................. 9-14 The alkaline earth elements, strontium, barium and radium..................................................................................... 9-14 Iodine ................................................................................................................................................................................................................................ 9-16 Caesium .......................................................................................................................................................................................................................... 9-16 Plutonium and americium ............................................................................................................................................................................ 9-18 Thorium .......................................................................................................................................................................................................................... 9-26 Uranium.......................................................................................................................................................................................................................... 9-27 Future research needs ....................................................................................................................................................................................... 9-29 References .................................................................................................................................................................................................................... 9-31
10 10.1 10.1.1 10.1.2 10.1.2.1 10.1.2.2 10.1.2.3 10.1.2.4 10.1.3 10.1.4 10.1.5 10.1.6 10.1.7 10.1.8 10.1.9 10.2 10.2.1
Measuring techniques .................................................................................................................................................................................. 10-1 Detectors for radiation protection ........................................................................................................................................................ 10-1 Overview and general characteristics of radiation detectors .................................................................................... 10-1 Gas-filled ionization detectors................................................................................................................................................................. 10-3 Ionization and gas amplification ........................................................................................................................................................... 10-3 Ionization chambers ........................................................................................................................................................................................... 10-4 Proportional counters........................................................................................................................................................................................ 10-8 Geiger-Müller counters ............................................................................................................................................................................... 10-10 Scintillation detectors .................................................................................................................................................................................... 10-11 Semiconductor detectors ............................................................................................................................................................................ 10-14 Thermoluminescence and radiophotoluminescence detectors ............................................................................. 10-18 Photographic films ........................................................................................................................................................................................... 10-20 Detectors for neutrons................................................................................................................................................................................... 10-21 Biological dosimetry ...................................................................................................................................................................................... 10-23 References for 10.1 .......................................................................................................................................................................................... 10-25 Radiological protection measurements: external exposure ..................................................................................... 10-27 Operational quantities ................................................................................................................................................................................... 10-27
9.2.4 9.2.5 9.3 9.3.1 9.3.2 9.4 9.4.1 9.4.2 9.4.3
Contents
XVII
10.2.2 10.2.3 10.2.3.1 10.2.3.2 10.2.4 10.2.4.1 10.2.4.2 10.2.4.3 10.2.5 10.2.5.1 10.2.5.2 10.2.6 10.2.6.1 10.2.6.2 10.2.6.3 10.2.6.4 10.2.7 10.3 10.3.1 10.3.1.1 10.3.1.2 10.3.1.3 10.3.1.4 10.3.2 10.3.2.1 10.3.2.2 10.3.2.3 10.3.2.4 10.3.2.5 10.3.2.6 10.3.2.7 10.3.2.8 10.3.2.9 10.3.2.10 10.3.2.11 10.3.3 10.3.3.1 10.3.3.2 10.3.3.3 10.3.3.4 10.3.3.5 10.3.3.6 10.3.3.7 10.3.3.8 10.3.3.9 10.3.3.10
Reference levels .................................................................................................................................................................................................. 10-29 Types of exposure ............................................................................................................................................................................................. 10-29 Occupational exposure ................................................................................................................................................................................. 10-30 Public exposure ................................................................................................................................................................................................... 10-30 Types of monitoring programs............................................................................................................................................................. 10-31 Individual monitoring for external exposure ......................................................................................................................... 10-31 Area monitoring for external exposure ....................................................................................................................................... 10-31 Calibration ................................................................................................................................................................................................................ 10-32 Requirements for individual monitoring of external exposure ........................................................................... 10-33 Operational requirements .......................................................................................................................................................................... 10-34 Accuracy requirement ................................................................................................................................................................................... 10-35 Personal dosimeters for individual monitoring in different radiation fields.......................................... 10-36 Photon dosimetry ............................................................................................................................................................................................... 10-36 Beta dosimetry ...................................................................................................................................................................................................... 10-37 Neutron dosimetry ............................................................................................................................................................................................ 10-37 Dosimetry in mixed field situations (photons and neutrons) ................................................................................ 10-38 References for 10.2 .......................................................................................................................................................................................... 10-39 Radiological protection measurements: internal exposure ...................................................................................... 10-41 Measurement of radon and its progeny ....................................................................................................................................... 10-41 Measurement of radon in air .................................................................................................................................................................. 10-44 Measurement of radon progeny in air........................................................................................................................................... 10-47 Measurement of radon in the ground and in water .......................................................................................................... 10-50 References for 10.3.1 ..................................................................................................................................................................................... 10-51 In vivo measurements ................................................................................................................................................................................... 10-52 Introduction ............................................................................................................................................................................................................. 10-52 Requirements ......................................................................................................................................................................................................... 10-53 Principles of γ spectrometry ................................................................................................................................................................... 10-56 Equipment ................................................................................................................................................................................................................. 10-58 Spectrum evaluation ....................................................................................................................................................................................... 10-66 Measuring geometries ................................................................................................................................................................................... 10-67 Calibration ................................................................................................................................................................................................................ 10-71 Uncertainties and detection limits..................................................................................................................................................... 10-77 Measurement procedure.............................................................................................................................................................................. 10-79 Quality assurance and control............................................................................................................................................................... 10-79 References for 10.3.2 ..................................................................................................................................................................................... 10-81 In vitro measurements: excretion analyses .............................................................................................................................. 10-83 Introduction ............................................................................................................................................................................................................. 10-83 Urine samples ........................................................................................................................................................................................................ 10-83 Faeces samples ..................................................................................................................................................................................................... 10-84 Exhalation ................................................................................................................................................................................................................. 10-84 Other biological samples............................................................................................................................................................................ 10-85 Radiochemical analyses .............................................................................................................................................................................. 10-85 Measuring techniques.................................................................................................................................................................................... 10-88 Quality assurance............................................................................................................................................................................................... 10-91 Examples for dose estimations from in vitro measurements ................................................................................. 10-93 References for 10.3.3 ..................................................................................................................................................................................... 10-97
11 11.1 11.2 11.2.1 11.2.2
Exposures from natural and man-made radiation sources ............................................................................. 11-1 Introduction ................................................................................................................................................................................................................ 11-1 Exposures by cosmic radiation and cosmogenic radionuclides............................................................................. 11-2 Origin and kinds of cosmic radiation ............................................................................................................................................... 11-2 Exposures by cosmic radiations............................................................................................................................................................. 11-2
XVIII 11.3 11.3.1 11.3.2 11.4 11.5 11.6
Contents Terrestrial radiation ............................................................................................................................................................................................ 11-4 External exposures .............................................................................................................................................................................................. 11-5 Internal exposures ................................................................................................................................................................................................ 11-6 Enhanced exposures form industrial activities ....................................................................................................................... 11-9 Worldwide average exposure from natural and man-made sources .............................................................. 11-10 References ................................................................................................................................................................................................................. 11-12
1 The development of the organizational and conceptual basis of radiological protection
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1 The development of the organizational and conceptual basis of radiological protection1
Within a few weeks of Roentgen's discovery of X-rays in 1895, the potential of X-rays for diagnosing fractures became apparent. However, the occurrence of acute adverse effects such as erythema and skin burns within the next few years made persons applying X-rays in medicine and technique aware of the need to avoid overexposure. Similar undesirable effects were reported after the discovery of natural radioactivity by H. Becquerel in 1896, specifically of radium by M. Curie, and medical application. The first organized coordinated effort for radiation protection came in 1921 when the British X-ray and Radium Protection Committee issued detailed recommendations and instructions. The American Roentgen Ray Society also proposed general recommendations in the early 1920s on the basis of avoiding acute effects. At the Second International Congress of Radiology held in Stockholm in 1928 [28B1], the unit “roentgen” (R) was recognized as the unit for X-ray dose. It was at this congress that the “International X-ray and Radium Protection Commission” was founded, the forerunner of the later (from 1950 onwards) “International Commission on Radioligical Protection (ICRP)”. The primary concern of the 1928 Commission was to elaborate recommendations designed to provide protection to members of the medical profession in their work with X-rays and gamma-rays from radium. In 1934 the Commission recommended 0.2 R per day as the “tolerance dose” [34I1]. Due to the great expansion in radiation protection work consequent upon nuclear energy developments in the period from 1940 to 1950, the International Congress of Radiology in 1950 [51I1] extended the scope of the Commission - now ICRP - and broadened its area of responsibility beyond the protection of the medical profession only. It was in 1950 that the ICRP spelt out the first time the various effects which were to be considered in making its recommendations. These recommendations were to deal primarily with the basic principles of radiation protection and to leave to the various international and regional agencies such as IAEA, EURATOM and national regulatory bodies the responsibility of introducing detailed technical regulations, codes of pratice or laws suited to the needs of their member countries or specific countries. The present Commission of ICRP is assisted by 4 Committees working in the following specialized fields [99I1]: • Committee 1 (Radiation Effects) considers the risk of induction of cancer and heritable disease together with the underlying mechanisms of radiation action; also, the risks, severity, and mechanism of induction of tissue/organ damage and developmental defects. • Committee 2 (Doses from Radiation Exposures) is concerned with the development of dose coefficients for the assessment of internal and external radiation exposure, development of reference biokinetic and dosimetric models, and reference data for workers and members of the public.
1
A concise and consolidated summary is given by A. Nagaratnam [95N1] in his handbook on the salient features of the information given in ICRP Publications on the concept of radiological protection. Landolt-Börnstein New Series VIII/4
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• Committee 3 (Protection in Medicine) is concerned with protection of persons and newborn children when ionizing radiation is used for medical diagnosis, therapy, or for biomedical research; also, assessment of the medical consequences of accidental exposures. • Committee 4 (Application of the Commission's Recommendations) is concerned with providing advice on the application of the recommended system of protection in all its facets for occupational and public exposure; it also acts as the major point of contact with other international organizations and professional societies concerned with protection against ionizing radiation. In its 1950 recommendations [51I1], ICRP replaced the 1934 concept of “tolerance dose” [34I1] by that of the “maximum permissible dose” with the recognition that there could be risk even at these levels: “Whilst the values proposed for maximum permissible exposure are such as to involve a risk which is small compared to other hazards of life, nevertheless, in view of the unsatisfactory nature of much of the evidence on which our judgement must be based, coupled with the knowledge that certain radiation effects are irreversible and cumulative, it is strongly recommended that every effort be made to reduce exposure to all types of radiation to the lowest possible levels ... and that any unnecessary exposure be avoided”. According to the 1958 recommendations [59I1] the basic permissible dose to gonads, bloodforming organs, and lenses of the eyes for persons occupationally exposed at any age over 18 years was 5 rem (50 mSv) per year or weekly 0.1 rem (1 mSv), used for purposes of planning and design. No recommendation was made for exposure of individual members of the public but it was suggested that the per capita dose should not exceed 5 rem (50 mSv) per generation excluding medical exposures and exposures to natural background radiation. A linear non-threshold response was assumed for genetic effects. In 1962 it was recommended by the ICRP that the dose to individual members of the population at large should be limited to 0.5 rem (5 mSv) per year [64I1]. ICRP made it explicit that doses from natural background and from medical exposures were excluded from the maximum permissible doses. However, ICRP “recognizes especially the importance of the gonad doses resulting from medical exposure and the attendant genetic hazard to the population”, and recommended that “the medical profession exercises great care in the use of ionizing radiation in order that the gonad dose received by individuals before the end of their reproductive periods be kept at the minimum value consistent with medical requirements”. In 1977 [77I1] ICRP published epoch-making recommendations giving a new philosophical and conceptual framework of radiological protection. It is characterized by 1. Statement of the aim of radiation protection as being to prevent detrimental non-stochastic effects and to limit the probability of stochastic effects to levels deemed to be acceptable. 2. Formulation of the basic tenets of the system of radiation protection as a: Justification: No practice shall be adopted unless its introduction produces a positive net benefit. b: Optimization: All (necessary) exposures shall be kept as low as reasonably achievable, economic and social factors being taken into account (ALARA principle). c: Dose limitation: The dose equivalents to individuals shall not exceed the limits recommended for the appropriate circumstances by the Commission (limitation of the effective dose equivalent for stochastic effects in workers to 50 mSv per year, for non-stochastic effects in specific organs to 500 mSv annually; limitation of the effective dose equivalent to control the risk from stochastic effects of individual members of the public (critical groups) to 5 mSv in a year, and to 50 mSv annually for non - stochastic effects). Subsequent to the publication of the 1977 recommendations there have been clarifications and amendments, the most important ones at the 1985 meeting of the Commission [85I1]: Considering the effective dose equivalent limits for members of the public, made in its 1977 recommendations, “the Commission's present view is that the principal (stochastic) limit is 1 mSv in a year. However, it is permissible to use a subsidiary dose limit of 5 mSv in a year for some years, provided that the average annual effective dose equivalent over a lifetime does not exceeed the principal limit of 1 mSv in a year”. Landolt-Börnstein New Series VIII/4
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Apart from changes in terminology and definitions: • • • •
“non - stochastic effects” are now called “deterministic effects”, “quality factor” is replaced by “radiation weighting factor”, “dose equivalent” is replaced by “equivalent dose”, “effective dose equivalent” is replaced by “effective dose”
the 1990 recommendations of the ICRP in its Publication 60 [91I1] have brought down significantly the dose limits for occupational exposure from 50 mSv for the annual effective dose to 100 mSv in 5 years corresponding to an average of 20 mSv annually. (Additionally the effective dose should not exceed 50 mSv in a single year). The concept of justification, optimization and individual dose limits has been retained, however, a distinction is made between the systems of protection for proposed and continuing practices, and intervention: While, as in the past, the system of protection in practices is following the general principles of justification of a practice, optimization of protection (the magnitude of individual doses, the number of people exposed, and the likelihood of incurring potential exposures should all be kept as low as reasonably achievable, economic and social factors being taken into account), and limitation of individual dose and risk, an additional system of protection in intervention has been introduced. This is based on the following principles: 1. The proposed intervention should do more good than harm, i.e. the reduction in detriment resulting from the reduction in dose should be sufficient to justify the harm and the costs, including social costs, of the intervention. 2. The form, scale, and duration of the intervention should be optimized so that the net benefit of the reduction of dose and consequently of the detriment should be maximized. Regarding hereditary effects, ICRP Publ.26 [77I1] added the hereditary risk to the first and second generation offspring to the stochastic risk to the exposed individual, the effects in later generations being considered as part of the consequences for society. ICRP Publ.60 [91I1] now attributes the whole hereditary detriment to the detriment suffered by the exposed individual, thus avoiding the need for a two - stage assessment. Under the motto “Radiological Protection at the Start of the 21th Century” ICRP in 2002 has started an initiative which represents a genuine attempt to simplify the system of protection to one that is more coherent and easily explicable [02C1; 02C2]. Since classical cost-benefit analysis based on an utilitarian ethical policy answering the question “how much does it cost to reduce a dose and how many lives are saved?”, is unable to consider the individual, the Commission already modified the principle of optimization by the introduction of the concept of a constraint. Constraint is an individual-related criterion, applied to a single source in order to ensure that the most exposed individuals are not subjected to excessive risk, and to limit the inequity introduced by cost-benefit analysis. Although in the future the process of taking all reasonable action to reduce exposures is still likely to be called the Optimization of Protection, optimization is intended to be replaced by a different requirement. Namely, residual doses, after the application of Constraints, should be kept “as low as reasonably achievable” (ALARA). In this context the emphasis of constraints should provide a basic level of health protection for individuals exposed to a particular controllable source. Since there is likely to be some risk to health even at small doses introduction of a moral requirement is discussed for each controllable source to take all reasonable steps to restrict both the individual doses to levels below the action level and the number of exposed individuals. In this context it should be emphasized that these Constraints are not intended to be applied to justified medical exposures. Under the aspect “common sense would be often more important than formal application of differential equations in optimization” stakeholder involvement is discussed to determine or negotiate for the best level of protection in the circumstances. This means that whilst the dose constraints thus represent a basic standard of individual health protection, stakeholder involvement determines how far Landolt-Börnstein New Series VIII/4
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below the action level is “as low as reasonable practicable”, and will avoid the previous formal costbenefit analyses. Consequently ALARA would represent the optimum level of protection from the source under control or for an uncontrolled source.
References 28B1 Bureau of Standards, Circular No. 374: X-Ray and Radium Protection; Recommendations of the 2nd International Congress of Radiology, 1928; Br. J. Radiol.1, (1928), 359 34I1 International X-ray and Radium Protection Commission: Br. J. Radiol. 7, (1934), 1 51I1 International Commission on Radiological Protection: Radiology 56, (1951), 431; Br. J. Radiol. 24, (1951), 46 59I1 International Commission on Radiological Protection: Publ. 1, Pergamon Press, Oxford (1959) 64I1 Recommendations of the International Commission of Radiological Protection, as amended 1959 and revised 1962. ICRP Publ. 6. Pergamon Press, London (1964). 77I1 International Commission on Radiological Protection: Publ. 26, Annals of the ICRP 1 (3) (1977) 85I1 Statement from the Paris Meeting of the ICRP: Annals of the ICRP 15 (1985) 91I1 1990 Recommendations of the International Commission on Radiological Protection: Publ. 60, Annals of the ICRP 21 (1991) 95N1 Nagaratnam, A.: Defence Research and Development Organisation, Ministry of Defense, New Delhi - 110 011 (1995) 99I1 International Commission on Radiological Protection; Annual Report (1999): 24-06-2000 02C1 Clarke, R. H.: Int. Zeitschr. f. Kernenergie 47,1, (2002), 20 02C2 Clarke, R. H.: Strahlenschutzpraxis 8, 1, (2002), 45
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2 Biological effects of ionising radiation
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2 Biological effects of ionising radiation
This Chapter describes the effects of ionising radiation on the body. It covers both effects at the subcellular and cellular level as well as on the whole body. Acute effects can result from high radiation doses and in extreme cases can cause severe tissue damage and even death. For most exposures of people to radiation it is low doses that are of most concern. These can give rise to radiation-induced cancer in those exposed and hereditary disease in future generations. The chapter discusses the sources of information on radiation damage and includes estimates of risk for these different effects.
2.1 Introduction Within a few weeks of Wilhelm Conrad Roentgen’s discovery of X-rays on 8 November 1895, for which he received the Nobel Prize for physics in 1901, the potential of the technique for diagnosing fractures and other medical problems had become apparent, but acute adverse effects (such as hair loss, erythema and dermatitis) were also found. Similar undesirable effects were reported shortly after the discovery of radium (by Henri Becquerel in 1896) and its subsequent medical applications. In 1904, the first death of a person exposed to X-rays was reported; X-ray burns had developed into cancer. This death was soon followed by a steady stream of ‘martyrs to science through roentgen rays’ to use the title of a book by a radiologist who subsequently died of cancer. The widespread use of X-rays and radium in treating disease in the early 1900s led to the recognition of a cancer risk in many organs and tissues following high radiation doses which caused gross tissue damage. There was, however, a delay of about 40 years before it became clear that there was a risk of radiation-induced cancer from irradiation at lower doses and that there is no apparent threshold dose below which exposure to radiation can be considered safe. This delay can be attributed to the fact that radiation-induced cancers do not differ in any known way from those occurring naturally or caused by other agents. For many cancers there is also a long interval between exposure and the appearance of the tumour. It is now believed that any radiation dose, whether from external radiation or from incorporated radionuclides, is capable of inducing cancer and that the probability of its occurrence, but not its severity, depends on the radiation dose. Animal studies have shown that an increased incidence of certain types of inherited disorders can also occur in the descendants of irradiated parents. For both cancer and inherited disorders the probability of their occurrence, but not their severity, depends on the radiation dose. In radiological protection terminology they are termed stochastic effects. A second type of damage is seen after exposure of the whole or parts of the body to high doses of radiation between a few gray and a few tens of gray. It is a reflection of impairment of the functional capacity of tissues and is referred to as a deterministic effect. Severity of the damage is related to the extent of radiation exposure and it is assumed that there is a threshold below which the clinically detectable damage does not occur. If damage is extensive death may result. Following radiation exposure in utero serious mental retardation has been observed in the children of the atomic bomb survivors in Japan. Current evidence suggests this phenomenon is deterministic with a threshold related to the minimum shift in intelligence quotient (IQ) that can be measured. This Chapter reviews the sources of information available on the response of the body to radiation damage, and considers the extent to which dose-response relationships can be determined and quantified. Landolt-Börnstein New Series VIII/4
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2.2 Cellular effects 2.2.1 Primary events following exposure to ionising radiation Ionising radiations, hereafter abbreviated to radiation(s), can be classified into directly or indirectly ionising. Charged particles such as alpha particles and beta particles emitted from radionuclides are directly ionising if they have sufficient kinetic energy to disrupt atomic structure. Other types of radiation such as X-rays (generated artificially) or gamma rays (from nuclear transitions) are indirectly ionising. When passing through matter, they give up their energy to the atoms with which they collide and high velocity charged electrons are ejected from these atoms leaving behind positive ions. These electrons move randomly along a trajectory and may ionise other atoms in their path. If this occurs, more electrons are ejected, while the incident electrons continue on their trajectory with decreased energy and velocity, having transferred some of their energy to the newly formed electrons and eventually come to rest. Neutrons also lose energy in various ways, an important means being through collisions with hydrogen nuclei, which are single protons. The protons are set in motion and, being charged, they again deposit energy through electrical interactions. The unique feature of ionising radiation, then, is the localised release of energy in sufficient amounts to alter atomic and molecular structure. The particle track is the ensemble of ionisations (and excitations) along the trajectory of the electron or proton. One way of expressing the amount of atomic disruption is to quote the average energy loss along the track. This is referred to as the unrestricted linear energy transfer (LET or L). LET quantities are given in terms of average energy lost per unit path length, expressed in terms of kiloelectronvolts per micron (keV µm−1). This physical quantity has been used extensively in experimental radiobiology as a simplistic approach in order to relate the quality of radiation to cellular damage [91I2]. The rate of energy loss in biological material can vary greatly along the particle track depending upon kinetic energy and charge. In general terms, photons and electrons have LET values in the range of about 0.2 to 10 keV µm−1; for example, 1 MeV, 100 keV, 10 keV and 1 keV electrons have LET values of 0.2, 0.5, 2, and 10 keV µm−1 respectively. Protons, alpha particles and neutrons have LET values between about 10 and 100 keV µm−1; and heavy charged particles (e.g. nuclei of elements such as C, Ne and Si) can have still higher values to about 2000 keV µm−1. LET does not address the magnitude of the individual energy-loss events that occur along the track; nor does it address the amount of energy lost to matter in the volume of interest. This can be expressed as mean lineal energy which, in concept is more meaningful than LET [93I4]. The random nature of the particle track can be simulated by computer analysis using Monte Carlo techniques. A two-dimensional clustering of ionisations is shown in Fig. 2.1. This is only an approximation of the more complex three-dimensional events that involve random clustering of ionisations on a sub-atomic scale. Nevertheless, the figure illustrates the concept that low energy electrons are sparsely ionising because the ionisations are well separated spatially. Alpha particles, in contrast, are densely ionising because the ionisations are closely packed together along the track. It has been calculated that a single particle track of low-LET radiation (e.g. 1 MeV gamma-rays) passing through an 8 µm diameter spherical nucleus delivers an absorbed dose of about 1 mGy [94G2]. The gamma-rays are about one hundred times less damaging than high-LET radiation, for example 1 MeV neutrons which deliver an absorbed dose of a few hundred mGy in the same shape of nucleus. Each ionisation can result in energy being deposited within the atoms of a target molecule in sufficient amounts to disrupt chemical bonds. Alternatively, it may indirectly break the chemical bonds in a nearby molecule. It is the predominant reaction in water molecules in cells after exposure to X-rays. Free hydroxyl and other related radicals are produced and during their short existence of about a microsecond, these highly reactive radicals are capable of diffusing a few micrometres to reach and damage a target molecule such as deoxyribonucleic acid (DNA).
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Fig. 2.1. Simulated low energy electron track (upper: initial energy 500 eV) and simulated short portion of an alpha particle track (lower: 4 MeV). Large circles are ionisations, small circles are excitations. A Section of DNA is shown to give a perspective on dimensions; [94G2].
2 nm
The temporal sequence of ionisations (and excitations) leading to biological effects is illustrated in Fig. 2.2. Physico-chemical events are completed rapidly, the repair of damage may be completed within tens of minutes while effects in cells can arise within hours or days. The biological manifestations in multi-cellular organisms, including man, can be delayed for many years or, as in the case of hereditary disease, only be manifest in future generations. Tissue and whole body effects Cellular effects
Enzymatic actions ( repair/fixation of damage ) Formation of radicals and radical interactions Ionisations and excitations
1 min
1 year
Fig. 2.2. Timescale of events leading to radiation effects following exposure to ionising radiations.
Seconds
2.2.2 Cellular damage and repair following the primary radiation events It is widely accepted that the most important cellular constituent to be damaged by radiation is nuclear DNA. The molecular structure consists of a double helix (Fig. 2.1), formed from two complementary strands of nucleotides. These are purine and pyrimidine bases linked to sugar molecules with phosphate Landolt-Börnstein New Series VIII/4
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molecules joined by ester linkages. The two strands are held together by hydrogen bonds between guanine-cytosine and adenine-thymine base pairs. The cell’s genetic information is carried in a linear sequence of nucleotides that make up the estimated 100,000 genes in the human genome. Each gene controls a discrete characteristic. Just as cells inherit genes, they also inherit a set of instructions that tell the genes when to become active. These gene regulatory proteins recognise short stretches of nucleotide sequences on the double helix and determine which of the genes in the cell will be transcribed. About two-thirds of genes provide instructions for cell division and for the synthesis of tens of thousands of proteins that provide the structural components of cells, as well as numerous enzymes promoting and controlling cellular activity. Ribonucleic acid (RNA) is the molecule that helps to transport, translate and implement the coded instructions from the genes in the nucleus to the body of the cell. All cell types contain the same genes, but encoding sets of genes is cell-specific. This uniqueness ensures that cells in each tissue produce their own proteins. Maintaining stability in the gene is essential for cell survival. This stability requires not only extremely accurate mechanisms for DNA synthesis and replication, but also mechanisms for repairing DNA damage before replication. Observations with proliferating cells in the laboratory indicate that DNA is subjected to only an estimated few tens of base-pair or nucleotide permanent changes per year during normal metabolism, despite the fact that metabolic processes alter thousands of bases and nucleotides every day. DNA single strand breaks, without base involvement, are effectively ligated enzymatically. Base excision repair pathways require different groups of enzymes that identify and excise the damaged base site, make a complementary copy of the information bases on the opposite undamaged strand, and seal the correct sequence of copied bases in the gap. If nucleotide damage occurs, nucleotide excision repair pathways are able to repair the more extensive damage on one strand. Once the lesion is identified along the strand, the damaged nucleotides are removed and repair proceeds thereafter as for base damage. DNA double strand damage with or without base damage, occurs much less frequently than damage to single strands during normal cellular activity. Recombination repair pathways are available, but they are not totally effective, since there is no undamaged strand to act as a template for base or nucleotide replacement. Damage to bases can result in their alteration or loss. When the repair processes fail, the resulting misrepair is referred to as a mutation. DNA damage due to radiation causes similar lesions to those occurring after normal metabolism, but double strand breaks, multiple gene losses and the translocation of gene sequences occur more frequently as a dose-related effect. The probability of misrepair is greater under these circumstances. Estimated yields of damage caused by low-LET radiation are shown in Table 2.1 [88W1]. Table 2.1. Examples of damage in a mammalian cell nucleus from 1 Gy of low-LET radiation. Initial physical damage Ionisations in cell nucleus ~ 100,000 Ionisations directly in DNA ~ 2,000 Excitations directly in DNA ~ 2,000 Selected biochemical damage DNA single-strand breaks ~ 1,000 Base (8-hydroxyadenine) damage ~ 700 Base (thymine) damage ~ 250 DNA double-strand breaks ~ 40 DNA-protein cross links ~ 150 Selected biochemical damage Lethal events ~ 0.2-0.8 Chromosome aberrations ~ 0.4 Hprt(1) gene mutations 0.6 × 10−5 Translocation frequency (2 loci) 1.2 × 10−4 (modified from 88W1) (1) hypoxanthine-phosphoribosyl transferase Landolt-Börnstein New Series VIII/4
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Recent investigations have revealed that DNA repair pathways may work in conjunction with other intracellular activities in order to minimise cell damage. These include delay in cell-cycling (as a means of maximising the chances of repair); and programmed cell death (apoptosis), whereby severely damaged cells are eliminated to stimulate cell proliferation.
2.2.3 Classification of radiation-induced damage Laboratory techniques have been available for many years to observe radiation effects in proliferating cells. These techniques include measuring changes in cell survival, in the frequency of chromosomal aberrations (deletions and translocations), in gene structure (mutations), and in oncogenic transformation (neoplasia). 2.2.3.1 Cell survival Cellular damage can be classified into three arbitrary categories: lethal damage which results in cell death; sublethal damage, which may be repaired; and potentially lethal damage, defined as damage that can be repaired by altering the growth conditions as for cells in culture. Cell lines of fibroblasts from rodent and human tissues have been used extensively to establish doseresponse relationships [93U6]. Expressed graphically as the logarithm of cell survival plotted against absorbed dose on a linear scale, the dose-response is linear for low-LET radiation at low doses, followed by a curvature at higher doses. Expressed mathematically, the relationship can be represented by a linear-quadratic equation: S = e − (α D+ β D
2
)
(1)
where S is the surviving fraction after exposure to dose D and α and β are coefficients representing the linear and quadratic components for cell killing. The initial slope of the relationship is determined by α, while the quadratic component, reflects the curvature in the dose-survival relationship (Fig. 2.3). The dose at which the linear and quadratic components are equal is the ratio of α and β. The response to highLET radiation is also shown in Fig. 2.3 where survival is best expressed as a linear function of dose passing through the origin. 1 aD b D2
Cell survival
10 -1
10 -2 High LET 10 -3 0 Landolt-Börnstein New Series VIII/4
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a/b 4
8 Dose D [Gy]
12
16
Fig. 2.3. Typical survival curves for cultured cells exposed at high dose-rate (>0.1 Gy min−1). The curves illustrate the linear-quadratic relationship for low-LET radiations and linear relationship for high-LET radiations.
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A plausible explanation of the linear component following exposure to low-LET radiation at low doses is that the majority of DNA interactions are single particle track events [94G2]. Under these circumstances, DNA damage can be effectively repaired. As the dose increases, multi-track events reflecting the quadratic component, and which are associated with clustered DNA damage, increasingly predominate with a consequent increase in the probability of misrepair and lethal events. At 1 Gy, for example, lethal events have a frequency of about 0.2 to 0.8 per cell (Table 2.1). Protracted exposure to low-LET radiation results in less damage, per unit of dose, compared with acute exposure [93U6, 00U8]. This is referred to as the dose rate effect and is due to the ability of cells to repair more sublethal damage as the dose rate is reduced. Below about 1 Gy min−1, the slope on the exponential portion of the survival curve typically becomes progressively shallower as more and more sublethal damage is repaired. Below about 0.01 Gy min−1, undamaged cells are able to proliferate at a sufficient rate to offset the reduction in cell numbers while repair is progressing. This response is illustrated in Fig. 2.4. A dose rate effect is not observed after exposure to high-LET radiation, suggesting little repair of damage. 1
Surviving fraction
0.004 Gy min -1 10 -1 Proliferation
0.01 Gy min -1
10 -2
Repair
1 Gy min -1
10 -3 0
4
8 Dose D [Gy]
12
16
Fig. 2.4. Dose-rate effect showing the influence of repair and repopulation on the dose-survival relationship for cells.
The relative biological effectiveness (RBE) of different types of radiation is defined as the ratio of a dose of a reference low-LET radiation to a dose of the test radiation that gives an identical biological endpoint [90N3]. RBE values are influenced by variations in LET, dose and dose rate. RBE values increase to a maximum at about 100 keV µm−1, decreasing thereafter because of an “overkill effect”. The absolute value of the RBE is not unique but depends on the level of biological damage and, therefore, on the absorbed dose [86B1]. For irradiation by alpha particles, for example, the RBE is generally taken to be 20 for stochastic effects (cancer and hereditary disease) but to have a lower value of around 5 for deterministic effects. 2.2.3.2 Damage to viable cells Chromosome aberrations and gene mutations The technique of culturing human lymphocytes in vitro has been available for many years. It provides a means of measuring the frequency of unstable and stable chromosome aberrations at various stages in the cell-cycle. In terms of unstable aberrations, their frequency increases from a background level of about 10−3 to a rate of about 4 × 10−2 Gy−1 after exposure to low-LET radiations. Dose-response relationships for different types of radiation are illustrated in Fig. 2.5 [89E1]. Neutrons are more damaging than X-rays Landolt-Börnstein New Series VIII/4
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or gamma-rays and low energy neutrons are more damaging than high energy neutrons. For low-LET radiations a linear-quadratic relationship is consistent with the data. That is: E = αD +βD2
(2)
where E is the frequency of chromosome aberrations (i.e. a stochastic effect), D is the dose, and α and β are the linear and quadratic coefficients for the induction of the aberrations. 1.0
Dicentrics per cell
Fission neutrons
0.5
250 kVp X - rays
0 0
1
2
3
Fig. 2.5. Dicentric yield in chromosomes per cultured human lymphocyte as a function of dose for selected radiations; [89E1].
Dose D [Gy]
A number of specific-locus mutation test systems using mouse, hamster and human fibroblasts have been developed to measure mutagenesis. One cell line, the human B-lymphoblastoid TK6, illustrates the use of the test [89K4]. Cultured cells were exposed to radiation and the mutation frequency at two loci (hgprt and tk) was measured under different exposure conditions. For acute radiation exposure, 100 kVp X-rays (0-2 Gy) and (Pu, Be) neutrons (0-0.2 Gy) both showed a linear dose-response relationship in terms of induced mutants. The induced mutant frequency per 0.01 Gy per surviving cell was 0.55 × l0−7 (CI 0.09) and 1.92 × l0−7 (CI 0.03) respectively. Protracted exposure to X-rays (0.01-0.1 Gy per day) for 5 to 20 days showed a slight increase in the mutation frequency (0.84 × l0−7 (CI 0.17)); while continuous exposure to neutrons (0-0.4 Gy) resulted in a substantial increase (6.00 × l0−7 (CI 0.7)). These data demonstrate an ‘inverse dose-rate effect’ for neutron-induced mutation in human cells. Syrian hamster embryo cells showed a similar effect, but other cell lines did not. It is concluded that there are a number of difficulties in interpreting the results of somatic cell mutations. Estimated yields of chromosome aberrations and mutation frequency are shown in Table 2.1. Cell transformation An established technique for studying carcinogenic potential is that of culturing cells that can grow indefinitely, provided that they are frequently transferred to fresh media. Under specific conditions, cells that have acquired this ability are said to be immortalised. A characteristic of these immortalised cells is that they stop dividing when they come into contact with similar cells in the culture medium (contact inhibition). They are not classified as malignant cells because they do not cause tumours when injected into immunologically-suppressed animals. Occasionally, an immortalised cell undergoes a spontaneous change, whereby it loses its contact inhibition and continues to proliferate by spreading over adjacent immortalised cells to form a recognised foci of cells. Such cells are said to have undergone Landolt-Börnstein New Series VIII/4
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transformation and when they are injected into animals, they develop into tumours. Spontaneous transformation is a rare event, occurring at a rate of about one in ten thousand to one in a hundred thousand per surviving cell. The mechanism is not fully understood, but it is thought to involve the mutation of two or more genes. Two classes of mutated genes in particular have been identified and characterised. These are ‘gain-of function’ mutations of proto-oncogenes, whereby the mutated genes (oncogenes) stimulate cell proliferation in an uncontrolled manner; and ‘loss-of function’ tumour suppressor genes, whereby cells are no longer prevented from proliferating in defiance of normal controls. The cell types used in transformation studies are mainly derived from fibroblasts. It is generally accepted that the sensitivity of the test is low, the detection limit being about 0.25 Gy. Ideally, human epithelial cells would be a better choice to represent human cancers. Future studies are in hand which aim to use this type of cell. To illustrate the technique, C3H/1OT½ fibroblasts derived from the prostate of the C3H mouse embryo, were irradiated with low-LET radiation [98M2]. Cell survival and transformation frequencies were simultaneously measured. The survival curve was consistent with a linear-quadratic dose-response relationship (Equation 1), while the transformation frequency per surviving cell following exposure to X-rays was consistent with a linear relationship (Fig. 2.6a). However, if the number of transformants per cell at risk was plotted, the relationship to intermediate doses was consistent with a linear-quadratic equation, the transformation frequency reaching a maximum at about 2 Gy (Fig. 2.6b). This doseresponse relationship is consistent with other results reported in the literature, although the maximum transformation frequency was usually in the 3 to 4 Gy range. 6
2.0
Transformation frequency [×10 -4 ]
Transformation frequency [×10 -4 ]
5 4 3 2 1
1.6
1.2
0.8
0.4 Cells at risk
Cells viable for survival 0 0
a
1
2 3 4 Absorbed dose D [Gy]
5
0
6
0
b
1
2 3 4 Absorbed dose D [Gy]
5
6
Fig. 2.6. Transformation frequencies per surviving C3H 10 T½ cell (a) and per cell at risk (b) as a function of absorbed dose after exposure to 250 kVp X-rays at 2 Gy min−1; [98M2].
Exposure to neutrons resulted in a higher transformation frequency than for low-LET radiation, with no evidence of a dose rate effect. One exception was a study of 5.9 MeV or fission neutrons where an inverse dose-rate effect was reported [93U6]. It is concluded that there are difficulties in interpreting data on cell transformation studies. Generalised dose-response relationships The conventional approach to representing the absolute biological effectiveness of a given radiation at low doses is based on the assumption derived from target theory in which the induction I of an effect as a function of dose D can be represented by Landolt-Börnstein New Series VIII/4
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I (D )= α1 D + β1 D 2 e − (α 2 D+ β 2 D
(
)
2
)
2-9
(3)
in which α1 and β1 are single and multihit components for a radiation effect and α2 and β2 represent single and multihit components for cell killing. At low doses the incidence from effect is determined by α1 with the response increasing linearly with dose. It is generally assumed that in this region α1 will be independent of dose rate. With increasing dose the amount of damage due to multihit effects increases, resulting in a quadratic component in the dose-response curve. At doses above a few gray β1 and β2 become significant resulting in a reduction in tumour yield due to the effect of cell killing. For high-LET radiation (α particles, neutrons) the dose-response curve is generally found to be linear up to the point at which cell killing starts to exert an effect and reduces the tumour yield.
2.2.4 Implications of cellular damage for whole or partial body exposure The outcome of cell damage in terms of human radiation detriment can be profoundly different according to the exposure conditions. Cellular studies can provide a sound basis upon which to judge these outcomes. After acute exposure to absorbed doses above a few gray, the cells at greatest risk are selfreplicating stem cells that supply functional cells. They are programmed to divide so that one daughter cell remains as a stem cell (in order to ensure that stem cell numbers in the tissue remain constant), while the other daughter cell proceeds to specialise (differentiate) by clonal expansion. If sufficient numbers of stem cells in a tissue are killed or are prevented from dividing at the appropriate rate, the tissue loses its ability to function effectively. The consequential effects are referred to as deterministic. Studies have established that cell survival is dose and dose-rate dependent for low-LET radiations, and that there is a tissue-specific dose threshold. At high risk are rapidly dividing bone marrow stem cells, and stem cells in the epithelium of the gastrointestinal tract, lungs, thyroid, gonads, skin and lens of the eye. The effects due to the proliferation of mutated cells at low doses are termed stochastic. There is sufficient radiobiological evidence for low-LET radiation to support the general assumption of an increasing risk of an effect with increasing dose at low to intermediate doses, with no threshold. Cellular techniques are providing insight into the way in which radiation can initiate the complex multistage process of carcinogenesis. However, there is still much to be learned about the molecular changes that lead to cells with the potential towards malignancy; and most importantly, any advances in knowledge at the cellular level have to be seen in the context of the living organism.
2.3 Deterministic effects 2.3.1 Tissue and organ development In the space of a few weeks, a single fertilised human egg gives rise to a complex multicellular organism consisting of embryonic cells arranged in a precise pattern, each in its proper place. In the subsequent period of fetal growth, the cells continue to proliferate in the developing tissues and organs. Growth of tissues and organs continues in childhood with increase in cell mass in many tissues, but growth essentially ceases in the adult when cell masses reach a predetermined size. The majority of cells in tissues of the adult are differentiated, that is, they have developed specific morphology and function which is usually irreversible, but these cells are predestined to die. In many tissues of the body, the rate of death of differentiated cells is rapid and, in a healthy state, must be balanced by proliferation from stem cells. These cells, by definition, are cells that have retained Landolt-Börnstein New Series VIII/4
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embryonic characteristics. They are able to divide during the lifetime of the organism, yielding progeny that are destined to differentiate by a process of clonal expansion. Stem cells also retain the ability of selfrenewal. These characteristics are illustrated in Fig. 2.7. The number of stem cells compared to differentiated cells varies according to the tissue, but they usually represent, at most, a few percent of the total cell numbers. Furthermore, only a small fraction of the stem cells are active at any one time under normal circumstances. It is not known how the balance between cell proliferation and cell death is achieved, but it is thought that all cells are genetically programmed to die, by apoptosis. When differentiated cells die, a feed-back mechanism is activated to stimulate the stem cells to divide and replenish the population.
Stem cell
S
Self - renewal
Clonal expansion Maturation without division Functional cells with finite lifetime
Fig. 2.7. Derivation of differentiated cells from a selfrenewing stem cell.
If enough stem cells in a tissue are killed or prevented from undergoing cell division, there will be loss of tissue function; termed deterministic by the International Commission on Radiological Protection (ICRP). The dose-response relationship is characterised by a frequency and severity that increases with dose above a threshold. Most tissues and organs of the body are able to compensate for small reductions in the number of differentiated cells. But if the decrease is large enough, there will be changes seen as loss of tissue or organ function and a consequential response to repair the damage.
2.3.2 Dose-response relationships for radiation damage The probability of detecting loss of tissue or organ function following exposure to radiation increases steeply above a threshold dose to a maximum. Expressed as a generalised dose-response relationship, the plot of the frequency of the effect versus dose expressed on linear axes is sigmoid (Fig. 2.8, upper panel). Above the threshold dose, the severity of the effect also increases with dose reflecting more cell loss and hence damage to tissue function (Fig. 2.8, lower panel). Protracting the dose results in a lower frequency and less severe symptoms at a given dose compared with acute exposure, demonstrating the importance of stem cell repopulation. There is individual variation in radiosensitivity in any exposed population. This variation reflects differences in the ability of individuals to cope with radiation-induced cellular damage. Any response is influenced by the age and state of health of the exposed individual.
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Frequency [%]
100
50
0
Seveity
Variation in population
Threshold
Fig. 2.8. Dose-response relationship for deterministic effects. Variation in frequency and severity; [based on 91I2]. Dose
2.3.3 Deterministic effects in humans following acute whole-body irradiation Evidence of the deterministic effects of radiation comes from several sources. These include retrospective studies on radiotherapy patients, radiologists in the early part of the 20th century, Japanese populations exposed to radiation from atom bombs, and individuals accidentally exposed to high doses following nuclear reactor accidents and radiographic sources. Understanding the effects of acute high doses is important as an aid to prognosis in the treatment of accidental over-exposure, and to ensure that deterministic effects are avoided in normal practices and minimised in accidents. Evidence on deterministic effects also comes from studies with animals. After exposure to doses of a few Gy, the depression in the numbers of circulating white blood cells (granulocytes) and blood platelets may be so severe as to result in death from septicaemia (infection) and haemorrhage. This is referred to as the haematopoietic syndrome. Recovery depends upon the radiation dose and the ability of the remaining stem cells in the marrow to recover. Loss and recovery of granulocytes and blood platelets follows a similar dose- and time-related pattern. Depression of the stem cells providing the protective mucosal cells lining the intestinal tract wall results in a denuding of the gut surface. This gastrointestinal syndrome is seen in individuals who have received doses to the gastrointestinal tract in excess of about 5 Gy. Leakage of blood from damaged blood capillaries results in severe anaemia and ingress of intestinal bacteria through the damaged blood vessels results in septicaemia. The haematopoietic syndrome will manifest itself concurrently at these higher doses. Landolt-Börnstein New Series VIII/4
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Damage to endothelial cells lining the alveolar air sacs may result in acute inflammation of the lungs (pneumonitis) at doses in the range 5 - 15 Gy. This may occur after radiotherapy or after the inhalation of high specific activity radioactive particles. If the individual survives the pneumonitis, lung fibrosis may later develop which can also be life-threatening through loss of lung function. At higher whole-body doses (>15 Gy), generalised shock occurs affecting the brain and the cardiovascular system. Coma and death develop rapidly thereafter. The range of doses associated with death from these syndromes after acute exposure to low-LET radiation is given in Table 2.2. The ranges are based upon human data, supplemented by knowledge of the form of the dose-response relationship derived from animal experiments. No individual would be expected to die after receiving absorbed doses below about 1 Gy. The dose range where half the exposed population would be expected to die without medical treatment is 3 to 5 Gy. Death would be likely at doses between about 6 Gy and 10 Gy, unless they receive treatment to prevent infection and bleeding. Above about 10 Gy death is assumed at present to be inevitable, even after attempts to stimulate the bone marrow or bone marrow transfusion from a suitable donor. These estimates of lethality do not take account of any concurrent radiation-induced damage (e.g. skin burns), or existing debilitating diseases. Table 2.2. Range of doses associated with acute radiation syndromes in adults exposed to low-LET radiation. Whole body absorbed dose 1-6 Gy 5-15 Gy >15 Gy
Principal effect contributing to death Damage to bone marrowa Damage to the gastrointestinal tract and lungsb Damage to nervous system and shock to the cardiovascular system
Time of death after exposure [days] 30-60 10-20 1-5
a) Dose range considered to result in 50 % of an exposed population dying (LD50) 3-5 Gy. b) Damage to vasculature and cell membranes especially at high doses is an important factor in causing death.
2.3.4 Deterministic effects following partial body irradiation 2.3.4.1 Tolerance doses in adults after radiotherapy Extensive experience in the treatment of patients undergoing radiotherapy has provided data upon which to determine the tolerability of healthy tissues and organs to radiation. Called the tolerance dose by clinicians, it is defined as the amount of radiation received during conventional treatment below which unacceptable effects do not occur in more than a few percent of patients within 5 years following treatment. The tolerance doses for some adult tissues are shown in Table 2.3 (children are usually less tolerant to exposure). It is evident that the gonads, lens of the eye and the bone marrow are the most radiosensitive. 2.3.4.2 Threshold doses in radiological protection The limitations of using data on tolerance doses to derive threshold doses for radiological protection purposes need to be recognised. In contrast to the precise exposure conditions of radiotherapy, exposure of workers to high doses of low-LET radiation is most likely to be non-uniform and resulting from mixed radiations. The tolerance dose therefore can at best be used as a cautious approximation to a threshold dose.
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Table 2.3. Tolerance doses for deterministic effects in adults after fractionated radiotherapy treatment Organ Effect Tolerance dose [Gy] Total bone marrow Blood cell depletion 1-2 Ovary Permanent sterilisation 2-6 Testis Permanent sterilisationa 3-4 5-10 Eye Cataractb Kidney Nephrosclerosis 23 Liver Loss of function, ascites 35 Lung Pneumonitisc 40 Heart Pericarditis 40 Lymph nodes Hypoplasia, fibrosis 35-45 Thyroid, pituitary Hypoplasia >45 Other organs Hypoplasia, fibrosis >45 a) A significant but reversible, depression of sperm count occurs after about 0.1 Gy brief exposure. b) About 2 Gy after a brief exposure. c) LD50 after brief exposure is about 10 Gy.
The threshold doses recommended by the ICRP for the most radiosensitive tissues and organs are summarised in Table 2.4. Thus the threshold dose for temporary sterility in the male for a single absorbed dose in the testes is about 0.15 Sv. Under conditions of prolonged exposure, however, the dose rate threshold is about 0.4 Sv y−1. The corresponding values for permanent sterility are about 3.5 Sv and 2 Sv y−1. The threshold dose for permanent sterility in women for a single absorbed dose is in the range from about 2.5 Sv. For protracted exposure, the dose rate threshold is about 0.2 Sv y−1. Clinically significant depression of the blood-forming process occurs above a single bone marrow dose of about 0.5 Gy. The dose rate threshold for protracted exposure is about 0.4 Gy y−1. The tolerance dose for death is in the range of 6 to 7 Gy if the radiation is spread over 30 fractions in a period of 6 weeks. Table 2.5 summarises the principal syndromes associated with whole body exposure. Table 2.4. Estimates of the thresholds for deterministic effects in adults recommended in radiological protection [91I2]. Equivalent dose rate Equivalent dose brief Tissue and effect exposure [Sv] protracted exposure [Sv y−1] Testes Temporary sterility 0.15 0.4a Permanent sterility 3.5-6.0 2.0 Ovaries Sterility 2.5-6.0 >0.2 Lens Detectable opacities 0.5-2.0 >0.1 Visual impairment (cataract) 5.0c >0.15 Bone marrow Blood cell depletion 0.5 >0.4b a)
This dose is higher because differentiating cells are more radiosensitive than the stem cells so the latter can replenish the differentiating cells at an adequate rate. b) Supported by evidence of effects after chronic radiation of Beagle dogs. c) Range 2-10 Sv.
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Table 2.5. Summary of acute radiation syndrome Syndromes Intestinal Cerebral
Bone marrow
Critical organ Latent period Syndrome threshold [Gy] Death threshold [Gy] Death occurring within Cause of death
Bone marrow 2-3 weeks 1 2 3-8 weeks Haemorrhage, infection
Prodromal vomiting Symptoms
Brain 20 min 20 50 2 days Cerebral oedema, heart failure Minutes Tremors, cramps, loss of coordination, lethargy, impaired vision, coma
Treatment
Palliative
Prognosis
Hopeless
Small intestine 3-5 days 3 10 2 weeks Sloughing of gut, shock
1 hour A few hours Loss of appetite, vomiting, Fever, breathlessness, internal diarrhoea with bleeding, bleeding, depletion of bone fever, electrolyte and fluid marrow leading to low blood balance counts Barrier nursing, fluid and electrolyte replacement, transfusions of blood cells, bone marrow transplants Very poor Dose-dependent and influenced by treatment
2.3.4.3 Skin irradiation Based upon extensive experience in the use of fractionated X and gamma radiation in radiotherapy, (typically, 20 to 30 fractions each of 2 to 6 Gy over several weeks), various degrees of skin damage can be observed according to the area and depth of skin involved, the absorbed dose and the duration and frequency of the exposure. The earliest observable change is a transient reddening within a few hours after exposure to doses above about 2 Gy; due to increased capillary permeability. This is followed after moderate doses (about 5 Gy) two to four weeks later by a persistent reddening (the main erythematous reaction) and peeling of skin (dry desquamation). This is due to secondary inflammation resulting from the death of basal (stem) cells of the epidermis. Hair loss also occurs. At higher doses (about 20 Gy), blistering (moist desquamation) occurs after about four to six weeks due to the inability of basal cells in the irradiated area to divide and for viable basal cells to migrate into the area at a sufficiently rapid rate. It is the health effect to be avoided in both radiotherapy and radiation protection practice. The threshold doses for moist desquamation depends upon the area irradiated and the penetrating powers of the radiation. Ulceration is the result of infection following moist desquamation and may occur after about 6 weeks. Necrosis due to irreversible damage to the basal cells of the dermis and the underlying blood vessels occurs within two to three weeks after doses of tens of Gy. Late effects developing months to years later include changes in pigmentation; atrophy of the epidermis, sweat glands and sebaceous glands and hair follicles; and fibrosis. Quantifying the threshold doses for these effects is complicated in practice by the multiplicity of targets at different critical cell depths, which makes it difficult to select a single depth at which to specify the dose to the skin. The depths at which the most serious effects arise are estimated to be in the range of 300 - 500 µm. However, a conservative approach for protection purposes is to use shallower depths (20 100 µm, typically 70 µm) for monitoring specifications. To prevent moist desquamation, the dose must be reduced as the radiation field is increased. To illustrate the importance of field size, the tolerance doses following a single treatment with orthovoltage X-rays was found to be 20 Gy for an area of 6 × 4 cm, and 11 Gy for an area of 15 × 20 cm. Following fractionated treatment, the tolerance doses were estimated to be about 50 Gy and 30 Gy respectively for the two field sizes. From experimental studies, the estimated dose threshold following exposure of large areas of skin is about 20 Gy; and no acute tissue breakdown was observed at a dose rate of 0.4 Gy h−1 with total doses of about 100 Gy. Accidental over-exposure of industrial radiographers is a cause for concern in radiation protection. In normal practices, ICRP recommends a limit on effective dose of 20 mSv per year, averaged over 5 years with the further provision that the effective dose should not exceed 50 mSv in a single year [91I2]. This limitation is on effective dose and is assumed to be adequate to prevent deterministic effects. However, an additional annual limit is recommended for localised exposures in order to prevent deterministic effects to the skin. It is 500 mSv averaged over any 1 cm2 regardless of the area exposed. Landolt-Börnstein New Series VIII/4
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2.4 Radiation-induced cancer 2.4.1 Cancer development The development of cancer is the major late effect resulting from exposure to radiation. Cancer is generally understood to develop in a number of stages. That is, for malignancies to be expressed a series of events must occur in cells and the rate at which they occur is thought to be reflected in the way cancers appear in the population over the course of time. The development of cancer in tissues is a complex, multi-stage process that can be sub-divided into four phases: neoplastic initiation; neoplastic promotion; conversion and progression. The sub-divisions are necessarily simplifications of the overall process which is, in any event, somewhat variable between different tumour types. However, they do provide a basis from which to interpret the cellular and molecular changes involved [93U6, 00U8]. Neoplastic initiation encompasses the essentially irreversible cellular damage, which although not necessarily expressed immediately, provides the potential in cells for the development of cancer. There is good evidence that this initiation process results from damage to DNA leading to gene mutations in single target cells in tissues. The critical damage is likely to be coincident damage to both DNA strands (DNA double strand breaks, Section 2.2.2). Although a proportion of such double strand damage will be repaired, completely error free repair of such damage, even at low doses, is not expected. Neoplastic promotion can be seen as a process whereby initiated cells receive an abnormal growth stimulus and begin to proliferate in a semi-independent manner. Conversion of these pre-neoplastic cells to a form in which they are committed to become fully malignant is a central feature of the process of neoplastic development. Such changes are now believed to be driven by further gene mutations accumulating within the expanding population of pre-neoplastic cells. Once the potential for full malignancy has been established, the subsequent progression of the disease may depend upon further cellular changes that allow invasion of adjacent normal tissues, the circulation of neoplastic cells in the blood and lymphatic systems and the establishment of metastases (secondary tumour growths) at other sites in the body. It is this invasive process that provides principally for the fatal effects of most common human tumours. On this basis, a single mutational event in a critical gene in a single target cell in vivo can create the potential for neoplastic development. Thus, a single radiation track traversing the nucleus of an appropriate target cell has a finite probability, albeit very low, of generating the specific damage to DNA that results in a tumour initiating mutation. These initiated cells can then develop by multistage processes into an overt malignancy. As a consequence, at the level of DNA damage, there is no basis for assuming that there is likely to be a dose threshold below which the risk of tumour induction would be zero. For radiation protection purposes, a progressive increase in risk with increasing dose, with no threshold, is therefore assumed [95C2]. Whilst such a multistage mechanism is considered to be the cause of many human tumours there are likely to be some tumours that may arise in tissues where there has been deterministic damage (fibrosis) for such tumour types a threshold dose may need to be exceeded before the tumour will occur. There are many examples of such tumour types in animals and the development of radiation-induced bone tumours in man may also require a threshold dose to be exceeded [00U8]. Radiation appears to be capable of causing tumours in nearly all tissues of the body, although the frequency of appearance following a unit dose may vary markedly from one tissue to another. Information on the dose related frequency of tumour induction by radiation is gained through follow-up of groups of persons exposed to radiation. The observed tumour frequency can then be compared with an age and sex matched control group, not exposed to radiation, to determine the increase in frequency due to the radiation exposure. Extensive follow-up studies have been carried out on groups of persons exposed to either external radiation or to internally incorporated radionuclides.
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Tumours induced by radiation are in general indistinguishable from those occurring spontaneously and since cancer is not uncommon (about one in five die as a result of it in Western Europe and North America), the problem of determining a relatively small excess due to radiation exposure is difficult. In general, large exposed populations are necessary to obtain statistically meaningful results. The chief sources of information on the risks of radiation-induced cancer are the A-bomb survivors exposed to whole-body irradiation in Hiroshima and Nagasaki, patients with ankylosing spondylitis and other patients who were exposed to partial-body irradiation therapeutically, either from external radiation or internally incorporated radionuclides, and various occupationally exposed populations, such as uranium miners and radium-dial painters. Some quantitative information on thyroid cancers may also be obtained following the Chernobyl accident. Increasingly information is becoming available from epidemiological studies on groups of persons occupationally exposed to radiation. In general, however, the radiation exposures in these populations is relatively low and there is limited power in the studies to obtain quantitative estimates of risks of radiation-induced cancer. Reports by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) provide a comprehensive review of the data available [94U7, 00U8]. In it’s 1990 recommendations, the ICRP re-assessed the epidemiological data and this resulted in an increase in estimates of the lifetime risk of radiation-induced cancer. Partly, this arose as a result of revised dosimetry for the A-bomb survivors and a longer follow-up of the population, but mainly it was attributed to a change in the model now used to project lifetime risks [91I2]. Similar calculated values of lifetime risk have been published by UNSCEAR [94U7, 00U8].
2.4.2 Dose-response relationships 2.4.2.1 Assessment of lifetime risk There is always a minimum period of time between irradiation and the appearance of a radiation-induced tumour. This period is termed the latent period and its length varies with age and from one tumour type to another. Some types of leukaemia and bone cancer have latent periods of only a few years but many solid tumours have latent periods of ten or more years. For leukaemia and bone cancer there is fairly good evidence that the risk is almost completely expressed within about twenty-five years following exposure. For solid tumours of longer latency, such as those of the GI tract , liver and lung it is not yet clear whether the incidence of these tumours passes through a maximum and declines with time following exposure, whether the risk levels out, or alternatively increases indefinitely during the remainder of life. To project the overall cancer risk for an exposed population, it is therefore necessary to use models that extrapolate over time data based on only a limited period of the lives of the individuals. Two such projection models have generally been used: (a) the additive (absolute) risk model which postulates that radiation will induce cancer independently of the spontaneous rate after a period of latency, variations in risk may occur due to sex and age at exposure as well as the tissue exposed. (b) the relative (multiplicative) risk model in which the excess (after latency) is given by a constant (or time-varying) factor applied to the age dependent incidence of natural cancers in the population. In most cases the spontaneous risk of cancer increases with age and therefore the relative risk model will predict an increasing incidence of radiation-induced cancer with age. This model also gives different risks of radiation-induced cancer in different populations, depending on the national cancer incidence. Data available from the A-bomb survivors in Japan and from studies on uranium miners suggest the relative risk projection model gives a better fit to the data, at least for some of the most common cancer types. Despite this there are indications from a number of exposed groups that the risk of cancer starts to decline many years after exposure. This has been well documented for leukaemia, but has also been Landolt-Börnstein New Series VIII/4
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observed in the case of bone cancers (patients in Germany given 224Ra), thyroid cancers (US follow-up study after thymus irradiation), solid cancers (patients treated for ankylosing spondylitics) and possibly lung cancers in the uranium miners exposed to radon and its decay products [88U5; 00U8]. These results suggest that for the Japanese population the excess risk may ultimately decrease with time and thus relative risk projection models applied over a lifetime could result in an overestimate of the cancer risk. 2.4.2.2 Effects of dose and dose rate The total radiation dose and the dose rate both influence cancer induction and are linked to the form of the dose-response relationship. For radiological protection purposes tumour induction is generally assumed to increase with increasing dose, with no threshold, as indicated above. However, studies using cells in culture reveal that for many endpoints, including mutation, the dose-response for exposure to low-LET radiation is not linear, but that the effectiveness of radiation, per unit dose, increases as the dose increases. At very low doses, where there is a low probability of more than one radiation event occurring in a cell nucleus it may be expected that the effect is linearly related to dose. At higher doses, where multiple ionising events within a single cell are commonplace, damage arising from interactions between two or more events becomes more probable. Ultimately, at high doses cell killing will progressively reduce the risk of tumour induction. For single (acute) radiation exposures cell killing starts to become significant at doses of a few gray. The generalised dose-response is given in Equation 3 (Section 2.2.3.2). The difficulty in assessing risks of cancer following exposures to low-LET radiation at low doses and dose rates is illustrated in Fig. 2.9. This gives, schematically, data points and possible dose-response curves for cancer induction. Frequently, as in this example, information is only available at relatively high doses. An approach commonly used in risk assessment is to fit a linear dose-response relationship to the data (curve B) a procedure usually considered to give an upper limit to the risk at low doses. This will be the case unless significant cell killing has occurred. If this linear relationship is due to single tracks acting independently then the effect per unit dose would be expected to be independent of dose magnitude and dose rate. In practice, however, this is not generally observed and the linear quadratic relationship (curve A) frequently gives a better fit to the data at low to intermediate doses implying that at higher doses damage is the result of both single and multiple tracks. At still higher doses cell killing becomes significant with a consequent reduction in tumour yield.
ì í Low LET î
A - High absorbed doses and high dose rates B - Linear, no threshold C - Low dose rate D - Limiting slope for low dose rate
Induced incidence of cancer
B
High LET
A
C D
Absorbed dose D [Gy]
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Fig. 2.9. Dose-response relationship for radiationinduced cancer: possible inferences are illustrated in extrapolating data available at high doses and high dose rates to response at low doses and dose rates for lowLET radiation; [based on 90N2].
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With a progressive lowering of the dose and the dose rate, allowing more opportunity for repair of damage, curve C might be obtained. A point may ultimately be reached at which multiple track events make a negligible contribution to tumour incidence and damage is produced only as a result of single tracks acting alone giving a linear response (curve D) with the effect directly proportional to dose (slope α1, the risk coefficient). A similar response would be obtained by lowering the dose rate alone as even with high total doses the rate of build up of lesions would be slower and the opportunity for multiple track events would decrease. Hence in the limit, curve D, could be achieved either by reducing the dose to very low values so that effects are independent of dose rate or by reducing the dose rate to very low values. The approach used for assessing risks at low doses and low dose rates of low-LET radiation is described in Sections 2.4.5 and 2.4.6. For high-LET radiation it is assumed that there is no dose rate effect and the response is proportional to dose for doses below those at which there is cell killing. The data on the A-bomb survivors provide information on risks of cancer in a range of tissues, although to date no quantitative information is available for radiation-induced cancers of the liver, cells on bone surfaces, thyroid and skin. Information on radiation-induced cancer in these tissues is, however, available from other epidemiological studies summarised in Table 2.6. The principal studies used to quantify the effects of both external radiation and internally incorporated radionuclides are summarised below. Table 2.6. Human populations available for risk estimation Atomic bombs Medical diagnosis Medical therapy
Occupational exposure Radiation accidents
Japanese survivorsa Marshall islandersa,b Multiple fluoroscopies (breast)a Prenatal irradiationa Thorotrast injectionsc Pelvic radiotherapy (cervix)a Spinal radiotherapy (ankylosing spondylitis)a Neck and chest radiotherapy (thyroid)a,b Scalp radiotherapya Radium treatmentc Uranium minersc Radium ingestion (dial painters)c Radiation workersa,b,c Chernobyla,b
a) Exposure to external radiation b) Internal exposure to β/γ internal emitters c) Internal exposure to α emitters
2.4.3 Exposures to external radiation 2.4.3.1 The A-bomb survivors in Japan The mortality experience of the Hiroshima and Nagasaki A-bomb survivors has been the single most important source of information on the risk of radiation-induced cancer. This population has been the subject of a comprehensive follow-up since 1950. Information is available on the exposure of individuals to whole body radiation at a range of ages. Data on mortality from radiation-induced cancer that became available in the 1980s on the population of more than 90,000 people in the Life Span Study (LSS) necessitated a revision of previous risk estimates [87P3, 90S3]. There were a number of components to this change. The first was a revision of the dosimetry (termed DS86) to allow, amongst other factors, for the high humidity in the air over the cities which has substantially reduced the neutron dose at Hiroshima from the earlier 1965 (T65) estimates which were based on measurements in the dry atmosphere of the Nevada desert. Improved estimates were also made of the yield of the Hiroshima bomb (increased from 12.5 to 15 ktonnes), the shielding provided by buildings and of tissue and organ doses. The second was Landolt-Börnstein New Series VIII/4
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that the number of excess fatal cancers in the population increased due to the longer period of follow-up (to 1985) and an estimate of the cancers occurring in the period 1945-1950 was made. The third, and most significant change, was that relative risk, rather than additive risk models appeared to provide a better basis for assessing lifetime risk of most solid cancers (Section 2.4.2.1). UNSCEAR [88U5] in a report to the General Assembly provided the first information on radiation-induced cancer risks for a number of tissues in the Japanese population based on relative risk projection models. The total cancer risk in the population, at high dose and high dose rate, was then estimated to be 7-11 × 10−2 Sv−1 using age-averaged and age-specific constant relative risk models. This compared with the Committee's 1977 assessment of 2.5 × 10−2 Sv−1 [77U2] at high dose rate using the additive model. Because children and young persons are more sensitive to radiation than adults. The application of age specific risk coefficients therefore increases the predicted numbers of radiation-induced cancers in the whole population compared with that for a working population (Section 2.4.6). These risk estimates for whole body radiation exposure were based on an extrapolation into the future which is somewhat uncertain for solid cancers because two-thirds of the Japanese survivors were still alive and two-thirds of the cancer risk had still to be expressed. Up to 1985 about 80 excess leukaemias and 260 excess solid cancers had occurred in the LSS population for whom DS86 doses were available out of a total of about 6000 cancer deaths [87P3]. The risk of radiation-induced leukaemia is more certain than that for solid cancers, however, as few more excess cases are now expected. There remain uncertainties in extrapolating the cancer risks based on the Japanese population exposed to radiation at high dose rates to the low doses and dose rates relevant for radiological protection purposes (see Section 2.4.5). In a more recent report on the LSS, Pierce et al [96P1] reported on five more years of follow-up (1986-1990). Their analysis included an additional 10,500 survivors (86,572 in total). During 1950-1990 there have been 7827 cancer deaths, of which it is estimated there are 87 excess leukaemias and 334 solid cancers. The mortality curve for all solid cancers combined shows essentially a linear dose-response in the range 0-3 Sv, whereas for leukaemia the trend in dose is non-linear with an upward curvature. The radiation-induced leukaemia risk seems to have been almost completely expressed during the follow-up period, and the lifetime excess absolute risk of leukaemia associated with an acute dose of 1 Sv has been estimated as being about 1 %. However, in contrast to leukaemia, nearly a quarter of the radiation-induced solid cancers are estimated to have arisen in the most recent five-year period of the mortality follow-up, i.e. 1986-90 [96P1]. Since most of the A-bomb survivors exposed at young ages are still alive, the future pattern of cancer risks in this group will be important in determining lifetime risks. A significant increase in the risk of solid cancers is now seen at doses down to about 50 mSv [00U8]. 2.4.3.2 Thyroid cancer A number of epidemiological studies provide information on cancer risks in individual tissues. Groups of children and young persons who received thyroid irradiation, and who can be used to derive risk coefficients for thyroid cancer, include children who received X-ray treatment for thymic enlargement, patients treated in US hospitals for thyrotoxicosis and other benign lesions of the neck and patients who received X-ray treatment for thyroid disease [85N1, 85S4, 00U8]. In the majority of cases, particularly in the young, thyroid cancer is not fatal. The mortality from radiation-induced thyroid cancer is expected to be about 10 % of the incidence. There is also evidence that the risk in adults is about half that in children and that the risk in females is about twice that in males. For a population uniformly exposed to external radiation the risk of fatal thyroid cancer is estimated to be 8.0 × 10−4 Sv−1 assuming a 5 year latent period [91I2]. In human populations given iodine-131 for non-therapeutic reasons, and who received doses well below 2 Gy, no significant excess of thyroid cancers has been observed. This suggests a risk coefficient 3 to 4 times less than that obtained following external radiation at high dose rates [85N1]. Data on thyroid cancer incidence in children in areas of the former Soviet Union that were contaminated with fall-out from Chernobyl indicate an increased risk of thyroid cancer in some areas. To date the data are insufficient to provide quantitative risk estimates [00U8]. Thyroid cancer risks from exposures to radioiodine are considered further in Section 2.4.4. Landolt-Börnstein New Series VIII/4
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2.4.3.3 Skin cancer An ICRP Task Group [91I3] reviewed data on the risks of skin cancer. Most of the data came from groups given partial body irradiation in the course of medical treatment, although some data were also available from occupationally exposed groups, in particular radiologists and radiation technicians and uranium mining populations. Little information is available from the A-bomb survivors. On the basis of a relative risk model, the Task Group calculated a risk of fatal skin cancer for exposure of a general population of 2 × 10−4 Sv−1 at low doses, on the assumption that 0.2 % of cases would be fatal. They stressed the uncertainty in assessing the temporal pattern of radiation-induced skin cancers. 2.4.3.4 Breast cancer Data are available on radiation-induced breast cancer from follow-up studies on the A-bomb survivors as well as from studies of patients in North America given fluoroscopy examinations for tuberculosis or treated for acute postpartum mastitis. Risks calculated from either population are little different, based on additive projection models. ICRP has based its risk estimate of 2 × 10−3 Sv−1, for a mixed population of men and women, on the data on the A-bomb survivors. The risk of breast cancer also varies considerably with age at exposure. Thus, for exposure in the first decade of life, the risk is about 4 times that at ages 40-50 years [93M4].
2.4.4. Exposure to internally incorporated radionuclides Human data on cancer induction from internally incorporated radionuclides are available for only a few radionuclides and have been reviewed by UNSCEAR [94U7, 00U8]. Quantitative data for risk estimation are available only for alpha particle emitting radionuclides. Limited data are available on humans exposed to β/γ emitters. A number of epidemiological studies have followed groups exposed to 131I. These studies cover a wide range of doses, varying from very high doses delivered in the treatment of hypothyroidism to the low doses received by patients exposed to diagnostic procedures or exposed to radiation from fallout in the Marshall Islands. The information available provides little evidence that exposure to 131I is associated with a risk of thyroid cancer, although in some cases the follow-up is relatively short. This lack of effect, compared with the effect of external radiation, may be due to an effect of dose rate or to differences in the distribution of dose within the gland. There may also be differences due to ages at exposure. As in the case of external radiation the groups were predominantly young persons. The extent to which exposures to 131I has contributed to the increased risk of thyroid cancer following the Chernobyl accident is still uncertain. Some very sparse data on tumour induction are available on a few individuals given 32P, 35S and 59Fe for medical reasons and there is some information on persons in the Southern Urals exposed to 90Sr who used water from the Techa River for drinking and irrigation [94K1]. A number of studies have also considered the effects of radionuclides in weapons fallout or in discharges to the environment from other nuclear facilities. These data do not at present provide a basis for assessing risks from intakes of β/γ emitting radionuclides. The available information on α-particle emitters covers groups exposed to radium isotopes (224Ra, 226 Ra, 228Ra) where bone tumours are the predominant late effect, and Thorotrast (colloidal 232ThO2) which principally results in irradiation of the liver, spleen and bone marrow, with tumours arising mainly in the liver and bone marrow (leukaemia). Information is also available in man on lung cancer following occupational exposure to radon and its decay products. A number of epidemiological studies of domestic exposure to radon have been published and others are presently under way, to date the data are generally consistent with risks obtained from worker studies although exposures are lower and have a reduced sensitivity for obtaining quantitative risk estimates. Twenty-six men who worked with plutonium in North America on the Manhattan project during the Second World War have also been studied (estimated body Landolt-Börnstein New Series VIII/4
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contents 52-3180 Bq). Seven individuals had died by 1991. The causes of death were lung cancer (2 cases), myocardial infarction, arteriosclerotic heart disease, accidental injury, respiratory failure due to pneumonia/congestive heart failure and osteosarcoma of the sacrum. Three men also had a history of skin cancer [91V3]. There is a high probability that the bone cancer was caused by exposure to plutonium as the spontaneous risk is about 1 in 2000. ICRP [91I2] has recommended the use of radiation weighting factors wR for calculating the equivalent dose to tissues and thus interpolating between the effects of high and low-LET radiation. The wR for αparticle irradiation is taken to be 20. 2.4.4.1 Radium-226/228 luminisers An increased incidence of bone cancer and of head sinus carcinoma has been observed in persons in the USA exposed to long-lived radium, particularly in painters of luminous dials, but also radium chemists or persons treated with radium salts for a possible therapeutic effect [86R2, 94R1]. These persons became internally contaminated with pure 226Ra (t1/2 = 1,600 years) in some cases, and in other cases with various mixtures of 226Ra and 228Ra (t1/2 = 5.8 years). Bone cancers and head sinus carcinomas have arisen in these populations. The majority of these cancers had appeared by 1969, although three bone tumours have appeared since then and more recently head cancers have appeared at a greater rate than bone cancers. The radium isotopes deposit principally in the skeleton and the bone sarcomas appear to have been induced by α particles from either the 226Ra or 228Ra decay series. The head sinus carcinomas are thought to be caused mainly by the accumulation of decay products of radon (222Rn) gas in the frontal sinuses and mastoid air cells. This radon is produced by the decay of 226Ra in the bone. Except for the bone sarcomas and head sinus carcinomas no definite excess in other types of malignancy, including leukaemia, is presently ascribed to the internal deposition of long-lived radium. The follow-up study on this population was essentially discontinued in the USA in the mid 1990s. 2.4.4.2 Radium-224 patients The effects of intakes of radium has also been studied in German patients injected with 224Ra shortly after World War II. The study group consists of a population of 682 adults and 218 juveniles (age at first injection varied between 1 and 20 years) who received weekly or twice weekly intravenous injections of 224 Ra, mainly for the treatment of bone tuberculosis or ankylosing spondylitis [86M1, 94S5]. The last bone tumour occurred in 1988, 41 years after the injection of 224Ra into a three-year-old boy and is the only bone sarcoma reported in this series since 1974. Very few new tumours are now expected and follow-up of the population is now limited. Based on the information on the incidence of bone cancers following intakes of 224Ra and average bone dose from its deposition in the skeleton, ICRP [91I2] has adopted a total risk estimate for fatal cancer of 5 × 10−4 Sv−1 (assuming a radiation weighting factor wR for α-particle irradiation of 20). 2.4.4.3 Miners exposed to radon An increased mortality from lung disease has been observed in under-ground miners working in Czechoslovakia, Canada, United States of America and Sweden exposed to radon (222Rn) and its decay products [88B3, 98B5]. The increase in mortality from lung cancer has been correlated with air concentrations of radon in different mines and the duration of exposure. Bronchial stem cells and secretory cells in the airways are considered to be the main target cell for the induction of lung cancer resulting from radon exposure. There are many difficulties in calculating the radiation dose to these cells as a result of exposure to radon
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decay products (expressed in working level months1). The radiation dose over the working life must be taken into account and the dust loading of the atmosphere known as it determines the extent of the uptake of radon decay products onto the respirable particles. In addition to any possible synergistic effects between smoking and radon exposure, the presence of dust, diesel fumes and other possible carcinogens in the mine atmosphere causes some uncertainty as to whether an excess of cancer can be attributed to radiation alone. The BEIR VI Committee [98B5] recommended two models for estimating radon risks based on its analysis of the data on radon-exposed miners, without expressing a preference for either. One of the BEIR VI models takes account of factors such as total exposure, age and average radon concentration. Risks predicted under the latter model are about 50 % greater than those based on the former model. The BEIR VI Committee also considered both multiplicative and submultiplicative versions of these models. The risk predicted for smokers under the submultiplicative form of each model is only slightly smaller than that based on the multiplicative version. In contrast, the risk for non-smokers under the submultiplicative assumption is about twice that under the multiplicative version of the corresponding model. Based on the various combinations of the BEIR VI models, the lifetime risk of lung cancer for smokers in the UK would lie in the range 10 %-15 %, while that for non-smokers would be in the range 1 %-3 %. For a general population of smokers and non-smokers, the range in lifetime risks would be about 3 %-5 %. The BEIR VI model can also be used to calculate total risks of lung cancer in a population in absolute terms. Thus in the UK population lung cancers attributable to the mean domestic radon concentration of 20 Bq m−3 would be in the range of 2000-3300 per year, based on the above models. Taking into account the proportion of non-smokers in the population, it can be estimated that about 500-1300 of radon-associated deaths would arise among non-smokers. For a working population, ICRP [91I2] have adopted a risk factor for lung cancer of 0.68 × 10−2 Sv−1 based on data from the the A-bomb survivors. 2.4.4.4 Thorotrast patients Thorotrast is colloidal thorium dioxide. In the late 1920s it began to be injected into the arteries of patients for use in diagnostic radiology as an X-ray contrast material. The average dose of about 25 ml of Thorotrast contained 5 g of thorium with an activity (from α-particles) of about 20 kBq 232Th with additional radioactivity from its decay products. The colloidal Thorotrast was cleared from the bloodstream by uptake into phagocytic cells depositing about 60 % in liver, 30 % in spleen and 10 % in red marrow. Extensive epidemiological studies in Portugal, Sweden, Denmark, the United States, the Federal Republic of Germany and Japan have shown that retention of thorium dioxide particles in the liver and in the bone marrow resulted in an increased risk of liver tumours and leukaemias as well as liver cirrhosis and other cardiovascular diseases [84V1, 94V2]. On the basis of an injected dose of 25 ml the dose to the liver is estimated to be 0.25 Gy y−1 (high-LET). Present estimates, based on a latent period of 20 years, suggest a lifetime risk of liver cancer following exposure to Thorotrast of about 0.15 × 10−2 Sv−1 (assuming a wR for α-particle irradiation of 20), about half this risk is expected to be expressed by 40 years after exposure [88B3, 91I2].
1
1 WL is any combination of the short-lived decay products of radon per litre of air which will result in the ultimate emission of 1.3 × 105 MeV of α particle energy. A WLM results from exposure to a concentration of decay products in air of 1 WL for an average working month of 170 hours.
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2.4.5 Dose and dose rate effectiveness factors (DDREFs) Risk coefficients for radiation-induced cancer are based mainly on population groups exposed at high doses and high dose rates as described above. Studies at the molecular, cellular, tissue and whole animal level have demonstrated that radiation damage increases with dose and that, at least for low-LET radiation, at high dose rates it is often greater per unit of exposure than at low dose rates. Thus, although the assumption normally made for radiation protection purposes is that the dose-response curve for cancer induction is linear, with the risk proportional to dose, in practice a dose and dose rate effectiveness factor (DDREF) has commonly been used to allow for a reduced effectiveness of radiation in inducing cancer in man at low doses and low dose rates. The choice of a suitable DDREF has caused considerable debate with relevant data being available from cellular and animal studies, as well as human epidemiology. ICRP in its 1990 recommendations based estimates of DDREF principally on an analysis by Pierce and Vaeth [89P2] of the data from the Japanese survivors. This analysis shows that the data do not allow for a reduction factor of much more than about 2. Other epidemiological data showed little evidence of dose rate effects although studies on thyroid cancer incidence [85S4] and breast cancer mortality [89M3] indicate possible reduction factors of up to 3 or 4. As a consequence ICRP adopted a DDREF of 2, recognising that ‘the choice is somewhat arbitrary and may be conservative’. In practice, the DDREF would be expected to vary with tissue and with exposure conditions although a single value had to be assigned for protection purposes. A better understanding of the mechanisms involved will be essential for improving understanding of the effects of both dose and dose rates on radiation-induced tumour induction in man. A summary of values of DDREF recommended by national and international bodies is given in Table 2.7. No DDREF is recommended for high-LET radiation (i.e. DDREF = 1). Table 2.7. Summary of dose and dose rate effectiveness factors for radiation-induced cancer Reference DDREF Source ICRP 1977 77I1 2 NCRP 1980 90N2 2-10 UNSCEAR 1986 86U4 up to 5 UNSCEAR 1988 88U5 2-10 BEIR 1990 90B4 2 ICRP 1991 91I2 2 UNSCEAR 1993 93U6 10 MeV the bremsstrahlung photons have the energies of γ-rays. Since neutrons carry no charge, they can interact with matter via the very short-range nuclear force, and they loose their energy stepwise mainly by collisions with atomic nuclei or they may induce nuclear reactions. This requires a large number of scattering processes and especially light materials to slow down and absorb neutrons. From range calculations it follows that α-radiation is easily absorbed quantitatively (e.g. by a sheet of paper), for absorption of β-radiation several millimetres or centimetres of material are necessary, and for absorption of γ-radiation thick layers of either lead or concrete are needed taking into account that the absorption of γ-rays follows an exponential law. The ratio of the absorption coefficients for α-, β- and γradiation of equal energy is about 104:102:1.
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3.5.1.2 Electromagnetic radiation
The types of electromagnetic radiation considered here are mostly produced during de-excitation processes in the electron shells or the nuclei of atoms. Due to historical reasons, this electromagnetic radiation − despite of similar fundamental properties − is differently denominated according to different origins: • • • •
emission from de-exciting outer electron shells: emission from de-exciting inner electron shells: emission from de-exciting nuclei: emission during deceleration of charged particles:
light X-rays γ-radiation bremsstrahlung.
The energy range of X-rays lies within about 100 eV to 100 keV (= wavelengths of about 10 nm to 10 pm), and that of γ-rays within about 10 keV to 104 MeV (= wavelengths of about 0.1 nm to 10−7 nm). That means there is an overlap in the energy ranges of X-rays and γ-rays. Electrons with energies >10 MeV decelerated in a substance of high atomic number induce the emission of high-energetic (“hard”) bremsstrahlung. In contrast to the γ-rays emitted from nuclei, this bremsstrahlung shows a continuous energy distribution. X-rays are emitted from the electron shell of the nuclides after formation of a hole in one of the inner shells either by an external process (bombardment with photons, electrons or heavy charged particles) or by an internal decay process like electron capture or internal conversion. The X-ray spectrum of an individual nuclide shows a distinct line structure corresponding to the different transition possibilities in the electronic shell and reflecting its structure. The γ-rays emitted by an excited nucleus have well-defined energies, which correspond practically to the differences in the excitation energies of the nuclei (the recoil energies transferred to the emitting nuclei are very small). Gamma spectroscopy is, therefore, the usual method to investigate the level structure and decay schemes of atomic nuclei. Generally, the γ-radiation is emitted immediately after a preceding α- or β-decay, since the lifetimes of excited states are of the order of 10−13 s. However, if immediate γ-transitions are “forbidden” because of high differences of the nuclear spins of the involved states and the ground state in combination with the conservation laws of nuclear momentum and parity (selection rules), a metastable or isomeric state results which decays with its own half-life different from that of the mother nuclide. The transition from the isomeric to the ground state is called isomeric transition (IT). Some long-lived nuclear isomers in isotopes with a stable ground state are of practical importance as pure γ-emitters. 3.5.1.3 Conversion electrons
If a γ-transition in an excited nucleus is hindered by the selection rules, a certain probability exists increasing with increasing “hindrance” to transfer the excitation energy directly to a shell electron, which is emitted monoenergetically. The energy Ee of these conversion electrons is given by Ee = Eγ − EB where Eγ is the respective γ-energy, and EB the binding energy of the electron in the respective shell. Since the higher shells L, M, ... have sub-shells with slightly different binding energies, the conversion electron spectra contain 3 lines for L-conversion and 5 lines for M-conversion. The internal conversion leaves a hole in the electron shell, which by recombination causes the emission of characteristic X-rays or Auger electrons. The latter − in contrast to conversion electrons − gain their energy from transitions between electron shells, which is in general much lower than the nuclear transition energies of conversion electrons. The relative abundance of internal conversion is given by the total conversion coefficient α = Ne/Nγ, which is composed from the partial conversion coefficients for the individual electron shells and subshells: α = αΚ + αLΙ + αLII + αLIII + αM..., each defined as the ratio of the number of the respective conversion electrons to the number of γ-rays emitted.
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3.5.2 Physical properties of radiation In order to investigate radionuclides and to develop protection methods against radioactive radiation, one has to deal with their properties and the interaction processes between the different radiation types and matter. In principle, the absorption of electromagnetic radiation like γ- and X-rays is different from that of particles. While the latter lose their energy by successive collisions, photons give off their energy mostly in one process. Because they are chargeless, their interaction with matter is small. The absorption of γrays follows an exponential law: I = I0 e−µd
(3.5.1)
where µ is the absorption coefficient and d the absorber thickness. The exact validity of this exponential law is, however, restricted to monoenergetic γ-radiation, a narrow beam of γ-rays and a thin absorber. 3.5.2.1 Interaction of charged particles with matter
The primary interactions of fast charged particles with matter can be classified as follows: a) b) c) d) e)
elastic collisions with atomic electrons inelastic collisions with atomic electrons elastic collisions with nuclei inelastic collisions with nuclei nuclear reactions and interaction with nuclear forces
All these interactions can contribute in principle to deflection and deceleration of incoming particles. Since more than 104 collisions are necessary to stop particles with energies in the MeV range, this is a multi-scattering process composed statistically in different ways of the individual processes. Heavy charged particles (p, d, α, ions) loose most of their kinetic energy by process (b). The electrons of the scattering atom are excited by the energy transfer, most of them to energies high enough to ionise the atoms. Elastic collisions of heavy particles (protons, deuterons, α-particles, ...) with nuclei (c) are rare in comparison to (b), for light particles (electrons, positrons), however, much more frequent. Inelastic scattering from nuclei (d) also occurs rarely and is negligible for deceleration of heavy particles, for light particles, however, it is important at higher energies giving rise to the emission of bremsstrahlung. Consequently, there are characteristic differences for the deceleration of heavy and light particles. On the other hand, the detection of all types of charged particles, as well as of electromagnetic radiation, is based in general on their ionisation effects. This holds also for the damaging effects in biological material. Deceleration of heavy particles
By the inelastic scattering of fast particles from shell electrons the atoms are excited or preferentially ionised. For the resulting loss of kinetic energy per unit path one obtains: −
dE 4π z 2 e 4 = N0B ds me v 2
(3.5.2)
In this expression − also called stopping power − z means the charge of the fast particle, v its velocity and N0 the number of nuclei per cm3 in the absorber material. Landolt-Börnstein New Series VIII/4
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The atomic stopping number B is a function, depending on the energy of the incoming particle. From a quantum mechanical calculation H. Bethe obtained the following expression for the atomic stopping power B under consideration of relativistic effects at high particle energies: ⎡ 2m v 2 C ⎤ B = Z ⎢ln e − ln(1 − β 2 ) − β 2 − K ⎥ I Z ⎦ ⎣
(3.5.3)
where Z means the atomic number of the absorber, β = v/c and I = 11.5·Z the average ionisation potential; CK is a correction factor depending on E and Z, which is of influence only at very low projectile energies and takes values in between 0 and ~1. For the stopping power it follows: −
⎡ 2m v 2 C ⎤ dE 4π z 2 e 4 = N 0 Z ⎢ln e − ln (1 − β 2 ) − β 2 − K ⎥ 2 I Z ⎦ ds me v ⎣
(3.5.4)
The schematic course of equ. (3.5.4) shown in Fig. 3.6 demonstrates qualitatively that in the intermediate region for energies much larger than the ionisation potential and much less than the particle rest mass (I « E « Mc2) the energy loss decreases approximately as 1/E, since the logarithmic term varies slowly and the relativistic terms are negligible. At high velocities, however, these terms give rise to a slight increase causing a flat minimum at E ≈ 3Mc2. At low energies the logarithmic term prevails and the curve decreases steeply below E ≈ 500·I. From this behaviour it follows that the ionisation density produced by the decelerated particle in the absorber increases markedly before the end of the path s and decreases steeply behind this point. Furthermore, the 1/v2 dependence of the first term in equ. (3.5.4) causes a stretching of the energy scale for different masses of the incoming projectiles (cf. Fig. 3.6).
Loss of kinetic energy per unit path - d E ds
Loss of kinetic energy per unit path - d E [ MeV/ cm ] dx
0.05
~ = 500 l
3 Mc 2 Particle energy E
0.04
µ
d
α
0.03 e
_
p
0.02
0.01
0 10 -2
10 -1
1 10 2 10 Particle energy E [MeV]
10 3
10 4
Fig. 3.6. Schematic course of the stopping power with energy (left) and dependence on the projectile type for electrons, muons, protons, deuterons and α-particles (right).
3.5.2.2 Interaction of neutrons with matter
Neutrons are emitted by radioactive decay during spontaneous fission and in rare cases after β-decay of a very neutron-rich nuclide to an excited level of the daughter nuclide. In nuclear reactions and in particular in nuclear fission neutrons are of high importance, especially for the production of radionuclides. Since from radioactive sources mostly only small fluxes can be obtained, neutron sources based on nuclear reactions, like 9Be(α,n)12C, 9Be(d,n)10B or 9Be(γ,n)2α, are used. The highest fluxes of neutrons are available in nuclear reactors. Landolt-Börnstein New Series VIII/4
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The interaction of neutrons with matter happens mainly with nucleons and nuclei via elastic and inelastic scattering and nuclear reactions, because they are electrically neutral and interact hardly with electrons. Thus, ionisation by neutrons is negligible. In elastic collisions the total kinetic energy remains constant and only deflection occurs, while in inelastic collisions part of the kinetic energy is used for excitation of the collision partner. Depending on the neutron energy different types of interactions occur and, correspondingly, several energy ranges are distinguished: • • • • •
0-0.1 eV: 0.1-100 eV: 1-10 keV 0.1-100 keV: 0.1-10 MeV:
thermal neutrons (energies similar to those of gas molecules at room temperature) slow neutrons epithermal neutrons neutrons of intermediate energies fast neutrons
In contrast to charged particles, neutrons do not undergo Coulomb interaction with nuclei. However, low-energy (thermal and slow) neutrons are very effectively absorbed by a great number of nuclei, giving rise to nuclear reactions. Elements such as B, Cd, Sm, Eu, Gd and Dy are used as good neutron absorbers. Epithermal neutrons are also called resonance neutrons, because there exist absorption maxima or resonances at defined energies for distinct absorber nuclei. Neutrons with energies corresponding exactly to the excitation energies of excited levels of the nucleus are absorbed with very high probability. Fast neutrons are decelerated mainly by elastic and inelastic collisions. The energy released in one elastic collision depends on the collision angle and the mass number of the target nucleus. The lighter the nucleus, the higher is the energy loss of the neutron. Consequently, hydrogen or hydrogen-containing substances like water or paraffin are very effective to reduce (“moderate”) the energy of neutrons. Graphite can also be used as moderating material, but needs larger material thickness due to its higher mass number. After deceleration the “slow neutrons” are captured by nuclei, giving rise to nuclear reactions. High-energy (fast) neutrons may also induce nuclear reactions, but the probability and consequently the contribution of this interaction type is relatively small. 3.5.2.3 Interaction of electromagnetic radiation with matter
Electromagnetic radiation passing a material experiences an intensity attenuation according to equ. (3.5.1) (see above) via interactions with the various components of matter. For instance, γ-rays interact with: • • • •
atomic electrons, nuclei, electrical fields of the electrons and nuclei, meson fields of the nuclei.
These interactions can result in energy losses, alterations of the propagation direction and polarisation. The effects on γ-rays are: • total absorption, • inelastic scattering (incoherent), • elastic scattering (coherent, Thomson scattering). Though all of the possible combinations of interactions and processes can occur, most of them result in very weak effects, which can normally be neglected for attenuation considerations. Important are: • the photoeffect, • the Compton effect, • the pair formation. In the photoeffect the incoming photon is totally absorbed by an atomic electron of the inner shells (preferably of the K- or L-shells). A free electron cannot take up the total energy of the photon because of the momentum conservation. In contrast, this is possible for the bound atomic electrons, because the atom Landolt-Börnstein New Series VIII/4
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takes up the residual momentum as recoil. Consequently, the stronger bound inner electrons exhibit the strongest absorption. The photoelectrons in turn are emitted with a kinetic energy Ekin corresponding to the total photon energy h·ν reduced by the electron's binding energy Eb, which depends on the atomic number of the respective atom: Ekin = h·ν − Eb
(3.5.5)
Consequently, for high photon energies and light absorbers the photoelectrons carry nearly the total photon energy. In addition, they exhibit a pronounced angular distribution favouring the forward direction with increasing photon energy. Additional (isotropic) radiation is emitted originating from the recombination process of the electron hole left by the emitted photoelectron. During this recombination characteristic X-rays and Auger electrons are released by the energy set free by an outer electron filling up the hole in the inner shell. Furthermore, bremsstrahlung is produced by the photoelectrons especially in heavy materials. All these radiation types connected with the absorption of photons together with the ionised residual absorber atoms can result in a high local energy density. The Compton effect describes the inelastic scattering of photons from outer shell electrons, where the photon looses only part of its energy, which is transferred to the electron. Both, the recoiling electron and the photon are scattered with respect to the original photon direction, and the scattering into forward directions is again favoured with increasing photon energy. The relative energy transfer to the electron increases with increasing photon energy and also with increasing photon scattering angle. On the other hand, the highest backscattering contributions occur at low photon energies. The pair formation occurs in the electric field of an atomic nucleus, where the γ-ray can be spontaneously transformed into an electron-positron pair, if its energy is larger than the sum of the rest masses of the pair: Eγ > 2mec2 = 1,022 MeV. The nucleus remains unchanged but is necessary for the conservation of momentum. The positron annihilates by recombination with an electron of the absorber atoms, whereby two annihilation γ-quanta are produced with an energy of 511 keV each.
Compton effect
Total
Pair formation
Absorption coefficient m [cm -1]
Absorption coefficient m
Photoeffect 1.5
1.0 Pb
0.5 Cu Fe
10 -1
1 10 γ - energy E γ [MeV ]
10 2
Fig. 3.7. Schematic representation of the partial and total absorption coefficients of a heavy absorber in dependence on the γ-ray energy.
0 0.1
Al 1 γ - energy E γ [MeV ]
10
Fig. 3.8. Schematic representation of the total absorption coefficients of different absorbers in dependence on the γ-ray energy.
For the attenuation of photons in matter, mainly the three processes photoeffect, Compton effect and pair formation have to be taken into account. Fig. 3.7 shows schematically their contributions to the total absorption coefficient µ in dependence on the photon energy. A primary photon beam is attenuated by scattering as well as by absorption. The main parameters for the attenuation are the photon energy and the Landolt-Börnstein New Series VIII/4
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atomic number and density of the absorber material (cf. Fig. 3.8). At low photon energies, classical coherent (Thomson) scattering predominates, at higher energies incoherent scattering via the Compton effect. Absorption of photon energy occurs mainly via the photoeffect and pair formation and partially via the Compton effect. The lower energy threshold for the photoeffect is given by the binding energy of the inner shell electrons, for the pair formation by the rest mass of the electron-positron pair (1022 keV). At high atomic numbers Z and low photon energies the photoeffect dominates, at high energies the pair formation. For energies between about 1 to 5 MeV the Compton effect dominates at all atomic numbers Z. The resulting Compton electrons give, therefore, the most important contribution to the energy dose in human tissue and to the biological effect of radiation.
3.6 Nuclear fission and fission products 3.6.1 Particle induced nuclear fission Besides the few heavy nuclides decaying by spontaneous fission many other heavy isotopes undergo fission after bombardment with particles, especially neutrons. Fission of uranium was first observed but misinterpreted in 1934 by Fermi [34Fer] in an attempt to produce transuranium elements by irradiation of uranium with slow neutrons. Similar experiments were performed by several groups, but only in 1937 Hahn, Meitner and Strassmann [37HMS] identified the observed radioactive products of the fission of uranium to have appreciably lower atomic mass, such as 140Ba. Fission of heavy nuclei always leads to products with a high neutron excess due to the much larger neutron-to-proton ratio of heavy nuclides. The primary fission products are formed in about 10−11 s by fission and emission of prompt and β−-delayed neutrons and γ-rays from the highly excited fragments. They always lie on the right hand side of the valley of β-stability and decay by several successive β−-decays following isobaric chains into nuclides of increasing atomic number Z ending up with the first stable isobar in the chain. The fission process exhibits different features depending mainly on the energy of the inducing particles and on the atomic number Z of the fissioning nuclide. For fission induced by low-energy neutrons with energies up to about 10 MeV two fission products with mass numbers in the range between about 70 and 160 and ≈ 2-3 neutrons are emitted. The energy ∆E released by fission is relatively high (∆E ≈ 200 MeV), since for the light fission products the binding energy per nucleon is higher than for the heavy fissioning nuclei. When comparing the naturally occurring uranium isotopes 234U (abundance: 0.0055 %), 235U (0.720 %), and 238U (99.2745 %), it is found that 235U undergoes fission by slow (i.e. thermal) neutron capture with a remarkably large cross section of σn,f = 586 × 10−24 cm2. This holds also for other odd-mass (even Z, odd N) nuclei like 233U, 235U, 239Pu and 241Pu. In all these cases the binding energy of an additional neutron is very high resulting in high cross sections σn,f for fission by slow (thermal) neutrons. To induce fission in 238U, the neutron must have an energy of 1 MeV. The natural isotope 232Th undergoes fission by 1.1 MeV neutrons. Using higher energetic particles, either neutrons, deuterons, or α-particles, it is possible to induce fission in any element with atomic number larger than 73.
3.6.2 Fission products One of the most characteristic features of the fission process − very important also for the activities that have to be regarded in nuclear technologies − is the resulting mass distribution of the various fission products. As an example, Fig. 3.9 displays the yield of fission products found after thermal fission of 235U Landolt-Börnstein New Series VIII/4
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as a function of the mass number A. The maximum yields are around A = 90 to 100 and A = 133 to 143, respectively. For these mass ranges the fission yields are about 6 %, whereas symmetrical fission around A = 117 occurs with much smaller probability ( ≈0.01 %). It should be reminded that the sum of the fission yields is 200 %, because each fission produces two fission products. When investigating the mass distributions from fission with thermal neutrons one observes that those of 233U and 239Pu are similar to that of 235U. In the case of 239Pu, however, the low mass maximum is slightly shifted to higher masses, while the maximum for heavy fission products remains nearly unchanged. This tendency continues with increasing mass of the fissioning nuclei, and in the case of 258 Fm the two maxima are superimposed. An increase of the energy of the neutrons leads to a strong increase of the probability for symmetric fission and gives rise to a flattening of the valley of the mass distribution by up to two orders of magnitude. Increase of symmetric fission is also observed for nuclides with lower atomic numbers Z. For 227 Ac (Z = 89) symmetric and asymmetric fission have nearly the same probability, resulting in three maxima in the mass distribution. 10
Fission yield [%]
1
10 -1
10 -2
10 -3
10 - 4 80
100
120 Mass number A
140
160
Fig. 3.9. Yields of fission products for the fission of 235 U by thermal neutrons.
It should be mentioned that the mass distribution curves as discussed before give the total yields of the decay chains of mass numbers A. The independent yields for individual members of the decay chains, i.e. the yields for direct formation in the fission process, are often difficult to determine, especially if in the precursor chain short half-lives occur. The total energy ∆E released during fission consists of kinetic energy and excitation energy of the primary fission fragments, where the kinetic energy resulting mainly from the Coulomb repulsion of the two fission fragments gives the main contribution. In the case of low-energy fission the kinetic energy Ekin is given by the empirical relation Ekin ~ Z2 / A1/3 where Z and A are the atomic number and the mass number of the fissioning nucleus, respectively. The primary fission fragments release their excitation energy by emission of prompt neutrons with energies between 0 and about 10 MeV (mean value ≈2 MeV) and of prompt γ-rays. The number of prompt neutrons emitted increases with the mass number of the fissioning nuclei and depends mainly on the excitation energy of the primary fission fragments. Furthermore, an average number of 7.5 γ-rays with a mean energy of about 1 MeV are emitted per fission, as well as several low-energy transitions in form of conversion electrons and X-rays. Landolt-Börnstein New Series VIII/4
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In very rare cases high-energy charged particles such as p, d, t, α-particles, 3He, 7Li, 8Li, 9Li, 9Be, 10Be and isotopes of B, C, N and O are also emitted at an early stage of the low-energy fission process, when the fission fragments are still very close to each other. The probability of this so-called ternary fission, i.e. formation of three fragments increases strongly with the excitation energy of the fissioning nuclei. For example, high-energy fission of 232Th with 400 MeV argon ions leads to a ratio of ternary to binary fission of about 1:30. In contrast to low-energy fission, high-energy fission induced by neutrons or other high-energy particles leads to marked changes in the mass distribution of fission products. Among others, the probability of symmetrical fission increases considerably with increasing excitation energy of the target nuclei, resulting in a single flat maximum of the fission yield curve slightly below half the target mass number.
3.6.3 Nuclear reactors When comparing the energies available from combustion of carbon or carbon compounds (or in general: from a chemical reaction like oxidation) with those from nuclear fission, one has to compare the orders of magnitude of the binding energies in the electron shells and in the atomic nuclei, which are in the order of eV and MeV, respectively. This means a difference of six orders of magnitude: 1 kg of carbon produces an energy of 9.4 kWh, 1 kg of uranium can produce a maximum of 1.85 × 107 kWh. This high energy release as already observed with the first fission experiments caused immediate considerations to make use of this enormous energy potential. In addition, the neutrons released in each fission process could initiate further fissions leading in principle to a continuous generation of energy. However, it soon turned out that a safe control of a continuously proceeding fission process required the solution of various technological questions and problems. Consequently, the first application of nuclear fission was the construction of a nuclear explosive (“atomic bomb”) on the basis of uncontrolled self-amplifying fission in the form of chain reactions. As mentioned above, the fission process in the first step results in two fragments with high neutron excess and in some prompt high energy neutrons. Furthermore, from the decay chain of the highly excited primary fragments besides β−-particles and γ-rays about l % of β-delayed neutrons are emitted which are delayed by at least 0.01 second and about 0.07 % which are delayed by as much as 1 minute. This would be a satisfactory condition for a chain reaction provided the neutrons originating from the fission process have a possibility to react with other atoms of 235U. The neutrons generated during fission are, however, no thermal neutrons but have energies up to the order of 1 MeV. If these neutrons are slowed down to thermal velocities they can excite other 235U atoms to fission, so that the reaction would proceed with increasing amplification as a chain reaction if only 235U would be present. In natural uranium this process is not possible because of the high concentration of 238U atoms which can capture higher energy neutrons without undergoing fission. In addition, there exist well-defined energies, where a resonance absorption (in a very sharp energy window) of neutrons in 238U occurs with very high cross sections (up to 1200 × 10−24 cm2). Consequently, the fast neutrons have to be effectively slowed down, whereby a sufficient number must escape the resonance absorption in 238U in order to reach thermal velocities. This is achieved by use of a moderator, an element of small atomic weight like deuterium, helium, beryllium, and carbon, whose atoms will not capture the neutrons but rather scatter them elastically.
3.6.4 Nuclear explosives The high amount of energy released by nuclear fission led very early to the production of nuclear explosives. Since 235U, 233U and 239Pu have sufficiently high fission cross sections for fast neutrons, they can be used as nuclear explosives if the respective critical masses are brought together. Without a neutron Landolt-Börnstein New Series VIII/4
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reflector increasing the number of neutrons within the explosive material, a sphere of about 50 kg uranium metal containing 94 % 235U or a sphere of about 16 kg plutonium metal (239Pu) reaches criticality. If a neutron reflector is used, the critical masses are about 20 kg for 235U and about 6 kg for 239 Pu. The critical masses for 233U are similar to those for 239Pu. The use of plutonium in nuclear weapons requires a low concentration of 240Pu in the plutonium, because its presence leads to the production of high numbers of neutrons by spontaneous fission. Consequently; a too high concentration of 240Pu would initiate the neutron multiplication too early with a relatively small multiplication factor and a relatively low energy release. Higher concentrations of 241Pu also interfere, because of its decay into 24lAm with a half-life of only 14.4 y. To minimise the formation of 240Pu and 241Pu, Pu for use in weapons is, in general, produced in special reactors with low burn-up (« 20000 MWthd per ton). Criticality can be reached by shooting two under-critical hemispheres onto each other by means of normal explosives (gun-type) or by compressing an under-critical spherical shell into a supercritical sphere (implosion-type). The bomb ignited over Hiroshima (energy release corresponding to ≈15 kilotons of TNT) was of the gun type using 235U, whereas that ignited over Nagasaki was of the implosion type using 239Pu (energy release corresponding to ≈22 kilotons of TNT). Generally, the fissile core is surrounded by a heavy material, in order to reflect the neutrons and to increase the inert mass and consequently the time in which the super-critical configuration is held together. The explosion of fissile material leads to temperatures of about 108 K which are sufficient to initiate fusion between deuterium and tritium. This is the basis of the development of hydrogen bombs, in which the energy of fission is used for ignition of fusion. LiD serves as a source of D and T, the latter being produced by thermal (6Li(n,αn)d) and by fast neutrons (7Li(n,αn)t). If the temperature is high enough, the D-D reaction can contribute to the energy production. The fast neutrons released by the fusion reactions react very effectively with natural or depleted U initiating fission of 238U. By these kinds of weapons large amounts of fission products are formed (“dirty weapons”). If a surrounding of non-fissile heavy material is used, fission products are released only by the ignition process (“clean weapons”).
3.6.5 Radioactive inventory and nuclear waste The radioactive inventory in a nuclear power reactor originates from: • fission products • uranium and transuranium elements formed by direct neutron induced reactions and their decay chains • isotopes produced by nuclear reactions in the cladding material of fuel rods, the reactor vessel components and the coolant. The dominating longer-lived fission products occurring in spent fuel elements are 85Kr, 131I, 133I, Xe, 135Xe, 134Cs and 137Cs. Radionuclides produced by nuclear reactions in the coolant are 3H, 14C, 13N, 16 N, 19O, 18F and 41Ar. Furthermore, fission products or actinides may leak into the cooling system from faulty fuel elements. Other radionuclides are produced by reactions with metals and their corrosion products in various reactor vessel components like 51Cr, 54Mn, 59Fe, 58Co, 60Co, 65Zn, 124Sb. 133
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10 3
Rel.activity
a 10 2 b
Fig. 3.10. Radioactive decay of the sum of all fission products (a), 106Ru (≈0.5 %, T = 373.6 d) (b), and 137Cs (≈ 6 %, T = 30.17 y) (c).
10 c 5 0
100 200 300 400 500 600 700 800 900 Decay time [d ]
Shortly after shut-down of a nuclear reactor the activity of the fuel is ≈1.7 × 1017 Bq per MW of thermal energy produced. 237U (from the reactions 235U(n, γ)236U(n, γ)237U and 238U(n, 2n)237U) causes a relatively high initial uranium activity. Since it decays with a half-life of 6.75 d, it vanishes rapidly after the necessary storage of the discharged fuel elements. A global composition of spent nuclear fuel from light water reactors after storage of 1 year is given in Table 3.5. The radioactive waste produced during the operation of nuclear reactors is usually classified according to the state of matter (gaseous, liquid or solid) and according to the activity level as low-active waste (LAW), medium-active waste (MAW), and high-active waste (HAW). The largest amount of radioactivity is concentrated in the spent fuel elements representing highly-active waste (HAW). Table 3.4. Main components of spent nuclear fuel from a light-water reactor with an initial enrichment of 3.3 % 235U, a burn-up of 34000 MWd per ton and a storage time of 1 year (from [97Lie]).
Nuclide
Weight percent
uranium and transuranium elements 235 U 0.756 236 U 0.458 237 U 3×10−9 238 U 94.2 237 Np 0.05 238 Pu 0.018 239 Pu 0.527 240 Pu 0.220 241 Pu 0.105 242 Pu 0.038 americium isotopes 0.015 curium isotopes 0.007
Nuclide
Weight percent
fission products 85 K 90 Sr 129 I 134 Cs + 137Cs others Stable fission products
0.038 0.028 0.09 0.275 0.19 3.0
Reprocessing of nuclear fuel transforms all waste types into liquid solutions and results in the following amounts per ton of U: ≈1 m3 HAW (fission products and actinides in HNO3 solution), ≈3 m3 MAW as organic solution, ≈17 m3 MAW as aqueous solution, ≈90 m3 LAW (aqueous solution). By further processing a volume reduction is achieved: ; ≈0.1 m3 HAW, ≈0.2 m3 MAW (organic), ≈8 m3 MAW (aqueous), ≈3m3 LAW (aqueous). After respective storage times the HAW solutions are transformed by calcination or vitrification into stable forms like ceramics or glasses suitable for long-term disposal and also in order to reduce the volume. Landolt-Börnstein New Series VIII/4
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After one year of intermediate storage and reprocessing, the initial activity of the HAW solutions is of the order of 1014 Bq/l from which the activity due to 90Sr and 137Cs is about 1013 Bq/l; after 10 y the activity of the HAW solution decreases approximately with the half-life of these nuclides (28.64 y and 30.17 y, respectively, cf. Fig. 3.10). After 1000 y the residual activity (of the order of 104 Bq/l) is determined by long-lived fission products like 99Tc, 129I and actinides. Solid MAW and LAW originate from structure material of the fuel elements, undissolved, dispersed and filtered particles of metals or metal oxides (e.g. Ru, Rh, Mo, Tc), and gaseous components like tritium (as T2 or HTO), 14C (as CO2), 85Kr, 129I or 106Ru (as RuO4) after adsorption in special adsorbents. Computer codes have been developed to calculate the inventory of all radionuclides in spent fuel rods after discharge and after several periods of storage. As an example in the Tables 3.6 and 3.7, taken from [95SSK], the activities of actinides, fission products and light elements from surrounding material are listed, which were obtained from a calculation with the code KORIGEN for a given reactor type, initial 235 U enrichment, and fuel burn-up at charge time, discharge time and several storage times. Further information on these calculations and their results as well as on general reprocessing, recycling and disposal concepts can be found e.g. in [83FiW], [93Wie], [95SSK], [80Clo] and [97Lie]. Table 3.5. Inventory of selected relevant radionuclides in Bq for a pressurized water reactor with a thermal power of 3.733 MW after an operation time of 1 day [95SSK].
Nuclide 85
Time after discharge 1h 6h
0h
24 h
120 h
Kr Kr 87 Kr 88 Kr 133 Xe 135 Xe Σ Kr-Xe
13
4.08×10 1.43×1018 2.86×1018 4.03×1018 2.89×1017 1.74×1018 1.03×1019
13
4.29×10 1.24×1018 1.68×1018 3.16×1018 3.10×1017 2.09×1018 8.47×1018
13
4.96×10 5.72×1017 1.10×1017 9.32×1017 4.04×1017 2.91×1018 4.93×1018
13
5.50×10 3.53×1016 6.02×1012 1.15×1016 6.15×1017 1.72×1018 2.38×1018
5.53×1013 1.25×1010 0.0 7.60×1005 5.82×1017 2.04×1015 5.84×1017
131
I I 133 I 134 I 135 I Σ Iodine
2.36×1017 9.66×1017 4.16×1018 8.76×1018 6.76×1018 2.09×1019
2.43×1017 9.61×1017 4.15×1018 6.53×1018 6.09×1018 1.80×1019
2.46×1017 9.26×1017 3.57×1018 2.63×1017 3.60×1018 8.61×1018
2.38×1017 7.92×1017 1.96×1018 2.23×1011 5.46×1017 3.54×1018
1.80×1017 3.38×1017 8.01×1016 0.0 2.32×1013 5.98×1017
89
7.29×1016 4.30×1014 5.51×1018 1.08×1014 4.11×1016 7.87×1016 4.28×1018 7.68×1014 1.58×1018 9.38×1017 6.80×1016 1.40×1018 1.22×1015 4.10×1017 2.27×1016 7.54×1017
7.42×1016 4.32×1014 5.14×1018 1.12×1014 4.37×1016 7.95×1016 4.11×1018 8.32×1014 l.56×1018 9.86×1017 6.81×1016 1.24×1018 1.22×1015 4.28×1017 2.32×1016 6.53×1017
7.41×1016 4.31×1014 3.57×1018 1.29×1014 5.47×1016 7.93×1016 3.35×1018 1.15×1015 1.48×1018 1.14×1018 6.78×1016 5.67×1017 1.22×1015 4.68×1017 2.36×1016 2.93×1017
7.33×1016 4.31×1014 9.59×1017 1.83×1014 7.28×1016 7.87×1016 1.60×1018 2.29×1015 1.23×1018 1.15×1018 6.69×1016 3.41×1016 1.22×1015 3.81×1017 2.09×1016 1.63×1016
6.94×1016 4.31×1014 8.71×1014 3.44×1014 7.60×1016 7.53×1016 3.12×1016 7.93×1015 4.48×1017 4.32×1017 6.24×1016 1.05×1010 1.21×1015 5.88×1016 1.02×1016 3.32×1009
85rn
132
Sr Sr 91 Sr 90 Y 91 Y 95 Zr 97 Zr 95 Nb 99 Mo 99m Tc 103 Ru 105 Ru 106 Ru 105 Rh 127 Sb 129 Sb 90
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Nuclide 127 Te 127rn Te 129 Te 129m Te 131m Te 132 Te 134 Cs 136 Cs 137 Cs 140 Ba 140 La 141 Ce 143 Ce 144 Ce 143 Pr 239 Np 238 Pu 239 Pu 240 Pu 241 Pu 241 Am 242 Cm 244 Cm Σ Aerosols
0h 1.07×1016 1.07×1013 6.70×1017 1.79×1015 1.72×1017 9.48×1017 3.67×1010 2.99×1014 4.56×1014 3.83×1017 7.20×1016 1.07×1017 2.66×1018 1.52×1016 7.23×1016 1.50×1019 8.38×1006 6.18×1011 5.25×1009 3.60×1009 3.87×1003 2.15×1002 9.62×10−04 3.52×1019
Time after discharge 1h 6h 1.13×1016 1.41×1016 1.16×1013 1.59×1013 17 6.43×10 3.42×1017 15 1.87×10 2.11×1015 17 1.70×10 1.52×1017 17 9.42×10 9.01×1017 10 3.77×10 4.02×1010 14 2.98×10 2.95×1014 14 4.58×10 4.58×1014 3.82×1017 3.78×1017 16 7.73×10 1.02×1017 17 1.13×10 1.30×1017 18 2.65×10 2.39×1018 16 1.52×10 1.52×1016 16 7.78×10 1.04×1017 19 1.51×10 1.43×1019 06 9.42×10 1.44×1007 11 6.67×10 9.09×1011 09 5.51×10 5.79×1009 09 3.60×10 3.60×1009 03 4.53×10 7.82×1003 02 2.67×10 4.95×1002 −03 1.26×10 2.46×10−03 19 3.46×10 3.00×1019
24 h 1.77×1016 3.06×1013 2.08×1016 2.27×1015 1.00×1017 7.68×1017 4.13×1010 2.83×1014 4.58×1014 3.63×1017 1.74×1017 1.40×1017 1.63×1018 1.52×1016 1.74×1017 1.15×1019 2.99×1007 1.67×1012 5.81×1009 3.60×1009 1.97×1004 1.01×1003 4.55×10−03 2.05×1019
120 h 9.85×1015 8.18×1013 1.37×1015 2.10×1015 1.09×1016 3.28×1017 4.12×1010 2.29×1014 4.58×1014 2.92×1017 2.90×1017 1.29×1017 2.18×1017 1.50×1016 2.68×1017 3.53×1018 7.04×1007 3.79×1012 5.81×1009 3.60×1009 8.28×1004 1.41×1003 5.40×10−03 6.37×1018
Total
6.27×1020
1.46×1020
3.29×1019
7.96×1018
7.09×1019
Table 3.6. Inventory of selected relevant radionuclides in Bq for a pressurized water reactor with a thermal power of 3.733 MW after an operation time of 333 days [95SSK]
Nuclide 85 Kr 85m Kr 87 Kr 88 Kr 133 Xe 135 Xe Σ Kr-Xe
0h 1.61×1016 1.19×1018 2.27×1018 3.22×1018 7.62×1018 1.83×1018 1.61×1019
Time after discharge 1h 6h 1.61×1016 1.61×1016 1.04×1018 4.78×1017 18 1.33×10 8.72×1016 18 2.53×10 7.45×1017 18 7.62×10 7.60×1018 2.20×1018 3.07×1018 19 1.47×10 1.20×1019
24 h 1.61×1016 2.95×1016 4.78×1012 9.19×1015 7.35×1018 1.82×1018 9.22×1018
120 h 1.61×1016 1.05×1010 0.0 6.08×1005 4.75×1018 2.16×1015 4.77×1018
131
3.50×1018 5.18×1018 7.63×1018 8.38×1018 7.14×1018 3.18×1019
3.50×1018 5.16×1018 7.50×1018 6.00×1018 6.43×1018 2.86×1019
3.25×1018 4.27×1018 3.52×1018 1.96×1011 5.77×1017 1.16×1019
2.33×1018 1.82×1018 1.43×1017 0.0 2.45×1013 4.30×1018
I I 133 I 134 I 135 I Σ Iodine 132
3.45×1018 4.99×1018 6.40×1018 2.33×1017 3.81×1018 1.89×1019
Landolt-Börnstein New Series VIII/4
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3 Physical fundamentals
3-37
Nuclide 89 Sr 90 Sr 91 Sr 90 Y 91 Y 95 Zr 97 Zr 95 Nb 99 Mo 99m Tc 103 Ru 105 Ru 106 Ru 105 Rh 127 Sb 129 Sb 127 Te 127m Te 129 Te 129m Te 131m Te 132 Te 134 Cs 136 Cs 137 Cs 140 Ba 140 La 141 Ce 143 Ce 144 Ce 143 Pr 239 Np 238 Pu 239 Pu 240 Pu 241 Pu 241 Am 242 Cm 244 Cm Σ Aerosols
0h 4.49×1018 1.26×1017 5.42×1018 1.30×1017 5.58×1018 6.70×1018 6.52×1018 6.55×1018 6.91×1018 6.05×1018 4.88×1018 2.80×1018 6.57×1017 2.66×1018 2.58×1017 1.01×1018 2.50×1017 2.80×1016 9.89×1017 1.47×1017 4.86×1017 5.12×1018 8.26×1016 6.90×1016 1.51×1017 6.87×1018 6.95×1018 6.45×1018 6.18×1018 3.25×1018 6.10×1018 6.25×1019 5.76×1014 8.83×1014 5.91×1014 1.10×1017 4.16×1013 4.73×1015 2.92×1013 1.67×1020
Time after discharge 1h 6h 4.48×1018 4.47×1018 1.26×1017 1.26×1017 18 5.05×10 3.50×1018 17 1.30×10 1.30×1017 18 5.57×10 5.57×1018 18 6.70×10 6.68×1018 18 6.26×10 5.10×1018 18 6.55×10 6.55×1018 18 6.84×10 6.49×1018 6.05×1018 5.95×1018 18 4.88×10 4.86×1018 18 2.47×10 1.13×1018 17 6.57×10 6.57×1017 18 2.66×10 2.57×1018 17 2.57×10 2.50×1017 17 8.69×10 3.89×1017 17 2.50×10 2.49×1017 16 2.80×10 2.80×1016 17 9.52×10 5.49×1017 17 1.47×10 1.46×1017 17 4.77×10 4.25×1017 18 5.08×10 4.86×1018 16 8.26×10 8.26×1016 16 6.89×10 6.81×1016 1.51×1017 1.51×1017 18 6.86×10 6.78×1018 18 6.94×10 6.93×1018 18 6.45×10 6.44×1018 18 6.09×10 5.49×1018 18 3.25×10 3.25×1018 18 6.10×10 6.10×1018 19 6.21×10 5.85×1019 14 5.76×10 5.77×1014 14 8.84×10 8.85×1014 14 5.91×10 5.91×1014 17 1.10×10 1.10×1017 13 4.16×10 4.17×1013 15 4.74×10 4.75×1015 13 2.92×10 2.93×1013 20 1.65×10 1.55×1020
24 h 4.43×1018 1.26×1017 9.42×1017 1.29×1017 5.54×1018 6.63×1018 2.44×1018 6.55×1018 5.37×1018 5.14×1018 4.80×1018 6.82×1016 6.56×1017 1.91×1018 2.19×1017 2.17×1016 2.32×1017 2.80×1016 1.20×1017 1.44×1017 2.81×1017 4.14×1018 8.25×1016 6.55×1016 1.51×1017 6.51×1018 6.85×1018 6.35×1018 3.76×1018 3.25×1018 6.04×1018 4.69×1019 5.80×1014 8.88×1014 5.91×1014 1.10×1017 4.21×1013 4.76×1015 2.94×1013 1.30×1020
120 h 4.19×1018 1.26×1017 8.55×1014 1.27×1017 5.29×1018 6.35×1018 4.75×1016 6.55×1018 1.96×1018 1.89×1018 4.47×1018 2.10×1010 6.51×1017 2.93×1017 1.07×1017 4.42×1009 1.30×1017 2.79×1016 8.65×1016 1.33×1017 3.05×1016 1.77×1018 8.22×1016 5.30×1016 1.51×1017 5.24×1018 5.91×1018 5.83×1018 5.01×1017 3.21×1018 5.21×1018 1.45×1019 5.88×1014 8.96×1014 5.91×1014 1.10×1017 4.40×1013 4.71×1015 2.95×1013 7.50×1019
Total
7.75×1020
3.12×1020
1.75×1020
9.69×1019
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2.33×1020
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3 Physical fundamentals
[Ref. p. 3-39
3.6.6 Release of radionuclides from the radioactive inventory of a nuclear reactor In case of an accident of a less dangerous category mainly gaseous or volatile fission or decay products are released. Heavier elements or compounds are more or less retained, since they are hardly transported over longer distances. Consequently, the respective risk assessment studies show the highest risk potentials for the gaseous fission or decay products like: noble gases, iodine, radon, volatile elements and compounds (cf. e.g. [95SSK]). Furthermore, it has to be taken into account that the chemical and physical behaviour of many fission products may change considerably after nuclear transformations within their radioactive decay chains (e.g.: Ra → Rn). This often leads to consequences for the retention behaviour, release of and filter effectiveness for various nuclides in the radioactive inventory.
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3.7 References 10Bat 34Fer 37HMS 79Ewb 80Clo 83FiW 93Wie 95AMC 95SSK
96Bec 97Lie 98Cur
98PKS 98Sch 99Mag 99MoT
Bateman, H.: Proc. Cambridge. Philos. Soc. 15 (1910) 423. Fermi, E.: Nature (London) 133 (1934) 898. Hahn, O., Meitner, L., Strassmann, F.: Z. Phys. 106 (1937) 249. Ewbank, W.B., Ellis, Y.A., Scmorak, M.R.: Nucl. Data Sheets 26 (1979) 1. Closs, K.D. (ed.): Report KfK 3000, 1980. Fischer, U., Wiese, H.W.: Report KfK 3014, 1983. Wiese, H.W.: Nucl. Technol. 102 (1993) 68. Adloff, J.P., MacCordick, H.J.: Radiochim. Acta 70/71 (1995) 13. Leitfaden für den Fachberater Strahlenschutz der Katastrophenschutzleitung bei kerntechnischen Notfällen, Veröffentlichungen der Strahlenschutzkommission Band 13; Herausgegeben vom Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit; 2. überarbeitete Auflage, Stuttgart, Jena, New York: Gustav Fischer Verlag, 1995. Becquerel, H.: C. R. Seances Acad. Sci. (Paris) 122 (1896) 501. Lieser, K.H.: Nuclear and radiochemistry: Fundamentals and applications, D-69451 Weinheim (Federal Republic of Germany): VCH Verlagsgesellschaft mbH, 1997. Sklodowska-Curie, M.: C. R. Seances Acad. Sci. (Paris) 126 (1898) 1101. Curie, P, Sklodowska-Curie, M.: C. R. Seances Acad. Sci. (Paris) 127 (1898) 175. Curie, P., Curie, M., Bémont, G.: C. R. Seances Acad. Sci. (Paris) 127 (1898) 1215. Sklodowska-Curie, M.: Rev. Gén. Sci. Pures Appl. Bull. Assoc. Fr. Av. Sci. 10 (1899) 41. Curie, P., Curie, M., Bémont, G.: Sci. Am. 80 (1899) 60. Curie, P., Curie, S.: C. R. Seances Acad. Sci. (Paris) 134 (1902) 85. Pfennig, G., Klewe-Nebenius, H., Seelmann-Eggebert, W.: Chart of the nuclides (Karlsruher Nuklidkarte), 6th ed., reprint 1998, Forschungszentrum Karlsruhe: Technik und Umwelt, 1998. Schmidt, G.C.: Verh. Dtsch. Phys. Ges. 17 (1898) 14; C. R. Seances Acad. Sci. Paris 126 (1898) 1264. Magill, J.: Nuclides 2000: An electronic chart of the nuclides; EUR 18737 EN, 1999. Mohr, P.J., Taylor, B.N.: J. Phys. Chem. Ref. Data 28 (1999) 1713.
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4 Radiological quantities and units
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4 Radiological quantities and units
In this Chapter the fundamental quantities and units for ionising radiation and in addition specific quantities used in radiological protection are described.
4.1 Introduction While radiation field quantities, quantities describing radioactivity and absorbed dose quantities are based on physical phenomena only, specific dose quantities in radiation protection as e.g. effective dose, include factors which are based on judgements about the biological response of tissues, e.g. due to cancer induction. These factors have been changed in the past in view of new research results and ideas. The definitions given are mainly based on ICRU Report 51 [93I1], ICRU Report 60 [98I1], ICRP Publication 60 [91I1] and the ISO Standards Handbook, Quantities and Units [93I2].
Stochastic and non-stochastic quantities Physical quantities are used to describe physical phenomena or objects. Many physical processes, e. g. the decay of radionuclides, the number of interactions in a small volume irradiated or the energy transferred, are subject to inherent fluctuations. This situation is described by stochastic quantities the values of which follow a probability distribution. Some times this may be a Poisson distribution which is uniquely determined by its mean value. In many other cases a quantity is defined by averaging in time or over a volume which results in a single value with no inherent fluctuation. Those quantities, e. g. fluence or absorbed dose, are called non-stochastic quantities.
Units A unit is a reference sample of a quantity with which other quantities of the same kind are compared. Every quantity is expressed as a product of a numerical value and a unit. Generally the use of the International System of Units (SI) as given by the BIPM [91BI] is recommended which is based on the 7 base units meter, kilogram, second, ampere, kelvin, mole and candela. Derived SI-units are often given special names like joule, becquerel or gray. Some other units are, however, generally used which are outside of the international system, e. g. the electron volt (eV) and the atomic mass unit (u) - and the time units minute, hour, day and year are also generally permitted. Nevertheless, other units even if not recommended are still in use in radiation measurements and radiation protection. Table 1 presents some numerical relationships between those units and the SI-units recommended. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
Table 4.1. Former units and its relations to SI-units Quantity
Symbol
SI-unit Name
Former units rad
1 rad = 0.01 Gy
roentgen
1 R = 2.58 10−4 J kg−1
rem
1 rem = 0.01 Sv
absorbed dose
D
J kg−1
exposure
X
C kg−1
dose equivalent
H
J kg−1
sievert (Sv)
activity
A
s−1
becquerel (Bq) curie
potential alpha energy
cp
J m−3
gray (Gy)
concentration potential alpha energy exposure
1 Ci = 3.7 1010 Bq
Working level 1 WL = 2.08 10−5 J m−3 = 1.30 108 MeV m−3
Ep
J h m−3
Working level month (T = 170 h) 1 WLM = 3.54 10−3 J h m−3 = 2.21 1010 MeV h m−3
4.2 Radiation field quantities Radiation field quantities are non-stochastic quantities defined at any point of a radiation field. Radiation fields may consist of various types of particles and the field quantities are always related to a specific particle type. This is usually expressed by adding the particle name to the quantity, e.g. photon fluence or neutron flux. There are two classes of radiation field quantities referring either to the number of particles or to the energy transported by them. A radiation field of a specific particle type can be fully described by the number N of particles, their distribution in energy as well as their spatial, directional and temporal distribution. This needs the definition of scalar and vectorial quantities. While in radiation dosimetry mostly scalar field quantities are used, vectorial quantities are often needed and applied in radiation transport theory and calculations. The radiation field quantities are defined in specifying the field in increasing detail.
4.2.1 Scalar radiation field quantities Particle number, radiant energy The particle number, N, is the number of particles that are emitted, transferred, or received. Unit: 1
The radiant energy, R, is the energy (excluding rest energy) of the particles that are emitted, transferred, or received. Unit: joule, J
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For particles of energy E (excluding rest energy), the radiant energy, R, is equal to N·E. The distributions, NE and RE, of the particle number and the radiant energy with respect to energy are given by NE = dN/dE
and RE = dR/dE
(4.2.1a+b)
where dN is the number of particles with energy between E and E + dE and dR is their radiant energy. Flux, energy flux The flux, N& , is the quotient of dN by dt, where dN is the increment of the particle number in the time interval dt. N& = dN/dt Unit: s−1
The energy flux, R& , is the quotient of dR by dt, where dR is the increment of the radiant energy in the time interval dt. R& = dR/dt Unit: W
The term flux has often been employed for the quantity fluence rate (see below). This usage should be avoided. Fluence, energy fluence The quantity fluence is based on the idea of counting the number of particles incident or passing a small sphere. It is defined by: The fluence, Φ, is the quotient of dN by da, where dN is the number of particles incident on a sphere of cross-sectional area da. Φ = dN /da Unit: m−2
The energy fluence, Ψ, is the quotient of dR by da, where dR is the radiant energy incident on a sphere of cross-sectional area da. Ψ = dR /da Unit: J m−2
The fluence is independent of the directional distribution of the particles passing the sphere. In calculations, fluence is often expressed in terms of the length of trajectories of particles passing a volume dV. The fluence, Φ, is given by
Φ = dl /dV
(4.2.2)
where dl is the sum of the lengths of trajectories through this volume. The distributions, ΦE and ΨE, of the fluence and energy fluence with respect to energy are given by
ΦE = dΦ/dE
and ΨE = dΨ/dE
(4.2.3a+b)
These quantities are often called spectral fluence and spectral energy fluence, respectively. Fluence rate, energy fluence rate The temporal distribution of the fluence and energy fluence is generally of interest. This results in the following definitions: The fluence rate, Φ& , is the quotient of dΦ by dt, where dΦ is the increment of the fluence in the time interval dt. Φ& = dΦ /dt Unit: m−2 s−1
Landolt-Börnstein New Series VIII/4
The energy fluence rate, Ψ& , is the quotient of dΨ by dt, where dΨ is the increment of the energy fluence in the time interval dt. Ψ& = dΨ /dt Unit: J m−2
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4 Radiological quantities and units
[Ref. p. 4-27
The fluence rate has often been termed particle flux density. Because the term density mostly characterises a mass density (kg−1), it is recommended to use the term fluence rate and not particle flux density. Particle radiance, energy radiance The energy radiance, Ψ&Ω , is the quotient of d Ψ& by dΩ, where dΨ& is the energy fluence rate of particles propagating within a solid angle dΩ around a specified direction. Ψ&Ω = dΨ& / dΩ Unit: W m−2 sr−1
The particle radiance, Φ&Ω , is the quotient of d Φ& by dΩ, where d Φ& is the fluence rate of particles propagating within a solid angle dΩ around a specified direction. Φ& Ω = dΦ& / dΩ Unit: m−2 s−1 sr−1
The specification of a direction Ω requires two variables. In a spherical coordinate system with a polar angle, θ, and an azimuthal angle, ϕ, dΩ is equal to sinθ dθ dϕ. The distribution of particle radiance and energy radiance with respect to energy are given by
Φ& Ω ,E = dΦ& / dΩ dE and Ψ&Ω ,E = dΨ& / dΩ dE
(4.2.4a+b)
4.2.2 Vectorial radiation field quantities Radiometric quantities are often used to describe the flow of radiation in specific directions. This needs the definition of vectorial quantities. For example, the scalar angular differential quantities like particle radiance and energy radiance are transferred to vectorial quantities by multiplication with the unit vector Ω in a specific direction. Vectorial quantities are vectorial particle radiance, Φ& Ω
with
vectorial energy radiance, Ψ& Ω vectorial fluence rate, Φ&
with
vectorial energy fluence rate, Ψ&
with
vectorial fluence, Φ
with
vectorial energy fluence, Ψ
with
with
& = Ω ⋅ Φ& Φ Ω Ω & & ΨΩ = Ω⋅ΨΩ Φ& = Φ& dΩ
∫ & Ψ = ∫ Ψ& dΩ Φ = ∫ Φ& dt Ψ = ∫ Ψ& dt Ω
Ω
unit: m−2 s−1 sr−1, unit: W m−2 sr−1, unit: m−2 s−1, unit: W m−2, unit: m−2, unit: J m−2.
A detailed description is given in ICRU Report 60 [98I1]. The distribution of a quantity with respect to energy of the particle considered is described by an index E similar to the scalar quantities. For example, the distribution of the vectorial particle radiance is given by Φ& Ω ,E = Ω ⋅ Φ& Ω ,E unit: m−2 s−1 sr−1 MeV−1.
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4.3 Interaction coefficients and quantities Ionising radiation is either charged (e.g. electrons, positrons, protons and α-particles) or uncharged (e.g. photons and neutrons). This dominates the main interaction with matter. While charged particles (called directly ionising particles) are mainly slowed down by electromagnetic interactions with electrons of the target atoms, the uncharged particles (indirectly ionising particles) interact with matter in separated events. Indirectly ionising particles are either absorbed or its energy and direction is altered. The probabilities of specific interactions between radiation and matter are characterized by interaction coefficients. They refer to specific interaction processes, type and energy of radiation and the matter involved. The definition of those coefficients important for dosimetry and related quantities are given in this Section.
4.3.1 Cross section The cross section is the most fundamental interaction coefficient. It is defined as follows. The cross section, σ, of a target entity, for a particular interaction produced by incident particles is the quotient of P by Φ, where P is the probability of that interaction for a single target entity when subjected to the particle fluence, Φ. It is
σ = P/Φ
unit: m2.
A special unit often used for the cross section is the barn (b) with 1 b = 10−28 m2. Cross sections mostly vary with the energy of the incident radiation (notation: σ(E)). The distribution of a cross section with respect to the energy and direction of the emitted radiation is often called differential cross section (dσ/dΩ : angular differential cross section, dσ/dE: energy differential cross section, d2σ/dEdΩ : energy and angular differential cross section). The total cross section, σT, is the sum of the cross sections of all possible interaction channels for an incident particle of a given type and energy and a given target material.
4.3.2 Mass attenuation coefficient and mass energy transfer coefficient For an infinite small parallel beam of uncharged radiation, the interaction of radiation with matter results in an attenuation of the incident beam with depth in material. This can be described by the relation dN = µ ⋅ dl N
(4.3.1)
where dN/N is the fraction of particles that experience interactions in traversing a distance dl in the material. µ is the linear attenuation coefficient. The reciprocal of µ is called the mean free path λ of an uncharged particle. In first order µ is proportional to the density ρ of a material. This leads to the definition of the mass attenuation coefficient, µ/ρ . µ 1 dN = ρ ρ dl N
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unit: m2 kg−1
(4.3.2)
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4 Radiological quantities and units
[Ref. p. 4-27
The mass attenuation coefficient is related to the total cross section by N µ NA = σ = A ρ M T M
∑σ
(4.3.3)
J
J
where NA is the Avogadro constant and M the molar mass of the material considered. σJ are the cross section related to the interaction of type J in this material. For uncharged particles the transfer of energy to charged particles in the material is of high interest in dosimetry. This is expressed by the mass energy transfer coefficient, µtr/ρ , which is defined by µ tr 1 dRtr = ρ ρ dl R
unit: m2 kg−1
(4.3.4)
where dRtr/R is the fraction of incident radiant energy that is transferred to kinetic energy of charged particles by interactions when traversing a distance dl in the material of density ρ. If incident uncharged particles of a given type and energy can produce several types of interactions in a material, µtr/ρ can be expressed in terms of the partial cross Sections, σ J, by the relation µ tr N A = ρ M
(4.3.5)
∑ f Jσ J J
where fJ is the average fraction of the incident particle energy that is transferred to kinetic energy of charged particles in an interaction of type J. The mass energy transfer coefficient is related to the mass attenuation coefficient by µ tr µ µ = f = ρ ρ ρ
∑ f Jσ J J ∑σ J
(4.3.6)
J
For µtr/ρ of a compound material the material is usually treated as consisting of independent atoms and the contributions from the different components are summed considering their partial density. A small part of the energy transferred to charged particles may not be locally absorbed in the material but further transferred to secondary photons (e.g. Bremsstrahlung). Therefore, an additional coefficient, the mass energy absorption coefficient, µen/ρ, is defined by the product of µtr/ρ and (1-g) where g is the fraction of the energy of charged particles that is lost in radiative processes in the material. Data of mass energy transfer and mass energy absorption coefficients are given by Seltzer [93Se]. For neutron radiation, the kerma coefficient K/Φ (kerma per unit neutron fluence, often called kerma factor) is mostly used instead of µtr/ρ for characterising the energy transfer (see 4.4.1). Data of kerma coefficients for biological important materials from thermal to 150 MeV neutrons are published by Chadwick et al. [99Ch] and in ICRU Report 63 [00I1].
4.3.3 Mass stopping power and linear energy transfer (LET) Charged particles passing matter loose energy by collisions with electrons, by emission of bremsstrahlung in the electric fields of nuclei or atomic electrons or by elastic Coulomb scattering and inelastic nuclear processes on atoms or nuclei. This effect is characterized by the mass stopping power S/ρ for charged particles in a material with density ρ. It is S
ρ
=
1 dE ρ dl
unit: J m2 kg−1
(4.3.7) Landolt-Börnstein New Series VIII/4
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where dE is the energy lost by a charged particle in traversing a distance dl in the material. S =dE/dl is called the linear stopping power. E may be given in eV and the unit of S/ρ may then expressed in eV m2 kg−1 or other multiples like MeV cm2 g−1, for example. The transfer of energy from the primary charged particle to secondary electrons is of specific interest in dosimetry, especially to those electrons receiving a kinetic energy less than a given value only. They will locally deposit their energy near to the track of the primary particle. This led to the definition of the quantity linear energy transfer (LET) or restricted linear electronic stopping power L∆ given by L∆ =
dE∆ dl
unit: J m−1, often used keV µm−1
(4.3.8)
where dE∆ is the energy lost by a charged particle due to electronic collisions when traversing a distance dl minus the sum of the kinetic energies of all electrons released with kinetic energies in excess of ∆. This definition given in ICRU Report 60 [98I1] differs from earlier ones [80I1] in a way that L∆ now includes the binding energies for all collisions and the threshold of the kinetic energy of the released electrons is now ∆ instead of ∆ minus the binding energy. ∆ is often given in eV and then the notation L100 means an energy cutoff of 100 eV. L∞ is often called unrestricted linear energy transfer L and is equal to Sel, the electronic stopping power due to collisions with electrons.
4.3.4 Mean energy expended in a gas per ion pair formed In dosimetry, where often charge measurements due to ionisation in gases are the basis of dose determinations, the kinetic particle energy necessary to create an ion pair is of general interest. This led to the definition of the mean energy expended in a gas per ion pair formed W. It is W=
E N
unit: J
(4.3.9)
where N is the number of ion pairs when the initial kinetic energy E of the charged particle is completely dissipated in the gas considered. This definition includes also the ions produced by secondary electrons or bremsstrahlung.
4.4 Quantities related to energy transfer 4.4.1 Stochastic quantities The energy transfer from incident particles to a target material is a stochastic process. For example, the energy deposition along a track of a charged particle is randomly distributed. The values of stochastic quantities are, therefore, subject to inherent fluctuations. They generally follow a probability distribution and mean values may be given. For example, a Poisson distribution is already uniquely determined by its mean value. Stochastic quantities are often used in microdosimetry in order to describe the energy transfer to very small volumes.
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4 Radiological quantities and units
[Ref. p. 4-27
4.4.1.1 Energy deposit and energy imparted The energy deposit ε i, is the energy deposited in a single interaction i, thus
ε i = ε in − ε out + Q
unit: J, often used eV
where ε in is the energy of the incident ionising particle (excluding rest energy), ε out the sum of energies of all ionising particles leaving the interaction (excluding rest energy) and Q the change in the rest energies of the nucleus and all particles involved in the interaction. Q > 0 means a decrease of rest energy, Q < 0 an increase. The total energy transferred to matter in a given volume is often of interest. The energy imparted ε to the matter in a given volume is the sum of all energy deposits in the volume
ε=
∑ε
unit: J, often used eV
i
The mean energy imparted ε to the matter in a given volume is a non-stochastic quantity and can be expressed in terms of the radiant energy Rin (sum of all radiant energies of the incoming particles) and Rout (sum of the radiant energies of all outgoing particles). It is
ε = Rin − Rout +
∑Q ,
unit: J, often used eV
4.4.1.2 Lineal energy and specific energy Corresponding to the non-stochastic quantity LET the stochastic quantity lineal energy y is defined by the quotient of ε s by l , where ε s is the energy imparted to the matter in a given volume by a single energy deposition event and l is the mean chord length of that volume, thus
y=
εs l
unit: J m−1, mostly used keV µm−1
This quantity is mainly used in microdosimetry, especially in measurements with low-pressure tissueequivalent proportional counters where single event distributions in terms of y are measured. The specific energy (imparted) z is the quotient of ε by m, where ε is the energy imparted to matter of mass m. It is
z=
ε m
unit: gray (Gy), 1 Gy = 1 J kg−1
The specific energy includes the energy transferred to the matter m from all events involved.
4.4.2 Non-stochastic quantities 4.4.2.1 Kerma, kerma rate The transfer of energy from uncharged particles (indirectly ionising particles, e.g. photons or neutrons) to matter is performed by the liberation and slowing down of secondary charged particles in this matter. This led to the definition of the quantity kerma. The kerma K is the quotient of dEtr by dm, where dEtr is the sum of the kinetic energies of all charged particles liberated by uncharged particles in a mass dm of material. It is given by K=
dE tr , dm
unit: gray (Gy), 1 Gy = 1 J kg-1
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Kerma is a non-stochastic quantity. For a very small mass element, however, the energy transfer dEtr underlies in principle stochastic fluctuations. In this case a non-stochastic quantity means that dEtr is seen to be the expectation value of the sum of energies of liberated charged particles. For monoenergetic uncharged particles of energy E the kerma is related to the fluence by K = Φ E (µtr/ρ)
(4.4.1)
For a given energy distribution ФE of the uncharged particles the kerma can be calculated by K = ∫ ΦE E (µtr/ρ) dE
(4.4.2)
For neutrons, the quotient of K by Φ, is called kerma coefficient (often also called kerma factor) where Φ is the neutron fluence (see 4.3.2). The kerma rate, K& , is the quotient of dK by dt, where dK is the increment of K in the time interval dt. dK , K& = dt
unit: Gy s−1.
4.4.2.2 Absorbed dose, absorbed dose rate The quantity absorbed dose is a basic quantity in radiation dosimetry and relevant to all types of ionising radiation whether directly or indirectly ionising. It is a non-stochastic quantity and defined by: D=
ε
unit: gray (Gy), 1 Gy = 1 J kg−1
dm
where ε is the mean energy imparted to the matter of mass dm. While kerma is related only to those secondary charged particles produced in dm but transferring their energy to matter partially also outside dm, absorbed dose includes all energy transferred to dm partially also from secondary charged particles produced outside but entering dm. Only under charged particle equilibrium and negligible radiation losses, however, the values of absorbed dose and kerma are equal in a homogeneous material. The absorbed dose rate, D& , is the quotient of dD by dt, where dD is the increment of the absorbed dose in the time interval dt. It is dD D& = dt
unit: Gy s−1
4.4.2.3 Exposure, exposure rate The quantity exposure is related to the production of charges in gas by ionising radiation. Historically its definition is elder than kerma or absorbed dose. Its use, however, is restricted to photons only. The exposure X is the quotient of dQ by dm, where dQ is the absolute value of the total charge of the ions of one sign produced in air when all the electrons and positrons liberated or created by photons in air of mass dm are stopped in air.
X =
dQ dm
unit: C kg−1 (former: roentgen, R)
It should be noted that in this definition the charges due to ionisation arising from the absorption of bremsstrahlung emitted by the electrons is not included in dQ. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
The exposure rate, X& , is the quotient of dX by dt, where dX is the increment of the exposure in the time interval dt. It is
dX X& = dt
unit: C kg−1 s−1
4.5 Dose quantities in radiation protection 4.5.1 Concept of radiation protection quantities The development of dosimetric concepts and the definition of specific quantities for use in radiation protection have a long history. An important basis for the present concepts was already provided in the 60's and 70's by both the International Commission on Radiological Protection (ICRP) and the International Commission on Radiation Units and Measurements (ICRU). In 1991 in its Publication 60 [91I1], the ICRP has published its most recent general recommendations for radiation protection including a system of quantities. The ICRP and ICRU have developed a hierarchy of quantities for radiation protection applications comprising primary limiting dose quantities (called “protection quantities”) taking account of human body properties and operational dose quantities for monitoring of external exposure. For monitoring of internal exposure other quantities than dose quantities are used. The basic idea of a primary limiting quantity is to relate the “risk” of exposure to ionising radiation (exposure by internal and external radiation sources) to a single (dose) quantity which takes account of the man as a receptor, the different radiation sensitivities of various organs and tissues and the different radiation qualities. Other influence parameters, however, e.g. the influence of dose and dose rate or sex and age of a person exposed on the biological response and the exposure risk, were not explicitly considered in the definition of these quantities. Operational quantities are dose equivalent quantities defined for use in radiation protection measurements related to external exposure (area or individual monitoring). They are needed for monitoring external exposures because • • • •
protection quantities are generally not measurable, for area monitoring a point quantity is needed, a non-isotropic human-body related quantity like effective dose is not appropriate in area monitoring, instruments for radiation monitoring need to be calibrated in terms of an operational quantity.
Operational quantities usually provide an estimate or upper limit for the value of the limiting quantities due to an exposed, or potentially exposed, person under most irradiation conditions. They are often used in practical regulations instead of the primary limiting quantities. For internal exposure, however, other methods are used and no similar dose quantities have been defined. In this case organ doses or effective dose are estimated from the information on intake or excretion of radioactive substances. Model based conversion coefficients exist for a large number of radionuclides relating the intake to organ doses and effective dose (see 4.6 and Chapter 7). Both, protection and operational quantities can be related to “radiation field quantities” (see Sect. 4.2) or air kerma (see Sect. 4.3) which are point quantities defined in any point of a radiation field and whose units are directly realised through primary standards at national standards laboratories since long time. The numerical relations (conversion coefficients) between those quantities and the protection or operational dose quantities are given in Chapter 6.
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4.5.2 Protection quantities In 1977 the ICRP [77I1] introduced the tissue (or organ) dose equivalent HT and the effective dose equivalent HE whose definition takes care of the relative variation of the tissue response with different types of radiation and different tissues or organs in the human body by introducing tissue weighting factors [73Jac]. Although in general this concept was not changed by ICRP 60 [91I1] in 1990, important modifications, however, were introduced e.g. replacing dose equivalent quantities by equivalent dose quantities. The present system of quantities is summarised in the following. 4.5.2.1 Absorbed dose and equivalent dose in a tissue or organ The absorbed dose in a tissue or organ DT is the absorbed dose averaged over the volume of a tissue or organ T (rather than at a point). While the absorbed dose at a point generally is the fundamental dose quantity, in radiation protection the mean dose in an organ becomes the basic protection quantity correlated with the exposure risk. This concept is obviously based on the linear dose-effect relationship and the additivity of doses for risk assessment as an appropriate approximation in the low dose range. The equivalent dose in a tissue or organ is defined by HT =
∑w D R
T, R
unit: sievert (Sv) (1 Sv = 1 J kg−1)
R
where DT,R is the mean organ dose in the tissue or organ T from radiation of type R incident on the human body and wR are radiation weighting factors characterising the biological effectiveness of the specific radiation R relative to photons. These factors have replaced the mean quality factors used in the concept of organ dose equivalent before [77I1]. The sum is taken over all types of radiation involved. 4.5.2.2 Radiation weighting factors For external irradiation, the values of the radiation weighting factors wR are given by the parameters of the external radiation field only (type and energy distribution of the radiation incident on the body). This means that wR is a body-averaged value representing a mean value for the relative biological effectiveness of all tissues of the body and any local variation of the radiation quality in the human body which may result from the generation of secondary radiation of different types in the body, is not explicitely considered. This effect is mainly important in the case of incident neutrons where at low energies secondary photons strongly contribute to the absorbed doses of various organs. The wR values for various types of radiation are specified in ICRP 60 in a table (see Table 4.2). For photons, electrons and muons of all energies a value of one is fixed with the exception of Auger electrons emitted from nuclei bound to DNA. For this case there exists no ICRP recommendation until now. The radiation weighting factor for neutrons depends on the neutron energy. Different wR values are given by either a step function or a continuous function as an approximation (see Fig. 4.1). In practice, neutron fields contain neutrons with a broad energy distribution. Because the use of a continuous wRfunction for effective dose estimation is more appropriate in these cases it is recommended to apply the continuous function in any case to avoid ambiguities. Then the weighting factor for neutrons ranges from 5 to 22 depending on neutron energy with its maximum value at 500 keV. All conversion coefficients for neutrons published in ICRP 74 [96I1] and ICRU 54 [98I2] are based on the continuous function only (see Chapt. 6).
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Table 4.2. Radiation weighting factors wR Radiation
wR
Photons Electrons1), muons
1 1
Neutrons: En En En En En
5 10 20 10 5
100 keV to 2 MeV >2 MeV to 20 MeV >20 MeV
Protons: Ep
5
> 2 MeV (unless recoil protons)
α-particles, fission fragments, heavy nuclei
20
As an approximation to the step function introduced for neutrons ICRP has specified a smooth wR function: wR = 5 + 17 exp (−[ln(2 En)]2/6) with En neutron energy in MeV.
1) With the exception of Auger electrons from atoms bound to DNA
The radiation weighting factor for incident external protons with energies above 2 MeV has been set to 5. It is, however, questioned if this value is appropriate for protons of all energies above 2 MeV. There exists a general opinion that a weighting factor of about 2 seems to be more realistic for high energy protons above about 5 to 10 MeV. External protons of lower energies have a small range in tissue and contribute to the skin dose only. 4.5.2.3 Effective dose The effective dose E is the weighted sum of the equivalent doses in tissues and organs T: E=
∑w H T
T
with
T
∑w
T
=1
unit: sievert (Sv)
(4.5.1)
T
where wT are tissue weighting factors characterising the relative sensitivity of the various tissues with respect to stochastic effects resulting from ionising radiation exposure and HT is the equivalent dose in one of the 13 specified tissues and organs (see Sect. 4.5.2.4). The effective dose is a quantity which is not sex specific or dependent on age of a person. In principle, the effective dose is determined by taking the dose values in all tissues and organs of an individual person. Those data, however, are never measurable. For external exposure, therefore, always calculated conversion coefficients are used which relate the external radiation field to the doses in the tissues and organs (see Chapt. 6). Following ICRP Report 74 [96I1], the effective dose is then calculated by E = wbreast H breast, female +
∑
T ≠ breast
wT
H T, male + H T, female 2
(4.5.2)
Under a given exposure condition (radiation field, direction of radiation incidence, exposure period), therefore, all persons are given the same effective dose value independent of sex and age. Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
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4.5.2.4 Tissue Weighting Factors The definition of effective dose takes care of the different radiosensitivity of the various organs and tissues in the human body with respect to cancer induction and mortality by introducing tissue weighting factors. Twelve tissues and organs are specified with individual weighting factors wT. The values have been developed from a reference population of equal numbers of both sexes and a wide range of ages. They are applied to workers, to the whole population, and to either sex including children and the unborn child (foetus). An additional “remainder” tissue with a weighting factor of 0.05 is also defined [91I1]. Its dose is given by the mean value from ten specified tissues and organs (see Table 4.3). The upper large intestine formerly included in the remainder, is now considered as part of the colon and has been replaced by the extrathoracic airways [93I3, 94I2]. While in the calculation of conversion coefficients for the intake of radionuclides the remainder dose is obtained from the mass-weighted doses to the single tissues and organs involved, the coefficients for external exposure are calculated giving identical weights to each of the remainder tissues [96I1]. Table 4.3. Tissue weighting factors wT Organ or tissue Gonads Bone marrow (red) Colon Lung Stomach Bladder Breast Liver Oesophagus Thyroid Skin Bone surface Remainder1)
wT 0.20 0.12 0.12 0.12 0.12 0.05 0.05 0.05 0.05 0.05 0.01 0.01 0.05
1) “Remainder” tissues are adrenals, brain, extrathoracic airways, small intestine, kidney, muscle, pancreas, spleen, thymus and uterus. The mean value of the equivalent doses of the ten remainder organs and tissues is to be multiplied by 0.05. If in a special case a single tissue or organ has an equivalent dose higher than each of the 12 individually defined organs and tissues, then this organ or tissue should get a weighting factor of 0.025 and the other 9 remainder tissues together a weighting factor of 0.025.
4.5.2.5 Committed or collective equivalent dose and effective dose Several subsidiary dosimetric quantities have been additionally defined. After an intake of radionuclides to a body these nuclides may give rise to equivalent doses in different tissues and organs of the body spread over long time depending on the physical and biological half-life of the radionuclides and their biokinetic behaviour in the body. The time integral of the equivalent dose rate is called the committed equivalent dose HT(τ ), in a tissue or organ T, where τ is the integration time (in years) following the intake at time t0. t0 +τ
H T (τ ) =
∫ H&
T
( t )dt
(4.5.3)
t0
If τ is not specified, it is implied that its value is 50 y for workers and from intake up to age 70 years for members of the public including children. For patients in nuclear medicine, the integration may run from t0 to ∞ because the biological and physical half-life of the radionuclides applied is much less than 10 y.
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4 Radiological quantities and units
[Ref. p. 4-27
The same specification holds for the quantity committed effective dose E(τ ) defined by the weighted sum of HT(τ ) over all specified tissues and organs T. E( τ ) =
∑w
T
⋅ H T (τ )
(4.5.4)
T
All dosimetric quantities referred before are related to a single tissue or organ of a single individual. Often it may be of interest to quantify the total dose a number of people received from one source or one release of radioactive material. The relevant quantity is called collective equivalent dose ST in a tissue or organ T and is defined by ∞
∫
ST = H T ⋅ 0
dN dH T dH T
unit: man sievert (man Sv)
(4.5.5)
where (dN/dHT)dHT is the number of individuals receiving an equivalent dose between HT and HT+dHT or by ST =
∑H
T, i
(4.5.6)
Ni
i
where Ni is the number of individuals in a subgroup i receiving a mean tissue equivalent dose HT,i . The summed effective doses of all members of a group or population is called the collective effective dose S defined in a similar way by ∞
∫
S = E⋅ 0
dN dE dE
or
S=
∑E ⋅N i
i
unit: man Sv
(4.5.7a+b)
i
where Ni is a subgroup i receiving a mean equivalent dose Ei .
4.5.3 Operational quantities 4.5.3.1 Dose equivalent and quality factor The radiation protection quantity dose equivalent H is defined by unit: Sv (1 Sv = 1 J kg−1)
H=QD
where D is the absorbed dose at the point of interest and Q a quality factor weighting the relative biological effectiveness of radiation. Q is defined as a function of the linear energy transfer L of a charged particle in water [77I1]. In principle, this concept of Q has not been changed by ICRP 60 [91I1], but the dose equivalent is now restricted to the definition of operational radiation protection quantities and the quality factor function Q(L) was modified in 1991 according to the following equation:
Q(L) =
1 0.32 L – 2.2 300/√L
for L < 10 keV/µm for 10 ≤ L ≤ 100 keV/µm for L > 100 keV/µm
(4.5.8)
Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
4 Radiological quantities and units
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The quality factor Q at a point in tissue is then given by [86I1]: ∞
Q=
1 Q( L )DL dL D L=0
∫
(4.5.9)
where DL is the distribution of D in L at the point of interest. This function is most important for neutrons because various types of secondary charged particles are produced in tissue in this case. 4.5.3.2 The concept of operational quantities The basic concept of the operational quantities is described in the ICRU Reports 39 and 43 [85I1, 88I1]. They have been introduced linking the external irradiation to the effective dose and the equivalent dose of the skin and eye lens in order to control their limits. The present definitions are given in ICRU Report 51 [93I1]. The operational quantities for radiation protection are dose equivalent quantities defined either for strongly penetrating or for weakly penetrating radiation incident on the human body (sometimes also the expressions penetrating and low penetrating are used instead of strongly and weakly penetrating radiation). The radiation is characterised as either weakly- or strongly penetrating depending on which dose (effective dose or skin equivalent dose) is closer to the corresponding limit. Weakly penetrating radiations are α-particles, β-particles with energies below 2 MeV and photons with mean energies below about 12 keV. Photons above this energy, electrons above 2 MeV and all neutrons are strongly penetrating radiation. Due to the different tasks in radiation protection monitoring − area monitoring for controlling the radiation at work places and definition of controlled or forbidden areas or individual monitoring for the control and limitation of individual exposures − different operational quantities were defined. While measurements with an area monitor are mostly performed free in air, an individual dosemeter is usually worn on the front of the body. As a consequence, in a given situation, the radiation field “seen” by an area monitor free in air differs from that “seen” by an individual dosemeter worn on a body where the radiation field is strongly influenced by the backscatter and absorption of radiation in the body. The operational quantities allows for this effect. They may be presented as follows:
Radiation type
Operational quantities for area monitoring individual monitoring
Strongly penetrating radiation Weakly penetrating radiation
ambient dose equivalent directional dose equivalent
personal dose equivalent personal dose equivalent
4.5.3.3 Operational quantities for area monitoring ICRU sphere phantom For all types of radiation the operational quantities for area monitoring are defined on the basis of a dose equivalent value at a point in a simple phantom, the ICRU sphere. It is a sphere of tissue-equivalent material (30 cm in diameter, density: 1 g cm−3, mass composition: 76.2 % oxygen, 11.1 % carbon, 10.1 % hydrogen and 2.6 % nitrogen). It adequately approximates the human body as regards the scattering and attenuation of the radiation fields under consideration.
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4 Radiological quantities and units
[Ref. p. 4-27
Aligned and expanded radiation field The operational quantities for area monitoring defined in the ICRU sphere should retain their character of a point quantity and the property of additivity. This is achieved by introducing the terms “expanded” and “aligned” radiation field in the definition of these quantities (see Fig. 4.1).
Φ
Φ ÄUHDOILHOG³
Φ Φ
Φ
ÄDOLJQHGILHOG³
ΦΦΦΦ Φ
ÄDOLJQHGDQGH[SDQGHGILHOG³
Fig. 4.1. Aligned and expanded field concept.
An expanded radiation field is a radiation field in which the spectral and the angular fluence have the same values in all points of a sufficiently large volume equal to the values in the actual field at the point of interest. The expansion of the radiation field ensures that the whole ICRU sphere is thought to be exposed to a homogeneous radiation field with the same fluence, energy distribution and directional distribution as in the point of interest of the real radiation field. If all radiation is (thought to be) aligned in the expanded radiation field so that it is opposed to a radius vector Ω specified for the ICRU sphere, the aligned and expanded radiation field is obtained. In this fictitious radiation field, the ICRU sphere is homogeneously irradiated from one direction, and the fluence of the field is the integral of the angular differential fluence at the point of interest in the real radiation field over all directions. In the expanded and aligned radiation field, the value of the dose equivalent at any point in the ICRU sphere is independent of the directional distribution of the radiation of the real radiation field. Ambient dose equivalent H*(d ) For area monitoring of strongly penetrating radiation the operational quantity is the ambient dose equivalent H*(10) defined by:
Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
4 Radiological quantities and units
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The ambient dose equivalent H*(d) at a point of interest in the real radiation field, is the dose equivalent that would be produced by the corresponding aligned and expanded radiation field, in the ICRU sphere at a depth d, on the radius vector opposing the direction of radiation incidence. For strongly penetrating radiation it is d = 10 mm and H*(d) is written H*(10). While this definition with the parameter d is given in ICRU [93I1] and ICRP [91I1] the most recent ICRU Report [01I1] dealing also with operational quantities defines ambient dose equivalent by H*(10), thus restricting its definition to strongly penetrating radiation only. In practice, however, this has already been realised because other values have never been used. As a result of the imaginary alignment and expansion of the radiation field, the contributions of radiation from all directions add up. The value of H*(10) is therefore independent of the directional distribution of the radiation in the actual field. This means that the reading of an area dosemeter for the measurement of H*(10) should be independent of the directional distribution of the radiation − an ideal detector should have an isotropic fluence response. Directional dose equivalent, H'(d ,Ω ) For area monitoring of weakly penetrating radiation the operational quantity is the directional dose equivalent H'(0.07,Ω ) or, in rare cases, H'(3,Ω ) defined by. The directional dose equivalent H'(d,Ω ) at a point of interest in the actual radiation field, is the dose equivalent that would be produced by the corresponding expanded radiation field, in the ICRU sphere at a depth d, on a radius in a specified direction Ω. For weakly penetrating radiation it is d = 0.07 mm and H'(d,Ω ) is written H'(0.07,Ω ). In case of monitoring the dose to the eye lens H'(3,Ω ) with d = 3 mm may be chosen. In practice H'(0.07,Ω ) is almost exclusively used in area monitoring for weakly penetrating radiation. For unidirectional radiation incidence the quantity may be written H'(0.07,α), where α is the angle between the direction Ω and the direction opposite to radiation incidence. The value of the directional dose equivalent can strongly depend on the direction Ω. The same is true for instruments for measuring weakly penetrating radiation − e.g. beta- or alpha-particle radiation – the reading of which can strongly depend on the orientation in space. In radiation protection practice, however, it is always the maximum value of H'(0.07,Ω ) at the point of interest which is of importance. It is usually obtained by rotating the dose rate meter during the measurement and looking for the maximum reading. 4.5.3.4 Operational quantities for individual monitoring Individual monitoring is usually performed with individual dosemeters worn on the body and the operational quantity defined for this application takes into account this situation. The true value of the operational quantity is determined by the irradiation situation near the point where the dosemeter is worn. For individual monitoring the operational quantity is the personal dose equivalent Hp(d). The personal dose equivalent Hp(d) is the dose equivalent in ICRU tissue at a depth d in a human body below the position where an individual dosemeter is worn. For strongly penetrating radiation a depth d = 10 mm is recommended. For weakly penetrating radiation a depth d = 0.07 mm is recommended. In special cases of monitoring the dose to the eye lens a depth d = 3 mm may be appropriate. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
The operational quantities for individual monitoring meet several criteria. They are defined for all types of radiation, additive with respect to various directions of radiation incidence, take into account the backscattering from the body and can be measured with a dosemeter worn on the body. The personal dose equivalent quantities, Hp(10) and Hp(0.07), are defined in the person, in the actually existing radiation field, and are measured directly on the person. Other requirements the quantities should satisfy can, however, be fulfilled only with additional specifications. An operational quantity for individual monitoring should allow the effective dose to be assessed or should provide a conservative estimate under nearly all irradiation conditions. This, however, requires that the personal dosemeter must be worn at a position on the body which is representative with respect to the exposure. For the usual dosemeter position in front of the trunk the quantity Hp(10) mostly furnishes a conservative estimate of E even in cases of lateral or isotropic radiation incidence on the body. In cases of exposure from the back, however, a dosemeter worn at the front side and correctly measuring Hp(10), will not provide a conservative estimate of E. A further requirement for an operational quantity is that it allows dosemeters to be calibrated under reference conditions in terms of that quantity. The personal dose equivalent is defined in the individual human body and obviously individual dosemeters cannot be calibrated in front of a real human body. For calibration, the human body must therefore be replaced by an appropriate phantom. Three standard phantoms have been defined by ISO for this purpose and the definition of Hp(10) and Hp(0.07) is extended to be defined not only in the human body but also in three phantoms of ICRU tissue (see Fig. 4.2) – a slab phantom (30 cm × 30 cm × 15 cm), a wrist phantom (a cylinder of 73 mm in diameter and 300 mm in length) and a finger phantom (a cylinder of 19 mm in diameter and 300 mm in length). In reference radiation fields used for calibration, the values of the quantities in these phantoms, Hp,slab(10) and Hp,slab(0.07) etc., are defined as the true values of the corresponding Hp-quantities (see also Sect. 6.1.2 and Sect. 10.2.1).
Fig. 4.2. Phantoms of ICRU tissue for the definition of Hp-quantities for calibration of individual dosemeters. a) slab phantom b) wrist phantom c) finger phantom
4.6 Radioactivity quantities The decay of a radionuclide is a stochastic process which means that the number of decays within a fixed time interval is described by a probability distribution. The expectation value of the number of decays is determined by the decay constant which is specific for each radionuclide and energy state (mostly the decay constant for the ground state is given). The decay constant λ of a radionuclide in a particular energy state is the quotient of dP by dt, where dP is the probability that a given nucleus undergoes a spontaneous transition from that energy state in the time interval dt. It is
λ=
dP dt
unit: s−1 Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
4 Radiological quantities and units
4-19
The half-life T1/2 of a radionuclide in a particular energy state is the mean time of the radionuclide in that state to decrease to one half of their initial number of nuclei. It is T1/2 = (ln 2)/λ.
4.6.1 Activity, specific activity, activity concentration, activity per area The activity A of an amount of a radionuclide in a particular energy state at a given time is the quotient of dN by dt, where dN is the expectation value of the number of spontaneous nuclear transitions from that energy state in the time interval dt. It is A=
dN dt
unit: becquerel (Bq), 1 Bq = 1 s−1
Radionuclides are mostly included in other solid, liquid or gaseous material and the amount is quantified by the quantities specific activity and activity concentration. The specific activity as is given by the quotient of the activity A by the mass m, where A is the activity of the radionuclide in the mass m. as =
A m
unit: Bq kg−1
The activity concentration cnuclide is given by the quotient of the activity A by the volume V, where A is the activity of the radionuclide in the volume V. c nuclide =
A, V
unit: Bq m−3
For the determination of contaminations the distribution of radionuclides on surfaces is of interest. The related quantity is the activity per unit area aa defined by the quotient of the activity A by the area F, where A is the activity of a radionuclide distributed on the surface area F. aa =
A F
unit: Bq m−2, often Bq cm−2
For decontamination of a surface from deposited radionuclides it is usually important if the radionuclides are removable or if they are diffused into the surface region of the material and are fixed near the surface in the material. If an aa-value is given it should be specified if this value is related to the removable part only or to the total activity at the surface.
4.6.2 Specific quantities for radon, thoron and their progeny Radon (222Rn) and thoron (220Rn) are gaseous radionuclides in the U- and Th-decay chain, respectively, occurring naturally (see 3.4.3). Their decay products are also radionuclides but metallic. While for radon the short-lived progeny 218Po, 214Pb, 214Bi and 214Po (see Table 4.4) are important in radiation protection, the important thoron progeny are 216Po, 212Pb, 212Bi and 212Po (see Table 4.5). In air there is usually a mixture of radon/thoron and short-lived radon/thoron progeny. These progeny are mostly attached to aerosols. A few percentages of them, however, are non-attached. The progeny may be deposited in the lung where its decay by alpha-particle emission is seen to be most important for lung cancer induction. Specific quantities have been defined taking care of this situation. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
Table 4.4. Data of radon (222Ra) progeny (nuclear data are from [NN98]) potential alpha energy number of Radionuclide half-life atoms per Bq per atom ε per Bq ε /λ A/λ i [MeV] [10−12 J] [MeV] T1/2 [10−12 J] 218
Po Pb 214 Bi 214 Po 214
1) 2)
1 2 3 4
3.10 min 26.8 min 19.9 min
268 2 320 1 723
164 µs
1)
13.69 7.69 7.69 7.69
2.19 1.23 1.23 1.23
3 670 17 800 13 100
589 2 860 2 100
0.106 0.513 0.381
2×10−3
2.9×10−4
6×10−8
no number is given because all atoms decay in less than 1 s and a calculated number would be much less than 1. factor k is defined in Eq. (4.6.1)
Table 4.5. Data of thoron (220Ra) progeny (nuclear data are from [NN98]) potential alpha energy number of Radionuclide half-life atoms per Bq per atom ε per Bq ε /λ A/λ i [MeV] [10−12 J] [MeV] [10−12 J] T1/2 216
Po
1
0.15 s
1)
Pb Bi 212 Po
2 3 4
10.6 h 60.6 min 304 ns
55 056 5 246
212
212
1) 2) 3)
k(2)
1)
14.6 7.8 7.82) 8.8
2.34
3.32
0.51
1.25 1.25 1.25
429 000 40 900
68 710 6 554
3.8×10−6
6×10−7
k(3)
6×10−6 0.913 0.087 8×10−12
no number is given because all atoms decay in less than 1 s and a calculated number would be much less than 1. mean value from decay of 212Bi and 212Po by α-particle emission. factor k is defined in Eq. (4.6.1)
Potential alpha energy The potential alpha energy εi of an atom i in the decay chain of radon or thoron is the total energy of alpha-particles emitted during the decay of this atom to the long-living 210Pb or stable 208Pb, respectively. The potential alpha energy of N atoms of type i is N·εi. The number of atoms N per Bq is equal to A/λ, where A is the activity of this radionuclide and λ its decay constant. The potential alpha energy per Bq is then given by ε /λ (unit: J Bq−1, often used MeV Bq−1). Concentration in air The potential alpha energy concentration cp,i of a short-lived radon (or thoron) progeny in air is the sum of the potential alpha energy εi of all atoms of this progeny present in a volume V divided by this volume. It is cp,i =
N iε i ε = ci i V λi
unit: J m−3 often MeV m−3
where Ni is the number of atoms of this progeny in the volume V, ci the corresponding activity concentration and λi the decay constant. The units are related by 1 J m−3 = 6.242 × 1012 MeV m−3. The potential alpha energy concentration (PAEC) cp of any mixture of short-lived radon (or thoron) progeny in air is the sum of the potential alpha energy concentrations of all progeny in the volume considered. Landolt-Börnstein New Series VIII/4
Ref. p. 4-27] cp =
4 Radiological quantities and units
∑ c = ∑ε ⋅ c p,i
i
i
i
λi
4-21
unit: J m−3 often MeV m−3
i
Historically, for the potential alpha energy concentration the unit working level (WL) has widely been used. While originally defined as the potential alpha energy concentration associated with the radon progeny in equilibrium with 100 pCi l−1, 1 WL is now accurately fixed equal to 1.300 × 108 MeV m−3 which equals 2.08 × 10−5 J m−3. Equilibrium equivalent concentration, equilibrium factor In case of radioactive equilibrium the activity concentration of radon cRn and of its progeny are equal. This, however, is usually not the case in air. For a non-equilibrium mixture a quantity equilibrium equivalent concentration ce has been defined. The equilibrium equivalent concentration (EEC) ce corresponding to a non-equilibrium mixture of progeny in air is the fictitious activity concentration of radon in radioactive equilibrium with its shortlived progeny that has the same potential alpha energy concentration cp as the actual non-equilibrium mixture. It is always ce ≤ cRn. The SI-unit for both quantities, ce and cRn, is Bq m−3. In order to avoid confusion, the values of ce are often marked Bq m−3 (EEC). The equilibrium equivalent concentration ce can be calculated from the activity concentrations of the progeny by the equation ce =
∑k ⋅c i
with ki = (ε i λi )
i
i
∑ (ε
i
λi )
(4.6.1)
The factors ki are given in Tables 4 and 5 and it is for radon progeny: and for thoron progenies:
ce = 0.106 cPo-218 + 0.513 cPb-214 + 0.381 cBi-214 + 6 × 10−8 cPo-214 −6
ce = 7 × 10 cPo-216 + 0.913 cPb-212 + 0.087 cBi-212 + 8 × 10
−12
cPo-212
(4.6.2) (4.6.3)
Obviously, the radionuclides 216Po, 214Po and 212Po can be ignored when calculating ce because of their very low ki–values. The equilibrium factor F is defined as the quotient of the equilibrium equivalent concentration and the activity concentration of the parent nuclide, radon, in air. F = ce / cRn
(4.6.4)
The value of F ranges from 0 to 1 and is a measure to what extent radioactive equilibrium between radon and its progeny is obtained. Mostly this is not the case and often a mean value of 0.4 is convenient for the situation in homes. The unattached progeny in air which are not attached to aerosols is also of special interest. The unattached fraction fp is defined by the relative fraction of the total potential alpha energy concentration which stems from progeny in air which are not attached to aerosols. It is fp =
cpf cp
=
cpf cpa + cpf
(4.6.5)
where c pa is the potential alpha energy concentration of the progeny attached to aerosols, c pf is that of the unattached fraction and cp is the sum of both parts. Landolt-Börnstein New Series VIII/4
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Inhalation exposure of individuals The exposure of an individual to radon progeny Pp is defined as the time integral of the potential alpha energy concentration cp in air to which the individual is exposed. T
Pp (T ) = cp (t ) dt
∫
(4.6.6)
0
where T is the period of the exposure. A similar integral is given if the equilibrium concentration ce(t) is taken for integration. It is then called the equilibrium equivalent exposure Pe(T): T
Pe (T ) = ce (t ) dt
∫
(4.6.7)
0
The potential alpha energy exposure Pp is often expressed in terms of working level month (WLM), even if not recommended for further use. This quantity has been introduced especially for specifying occupational exposure and a fixed time period T of 170 hours has therefore been chosen equal to a mean monthly working time. The relation to SI-units (see Table 4.1) is given by 1 WLM = 3.54 × 10−3 J h m-3 = 2.21 × 1010 MeV h m−3.
4.7 Quantities for internal dosimetry Internal exposure means an exposure by ionising radiation emitted from radionuclides incorporated and distributed in the body. A direct measurement of doses in a human body is not possible. For internal exposure there are, therefore, no specific operational dose quantities defined. In contrast to external monitoring usually the committed tissue or organ equivalent doses and committed effective dose are determined (period of 50 y for workers and a period up to the 70th year of life for members of the public including children) and complex compartment models are used to describe the long term biokinetic behaviour of the radionuclides in the human body (see Chapt. 7). The committed tissue and organ equivalent doses of an individual are usually determined by external measurements, e.g. activity concentration of specific radionuclides in the air, specific activity of food and water or contamination of the skin, and the application of calculated dose conversion coefficients (often called dose coefficients) which have been published for inhalation, ingestion and intake through the skin for a large number of radionuclides (see Chapter 7). Measured excretion data are usually used to estimate the intake of radionuclides subsequently and then conversion coefficients are applied to evaluate doses. The intake of radionuclides by inhalation, ingestion or through intact or wounded skin or the excretion by exhalation, urine, faeces etc, is determined in terms of measurable quantities. These are often activity concentrations in air , cnuclide (in terms of Bq m-3), and inhalation frequencies or inhaled activities, the specific activity of solids and liquids, as (in terms of Bq kg-1), the amount of ingested radioactive substances or their specific activity in excretions. Further details are given in Chapter 7. In cases where radionuclides emit high energy γ-rays their distribution in the body may be determined by external measurements of γ-rays with a whole-body counter (large γ-detectors well shielded against radiation from the environment) in combination with computer codes simulating the photon absorption in the body.
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4.8 Limits, constraints, action levels The system of radiation protection as recommended by the ICRP [91I1] is based on the following principles (see also Chapter 1): a) No practice involving exposure to radiation should be adopted unless it produces sufficient benefit to the exposed individuals or to society to offset the radiation detriment it causes (principle of justification). b) In relation to any particular source within a practice, the magnitude of individual doses, the number of people exposed, and the likelihood of incurring exposures where these are not certain to be received should all be kept as low as reasonably achievable, economic and social factors being taken into account. This procedure should be constrained by restrictions on the doses to individuals (dose constraints), or the risks to individuals in the case of potential exposures (risk constraints), so as to limit the inequity likely to result from the inherent economic and social judgements (the optimisation of protection). c) The exposure of individuals resulting from the combination of all the relevant practices should be subject to dose limits, or to some control of risk in the case of potential exposures. These are aimed at ensuring that no individual is exposed to radiation risks that are judged to be unacceptable from these practices in any normal circumstances. Not all sources are susceptible of control by action at the source and it is necessary to specify the sources to be included as relevant before selecting a dose limit (individual dose and risk limits). Generally, radiation protection takes care of exposure situations and doses which are relevant for the health of the persons involved or may not be ignored compared to the normal exposure from natural radiation sources. This means that there should exist a dose level below which exposures from radionuclides or other radiation sources may not be taken care of and where no regulations are necessary, independent of the fact that, in principle, any radiation may induce cancer or genetic defects. The ICRP sees such a dose range below a few tens of µSv (committed dose or dose per year) for a single individual which is about 1/100 of the normal exposure from natural sources in the environment. Often an upper boundary of 10 µSv (committed dose or dose per year) is used to decide if further investigations or actions are necessary. Usually the human exposures are classified in three different categories. The first is called occupational exposure which means any exposure incurred at work and principally as a result of situations which can be reasonably regarded as being in the responsibility of the operating management. It also includes potential exposures where the probability of a future exposure due to planned work forces may be estimated [97I1]. Medical exposures describe the exposure of patients during diagnostic and treatment. While medical exposures are intended to provide a direct benefit to the patient, the practice should be justified and optimized with respect to applied doses and the medical benefit. Public exposures are all exposures other than occupational or medical exposures. Public exposures include environmental exposures due to natural sources in the environment, e.g. natural actinides, radon, potassium-40 and cosmic radiation, but also those exposures due to artificial sources where the target group is the general population (details are given in Chapter 11). Examples are the broadly distributed radionuclides from the nuclear bomb test in 1950 to 1970 and the contamination due to the Chernobyl accident.
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Dose limits Dose limits have been recommended by the ICRP [91I1] for occupational exposure and for exposures to the public. They are given in terms of effective dose and few organ equivalent doses, always summed over a given period. The limits apply to the sum of the relevant doses from external exposure in the specified period (often one year) and from intakes of radionuclides in the same period. The corresponding internal dose is the 50-year committed dose (for occupational exposure) or the committed dose up to the age of 70 years (for members of the public). For public exposures the scope of these limits are restricted to doses incurred as the result of practices. Doses incurred in situations where the only protective action takes the form of an intervention are not included in system of dose limits (see action levels). Radon, thoron and their progeny in open air or in houses, natural radionuclides already in the environment and cosmic radiation on ground, are examples of those situations. While these dose limits are called primary limits, for practical reasons further limits (secondary or derived limits) are specified which are given in terms of operational or other quantities and derived from the primary limits. They are applied, for example, to define control or prohibited areas or annual limits of intake (ALI) of radionuclides. The ALI values which are specific for each radionuclide considered are also based on the committed dose for the same periods as mentioned above. The primary dose limits internationally recommended by ICRP [91I1] and the IAEA [96IA] are given below. Many, but not all countries have transferred these values into their national legislation and regulations. Some countries have either less or more restrictive regulations. As a consequence, the legal dose limits may, therefore, be different in different countries. For occupational exposure the effective dose is limited to 20 mSv per year averaged over 5 years (100 mSv in 5 years) with the further provision that the effective dose should not exceed 50 mSv within any single year. This limit avoids any deterministic effects of exposures and limits the stochastic effects to a risk level of about 10−3. For internal exposure the committed dose limit is restricted to 20 mSv in each year and the annual limit of intake (ALI) is related to this value. For women, when pregnancy has been declared, the embryo and foetus should be protected by applying for external exposure an additional equivalent dose limit to the surface of its abdomen of 2 mSv for the remaining period of the pregnancy and limiting the intake of radionuclides to 1/20 of the annual limit of intake. The detriment due to external weakly penetrating radiation mainly concerns the skin or the eye lens. In order to avoid deterministic effects the skin dose is, therefore, additionally limited to 500 mSv per year (averaged over any 1 cm2, regardless of the area exposed) and the dose to the eye lens to 150 mSv per year. For the same reason, the annual equivalent dose to the extremities (hands, feet) is also limited to 500 mSv. The approach for choosing dose limits to the public may be either based on the same ideas as for occupational exposure considering, however, the fact of the large number of persons involved or on the judgement on the existing dose level from natural radiation sources and its variation in different places where no influence on the health detriment of the population has been observed. For public exposure from sources given in practices, the ICRP has recommended a limitation of the effective dose to 1 mSv per year. In special circumstances, however, a higher value may be allowed in a single year if the average over 5 years does not exceed 1 mSv per year. The ICRP has also defined additional annual limits for the skin and the eye lenses which are 1/10 of the value recommended for workers (50 mSv averaged over any 1 cm2 of the skin and 15 mSv for the lens of the eye).
Dose constraints The control of public exposure in normal situations is usually performed by the application of controls at the different sources applying procedures of constrained optimisation and the use of prescriptive limits. A dose constraint, which is a value of individual dose from a defined source, should be used in the optimisation of protection to exclude protection options that would result in individual doses exceeding the constraint. Dose constraints are an integral part of the optimisation of protection and are thus prospectively. They are not, however, limits to be applied retrospectively.
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The above mentioned annual limit for occupational exposure means implicitly that the dose constraint for optimisation should not exceed 20 mSv per year. It is often convenient to define a homogeneous group of persons – a critical group – which are assumed to be most highly exposed by the single source considered and to apply the dose constraint to the mean dose in that critical group. For medical exposures no dose limits have been recommended because a radiotherapy or diagnostic treatment should always provide a direct benefit to the patient. The choice of the practice and its performance should be optimized with respect to applied doses and the medical benefit. In order to characterize good medical practice and to enable quality assurance programs for use in these cases, it is helpful to define constraints or reference values based on the actual state of the art of the various investigations and procedures. Such values are especially given in the various practices of diagnostics with X-rays and radionuclides.
Action levels While for the situation of occupational exposures generally dose limits are defined, there are other situations where the only protective action takes the form of an intervention, e.g. in cases of public exposure in areas of high level of natural radiation or in areas contaminated because of former human activities or accidents like e.g. nuclear bomb testing or the Chernobyl accident. For intervention situations action levels may be defined which specify dose levels or an activity concentration in air or the specific activity in materials of the environment which are of concern with respect to public exposure. If such a level is exceeded, this should initiate measures for a reduction of the exposure and different action levels may define different measures characterising the strength of the necessary intervention. While for public exposure from sources given in practices, a dose limit of 1 mSv per year is recommended which is in the order of the natural background exposure excluding radon, action levels for initiating protection measures to the public are mostly higher, depending on the strength of the recommended measures. A typical case for the definition of action levels is the exposure by radon and its progenies. Radon is always present in the environment and may appear in higher concentrations at specific work places or in homes. For radon, however, the specification of a dose level for actions is relatively complex because the dose coefficient relating a mean radon concentration at a place to an effective dose value depends on the mean equilibrium factor F (see 4.6.2) and the mean annual time people stay at this place. Furthermore, the coefficient is mainly based on modelling and includes a large uncertainty (see Chapt. 7). Mostly the radon action levels are given in terms of Bq m−3 or Bq m−3 h. For occupational exposure with the assumption of 2000 h at work and a mean equilibrium factor of 0.4 a conversion factor of 156 Bq m−3 mSv−1 or 62 Bq m−3(EEC)mSv−1, respectively, is given [94I1]. Actions for reducing the radon concentration at a work place may be performed if the mean annual radon concentration is in the range from 500 Bq m−3 to 1500 Bq m−3 (1000 Bq m−3 corresponds to about 6 mSv per year for 2000 h at work and F = 0.4). A mean annual concentration of 3000 Bq m−3 corresponds to about 20 mSv per year which is equal to the dose limit for occupational exposure. For public exposure in dwellings a conversion factor of 58 Bq m−3 mSv−1 or 23 Bq m−3(EEC)mSv−1, respectively, is given [94I1] under the assumption that a person stays 7000 h per year in the house (with F = 0.4 in the house) and the other time in free air (with a low radon concentration). Action levels are recommended by the ICRP also for this case. They are based on the following ideas. There exists a range of normal mean radon concentrations in dwellings where no actions are necessary or useful. For existing houses with higher mean concentrations actions for reducing such values should be considered. Future houses should be designed to stay within the normal range. Because of the very different situation in the various regions regarding the natural radon concentration in the ground and hence in houses, local recommendations or regulations may differ strongly in the different countries. The ICRP has specified an upper boundary of the normal mean radon concentration in dwellings with 200-400 Bq m−3 depending on the regional situation. Landolt-Börnstein New Series VIII/4
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Another situations which may occur are emergencies where people or the government needs advice what type of actions are necessary under given or expected exposures, e.g. staying at home, avoiding to eat fresh vegetables or drink fresh milk or leaving a defined area for some time. Such a situation is also a type of intervention where action levels may be defined in national regulations like those mentioned above. Principles and more detailed information are given in ICRP Publication 63 [93I4]. For immediate emergency situations there may also dose values be given for fireman and other rescue personnel in order to restrict their risks due to an exposure. Dose levels may be defined which should not be exceeded in one case or annually or in life time.
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4.9 References 73Jac 77I1 80I1 85I1 86I1 88I1 91BI 91I1 93I1 93I2 93I3 93I4 93Se 94I1 94I2 96IA 96I1 97I1 98I1 98I2 98NN 99Ch
00I1 01I1
Jacobi, W.: The concept of effective dose - a proposal for the combination of organ doses. Radiat. Environ. Biophys. 12 (1975) 101. ICRP: Recommendations of the international commission on radiological protection. ICRP Publication 26, Ann. ICRP 1 (3) (1977). ICRU: Radiation quantities and units. ICRU Report 33, Washington, 1980. ICRU: Determination of dose equivalents resulting from external radiation sources. ICRU Report 39. Bethesda, MD: ICRU Publications, 1985. ICRU: The quality factor in radiation protection. ICRU Report 40. Bethesda, MD: ICRU Publications, 1986. ICRU: Measurement of dose equivalents from external radiation sources, Part 2. ICRU Report 43. Bethesda, MD: ICRU Publications, 1988. Bureau International des Poids et Mesures: Le Système International d´Unités (SI). 6th edition. Pavillon de Breteuil, Sevres, 1991. ICRP: Recommendations of the international commission on radiological protection. ICRP Publication 60, Ann. ICRP 21 (1-3) (1991). ICRU: Quantities and units in radiation protection dosimetry. ICRU Report 51. Bethesda, MD: ICRU Publications, 1993. International Organisation for Standardisation: ISO Standards Handbook, Quantities and Units, 3rd edition, Geneva: International Organisation for Standardisation, 1993. ICRP: Age-dependent doses to members of the public from intake of radionuclides: Part 2 Ingestion Dose Coefficients. ICRP Publication 67, Ann. ICRP 23 (4) (1993). ICRP: Principles for intervention for protection of the public in a radiological emergency. ICRP Publication 63, Edited by ICRP, 1993. Seltzer, S.M.: Calculation of photon mass energy-transfer and mass energy-absorption coefficients. Radiat. Res. 136 (1993) 147. ICRP: Protection against Radon-222 at home and at work. ICRP Publication 65, Ann. ICRP 23 (2) (1994). ICRP: Dose coefficients for intakes of radionuclides by workers. ICRP Publication 68, Ann. ICRP 24 (4) (1994). IAEA: International basic safety standards for protection against ionizing radiation an for the safety of radiation sources. Safety Series No. 115, International Atomic Agency, Vienna, 1996. ICRP: Conversion coefficients for use in radiological protection against external radiation. ICRP Publication 74, Ann. ICRP 26 (3-4) (1996). ICRP: General principles for the radiation protection of workers. ICRP Publication 75, Ann. ICRP 27 (1) (1997). ICRU: Fundamental quantities and units. ICRU Report 60. Bethesda, MD: ICRU Publications, 1998. ICRU: Conversion coefficients for use in radiological protection against external radiation. ICRU Report 57. Bethesda, MD: ICRU Publications, 1998. NNDC: Nuclear Data, Decay Radiations. National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY, 1998. Chadwick, M.B., Barshall, H.H., Caswell, R.S., DeLuca, P.M., Hale, G.M., Jones, D.T.L., MacFarlane, R.E., Meulders, J.P., Schuhmacher, H., Schrewe, U.J., Wambersie, A., Young, P.G. A.: Consistent set of neutron kerma coefficients from thermal to 150 MeV for biologically important materials. Med. Phys. 26 (6) (1999) 974. ICRU: Nuclear data for neutron and proton radiotherapy and for radiation protection. ICRU Report 63. Bethesda, MD: ICRU Publications, 2000. ICRU: Determination of operational dose equivalent quantities for neutrons. ICRU Report 66, J. ICRU 1 (3) (2001).
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5 Shielding against ionizing radiation
In the shielding calculation for the radiation facility, simple dose estimation methods by using the shielding calculation constants are effective and widely used. These shielding calculation constants depend on the dose quantity to be estimated. Chapter 5 presents simple calculation methods and related constants for charged particles, photons and neutrons.
5.1 Introduction Shielding is an essential element of practical radiation protection. The necessary precautions depend especially on the kind of radiation. Charged particles as electrons and alpha particles can be absorbed in matter totally, as they have a maximum penetration depth, depending on their charge, mass, energy and on the properties of the absorbing media. A shielding slab will attenuate photons and neutrons just to a fraction of their primary intensity on the other hand, without enabling total absorption. Thus shielding calculations for charged particles mainly concern the evaluation of maximum penetration depths. As to photons an analytical calculation method has proved successful for simple geometries. It is based on the exponential law of attenuation of the unscattered photon component and a build-up portion for the scattered radiation component. The treatment of neutrons is more complicated than photon calculations. Simple methods to estimate dose rates can only be described for typical neutron sources and shielding materials, using results of more powerful shielding codes.
5.2 Stopping power and range To calculate the penetration of charged particles in matter, it is necessary to have information on the basic interactions that govern the passage through the shield. The predominant effects of protons and alpha particles are the elastic and inelastic collisions with electrons. While elastic collisions are resulting in a change of direction for the incident particle, inelastic collisions lead to energy loss and production of secondary radiation. Electrons traversing some distance in matter lose energy in numerous inelastic collisions with bound atomic electrons along their track. Furthermore Bremsstrahlung production becomes important in electron transport especially for high-Z media and high energies. The basic quantity for shielding purposes of charged particles is the stopping power, which is defined as the average energy loss per unit path length. It can be separated into the components collision stopping power Scol due to Coulomb collisions and radiative stopping power Srad due to Bremsstrahlung production. The range of charged particles is usually estimated on the basis of the continuous–slowing– down−approximation (csda). In this approximation particles are assumed to loose their energy continuously in the course of slowing down, with a fixed energy loss per path length given by the stopping power. That means energy-loss fluctuations are neglected. Integrating the reciprocal of the total
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stopping power with respect to kinetic energy E gives the csda-range r0, which is a rather good approximation of the mean path length by a particle on its course to rest. Generally the csda-range of a particle of initial energy E0 slowing down in matter to the rest energy Er is evaluated from the expression E0
r0 =
∫ [S (E )
col
+ S ( E ) rad ]−1 dE
(5.2.1)
Er
Because of the numerous scattering processes with angular deflections the csda-range is in most cases much larger than its projection on the initial direction of the particle track or the penetration depth. Tables of stopping powers and csda-ranges for protons and alpha particles are given in ICRU Report 49 [93ICR]. Corresponding values for electrons are presented in ICRU Report 35 [84ICR1], ICRU Report 37 [84ICR2], and ICRU Report 56 [97ICR]. An early review is given by Knop and Paul [64Kno].
5.3 Penetration depths of charged particles 5.3.1 Heavy charged particles Protons and alpha particles keep their initial direction rather far on their way, except near the end of the track. Because of this deviation from linear tracks, caused by multiple scattering, the average penetration depth R is a more useful quantity than the csda-range. It gives the expectation value of the distance in the initial direction of motion to that point, where the particle has slowed down to rest. The deviation from linearity of the particle track is described by the detour factor. It is practically equivalent to R/r0. The detour factor increases with increasing energy and decreasing atomic number. For this reason detour factors become important for low energies. The values are about 0.8 and 0.65 for protons and alpha particles of 1 MeV in lead, respectively. Fig. 5.3.1 gives average penetration depths R for common used materials of density ρ, estimated on the basis of the csda-range and detour factor tables of ICRU Report 49 [93ICR]. Table A5.3.1 gives the corresponding numerical values for some further materials as well (CD-ROM). As collision stopping powers at a given particle velocity are the same for all particles with the same charge number, proton ranges can be used as well for the estimation of the ranges of deuterons and positively charged pions and muons. At a given velocity the kinetic energy E of a particle with mass m is related to the kinetic energy Ep of a particle with mass mp by E=
m Ep mp
(5.3.1)
The relation between the range R(E) of particles with charge number z and the proton-range Rp(Ep) is given by R( E ) = R p (m p / m ⋅ E )
m 1 Fcorr mp z2
(5.3.2)
The correction factor Fcorr can be assumed to be unity for deuterons and positively charged pions and muons. It takes into account above all uncertainties of the particle charge. It is near unity for light ions at high particle energies. As slow ions can capture and lose electrons the effective charge may become much smaller than the nominal charge, resulting in a reduced stopping power. For alpha particles Fcorr approaches unity above 1 MeV and may increase to about 2 at lower energies. Fig. 5.3.2 gives further average penetration depths for protons and alpha particles in air on the basis of the csda-approximation. The average penetration depths for deuterons in air are calculated by Eq. (5.3.2). Table A5.3.2 gives the corresponding numerical values (CD-ROM). Landolt-Börnstein New Series VIII/4
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1
10
Average penetration depth Rρ [g cm−2]
0
10
lead -1 10 iron water -2 10 aluminum -3 10
-4 10 0.1
Fig. 5.3.1. Average penetration depth Rρ of protons in water, aluminium, iron and lead. 1
10
100
Proton energy E [MeV]
Fig. 5.3.2. Average penetration depth R of protons, deuterons and alpha particles in air. (ρ = 1.205×10−3 g/cm3).
Sample problem Average penetration depth of 3 MeV tritons in air. From Eq. (5.3.2) and Fig. 5.3.2 results: R(3 MeV) = Rp(1/3 · 3 MeV) 3/1 · 1/12 = 3 Rp(1 MeV) = 3 · 2.35 cm = 7.05 cm
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5.3.2 Electrons and positrons The range of electrons depends on the total (collision plus the radiative) stopping power. Monoenergetic electrons and beta rays suffer similar energy and angle straggling as heavy charged particles while slowing down. The deflections are however comparatively large because of their small mass. Thus the range resulting from the continuous-slowing-down-approximation (csda) is always larger than the projection of the particle track on the initial direction. The detour factor is near unit for low-Z media but reaches values up to 4 for lead. Different range definitions are used for practical purposes. In shielding calculations the maximum range Rmax is the most adequate quantity. It is defined as the depth at which the extrapolation of the tail of the transmission curve (versus electron beam axis depth) meets the Bremsstrahlung background. A frequently used numerical expression for the maximum range Rmax of beta rays of maximum beta energy Emax in matter of density ρ is given by Rmax ρ = −0.11 + 0.0121 + ( Emax / 1.92) 2
(5.3.3)
where Emax is in MeV and Rmax ρ in g cm−2 [97ICR]. The relation is useful in the energy range between 0.05 MeV and 5 MeV, for monoenergetic electrons as well. Fig. 5.3.3 gives maximum ranges for six common used materials, calculated by Eq. (5.3.3). Table A5.3.3 gives the corresponding numerical values (CD-ROM). 2
10
1
Maximum range Rmax [cm]
10
air*
0
10
glass
water
Al -1
Fe
10
Pb -2
10
-3
10
0.1
1
10
Fig. 5.3.3. Electron ranges Rmax in air (1.205×10−3), water (1.0), glass (2.23), Al (2.7), Fe (7.87) and Pb (11.35). Values in parentheses: densities in g/cm3. *Maximum range Rmax in m.
Electron energy E [MeV]
A rather simple empirical expression for the maximum range of electrons of energy E in matter of density ρ is given by Rmaxρ = E/2
(5.3.4)
where E is in MeV and Rmax ρ in g cm−2. Maximum ranges evaluated from Eq. (5.3.3) and (5.3.4) are shown in Fig. 5.3.4 together with csda-ranges for water and lead [84ICR2]. Eq. (5.3.4) turns out to be a cautious overall approximation of the maximum range overestimating for energies below 0.3 MeV and above 20 MeV. Eq. (5.3.3) provides considerable overestimation only in the energy range above 20 MeV. Table A5.3.4 gives the numerical values of the csda-ranges of air, water, Be, Al, Fe, Pb (CD-ROM). Landolt-Börnstein New Series VIII/4
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Electron range Rρ [gcm−2]
water(csda) lead (csda)
1
Eq. (5.3.4)
0.1
Eq. (5.3.3) 0.01 0.1
1
10
100
Fig. 5.3.4. Electron ranges Rρ evaluated from Eq. (5.3.3), Eq. (5.3.4) and by csda-approximation for water and lead.
Electron energy E [MeV]
Positrons undergo the same interactions in matter as electrons. Because of differences in collision and radiative stopping powers the positron csda-range varies between shortening at low energies and prolongation at high energies. The shortening of the positron csda-range (with respect to the corresponding electron csda-range) will amount up to 7 % at 0.1 MeV. At energies near 100 MeV the prolongation will be up to 11 % in lead and up to 2 % in water. On the assumption of positrons being annihilated before being slowed down to rest a further shortening of the positron range of up to 4 % in lead has to be considered [84ICR2]. In shielding of beta sources the range curves will yield a sufficient estimate of the necessary slab thickness. With high source activities Bremsstrahlung resulting from the deceleration of the beta particles in the material may need to be shielded as well.
5.4 Photons 5.4.1 Basic shielding concept For most gamma shielding studies photon energies of 10 keV to 10 MeV are important. In this energy range, the photoelectric effect, pair production and Compton scattering mechanisms of interaction predominate over all others. Of these three interactions, the photoelectric effect predominates at the lower photon energies; pair production is important only for higher-energy photons, while Compton scattering predominates at intermediate energies. In a few cases the shielding analyst may need to account also for coherent (Rayleigh) scattering, annihilation and fluorescence radiation. Most shielding analysis involves a study of the fluence field at pertinent locations with respect to the outside or inside shield. The purpose of such analysis is to predict the corresponding responses of some type of detector, and therefore the field information must be converted into the detector responses. These relate to the fluence by a multiplier called the detector response function. The fluence Φ of photons is the quotient of ∆N by ∆a, where ∆N is the number of photons, which enter a sphere of cross-sectional area ∆a. Detector responses for photons of interest are exposure, air kerma, absorbed dose and ambient dose equivalent.
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5-6
5 Shielding against ionizing radiation
[Ref. p. 5-32
In this chapter the ambient dose equivalent is used in accordance with ICRP Report 74 [96ICR]. The ambient dose equivalent rates are obtained by a shielding code that satisfies the Boltzmann transport equation. Using these calculated results, a simple method to estimate the ambient dose equivalent rate for typical radioactive sources and typical shielding media is introduced.
5.4.2 Attenuation data of radioactive sources in shielding materials The radionuclides dealt with in this chapter are shown in Table 5.4.1. The value of the ambient dose equivalent rate depends on the photon energy, the geometric configuration of the source, the nature and the thickness of the shielding material. The photon fluences emitted from a radioactive source with ordinary concrete (ρ = 2.10 g/cm3), iron, lead and water as shielding materials are calculated using the BERMUDA code [92Suz, 93Suz]. Densities and elemental weight fractions of ordinary concrete (ρ = 2.10 g/cm3 and 2.35 g/cm3) are shown in Table A5.4.1 (CD-ROM). The mass attenuation coefficients used were taken from the photon library PHOTX [88DLC], which are shown in Table A5.4.2 (CD-ROM). * (t ) , where t is the thickness In this text the ambient dose equivalent rate H& * (10) is represented as H& 10 of the shield layer. The purpose of shielding analysis is to predict the corresponding responses of certain types of detectors as defined by the symbol R, and to relate R to the fluence Φ (E, t) or to the flux density φ (E, t ) of photons of energy E by a multiplier to be called the conversion coefficient; where the flux density φ (E, t ) is defined as quotient of the incremental fluence ∆Φ that occurs at a specified position and the time interval ∆ t’. The conversion coefficients of the exposure dose rate ( X& / φ ) E , of the air kerma rate ( K& / φ ) and of the ambient dose equivalent rate ( H& * / φ ) for the flux density [cm−2 s−1] are given a
E
10
E
in Table A5.4.3 (CD-ROM). While the conversion coefficients of exposure rate ( X& / φ ) E [µR h−1 cm2 s] are taken from the third column of Table A.1 in ICRU Report 47 [92ICR], the ones of the air kerma rate * ( K& a / φ ) E [nGy h−1 cm2 s] and the ambient dose equivalent rate ( H& 10 / φ ) E [nSv h−1 cm2 s] are taken from the fourth and fifth column of Table A.21 in ICRP Report 74 [96ICR]. Here, the three conversion coefficients are given in special units for convenience’ sake of calculation. Consider a point source of activity A [Bq] and a point detector P located at a distance r + t [m] from the source, as illustrated in Fig. 5.4.1, where a shield layer of thickness t [m] is placed between the source and the detector. The energy spectrum of the photon flux density at point P is represented by φ (E, t ). Then the air kerma rate and ambient dose equivalent rate at point P are represented by the following formulas.
∫ (K& φ ) ⋅ φ (E ,t ) dE (t ) = ∫ (H& φ ) ⋅ φ (E ,t ) dE
K& a (t ) = * H& 10
a
(5.4.1)
E
* 10
(5.4.2)
E
By introducing the new constant Γ10* , the ambient dose equivalent transmission factor T(t), and effective conversion coefficient f * ( E ,t ) , the ambient dose equivalent rate H& * (t ) is simply obtained. 10
0
10
* 5.4.2.1 Simple method of calculating the ambient dose equivalent rate H& 10 ( t ) for radionuclides listed in Table 5.4.1
The flux density φi of photon energy group ‘i’ emitted from a radioactive point source of activity A [Bq] is represented by the following formula, where the detector P is located at distance r [m] from the source.
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
φi =
A 4π r 2
5 Shielding against ionizing radiation
5-7
(5.4.3)
Ii
Ii represents the transition yield (number of gamma rays per decay). Then, the ambient dose equivalent rate is obtained by the following summation. H& 10* =
∑ ( H&
/ φ ) i ⋅ φi
* 10
(5.4.4)
i
Constant: Γ10* With regard to Eq. (5.4.3) and Eq. (5.4.4), the constant Γ10* [µSv m2 MBq−1 h−1] is defined as the sum of components Γ10* ,i for specified photon energy groups ‘i’ of a gamma-emitting nuclide, as given in Eq. (5.4.5).
Γ10* =
∑Γ
* 10,i
=
i
1
∑ 4π I ⋅ ( H& i
* 10
/ φ ) i ⋅ 0.1
(5.4.5)
i
That means, the value of Γ10* represents the ambient dose equivalent rate H& 10* [µSv h−1] for A = 1 MBq and r = 1 m of a gamma-emitting nuclide. The values of Γ10* for 33 radionuclides are given in Table 5.4.1 [01JRIA]. With the constant Γ10* the ambient dose equivalent rate of a point source of a gamma-emitting radionuclide of activity A at distance r becomes A H& 10* = 2 Γ10* r
(5.4.6)
Ambient dose equivalent transmission factor: T(t) The ambient dose equivalent transmission factor T(t) is the quotient of the ambient dose equivalent rate * H& 10 (t ) by the ambient dose equivalent rate in the absence of shielding material H& 10* (0) . T (t ) = H& 10* (t ) / H& 10* (0)
(5.4.7)
The values of the ambient dose equivalent transmission factor T(t) for 33 radionuclides are given in Table A5.4.4 through A5.4.36 (CD-ROM) for four shielding materials, namely iron, lead, concrete (ρ = 2.10 g/cm3) and water [01Sak]. Furthermore, the transmission factors T(t) are presented in Fig. 5.4.2 through 5.38, as a function of the thickness t of shield layers [01Sak]. Using the constant Γ10* and the ambient dose equivalent transmission factor T(t), the ambient dose equivalent rate H& * ( t ) at a point detector P located a distance r + t [m] from the source, as illustrated in 10
Fig.5.5, is obtained by Eq.(5.4.8). H& 10* (t ) =
A ⋅ Γ10* ⋅ T (t ) ( r + t )2
Landolt-Börnstein New Series VIII/4
(5.4.8)
5-8
5 Shielding against ionizing radiation
[Ref. p. 5-32
* 5.4.2.2 Simple calculation of ambient dose equivalent rate H& 10 ( t ) for radionuclides unlisted in Table 5.4.1
Effective (averaged) conversion coefficient: f10* ( E0 ,t ) When f10* ( E ) defines the quotient of conversion coefficient ( H& 10* / φ ) E by ( K& a / φ ) E for a specified energy E, the ambient dose equivalent rate in Eq. (5.4.2) is given by the following formula:
∫f
* H& 10 (t ) =
* 10
(E ) ⋅ (K& a
φ )E ⋅ φ (E ,t ) dE
(5.4.9)
Equation (5.4.9) averaged by air absorbed dose rate spectrum at the point P in Fig. 5.4.1 gives Eq. (5.4.10). This formula was introduced as the effective conversion coefficient f10* = f10* ( E0 , t ) for photons of primary energy E0 and for shield layer thickness t is represented by Tanaka and Suzuki [91Tan]:
f
* 10
∫f =
* 10
(E ) ⋅ (K& a
φ )E ⋅ φ (E , t ) dE
(5.4.10)
∫ (K& φ ) ⋅ φ (E ,t ) dE a
E
Then, the relationship between the ambient dose equivalent rate and the air kerma rate becomes * H& 10 ( t ) = f 10* ⋅ K& a (t )
(5.4.11)
Effective conversion coefficients f10* = f10* ( E0 , t ) for iron, lead, concrete and water are given in Table A5.4.37 through A5.4.40 for photon energies from 15 keV to 10 MeV and shield layers up to 40 mfp (CD-ROM) [01Sak].
* 5.4.2.3 Calculation method of ambient dose equivalent rate H& 10 ( t ) using exposure dose rate and effective conversion coefficient
Step 1: The flux density of uncollided photons of energy E0 for source intensity S = A · I at distance r + t behind a shield layer of thickness t is
φ 0 (E0 ) =
S exp(− µt ) 2 4π (r + t )
(5.4.12)
The mass attenuation coefficients µ are listed in Table A5.4.2 (CD-ROM). Step 2: The exposure dose rate for uncollided photons is
(
X& 0 (E0 ) = φ0 (E0 ) ⋅ X& φ
)
E0
(5.4.13)
The conversion coefficients ( X& / φ )E are listed in Table A5.4.3 (CD-ROM). Step 3: The total exposure dose rate including collided photons is X& = B ⋅ X& 0
(5.4.14)
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
5-9
Exposure buildup factors B = B(E0, t ) for a point isotropic source in iron, lead, concrete (ρ = 2.35 g/cm3) and water are given in Table A5.4.41 through A5.4.44 for photon energies from 15 keV to 10 MeV and shield layers up to 40 mfp (CD-ROM), which are taken from ANSI/ANS6-4-3 [91ANS]. Step 4: The air kerma rate obtained from the total exposure dose rate is K& a = 8.764 ⋅ 10 −3 ⋅ X&
(5.4.15)
where 8.764·10–3 is the value of a typical conversion coefficient from exposure to air kerma, X& is in R h−1 and in Gy h−1. Step 5: Using the effective conversion coefficient f10* the ambient dose equivalent rate follows from Eq. (5.4.11). Values for not tabulated energies are obtained by interpolation.
5.4.3 An example of the calculation of an ambient dose equivalent rate Sample problem: 60Co source of A = 3.7⋅1013 Bq (1,000 Ci) is treated in a room enclosed by concrete wall of 100 cm thickness. Calculate the ambient dose equivalent rate at a point P on the outside wall. The calculation model is illustrated in Fig. 5.4.1, where the distance r between the source and the front of the concrete wall is 500 cm.
5.4.3.1 The method using the constant Γ10* and ambient dose equivalent transmission factor T(t ) Using the value Γ10* of 60Co in Table 5.4.1 and the one of T(t) at concrete 100 cm in Table A5.4.11 (CD-ROM), the ambient dose equivalent rate in Eq.(5.4.8) is obtained, as follows. 3.7 ⋅ 10 7 Γ10* T (t ) * (t ) = = 1.028 ⋅ 0.354 ⋅ 1.55⋅10−4 Sv h−1 = 56.4 µSv h−1 Eq. (5.4.8): H& 10 2 (5 + 1)
Values for not tabulated thicknesses of the shield layer are obtained by linear interpolation of log T(t ) – t [cm].
5.4.3.2 The method using the effective conversion coefficient f10* and the exposure dose rate conversion coefficient ( X& / φ ) Simplifying assumption: 60Co emits two gamma rays of E0 = 1.25 MeV per disintegration. Step 1: The photon flux density for the uncollided photons is: Eq. (5.4.12):
φ0 (E0 ) =
3.7 ⋅ 1013 ⋅ 2 ⋅ exp(− 12.02 ) cm−2 s−1 = 98.56 cm−2 s−1 4 ⋅ 3.14 ⋅ 6002
The linear attenuation coefficient of concrete (ρ = 2.10 g/cm3) for 1.25 MeV photons is µ = 0.1202 cm−1, which is obtained by the linear interpolation of log µ – log E, using the concrete mass attenuation coefficients of E0 = 1 MeV and 1.5 MeV given in Table A5.4.2 (CD-ROM). With t = 100 cm results µ t = 12.02.
Landolt-Börnstein New Series VIII/4
5-10
5 Shielding against ionizing radiation
[Ref. p. 5-32
Step 2: Calculation of the exposure dose rate of the uncollided photons:
(
X& 0 (E0 ) = φ0 (E0 ) ⋅ X& φ
Eq. (5.4.13):
)
E0
= 98.56 ⋅ 2.182 ⋅ 10−6 R h−1 = 2.151⋅10−4 R h−1
The conversion factor ( X& / φ ) E for 1.25 MeV is obtained by linear interpolation of log ( X& / φ ) E – log E, using ( X& / φ ) for 1 MeV and 1.5 MeV given in Table A5.4.3 (CD-ROM). E
Step 3: Calculation of total exposure dose rate including collided photons: X& = B ⋅ X& 0 = 21.84 ⋅ 2.151 ⋅ 10−4 R h−1 4.698⋅10−3 R h−1
Eq. (5.4.14):
The exposure buildup factor for E0 = 1.25 MeV in infinite concrete material is obtained by the following procedure. At first, concrete buildup factors B = 27.01 and 18.36 are obtained for E0 = 1 MeV and 1.5 MeV for µ t = 12.02 mfp. They result from the linear interpolation of log B(E) – µt [mfp], using buildup factors for µ t = 10 and 15 mfp from Table A5.4.43 (CD-ROM). Next, buildup factor B = 21.84 is obtained for E0 = 1.25 MeV by the linear interpolation of log B(E) – log E, using buildup factors of E0 = 1 and 1.5 MeV at µ t = 12.02. Step 4: Conversion from exposure dose rate X& to air kerma rate K& a : K& a = 8.764 ⋅ 10 −3 ⋅ 4.698 ⋅ 10 −3 Gy h−1 = 4.117⋅10−5 Gy h−1
Eq. (5.4.15):
* Step 5: Conversion from air kerma rate K& a to ambient dose equivalent H& 10 (t ):
The effective conversion coefficient f 10* in Eq. (5.4.11) is obtained by the following procedure. At first, the effective conversion coefficients of concrete f 10* = 1.308 and 1.256 are obtained for E0 = 1 MeV and 1.5 MeV and for µ t = 12.02. They result from the linear interpolation of log f 10* – µt [mfp], using f 10* for
µ t = 10 and 15 mfp from Table A5.4.39 (CD-ROM). Next, f 10* = 1.279 is obtained for E0 = 1.25 MeV by the linear interpolation of log f 10* – log E, using effective conversion coefficients of E0 = 1 and 1.5 MeV at µ t = 12.02. H& 10* (t ) = f 10* ⋅ K& a = 1.279 ⋅ 4.117 ⋅ 10−5 Sv h−1 = 52.7 µSv h−1
Eq. (5.4.11):
* The values of H& 10 (t ) obtained by the methods in 5.4.3.1 and 5.4.3.2 agree within 7 %.
Table 5.4.1 Constant Γ10* [µSv m2 MBq−1 h−1] Radionuclide 18
F Na 51 Cr 54 Mn 59 Fe 56 Co 57 Co 60 Co 64 Cu 24
Γ10*
Radionuclide
0.166 0.492 0.00547 0.130 0.171 0.492 0.0206 0.354 0.0307
65
Zn Ga 68 Ge* 75 Se 81 Rb* 85 Kr 85 Sr 99 Mo* 99m Tc 67
Γ10*
Radionuclide
0.0847 0.0268 0.158 0.0660 0.104 0.00037 0.0826 0.0444 0.0214
103
Pd* Ag 111 In 124 Sb 123 I 125 I 131 I 133 Xe 137 Cs* 110m
Γ10*
Radionuclide
Γ10*
0.0254 0.416 0.0663 0.266 0.0288 0.0357 0.0650 0.0170 0.0927
192
0.139 0.0685 0.0193 0.0175 0.251 0.00529
Ir Au 197 Hg 201 Tl 226 Ra* 241 Am 198
1) Radionuclides attached the mark * includes the effects of daughter’s radiation equilibrium. As the daughter of only considered. 2)
Γ10*
does not include effects by photon energy of below 30 keV and emission ratio within 0.1 %. However
226
Ra,
125
222
I and
Rn is 103
Pd
include the effect by photon energy of above 10 keV.
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
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5-16
5 Shielding against ionizing radiation
[Ref. p. 5-32
5.5 Neutrons 5.5.1 Basic shielding concepts In passing through shielding material, neutrons attenuate by elastic scattering, inelastic scattering and absorption. For high-energy neutrons over 10 MeV, inelastic scattering reaction is effective to decrease energy. Therefore iron is a suitable material because of its relatively big inelastic cross section. For neutrons with energy lower than 10 MeV, materials that contain hydrogen are used to utilize its elastic scattering reaction and absorption process. Dose calculations of neutrons are more complicated than photons because secondary gamma-ray dose by neutron capture and inelastic scattering should be taken into account. Moreover, reaction type and cross section depends strongly on the neutron energy and the shielding material. Therefore, dose evaluation is generally done by using a shielding code that solves the Boltzmann transport equation. In the present chapter a simple method to estimate dose rates from typical neutron sources and shielding materials is described using results of the shielding code ANISN [73Eng].
5.5.2 Attenuation data of various neutron sources in shield material Fig. 5.5.1 shows the geometry of the transmission calculations by ANISN. A neutron point isotropic source with source intensity S = 1 s−1 is surrounded by a thick spherical shield. The JENDL3.2 [90Shi] cross-section library is used in this calculation. The elemental composition and the density of each shield material are shown in Table 5.5.1. Dose rates were estimated using the calculated neutron and secondary gamma-ray flux and the flux-to-dose conversion factor of ambient dose equivalent H ∗ (10) in ICRP Report 74 [96ICR]. Table 5.5.3 - 5.5.6 give the calculated neutron and secondary gamma-ray ambient dose rates for 252Cf, Am-Be, D-D and D-T sources [01Sak]. The first column shows the distance r [cm] from the centre, the second column shows the equivalent dose rate at distance r when no shield material is present: H& 0 [µSv h−1], the third to fifth columns show the neutron, secondary gamma-ray and total equivalent dose rates in water at distance r: H& n [µSv h−1], H& g [µSv h−1] and H& t [µSv h−1], and the sixth to eighth columns show the neutron, secondary gamma-ray, and total transmission factors: Fn (= H& n / H& 0 ), Fg (= H& g / H& 0 ) and Ft (= H& t / H& 0 ) for water. Fn, Fg, Ft values for polyethylene, ordinary concrete and heavy concrete are also described. As for the simple shielding estimation, the constant Γ for various sources is shown in Table 5.5.2, representing the bare ambient dose equivalent rate at 1 m for unit source intensity. Similar tables for the effective dose rates for AP (anterior - posterior) exposure geometry are shown in Table A5.5.1 - A5.5.4 [01Sak] of the accompanying CD-ROM. Neutron and secondary gamma-ray ambient dose equivalent rates and effective dose rates are calculated by the following equation.
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
5-17
S H& = Γ ⋅ F (t ) 2 d
(5.5.1)
constant [µSv h−1 m2 s] (Table 5.5.2 and A5.5.5 (CD-ROM)) (bare ambient dose equivalent rate or effective dose rate at 1m for unit source intensity) F(t ) transmission factor (Fn, Fg, Ft ) for layer thickness t (Table 5.5.3 - 5.5.6 and A5.5.1 - A5.5.4, Fig. 5.5.2 - 5.5.17 and 5.5.20 - 5.5.35 (appendix)) S neutron source intensity [s−1] d distance from source to detector [m] H& Here H& means ambient dose equivalent rate for table 5.5.3 - 5.5.6, and effective dose rate for table A5.5.1-A5.5.4
Γ
Fig. 5.5.2 - 5.5.5 show the neutron and secondary gamma-ray dose transmission curves for a 252Cf source in water, polyethylene, ordinary concrete, and heavy concrete. As shown in these figures, secondary gamma-rays play dominant role from about 50 cm in water and polyethylene, and about 150 cm in ordinary concrete. The total transmission factor for heavy concrete is lower than that for the other materials, because it suppresses secondary gamma-rays efficiently. Fig. 5.5.6 - 5.5.9 show the transmission curves for Am-Be source in each shield material. The neutron source spectrum from (α, n)-sources depends on the grain radius and mixing ratio. This transmission factor was calculated using the source spectrum measured by Greiss [68Gre]. Transmission factors for Am-Be are similar to these for 252Cf source. Fig. 5.5.10 - 5.5.13 show the transmission curves for D-D source in each shield material. The transmission factor is low because of the low source neutron energy (2.45 MeV). D(d, p)T and D(d, n)3He reactions occur with almost the same probabilities. Therefore the ordinary D-D source produces a small amount of D-T neutrons by accumulated tritium. The D-T neutrons must be considered simultaneously. Fig. 5.5.14 - 5.5.17 show the transmission curves for D-T source in each shield material. The transmission factor is very high for D-T neutrons because of the neutron energy being high (14.1MeV). Similar figures for the effective dose rate for AP (anterior - posterior) exposure geometry are shown in Fig. 5.5.20 - 5.5.35 of the accompanying CD-ROM. These attenuation calculations were done in sufficiently thick material, called infinite geometry. This means, calculated dose rates at every point contain the backscattered components. This causes unrealistic results, so that dose attenuation factor exceeds one. Backscattered dose contributions depend on the shield material and thickness. Attention has to be paid as well to a more realistic geometry, which usually shows a certain distance between the source and the shield wall. This effect increases the transmission factor to a certain extent. Considering these effects, overestimation results at 0.5 m to about 1.4 for water and to about 1.2 for ordinary concrete. Although the attenuation factor is conservative, it is sufficient for easy evaluation. Table 5.5.1 Elemental composition and densities of shield material. *) Type 02-a concrete from ANL5800, p.660 (1963); **) From JAERI-M 6928, p.36 (1977). Material −3
Density [g cm ] Element H C O Mg Al Si Ca Fe Landolt-Börnstein New Series VIII/4
Water Polyethylene 1.0 0.93 Atomic densities [1024 cm−3] 6.6738·10–2 7.9793·10–2 3.9930·10–2 3.3370·10–2
Ordinary concrete* 2.1
Heavy concrete** 3.715
1.2985·10–2 1.0821·10–4 4.3051·10–2 1.1614·10–4 1.6321·10–3 1.5582·10–2 1.4086·10–3 3.2353·10–4
9.9885·10–3 4.3899·10–2 4.3265·10–5 1.6502·10–4 3.7264·10–3 2.2782·10–3 2.3602·10–2
5-18
5 Shielding against ionizing radiation
[Ref. p. 5-32
Table 5.5.2 Constant Γ (ambient dose equivalent) for various neutron sources Source 252
Cf Am-Be D-D D-T
Γ [µSv h−1 m2 s] 1.11·10–5 1.13·10–5 1.19·10–5 1.49·10–5
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Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
5-19
(1) Dose calculation at the surface of storage container Neutron dose rate The shielding effect by the 0.1 cm thick iron layer is small, therefore the iron is neglected, and polyethylene is the only shielding material to be considered. With Γ from Table 5.5.2 and the neutron transmission factor Fn(40) for polyethylene from Table 5.5.3 (t → r = 40 cm) the neutron ambient dose equivalent rate is 4.32 ⋅ 10 6 Eq. (5.5.1): H& n = 1.11·10−5 · 5.74·10−3 · µSv h−1 = 1.10 µSv h−1 0.52 Secondary gamma-ray dose rate The shielding effect by the 0.1 cm thick iron layer is small, therefore the iron is neglected, and polyethylene is the only shielding material to be considered. With Γ from Table 5.5.2 and the secondary gamma-ray transmission factor Fg(40) for polyethylene from Table 5.5.3 (t → r = 40 cm) the secondary gamma ambient dose equivalent rate is 4.32 ⋅ 10 6 Eq. (5.5.1): H& g = 1.11·10−5 · 1.26·10−2 · µSv h−1 = 2.42 µSv h−1 0.52
Also, the primary gamma-rays of 252Cf source should be considered. (2) Dose calculation outside the exposure room Neutron dose rate With Γ from Table 5.5.2 and the neutron transmission factor Fn(50) for ordinary concrete from Table 5.5.3 (t → r = 50 cm) the neutron ambient dose equivalent rate is 4.32 ⋅ 10 6 µSv h−1 = 0.299 µSv h−1 Eq. (5.5.1): H& n = 1.11·10−5 · 1.56·10−1 · 5.0 2
Secondary gamma-ray dose rate With Γ from Table 5.5.2 and the secondary gamma-ray transmission factor Fg(50) for ordinary concrete from Table 5.5.3 (t → r = 50 cm) the secondary gamma-ray ambient dose equivalent rate is 4.32 ⋅ 10 6 µSv h−1 = 0.0313 µSv h−1 Eq. (5.5.1): H& g = 1.11·10−5 · 1.63·10−2 · 5.0 2
5.5.4 Induced activity Structure materials, air, coolant waters etc. are activated in neutron fields. The induced activity has to be considered in radiation protection design of nuclear reactors, fusion experimental reactors and high energy accelerators. Examples of well-known activation reactions, half-lives, and gamma-ray energies of produced nuclides – typical for nuclear reactors − are listed in Table 5.5.7. These reaction cross sections strongly depend on neutron energy. The induced activity and gamma-ray dose rate can be estimated by the following three steps: 1) Estimation of the neutron energy spectrum with a computer code such as ANISN [73Eng]. 2) Estimation of the induced activity and gamma-ray source strength for a given irradiation and decay time with a computer code such as ORIGEN [73Bel], using the neutron energy spectrum data 3) Estimation of the gamma-ray dose at a given point with a computer code such as ANISN.
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5-20 5 Shielding against ionizing radiation [Ref. p. 5-32
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Ref. p. 5-32]
Landolt-Börnstein New Series VIII/4
5 Shielding against ionizing radiation 5-21
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5-22 5 Shielding against ionizing radiation [Ref. p. 5-32
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Ref. p. 5-32]
Landolt-Börnstein New Series VIII/4
5 Shielding against ionizing radiation 5-23
5-24
5 Shielding against ionizing radiation
[Ref. p. 5-32
Table 5.5.7 Typical activation reactions and cross section data, half-lives, gamma-ray energies of produced radionuclides [81Mug], [88Mcl], [02Nak]. Material Stainless steel Water
Reaction Ni(n, p)58Co
Half life 70.8 d
γ-ray energy [MeV] 0.811, 0.511
54
312 d
0.835
59
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5.27 y
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8.2·10−2 (Fission spectrum averaged) 3.7·101 (at 0.0253 eV)
7.13 s
6.13
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40
1.83 h
1.29
6.6·10−1 (at 0.0253 eV)
58
Fe(n, p)54Mn
Ar(n, γ)41Ar
Air 10
Cross section [barn] 1.1·10−1 (Fission spectrum averaged)
10 252
252
Cf/water
Cf/Polyethylene
–1
–1
10
–3
10
–3
10
Transmission factor
Transmission factor
10
–5
10
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10
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10
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10
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10
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10
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10
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0
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100 150 Radius [cm]
200
250
Fig. 5.5.2. Transmission factor of water for 252Cf source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.3. Transmission factor of polyethylene for 252Cf source.
10
10 252
252
Cf/ordinary concrete
1
Cf/heavy concrete
1 –1
10
Transmission factor
Transmission factor
–2
–2
10
–3
10
–4
10
–5
Fn Fg Ft
10
–6
10
–4
10
–6
10
Fn Fg Ft
–8
10
–7
10
10
–10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.4. Transmission factor of ordinary concrete for 252 Cf source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.5. Transmission factor of heavy concrete for 252 Cf source.
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10
Am-Be/water
Am-Be/Polyethylene –1
1
10
–3
Transmission factor
Transmission factor
–2
10
–4
10
–6
10
Fn Fg Ft
–8
10
10
–5
10
–7
10
Fn Fg Ft
–9
10
–10
10
5-25
–11
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.6. Transmission factor of water for Am-Be source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.7. Transmission factor of polyethylene for Am-Be source.
10
10
Am-Be/heavy concrete
Am-Be/ordinary concrete 1
1 –1 –2
Transmission factor
Transmission factor
10
–2
10
–3
10
–4
10
Fn Fg Ft
–5
10
–4
10
–6
10
Fn Fg Ft
–8
10
–6
10
10
–10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.8. Transmission factor of ordinary concrete for Am-Be source.
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0
50
100 150 Radius [cm]
200
250
Fig. 5.5.9. Transmission factor of heavy concrete for Am-Be source.
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10
[Ref. p. 5-32
1
D-D/water
–1
10
D-D/Polyethylene
–2
10
–4
10
–3
10
–6
10 10
Transmission factor
Transmission factor
–5 –7
10
–8
10
–10
10
–9
–12
10
10
–14
10
–11
10
–16
10
Fn Fg Ft
–13
10
–15
10
10
–20
10
–17
10
Fn Fg Ft
–18
–22
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.10. Transmission factor of water for D-D source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.11. Transmission factor of polyethylene for D-D source.
10
10
D-D/heavy concrete
D-D/ordinary concrete
1
1 –1
10
Transmission factor
Transmission factor
–2
–2
10
–3
10
–4
10
–5
Fn Fg Ft
10
–6
10
–4
10
–6
10
Fn Fg Ft
–8
10
–7
10
10
–10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.12. Transmission factor of ordinary concrete for D-D source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.13. Transmission factor of heavy concrete for D-D source.
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10
D-T/water
1
D-T/Polyethylene
1
–1
10
Transmission factor
Transmission factor
–2
–2
10
–3
10
–4
10
–5
–6
10
Fn Fg Ft
10
–8
10
–7
10
–4
10
–6
Fn Fg Ft
10
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.14. Transmission factor of water for D-T source.
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.15. Transmission factor of polyethylene for D-T source.
1
10
D-T/heavy concrete
D-T/ordinary concrete 1
–1
10 –1
Transmission factor
Transmission factor
10
–2
10
–3
10
–3
10
–5
10
–4
10
Fn Fg Ft
–5
10
–9
10
–6
10
Fn Fg Ft
–7
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.16. Transmission factor of ordinary concrete for D-T source.
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50
100 150 Radius [cm]
200
250
Fig. 5.5.17. Transmission factor of heavy concrete for D-T source.
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[Ref. p. 5-32
5.6 Computer codes and online nuclear data services Table 5.6.1 gives a collection of typical computer codes for transport calculations of gamma-rays, neutrons and charged particles with regard to shielding problems. The programs are collected and distributed by different Data Centres, as for example NEA or NNDC. The necessary data in specified formats for program testing and evaluation are provided by the centres as well. Some Online Nuclear Data Services for basic nuclear data and evaluated nuclear data are listed in Table 5.6.2. Table 5.6.1 Computer codes for shielding and source calculations Computer Codes ANISN
Engle Jr., W.W.: A Users Manual for ANISN, A One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering. K-1693, 1973
BERMUDA
Suzuki, T. et al.: Development of: A Radiation Transport Code System Part I. Neutron Transport Codes, JAERI 1327, JAERI, 1992 A Radiation Transport Code System Part II. Gamma Rays Transport Codes. JAERI-M 93-143, JAERI, 1993
DOORS
DOORS includes: TORT Three-dimensional neutron/photon transport DORT Two-dimensional neutron/photon transport ANISN One-dimensional neutron/photon transport Rhoades, W.A., D.B. Simpson: The TORT Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code. ORNL/TM-13221, 1997
DOT-4.2
Rhoades, W.A., D.B. Simpson, R.L. Childs, W.W. Engle Jr.: The DOT-4 Two Dimensional, Discrete-Ordinates Transport Code with Space-Dependent Mesh and Quadrature. ORNL/TM-6529, 1978
DUCT-III
Tayama, R., H. Nakano, H. Handa, K. Hayashi, H. Hirayama, K. Shin, F. Masukawa, H. Nakashima, N. Sasamoto: DUCT-III, A Simple Design Code for Duct-Streaming Radiations. KEK Internal 2001-8, 2001
EGS4
Nelson, W.R., H. Hirayama, W.O. Rogers: The EGS4 Code System. SLAC265, 1985 (Electron Photon Shower Simulation by Monte-Carlo)
ETRAN
Berger, M.J., S.M. Seltzer: Electron and Photon Transport Programs – 1. Introduction and Notes on Program DATAPAC-4. NBS 9836, 1968; 2. Notes on Program ETRAN-15. NBS 9837, 1968 (Monte Carlo Code System for Electron and Photon Transport Through Extended Media)
ISO-PC
Revision of ISOSHLD Engle, R.L., J. Greenborg, N.M. Hendrickson: ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis. BNWL-236, 1966 (X-Ray, Gamma-Ray, Bremsstrahlung Dose-Rates)
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Computer Codes MCNP-4C
Briesmeister, J. F.: MCNP - A General Monte Carlo N-Particle Transport Code. Version 4C, LA-13709-M, 2000 (Coupled Neutron, Electron, Gamma 3-D Time-Dependent Monte Carlo Transport Calculation Code)
MORSE-CGA
Emmett, M.B.: MORSE-CGA, A Monte Carlo Radiation Transport Code with Array Geometry Capability. ORNL-6174, 1985
NAC
Weinstein, Suzanne T.: NAC - Neutron Activation Code. NASA TM X-5260, 1968
ORIGEN2
Croff, A.G.: A User’s Manual for the ORIGEN2 Computer Code. ORNL/TM-7175, 1980
PALLAS
Takeuchi, K., S. Tanaka: PALLAS-1D(VII): A Code for Direct Integration of Transport Equation in One-Dimensional Plane and Spherical Geometries. JAERI-M 84-214, 1984
PENELOPE2001
Salvat, F., J.M. Fernandez-Varea, E. Acosta, J. Sempau: PENELOPE, A Code System for Monte Carlo Simulation of Electron and Photon Transport. Proceedings of a Workshop/Training Course, OECD/NEA 5-7 November, 2001, NEA/NSC/DOC, 2001
PUTZ
Ingersoll, D.T.: User's Manual for PUTZ - A Point-kernel Photon Shielding Code. ORNL/TM-9803, 1986
QAD-CGGP
Sakamoto, Y., S. Tanaka: QAD-CGGP2 and G33-GP2- Revised Versions of QAD-CGGP and G33-GP Codes with Conversion Factors from Exposure to Ambient and Maximum Dose Equivalents. JAERI-M 90-110, 1990
RAID
Moore, J.A., J.B. Eggen, F.O. Leopard: Monte Carlo Procedure for Analysis of Radiation in Ducts (RAID). AFWL-TR 67-9, 1967 (Gamma, Neutron Scattering in Cylindrical or Multibend Ducts)
SAM-CE
Steinberg, H.A. et al.: SAM-CE - A Monte Carlo Code for Three Dimensional Neutron, Gamma Ray and Electron Transport (Revision 5). MR-7052-5, 1977
SKYSHINE
Lampley, C.M., M.C. Andrew, M.B. Wells: The SKYSHINE-III Procedure: Calculation of the Effects of a Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. RRA-T8209A, 1988
SRNA-2KG
Ilic, R.D.: SRNA, Protons Transport Simulation by Monte Carlo Techniques User's Guide. Version 2KG, 2001
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Table 5.6.2 Online Nuclear Data Services and Code Services Address
Information
Atomic Mass Data Center (AMDC) http://csnwww.in2p3.fr/AMDC/
Atomic Mass Data, Q-Values
Gesellschaft für Schwerionenforschung (GSI) http://www.gsi.de/
Links to Data Banks, Literature-Research
International Atomic Energy Agency (IAEA) Nuclear Data Center http://www-nds.iaea.or.at
Nuclear Structure and Decay Data, Cross Section Data (Photons, Neutrons, charged Particles) (ENSDF, ENDF), Nuclear Science References
Nuclear Data Center Japan Atomic Energy Research Institute http://wwwndc.tokai.jaeri.go.jp/
Chart of Nuclides, Evaluated Nuclear Data Library, Tables of Nuclear Data
Korea Atomic Energy Research Institute (KAERI) Nuclear Structure and Decay Data, Nuclear Data Evaluation Lab Photon Cross Section Data http://atom.kaeri.re.kr Lawrence Berkeley National Laboratory (LBNL) Isotopes Project http://isotopes.lbl.gov
Nuclide-Table, Nuclide-Chart, Isotope Explorer, Internet Isotope Explorer
Los Alamos National Laboratory (LANL) T-2 Nuclear Information Service http://t2.lanl.gov/data/decayd.html
Nuclear Structure and Decay Data, Cross Section Data (Photons, charged Particles, thermal Neutrons) (ENSDF, ENDF) Nuclide-Chart, Nuclear Data Viewer
Lunds Universitet LUND Nuclear Data Service http://nucleardata.nuclear.lu.se/nucleardata
Nuclear Structure and Decay Data (ENSDF), Literature, References, Isotope Explorer, Internet Isotope Explorer
www-tech.mit.edu/Chemicool
Periodical System of Elements
Nuclear Energy Agency (NEA) http://www.nea.fr/html/dbdata/
Nuclear Data, Computer Codes, Experimental Nuclear Reaction Data Retrievals, Evaluated Nuclear Data Retrievals, Bibliographical Research
National Institute of Standards and Technology (NIST) http://physics.nist.gov/PhysRefData/contents.html
Physical Constants, X-ray and gamma ray data, X-Ray Attenuation and Absorption for Materials of Dosimetric Interest, XCOM: Photon Cross Sections Database, Stopping-Power and Range Tables for Electrons, Protons, and Helium Ions, Radionuclide Half-life Measurements Made at NIST, Atomic Weights and Isotopic Compositions
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Address
Information
National Nuclear Data Center (NNDC) http://www.nndc.bnl.gov
Nuclear Structure and Decay Data, Neutron Cross Sections, Online-Service, Literature, References, Documentation of Data Banks and Computer Codes (ENSDF, ENDF,...), Nuclear Science References
PhysicsWeb http://physicsweb.org/TIPTOP/paw/
Information, Links
Radiation Safety Information Computational Center (RSICC) http://www-rsicc.ornl.gov/rsicc.html
Codes and Data, Newsletter, Workshops
Triangle Universities Nuclear Laboratory (TUNL) http://www.tunl.duke.edu/NuclData
Nuclear Data for light Nuclides (A = 3 to 20)
Department of Computer Science University of Columbia (UBC) http://www.cs.ubc.ca/elements/periodic-table
Periodical System of Elements
US Nuclear Data Program (USNDP) http://www.nndc.bnl.gov/usndp/
Links to Nuclear Data Banks
WebElements http://www.webelements.com/
Periodical System of Elements
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5.7 References 64Kno 68Gre 73Bel 73Eng 74Jae 81Mug 84ICR1 84ICR2 84Mug 88DLC 88Mcl 90Shi 91ANS 91Tan 92ICR 92Suz 93Suz 93ICR 95Hub
96ICR
Knop, G., Paul,W.: Interaction of electrons and α-particles with matter; Alpha-, beta- and gamma-Ray spectroscopy, Vol. 1, Siegbahn, K. (ed.), North-Holland Publishing Company, 1964. Greiss, H.B.: Nukleonik 10 (1968) 283. Bell, M.J.: ORIGEN-The ORNL isotope generation and depletion code, ORNL-4628, 1973. Engle, jr., W.W.: A user manual for ANISN; A one dimensional discrete ordinates transport code, ORNL-TM-4280, 1973. Jaeger, R.G., Hübner, W.: Dosimetrie und Strahlenschutz, G. Thieme Verlag, 1974, p. 378. Mughabghab, S.F., Divadeenam, M., Holden, N.E.: Neutron cross sections, Vol 1, Neutron Resonance Parameters and Thermal Cross Sections, Part A, Z=1-60. New York: Academic Press, 1981; NNDC: Online-Datenbank: http://www.nndc.bnl.gov/nndc/ensdf/ensdfindex.html International Commission on Radiation Units and Measurements, ICRU Report 35, Radiation dosimetry: Electron beams with energies between 1 and 50 MeV, ICRU Publications, 1984. International Commission on Radiation Units and Measurements, ICRU Report 37, Stopping powers for electrons and positrons, ICRU Publications, 1984. Mughabghab, S.F.: Neutron cross sections, Vol 1, Neutron resonance parameters and thermal cross sections, Part B, Z=61-100. New York: Academic Press, 1984; NNDC: OnlineDatenbank: http://www.nndc.bnl.gov/nndc/ensdf/ensdfindex.html Radiation Shielding Information Center Data Package DLC-136/PHOTX, Photon Interaction Cross Section Library, contributed by National Institute of Standards and Technology, 1988. McLane, V., Dunford, C.L., Rose, P.F.: Neutron Cross Sections, Vol. 2. Boston: Academic Press, 1988. Shibata, K., Nakagawa, T., Asami, T., Fukahori, T., Narita, T., Chiba, S., Mizumoto, M., Hasegawa, , Kikuchi, Y., Nakajima, Y., Igarasi, S.: “Japanese Evaluated Nuclear Data Library, Version-3, -JENDL3-, JAERI1319” Japan Atomic Energy Research Institute, 1990. American National Standard for Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials, ANSI/ANS-6.4.3-1991, 1991. Tanaka, T., Suzuki, T.: A calculation method of photon dose equivalent based on the revised technical standards of radiological protection, ORNL/TR-90/29, Oak Ridge National Laboratory, 1991. International Commission on Radiation Units and Measurements, ICRU Report 47, Measurements of dose equivalents from external photon and electron radiations, ICRU Publications, 1992. Suzuki, T., Hasegawa, A., Tanaka, S., Nakashima, H.: Development of BERMUDA: A radiation transport code system, Part I: Neutron transport codes, JAERI 1327, Japan Atomic Energy Research Institute, 1992. Suzuki, T., Hasegawa, A., Tanaka, S., Nakashima, H.: Development of BERMUDA: A radiation transport code system, Part II: Gamma rays transport codes, JAERI-M 93-143, Japan Atomic Energy Research Institute, 1993. International Commission on Radiation Units and Measurements, ICRU Report 49, Stopping powers and ranges for protons and alpha particles, ICRU Publications, 1993. Hubbell, J.H., Seltzer, S.M.: Tables of X-ray mass attenuation coefficients and mass energyabsorption coefficients from 1 keV to 20 MeV for Elements Z=1 to 92 and 48 additional substances of dosimetric interest. Online: http://physics.nist.gov/PhysRefData/Xcom/Text/XCOM.html International Commission on Radiological Protection, Publication 74: Conversion coefficients for use in radiological protection against external radiation, Ann. ICRP 26, No.3/4 (1996).
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02Nak
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International Commission on Radiation Units and Measurements, ICRU Report 56, Dosimetry of external Beta-rays for radiation protection, ICRU Publications, 1997. The Japan Radioisotope Association, Radioisotope Pocket Data Book (Revised Edition 10), 2001 (in Japanese). Sakamoto, Y., Endo, A., Tsuda, S., Takahashi, F., Yamaguchi, Y.: Shielding calculation constants for use in effective dose evaluation for photons, neutrons and bremsstrahlung from Beta-ray, JAERI-Data/Code 2000-044, Japan Atomic Energy Research Institute, 2001, (in Japanese). Nakagawa, T, Kawasaki, H., Shibata, K. (Eds.):Curves and tables of neutron cross sections in JENDL-3.3, Part I (Z = 1 - 50) and Part II (Z = 51 - 100), JAERI-Data/Code 2002-020, Japan Atomic Energy Research Institute, 2002.
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6-1
6 External dosimetry
This chapter introduces the main protection and operational quantities in external dosimetry and describes the anthropomorphic models used for their calculations. Conversion coefficients i.e. mean organ equivalent doses normalised to the measurable quantity “air kerma free-in-air” are given for idealized geometries representing occupational exposures and for environmental source geometries.
6.1 Protection and operational quantities 6.1.1 Protection quantities The International Commission on Radiological Protection (ICRP) for more than 50 years supports a system for radiological protection, based on concepts, quantities, and basic recommendations. The concept of radiation protection is based on the justification, optimisation and limitation of radiation exposure. This concept includes a dose limitation system for occupational and man-made environmental radiation exposures to ensure that the radiation risk would not exceed reasonable limits. The most recent set of protection quantities recommended in ICRP60 [91ICR] includes the organ or tissue equivalent doses HT and the effective dose E (see Chap. 4). These quantities are not measurable but can be calculated if the exposure conditions are known. The quantity to be limited in radiation protection of occupationally exposed persons and members of the public is the effective dose E which is the weighted mean of equivalent doses of several organs and tissues of the body that are considered to be most sensitive. E=
∑w H = ∑w ∑D T
T
T
T
T
T ,R
wR
R
where HT is the mean organ equivalent dose and wT is the tissue weighting factor (with Σ wT = 1) which takes into account the differences in the stochastic radiation risk of the different organs. It is derived from the mean organ absorbed dose DT, i.e. the total amount of energy deposited in an organ (or tissue) T per mass of the organ, by multiplying with a radiation weighting factor wR reflecting the relative biological effectiveness of the radiation incident on the body or emitted from radionuclides in the body. The sensitive organs and tissues together with their respective tissue weighting factors wT were defined in ICRP Publication 60 [91ICR] (see Chap. 4).
6.1.2 Operational Quantities The International Commission on Radiation Units and Measurements (ICRU) has defined a set of operational quantities for area and individual monitoring [85ICR, 92ICR1, 93ICR] in response to the recommendations of the International Commission on Radiological Protection [77ICR] which were Landolt-Börnstein New Series VIII/4
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[Ref. p. 6-42
designed to provide an estimate of the protection quantities defined by ICRP and to serve as calibration quantities for dosimeters used in monitoring. For area monitoring, the appropriate operational quantities are the ambient dose equivalent H*(d), and the directional dose equivalent, H'(d,Ω), both defined at a depth d, on the principal axis of the 30 cm diameter ICRU sphere. The recommended value of d for strongly penetrating radiation is 10 mm and for weakly penetrating radiation it is 0.07 mm (see Sect. 4.5.3.3). For individual monitoring, the quantity “personal dose equivalent” Hp(d) was proposed, which is the dose equivalent in soft tissue, at an appropriate depth d below a specified point on the body [92ICR1, 93ICR]. For weakly penetrating radiation, depths of 0.07 mm for the skin and 3 mm for the eye lens are used, denoted by Hp(0.07) and Hp(3), respectively; for strongly penetrating radiation, a depth of 10 mm is currently recommended by the ICRU, denoted by Hp(10) (see Sect. 4.5.3.4). Personal dose equivalent is defined in the human body and may, therefore, vary between individuals; furthermore, the depth d is specified but the position of the point below which it is defined is not fixed but only correlated to the position of the dosemeter worn on the body. Consequently, the personal dose equivalent can be expected to vary also between locations on any given individual and is, hence, anticipated to be a multi-valued quantity [96ICR, 98ICR, 99Zan]. To make this quantity single-valued in a given exposure situation, both a particular location on the human body and a particular phantom of the body need to be specified for evaluation. For calibration purposes, “surrogate quantities” for Hp(d) have been introduced: it is recommended that personal dosimeters normally worn on the trunk are calibrated on an ICRU tissue slab or PMMA (polymethylmethacrylate) slab with dimensions 30 × 30 × 15 cm3 [92ICR1]. Conversion coefficients for personal dose equivalent at the relevant depths d in the ICRU tissue slab Hp,slab(d) have been calculated for calibration purposes [91Gro, 95ISO, 95Til] and have been recommended for use [96ICR, 98ICR, 98Cla]. The operational quantities used in measurement were designed to provide a reasonable estimate of the appropriate protection quantity. For external exposures of the body in a given field, it is desirable that the ratio of the value of the appropriate protection quantity to the value of the corresponding operational quantity is less than unity, i.e. the operational quantity should always provide a “conservative” estimate of the protection quantity. More about the definitions of the operational quantities can be found in Chap. 4.
6.2 Dosimetric models 6.2.1 Models and phantoms of the human body To estimate the protection quantities organ and tissue equivalent doses HT there are two approaches, an experimental and a theoretical one. The experimental determination is very difficult whereas the mathematical modelling of an exposure has been proved to be extremely flexible and powerful. For this purpose, a series of computer models of the human body were designed in the past, together with computer codes simulating the radiation transport and energy deposition in the body. The computer models used for the representation of the human body in dose calculations can range from simple geometric forms such as spheres, cylinders or slabs to complex representations of detailed anatomical features. Such complex models, used since 1966 for the estimation of organ doses are the socalled mathematical phantoms, which are models whose body organs and tissues are described by mathematical expressions representing planes or cylindrical, conical, elliptical or spherical surfaces. The mostly used model was the “MIRD” one, named after the initials of the Medical Internal Radiation Dose Committee of the US Society of Nuclear Medicine where it was initially developed [69Sny, 78Sny]. From this, several paediatric models were derived to represent infants and children of various ages, for example those from Cristy [80Cri]. As an improvement to these hermaphrodite models, separate male and female
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adult mathematical models have been introduced by Kramer et al. [82Kra] called Adam and Eva. For these models, the organ masses and volumes are in accordance with the ICRP data on Reference Man [75ICR]; in addition to a separation of the gender-specific organs, the phantom Eva is smaller than Adam, according to the difference in size of the male and female Reference Man. The oesophagus, an organ which had not originally been defined in these models, was incorporated in the form of an elliptical cylinder ranging in height from within the neck down to the top of the stomach and lying in front of the spine, slightly shifted to the left side [92Zan]. Adam and Eva contain all organs and tissues relevant for the evaluation of effective dose, with only few exceptions: since there is no specific representation of the “bone surface”, the skeleton is modelled as a homogeneous mixture of all skeletal constituents, i.e., hard bone, bone marrow and certain peri-articular tissues. Commonly, the dose to this representation of the entire skeleton is taken to represent the dose to the bone surface. Although there may be certain differences, these are usually considered to be small in view of the small weighting factor (wT=0.01) assigned to this tissue. The muscles were represented by that part of the body volume not attributed to any other organ or tissue of the models. More recently, four models representing the adult female, non-pregnant and at 3 stages of pregnancy were elaborated by Stabin et al [99Sta]. A comprehensive review of models and phantoms of the human body can be found in ICRU Report 48 [92ICR2]. The term “model” refers to computational models, whereas “phantom” implies either a physical phantom or a computational one. Spherical and slab phantoms are convenient and simple approximations of the human body. A spherical model of 30 cm diameter made of ICRU tissue-equivalent material (see Sect. 4.5.3.3) is used for the definition of the operational quantities. Various tissue substitutes are available for fabrication of corresponding physical phantoms, including tissue-equivalent material, water and perspex. For calibration purposes, slab tissue-substitute phantoms of 30 × 30 × 15 cm3 are used. Recently a new generation of computational phantoms has become available which offer the prospect of increased realism and accuracy in dose calculations. These models use computed (CT) or magnetic resonance (MR) tomographic data of real persons to provide three-dimensional representations of the human body and comprise a large number of volume elements (voxels) all of the same size but with differing composition according to the organ to which they belong. The GSF-National Research Centre in Germany started since the mid eighties the development of voxel models covering various ages and anatomical statures [88Zan, 01Zan, 02Pet]. Due to their anatomical realism, such models have been the subject of increasing interest and acceptance, and others have been developed also elsewhere [94Zub, 96Dim, 00Xu]. Both MIRD-type and voxel models incorporate different densities and atomic compositions for the various body tissues. The number of organs simulated varies from model to model, however, the latest versions include all organs defined to be important. Fig. 1 shows views of selected organs of the mathematical model Eva [82Kra] and the adult female voxel model Donna, developed at GSF [02Pet].
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[Ref. p. 6-42
Fig. 6.1. View of selected organs of the mathematical model Eva [82Kra] and the adult female voxel model Donna, developed at GSF [02Pet].
6.2.2 Idealized geometries representing occupational exposures To simulate occupational exposure conditions, whole-body irradiation with idealised geometries are conventionally taken into account. These include broad parallel beams and fully isotropic radiation incidence. The directions of incidence for the parallel beams considered are: anterior-posterior (AP), posterior-anterior (PA), left lateral (LLAT), right lateral (RLAT) and a full 360° rotation around the phantoms’ longitudinal axis (ROT). Although these geometries are idealised, they may be taken as acceptable approximations to actual conditions of exposure. The AP, PA and both lateral geometries are supposed to approximate radiation fields from single sources and particular body orientations. The ROT geometry approximates the exposure of a person who moves randomly in the field of a single source irradiating at right angles to the longitudinal axis of the body. The fully isotropic (ISO) source simulates the geometry of a body suspended in a large cloud of radioactive gas.
6. 2. 3 Environmental source geometries For external exposures to environmental sources the dosimetric quantities of interest are the radiation doses received by the radiosensitive organs and tissues of the body due to photons and electrons emitted by radionuclides distributed in soil and air. The radiation dose depends strongly on the temporal and spatial distribution of the radionuclide to which a human is exposed. The situation of radioactive release in water is more rare and is not covered here. The kinds of radiation of concern are those sufficiently Landolt-Börnstein New Series VIII/4
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penetrating to traverse the overlying tissues of the body and which deposit their energy in organs and tissues of the body. Penetrating radiations are limited to photons and electrons. Neutrons from cosmic radiation are not dealt here. For simulating the exposure to environmental gamma-rays, the following three typical cases of environmental sources are considered here: (1) semi-infinite volume source in the air; (2) infinite plane source in the ground; (3) semi-infinite volume source in the ground. The first source configuration models the gaseous radioactive release into the atmosphere at locations which are not too near to the release point, by assuming a homogeneous contamination of the air up to a height of 1000 m above a smooth airground interface. The second source simulates the deposition of radionuclides in the ground, by assuming an infinite plane source in the soil. The third source simulates the natural radioactivity in the ground (the dominant radionuclides of the 238U series, the 232Th series and 40K) being homogeneously distributed to a depth of 1 m in the soil.
6.2.4 Methods of calculating protection quantities in computational models Today the predominant method for assessment of absorbed doses in the body is the application of Monte Carlo methods to simulate the transport of radiation in the body. The organ and tissue doses are then estimated in the form of conversion coefficients giving organ doses per unit of a measurable quantity. The Monte Carlo method is a computational model in which physical quantities are calculated by simulating the transport of particles. In the computer program, single particles are followed through their histories of inelastic and elastic scattering or absorption within the anthropomorphic model. Depending on their energy and on the material they are passing through, the particles interact differently and each mode of interaction has a certain probability of occurring, which can be selected by appropriate use of random numbers and probability distributions. Individual particles have different energies, directions and path lengths modelled randomly from probability distributions. By averaging over large numbers of random paths, good estimates of the quantities of interest can be made. Basic elements of Monte Carlo simulation include the choice of random number generator which provides the method of sampling the cross-section data and coordinate transformations from probability distributions. The mean absorbed dose in a defined volume of material is computed from the incident and emerging energies from the volume by dividing the energy imparted by the mass of the volume material. Most codes dealing mainly with photon interaction assume that electrons generated through different interactions are absorbed “on the spot”. The energy transferred at a point of inelastic photon interaction is then modelled as being deposited at that point, without considering the energy transport by secondary charged particles (“kerma approximation”). This approach is valid as long as there is approximate secondary charged particle equilibrium, which can be supposed in most cases due to the macroscopic approach considering mean organ and tissue doses. This is acceptable for photon energies up to 3 MeV. For the skeleton, however, the boundary effects do have an impact on the tissue dose and corrections for secondary electron effects in the skeleton have to be applied. For neutrons up to 20 MeV the kerma approximation introduces no significant error due to the short range of the recoil protons and heavier charged particles. For the estimation of absorbed doses distributions in the body, several transport codes are used and their description is beyond the scope of this book. General Monte Carlo codes are available, such as EGS4 [85Nel, 94Hir], ITS [92Hal] and MCNP4 [91Bri]. Various other research institutes have also developed their Monte Carlo codes like the GSF-National Research Center code [82Kra, 89Vei], PTB/PG [86Gro, 94Gro], etc. The anthropomorphic models used for the collection of the data shown below were the mathematical Adam and Eva and the mathematical MIRD hermaphrodite for photons and neutrons; for electrons, the ICRU tissue sphere and slab as well as Adam and Eva were employed.
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A joint task Group of ICRP and ICRU reviewed the conversion coefficients reported by different researchers. Conversion coefficients for external photon, neutron and electron exposures and the idealised geometries AP, PA, LLAT, RLAT and ROT were evaluated and compiled to be used as “reference data”; they can be found at the ICRP Report 74 [96ICR] and ICRU Report 57 [98ICR].
6.3 Conversion coefficients for photons 6.3.1 Occupational This Section provides organ doses in the form of conversion coefficients, i.e. mean organ equivalent doses normalised to “air kerma free-in-air” which is a measurable quantity; the conversion coefficients are given in the unit Sv·Gy−1. No location for a measurement of the normalisation quantity has to be specified since for the parallel and ideally isotropic geometries the photon fluence is invariant throughout the field. The organ equivalent dose conversion coefficients were calculated for the male Adam and female Eva models [82Kra] separately. Average organ equivalent dose conversion coefficients were computed as the arithmetic mean of those for the male and female models. The gonad equivalent dose conversion coefficients are the arithmetic mean values of the respective coefficients for testes and ovaries. For these calculations, the GSF Monte Carlo code was used which computes the dose deposited by photons from an external or internal source in various sections of a different media model of the body. The code is based on the fractional photon technique and uses the kerma approximation. The latter is valid only when charged particle equilibrium is established which can be supposed in most cases due to the moderate differences of the photon cross sections for the tissues in the human body and the macroscopic approach of energy deposition. There are two exceptions where boundary effects do have an impact on the tissue dose. One is the red bone marrow, where a moderate increase in energy deposited in the marrow cavities is expected from increased photoelectron emission from the surrounding bone. For photon exposures this effect was accounted for by applying appropriate correction factors [69Spi, 97Zan] to the energy deposited to the red bone marrow calculated using the kerma approximation. The other tissue where boundary effects could be of consequence is the bone surface, a very thin soft tissue layer enveloping the bones. Here secondary electron equilibrium is not valid for energies below approximately 300 keV as there the bone cross section values are considerably higher than those for soft tissues, resulting in an increased production of secondary electrons in the bones and, consequently, a dose enhancement at the interface between the bones and the adjacent soft tissues compared to the dose to tissue beyond the range of these secondary electrons. This enhanced dose to the tissue adjacent to bones is, however, not as high as the mean dose to the homogeneous mixture of skeletal tissues [68Dre]. Consequently this can be taken as a conservative estimate of the dose to the bone surface in this photon energy range. Above 300 keV, the cross sections of bone and soft tissue per mass density have a similar magnitude, and approximate secondary electron equilibrium is established. To the calculated values of the conversion coefficients for monoenergetic photons, a fitting procedure using cubic spline functions was applied. With these fitting functions, values were also evaluated for 200 photon energies distributed equidistantly on a logarithmic scale between 10 or 15 keV and 10 MeV. Figures 6.2-6.8 show the conversion coefficients as a function of photon energy for 8 selected organs and tissues and for AP, PA, LLAT, RLAT, LAT, ROT and ISO geometries, respectively. These are the average values evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model. The complete sets of organ equivalent doses for the male, female and average can be found in Zankl et. al. [97Zan] shown graphically as well as in tabular form. The average values of the male and female models denoted as “adult” are also adopted as reference values for conversion coefficients and are presented in detail in ICRP Report 74 [96ICR] and ICRU Report 57 [98ICR] for those specific organs for which the ICRP recommends tissue weighting factors (see Chap. 4, Table 3).
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Organ equivalent dose per unit air kerma [Sv/Gy]
The energy dependence of the conversion coefficients for single organs is determined by the photon interaction cross sections in tissues, the location of the organ in the body and the irradiation geometry. The cross sections decrease with increasing photon energy and the conversion coefficients correspondingly increase due to the increasing range of photons in the body. With further increasing energy and range of the photons, the conversion coefficients decrease. This leads to more or less pronounced peak in the energy range between 80-100 keV. The more pronounced peak of the conversion coefficients of skeleton is due to the large values of the ratio of the attenuation coefficients of bone and air respectively. Table 6.1 shows the effective dose per unit air kerma as a function of energy for the six irradiation geometries, calculated for adults using the models Adam and Eva. The different forms of the energy dependence of the conversion coefficients for effective dose with irradiation geometry result from the different locations of the organs relative to the incoming photon beam and the value of their tissue weighting factors. As it can be seen from Table 6.1, the conversion coefficients of E for AP irradiation are always higher than the corresponding ones for other irradiation geometries. This is due to the fact that for AP photon incidence most organs with large tissue weighting factors are anteriorly located.
gonads
Parallel AP
lungs RBM stomach liver thyroid skin skeleton
2
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.2. Organ equivalent doses per unit air kerma free-in-air for selected organs in AP irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model;[97Zan].
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Organ equivalent dose per unit air kerma [Sv/Gy]
3 gonads
Parallel PA
lungs RBM stomach liver thyroid skin skeleton
2
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.3. Organ equivalent doses per unit air kerma free-in-air for selected organs in PA irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads lungs RMB stomach liver thyroid skin skeleton
Parallel LLAT
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.4. Organ equivalent doses per unit air kerma free-in-air for selected organs in LLAT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads
lungs RBM stomach liver thyroid skin
Parallel RLAT
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.5. Organ equivalent doses per unit air kerma free-in-air for selected organs in RLAT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads
lungs RBM stomach liver thyroid skin
Parallel LAT
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.6. Organ equivalent doses per unit air kerma free-in-air for selected organs in LAT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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[Ref. p. 6-42
gonads
Parallel ROT
lungs RBM stomach liver thyroid skin
2
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Organ equivalent dose per unit air kerma [Sv/Gy]
Fig. 6.7. Organ equivalent doses per unit air kerma free-in-air for selected organs in ROT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
gonads
Parallel ISO
lungs RBM stomach liver thyroid skin
2
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.8. Organ equivalent doses per unit air kerma free-in-air for selected organs in ISO irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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Table 6.1. Effective dose E per unit air kerma free-in-air Ka for monoenergetic photons and various irradiation geometries. Data are from Zankl et. al. [97Zan], calculated using the male (Adam) and female (Eva) model. Photon energy [MeV] 0.010 0.015 0.020 0.030 0.040 0.050 0.060 0.070 0.080 0.100 0.150 0.200 0.300 0.400 0.500 0.600 0.800 1.000 2.000 4.000 6.000 8.000 10.000
E/Ka [Sv·Gy−1] AP 0.00654 0.0402 0.122 0.416 0.787 1.104 1.306 1.405 1.431 1.392 1.255 1.172 1.091 1.055 1.035 1.024 1.010 1.002 0.992 0.992 0.993 0.991 0.990
PA 0.00248 0.00590 0.0183 0.129 0.372 0.641 0.847 0.968 1.020 1.031 0.960 0.916 0.881 0.872 0.870 0.871 0.875 0.881 0.901 0.918 0.924 0.928 0.929
LLAT 0.00173 0.00550 0.0156 0.0907 0.242 0.406 0.529 0.599 0.630 0.643 0.622 0.616 0.616 0.624 0.636 0.648 0.671 0.692 0.758 0.813 0.836 0.850 0.860
RLAT 0.00172 0.00551 0.0151 0.0782 0.204 0.344 0.454 0.521 0.553 0.570 0.551 0.548 0.556 0.570 0.585 0.600 0.627 0.651 0.728 0.796 0.827 0.846 0.860
ROT 0.00326 0.0154 0.0463 0.191 0.427 0.661 0.828 0.924 0.961 0.960 0.893 0.854 0.824 0.814 0.812 0.814 0.821 0.831 0.871 0.909 0.925 0.934 0.941
ISO 0.00271 0.0123 0.0362 0.144 0.326 0.511 0.642 0.720 0.749 0.748 0.700 0.679 0.664 0.667 0.675 0.685 0.703 0.719 0.774 0.824 0.846 0.859 0.868
Figures 6.8 and 6.9 demonstrate the age dependence for effective dose, by showing the effective dose for adults as well as for 0-, 1-, 5-, 10-, and 15-year old children for AP and ROT geometry respectively. The data of this figure stem from Yamagushi [94Yam] who used hermaphrodite paediatric and adult mathematical models of different body sizes developed by Cristy [80Cri]. It can be seen that the smaller body size results in higher organ dose conversion coefficients, and consequently higher effective doses, particularly at low photon energies. The largest variation of effective dose with age was found for the LAT and ISO geometries. Similarly, Zankl et. al. [97Zan] have shown that the conversion coefficients for the female model Eva are approximately 2 % to 20 % higher than those for the male model, depending on photon energy, due to the slightly smaller body size of the female model. For AP irradiation, the lung dose conversion coefficients for the female model are between 5 % and 20 % lower than those for the male, as for this geometry the lungs of the female model are partially shielded by the breast. Furthermore, some differences were observed for the organ conversion coefficients calculated by different authors and are mainly due to the different human models used in the calculations and occur at low photon energies: the female model Eva, used for the calculations of Zankl et. al. has a smaller body size than the hermaphrodite model used by Yamaguchi, resulting in higher organ conversion coefficients particularly for low energies. Consequently, differences (up to 20 %) between the effective dose coefficients from adults were observed. For energies above 70 keV there was general agreement within the statistical uncertainties.
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Parallel AP
1
0 year 1 year 5 years 10 years 15 years adult
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.9. Effective dose per unit air kerma free-in-air for AP irradiation geometry calculated for MIRD-type hermaphrodite phantoms of various ages; [94Yam].
Effective dose per unit air kerma [Sv/Gy]
Parallel ROT 1
0 year 1 year 5 years 10 years 15 years adult
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.10. Effective dose per unit air kerma free-in-air for ROT irradiation geometry calculated for MIRD-type hermaphrodite phantoms of various ages; [94Yam].
Conversion coefficients for the operational quantities ambient dose equivalent and directional dose equivalent are shown in Table 6.2. These are values recommended by the ICRP [92ICR1] and stem from calculations by several groups using Monte Carlo methods on the ICRU sphere assuming electronic equilibrium. Landolt-Börnstein New Series VIII/4
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Table 6.2. Conversion coefficients for air kerma free-in-air Ka, directional dose equivalent H´(0.07,0°), and ambient dose equivalent H*(10), per unit fluence of monoenergetic photons; [92ICR1]. Photon energy Ka/Φ H´(0.07,0°)/Φ H* (10)/Φ [MeV] [pSv cm2] [pGy cm2] [pSv cm2] 0.010 7.60 7.20 0.061 0.015 3.21 3.19 0.83 0.020 1.73 1.81 1.05 0.030 0.739 0.90 0.81 0.040 0.438 0.62 0.64 0.050 0.328 0.50 0.55 a 0.060 0.292 0.51 0.080 0.308 0.53 0.100 0.372 0.61 0.150 0.600 0.89 0.200 0.856 1.20 0.300 1.38 1.80 0.400 1.89 2.38 0.500 2.38 2.93 0.600 2.84 3.44 0.800 3.69 4.38 1.0 4.47 5.20 1.5 6.12 6.90 2.0 7.51 8.60 3.0 9.89 11.1 4.0 12.0 13.4 5.0 13.9 15.5 6.0 15.8 17.6 8.0 19.5 21.6 10.0 23.2 25.6 a
H´(0.07,0°) is not accurately determined at energies above 60 keV since there is no electronic equilibrium. Comparing tables 6.1 and 6.2, it can be seen that for photons with energies up to 10 MeV and irradiation geometries AP, PA and ROT, the operational quantity H*(10) always overestimates the protection quantity E, i.e. E/H*(10) < 1.
6.3.2 Conversion coefficients for environmental gamma ray fields 6.3.2.1 Calculation of doses for monoenergetic photons As already mentioned above, for simulating the exposure to environmental gamma-rays, the following three typical cases of environmental sources are considered to be representative: (a) semi-infinite volume source in the air; (b) infinite plane source in the ground; (c) semi-infinite volume source in the ground. The first source configuration models the gaseous radioactive release into the atmosphere at locations which are not too near to the release point, by assuming a homogeneous contamination of the air up to a height of 1000 m above a smooth air-ground interface. The second source simulates the deposition of radionuclides in the ground, by assuming an infinite plane source in the soil. The source is shielded by a soil slab of 0.5 g cm−2, allowing for some surface roughness and initial migration into soil with precipitation. The third source simulates the natural radioactivity in the ground (e.g. radionuclides of the 238 U series, the 232Th series and 40K) being homogeneously distributed to a depth of 1 m in the soil. Landolt-Börnstein New Series VIII/4
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In source (a), the dominant gamma rays come almost isotropically from the upper 2π directions, while only a small amount of scattered gamma-rays comes from the lower 2π directions. Source (c) shows the inverse tendency: the angular distribution is nearly uniform in the lower 2π directions with small components of scattered gamma-rays stemming from the upper 2π directions. In source (b), quite a large portion of the gamma-rays comes from horizontal directions. When the source distribution in the environment varies from the three typical source distributions, the angular and energy distributions also change. To estimate the organ doses from environmental photon sources presented in this book, a three-step procedure [91Pet] was followed: (1) Calculation of the gamma-ray transport in the environment (monoenergetic gamma-rays and natural radionuclides); (2) Simulation of a secondary source around the phantom; (3) Calculation of organ doses due to the secondary sources. The result of this procedure, is a set of dose conversion coefficients for monoenergetic photons. Using those, and considering the energies and intensities of the radiations emitted during nuclear transformations of these nuclides, conversion coefficients for specific radionuclides can be computed to relate a measurable quantity i.e. activity concentration or air kerma to the non-measurable quantities of organ dose. The photon transport in the environment was simulated with the Monte Carlo code YURI [85Sai], a code specially developed for environmental problems. Compton scattering, photoelectric absorption and pair production were considered as photon interaction processes. Air and ground were assumed to contact each other with an infinite plane. The cross Sections used were from Storm and Israel [70Sto]. Air was assumed to have a constant density of 1.2×10−3 g cm−3, corresponding to a temperature of 20 ºC and an air pressure of 0.1 MPa and to consist of N2, O2 and Ar having weight fractions of 75.5 %, 23.2 % and 1.3 %, respectively. Soil was taken to consist of SiO2, Al2O3, Fe2O3 and H2O with weight fractions of 58.3 %, 16.7 %, 8.3 % and 16.7 %, respectively. A soil density of 1 g cm−3 has been assumed in the calculations, since this value represents reasonably well the upper 2 cm of soil. It should be noted that the environmental transport calculations were performed without the presence of the phantom; however, the perturbation caused by the human body was investigated and found to be insignificant. From the transport calculations in the environment, double differential fluences, currents (i.e. fluences multiplied by the cosine of the angle of incidence) and air kerma values are obtained for points from 0 to 2 m above ground in steps of 20 cm. Table 6.3 shows calculated values of the air kerma rate free-in-air at 1 m height above the ground per unit activity concentration for a semi-infinite volume source in air and per unit activity per area for an infinite plane source in the ground; Table 6.4 shows the air kerma free-inair at 1 m height above the ground per disintegration/kg for the semi-infinite volume source in the ground due to the natural radionuclides. Table 6.3. next page Table 6.4. Calculated air kerma at 1 m height above the ground per disintegration/kg for a semi-infinite volume source in the ground due to natural radionuclides; [95Sai]. Radionuclides Air kerma per unit source intensity [Gy / (disintegration/kg)] 238 U series 1.29×10−13 232 Th series 1.68×10−13 40 K 1.16×10−14
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Table 6.3. Calculated air kerma rate at height 1 m above the ground per unit of activity concentration for a semi-infinite volume source in air and per unit of the activity per unit area for an infinite plane source in the ground; [95Sai]. Volume source in air Plane source in ground Energy Air kerma rate per unit Air kerma rate per unit [MeV] of activity concentration of activity per unit area [(Gy s-1)/ (Bq m-3)] [(Gy s-1)/ (Bq m-2)] −15 0.015 1.47×10 8.06×10−19 −15 0.020 7.72×10−18 1.71×10 −15 0.030 2.12×10 2.63×10−17 −15 0.040 2.40×10 3.59×10−17 −15 0.050 2.81×10 4.14×10−17 −15 0.060 3.31×10 4.65×10−17 −15 0.070 3.79×10 5.35×10−17 0.080 4.36×10−15 5.98×10−17 −15 0.100 5.55×10 7.54×10−17 −15 0.150 8.68×10 1.21×10−16 −14 0.200 1.20×10 1.68×10−16 −14 0.300 1.87×10 2.61×10−16 −14 0.500 3.21×10 4.34×10−16 −14 0.700 4.56×10 5.90×10−16 −14 1.000 6.58×10 8.09×10−16 −14 1.500 9.08×10 1.13×10−15 −13 2.000 1.32×10 1.41×10−15 −13 3.000 1.96×10 1.91×10−15 −13 6.000 3.85×10 3.19×10−15 −13 10.000 6.26×10 4.81×10−15 These height-dependent double differential (with respect to angle of incidence and photon energy) gamma ray fields were then incorporated into the organ dose calculation with anthropomorphic models, by establishing a secondary cylindrical source around the model to simulate the gamma-ray fields after the results of the transport calculation in the environment (step 2 of the procedure mentioned above) [91Pet]. The anthropomorphical models are standing on the soil modelled as a planar air/ground interface. Scatter and absorption of the radiation in both air and ground was considered in the calculation. The Monte Carlo code used for the transport calculation in the body was the GSF code mentioned above. The interactions considered were photoelectric absorption, Compton scattering and pair production and the cross section data were taken from ORNL [83Rou]. Dose conversion coefficients for Adam and Eva (see previous Section) and several organs, including the critical ones, were estimated and can be found in Zankl et. al. [97Zan]. From the dose conversion coefficients of the sex-specific models Adam and Eva, conversion coefficients for an “adult” were derived as arithmetic average. In Fig. 6.10 dose conversion coefficients are given for some selected organs for submersion in a radioactive cloud (volume source in air) and in Fig. 6.11 for surface contamination. The conversion coefficients are for an adult and are expressed as equivalent doses normalised to air kerma free-in-air at height 1 m above the ground in Sv Gy−1 as a function of photon energy. For the volume source in the ground, conversion coefficients for the photon energy distributions corresponding to the natural radionuclides of the decay series of 238U, 232Th and 40K are tabulated in Table 6.5.
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Table 6.5. Organ equivalent dose conversion coefficients for the natural radionuclides for an adult, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan]. Organ
Adrenals Bladder Brain Colon Eye lenses Gonads Kidneys Liver Lungs Muscle Oesophagus Pancreas Red bone marrow Skeleton Skin Small intestine Spleen Stomach Thymus Thyroid Effective dose
Organ equivalent dose per unit air kerma free-in-air at 1 m above ground [Sv Gy−1] 238 232 40 U series Th series K 0.589 0.617 0.634 0.648 0.681 0.692 0.689 0.715 0.727 0.627 0.655 0.659 0.872 0.876 0.947 0.682 0.681 0.738 0.674 0.700 0.700 0.658 0.684 0.692 0.709 0.732 0.740 0.737 0.761 0.767 0.608 0.635 0.638 0.600 0.627 0.662 0.656 0.680 0.687 0.770 0.792 0.764 0.849 0.863 0.861 0.620 0.651 0.652 0.646 0.699 0.703 0.660 0.671 0.684 0.675 0.753 0.733 0.659 0.776 0.731 0.672 0.695 0.709
For the environmental irradiation geometries, the dependence of the organ equivalent dose conversion coefficients on photon energy is much more uniform than for the unidirectional geometries considered for occupational radiation exposures, and depends less on the position of the organ in the body. As the radiation comes from all directions, every organ is quasi deep-lying relative to at least a considerable part of the incoming photons. The conversion coefficients for the female model were found to be up to 5 % higher than those for the male model, due to the slightly smaller body size of the female model. Considering the two different source types, it can be seen that the equivalent dose conversion coefficients for the volume source in air are generally lower than those for the plane source in the ground. This results from the different angular distribution of the radiation impinging on the body: the gamma-ray field from a source in the air is nearly isotropic with respect to directions from the upper hemisphere, while the incident directions of the gamma rays from a plane source have strong horizontal bias, and most photons come from horizontal directions. Since the human body standing vertically has a reduced shielding effect for photons coming from horizontal directions, this leads to the higher doses resulting from this geometry. However, in most cases, the differences in the conversion coefficients were found to be less than 20 %. Saito et. al. [98Sai] investigated the variation of effective dose for environmental gamma-rays for source distributions other than these three typical ones and for a lying posture further to the standing one. The change of posture of a human body and the biases of environmental sources were found to affect the effective dose by some tens percent. A similar trend is anticipated for the individual organ doses. Therefore, it could be concluded that the conversion coefficients for the three typical environmental sources can be used as a reference set of values to derive the organ doses and effective doses of adults from air kerma or source activity obtained by measurement for a variety of environmental source configurations.
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Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads
lungs RBM stomach liver thyroid skin
Volume source in air
skeleton effective dose
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.11. Organ equivalent doses per unit air kerma at 1 m above the ground for some selected organs of an adult for a volume source in air, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Organ equivalent dose per unit air kerma [Sv/Gy]
2
gonads
lungs RBM stomach liver thyroid skin
Plane source in ground
skeleton effective dose
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.12. Organ equivalent doses per unit air kerma at 1 m above the ground for some selected organs of an adult for plane source in ground, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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6.3.2.2 Calculation of doses for radionuclides Kerma rates in air and equivalent dose rates in organs for radionuclides are obtained from dose conversion coefficients for the monoenergetic sources hT,i by multiplication with the yield yiN (in number of photons Bq−1) of photons with energy i per disintegration and summing over the photon energies of the emission spectrum of radionuclide N: gN =
∑y
N i
⋅ hT,i
i
Kerma rates in air calculated with Monte Carlo methods in the environment due to monoenergetic photon sources distributed exponentially in the soil or homogeneously in the air are given in Tables 6.3 and 6.4 respectively. Radionuclide-specific results are given in Table 6.6. Table 6.6. Kerma-rates in air at 1 m above ground per unit activity per unit area (nGy h-1 per kBq m-2) and per activity concentration in air; [94ICR]. Radionuclide Source in soil Volume Radionuclide Source in soil Volume source in air source in air at a depth of at a depth of 0.5 g cm−2 0.5 g cm−2 Kerma rate Kerma rate Kerma rate Kerma rate −1 −1 −1 [(nGy h )/ [(nGy h−1)/ [(nGy h )/ [(nGy h )/ −3 (kBq m )] (kBq m−3)] (kBq m−2)] (kBq m−2)] Be-7 Nb-93m 1.67 × 10−1 1.29 × 10−2 1.10 × 10−2 6.19 × 10−4 0 −1 0 Na-22 Nb-95 6.97 × 10 2.49 × 10 5.11 × 10 1.80 × 10−1 1 −1 −1 Na-24 Nb-95m 1.10 × 10 2.40 × 10 9.72 × 10 1.59 × 10−2 −1 −2 0 K-40 Nb-97 4.62 × 10 2.21 × 10 3.78 × 10 1.52 × 10−1 −1 −2 −2 K-42 Mo-93 8.53 × 10 7.58 × 10 6.52 × 10 3.54 × 10−3 0 −1 −1 Sc-46 Mo-99 6.31 × 10 4.85 × 10 4.75 × 10 3.43 × 10−2 −1 −3 −1 Cr-51 Tc-99m 1.09 × 10 3.93 × 10 7.06 × 10 2.64 × 10−2 0 1 0 Mn-54 Ru-103 2.69 × 10 1.64 × 10 1.97 × 10 1.08 × 10−1 0 −1 0 Mn-56 Ru-105 5.05 × 10 2.45 × 10 3.96 × 10 1.82 × 101 0 −1 −2 Fe-59 Rh-103m 3.64 × 10 8.82 × 10 2.81 × 10 5.00 × 104 0 −1 −1 Co-56 Rh-105 9.86 × 10 2.62 × 10 8.39 × 10 1.73 × 10−2 Co-57 Rh-106 3.88 × 10−1 6.90 × 10−1 2.53 × 10−2 4.75 × 10−2 0 −1 0 Co-58 Ag-110m 3.17 × 10 8.76 × 10 2.28 × 10 6.14 × 10−1 0 −1 −2 Co-60 Ag-111 7.59 × 10 8.91 × 10 5.90 × 10 5.90 × 10−3 0 −1 −1 Ni-65 Sn-117m 1.65 × 10 5.52 × 10 1.31 × 10 3.41 × 10−2 0 −1 −1 Zn-65 Sn-126 1.82 × 10 1.92 × 10 1.39 × 10 1.18 × 10−2 0 −2 0 Zn-69m Sb-124 1.41 × 10 5.67 × 10 9.58 × 10 4.25 × 10−1 0 −2 0 Se-75 Sb-125 1.30 × 10 1.48 × 10 8.68 × 10 9.97 × 102 0 −1 0 Br-84 Sb-126 4.90 × 10 9.07 × 10 4.18 × 10 6.66 × 101 −1 −2 0 Rb-86 Sb-127 2.94 × 10 2.20 × 10 2.25 × 10 1.61 × 10−1 0 −1 1 Sr-92 Sb-128 4.01 × 10 1.01 × 10 3.14 × 10 7.20 × 101 0 −1 0 Y-90m Sb-129 2.10 × 10 4.58 × 10 1.42 × 10 3.36 × 10−1 −2 −4 1 Y-91 Sb-130 1.10 × 10 1.04 × 10 8.50 × 10 7.63 × 10−1 0 −1 −1 Y-91m Te-123m 1.77 × 10 4.96 × 10 1.23 × 10 3.16 × 10−2 −1 −2 −1 Y-92 Te-125m 7.91 × 10 1.87 × 10 5.94 × 10 9.07 × 10−3 −1 −2 −2 Y-93 Te-127 2.69 × 10 1.65 × 10 2.06 × 10 1.11 × 10−3 Zr-95 Te-127m 2.40 × 100 5.91 × 10−2 1.73 × 10−1 2.78 × 10−3 −1 −2 −1 Te-129 Zr-97 2.07 × 10 6.06 × 10 4.14 × 10 1.38 × 10−2 Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide
Te-129m Te-131m Te-132 Te-133m Te-134 I-129 I-130 I-131 I-132 I-133 I-134 I-135 Cs-134 Cs-134m Cs-136 Cs-138 Ba-137m Ba-139 Ba-140 La-140 La-141 La-142 Ce-141 Ce-143 Ce-144 Pr-145 Nd-147 Pm-148 Pm-148m Pm-149 Pm-151 Eu-152 Eu-152m Eu-154 Eu-155 Eu-156 Hf-181
Landolt-Börnstein New Series VIII/4
6 External dosimetry Source in soil at a depth of 0.5 g cm−2 Kerma rate [(nGy h−1)/ (kBq m−2)] 1.65 × 10−1 4.51 × 100 8.05 × 10−1 5.79 × 100 2.84 × 100 1.14 × 10−1 7.05 × 100 1.29 × 100 7.35 × 100 2.01 × 100 8.26 × 100 4.79 × 100 5.09 × 100 1.08 × 10−1 6.75 × 100 6.96 × 100 1.98 × 100 1.45 × 10−1 6.21 × 10−1 6.93 × 100 1.27 × 10−1 6.57 × 100 2.44 × 10−1 9.28 × 10−1 6.37 × 10−2 8.32 × 10−2 4.71 × 10−1 1.78 × 100 6.55 × 100 1.23 × 10−3 1.09 × 100 3.53 × 100 9.47 × 10−1 3.85 × 100 1.88 × 10−1 3.98 × 100 1.79 × 100
Volume source in air Kerma rate [(nGy h−1)/ (kBq m−3)] 9.04 × 103 3.22 × 10−1 5.15 × 10−2 5.36 × 10−1 2.03 × 10−1 6.01 × 10−3 4.93 × 101 8.68 × 10−2 5.26 × 10−1 1.40 × 101 6.05 × 10−1 3.67 × 10−1 3.64 × 10−1 5.90 × 10−3 5.08 × 10−1 5.44 × 10−1 1.40 × 10−1 9.18 × 10−3 4.14 × 10−2 5.44 × 10−1 9.97 × 10−3 6.26 × 10−1 1.59 × 10−2 6.34 × 10−2 4.36 × 10−3 3.11 × 10−3 3.11 × 10−2 1.35 × 10−1 4.64 × 10−1 2.36 × 10−3 6.88 × 10−2 1.82 × 10−1 6.91 × 10−2 2.89 × 10−1 1.22 × 10−2 3.09 × 10−1 1.25 × 10−1
Radionuclide
Ta-182 W-187 Pb-210 Pb-212 Bi-212 Ra-224 Ra-226 Ac-228 Th-228 Th-231 Th-232 Th-234 Pa-233 U-232 U-234 U-235 U-236 U-237 U-238 Np-237 Np-238 Np-239 Pu-236 Pu-238 Pu-239 Pu-240 Pu-242 Am-241 Am-242 Am-242m Am-243 Cm-242 Cm-243 Cm-244 Cm-245 Cm-247
6-19 Source in soil at a depth of 0.5 g cm−2 Kerma rate [(nGy h−1)/ (kBq m−2)] 4.00 × 100 1.70 × 100 2.71 × 10−2 4.74 × 10−1 3.38 × 10−1 3.33 × 10−2 2.19 × 10−2 3.05 × 100 1.43 × 10−2 1.31 × 10−1 7.89 × 10−3 3.33 × 10−2 7.37 × 10−1 1.06 × 10−2 9.47 × 10−3 5.55 × 10−1 7.90 × 10−3 4.80 × 10−1 8.33 × 10−3 1.42 × 10−1 2.05 × 100 6.02 × 10−1 1.08 × 10−2 1.06 × 10−2 4.38 × 10−3 9.25 × 10−3 8.56 × 10−3 1.12 × 10−1 6.76 × 10−2 3.25 × 10−2 1.81 × 10−1 9.38 × 10−3 4.47 × 10−1 8.87 × 10−3 3.10 × 10−1 1.03 × 100
Volume source in air Kerma rate [(nGy h−1)/ (kBq m−3)] 2.99 × 10−1 1.09 × 10−1 1.51 × 10−3 3.24 × 10−2 4.36 × 10−2 2.17 × 10−3 1.46 × 10−3 2.19 × 10−1 8.75 × 10−4 7.02 × 10−3 4.54 × 104 2.07 × 10−3 4.54 × 10−2 7.24 × 10−4 5.80 × 10−4 3.34 × 10−2 5.44 × 10−4 3.11 × 10−2 6.88 × 10−4 8.42 × 10−3 1.31 × 10−1 3.78 × 10−2 7.06 × 10−4 6.19 × 10−4 2.41 × 10−4 5.83 × 10−4 5.04 × 10−4 7.96 × 10−3 4.39 × 10−3 1.61 × 10−3 1.14 × 10−2 6.05 × 10−4 2.97 × 10−2 5.62 × 10−4 2.07 × 10−2 7.16 × 10−2
6-20
6 External dosimetry
[Ref. p. 6-42
6.4 Conversion coefficients for neutrons Significant radiation exposures to neutrons occur primarily at workplaces and not in the environment. In the natural environment neutrons are found mainly in secondary cosmic ray fields [00Pel, 02Roe]; accidental exposures to neutron emitting radionuclides in clouds or on the soil are extremely unlikely and therefore not considered here. Neutron fields are in practice mixed radiation fields of wide neutron energy range, almost always associated with photons. To obtain the conversion coefficients for such fields, appropriate averaging of coefficients over the relevant spectra should be performed. The calculation of deposition of energy at any point in a body resulting from external exposure in mixed fields is a complex process of summation over all primary and secondary particle deposition. Several authors calculated for incident neutrons the protection quantities organ absorbed dose and effective dose using anthropomorphic phantoms such as the hermaphrodite MIRD-5 phantom or the sexspecific MIRD-type phantoms Adam and Eva (see Sect. 6.2 and 6.3). For the operational quantities, the ICRU sphere and slab phantoms were used. Several Monte Carlo codes were applied such as the MCNP [91Bri], SAM-CE [79Lic], MORSE-CG [75Emm], the JAERI code [Yam93], the PTB-code [90Hol] etc. Extensive tables of organ dose conversion coefficients data are given in ICRP Report 74 [96ICR] and ICRU Report 57 [978ICR] derived as sets of “best estimates” of the data of various authors. Table 6.7 lists the effective dose per unit of neutron fluence for idealized whole-body irradiation geometries and for energies ranging from thermal up to 180 MeV. In the same table, the coefficients for ambient dose equivalent are also given. Table 6.7. Effective dose per unit neutron fluence E/Φ for monoenergetic neutrons incident in various geometries on an adult anthropomorphic computational model. The last column of the table shows the coefficients for ambient dose equivalent; [978ICR]. Energy E/Φ [pSv cm2] H*(10)/Φ [MeV] AP PA RLAT LLAT ROT ISO [pSv cm2] 5.24 3.52 1.36 1.68 2.99 2.99 6.60 1.0 × 10−9 6.55 4.39 1.70 2.04 3.72 2.89 9.00 1.0 × 10−8 7.60 5.16 1.99 2.31 4.40 3.30 10.6 2.5 × 10−8 9.95 6.77 2.58 2.86 5.75 4.13 12.9 1.0 × 10−7 11.2 7.63 2.92 3.21 6.43 4.59 13.5 2.0 × 10−7 12.8 8.76 3.35 3.72 7.27 5.20 13.6 5.0 × 10−7 13.8 9.55 3.67 4.12 7.84 5.63 13.3 1.0 × 10−6 14.5 10.2 3.89 4.39 8.31 5.96 12.9 2.0 × 10−6 15.0 10.7 4.08 4.66 8.72 6.28 12.0 5.0 × 10−6 15.1 11.0 4.16 4.80 8.90 6.44 11.3 1.0 × 10−5 15.1 11.1 4.20 4.89 8.92 6.51 10.6 2.0 × 10−5 −5 14.8 11.1 4.19 4.95 8.82 6.51 9.90 5.0 × 10 −4 14.6 11.0 4.15 4.95 8.69 6.45 9.40 1.0 × 10 −4 14.4 10.9 4.10 4.92 8.56 6.32 8.90 2.0 × 10 −4 14.2 10.7 4.03 4.86 8.40 6.14 8.30 5.0 × 10 14.2 10.7 4.00 4.84 8.34 6.04 7.90 1.0 × 10−3 14.4 10.8 4.00 4.87 8.39 6.05 7.70 2.0 × 10−3 15.7 11.6 4.29 5.25 9.06 6.52 8.00 5.0 × 10−3 18.3 13.5 5.02 6.14 10.6 7.70 10.5 1.0 × 10−2 23.8 17.3 6.48 7.95 13.8 10.2 16.6 2.0 × 10−2 29.0 21.0 7.93 9.74 16.9 12.7 23.7 3.0 × 10−2 38.5 27.6 10.6 13.1 22.7 17.3 41.1 5.0 × 10−2 47.2 33.5 13.1 16.1 27.8 21.5 60.0 7.0 × 10−2
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
Energy [MeV] 1.0 × 10−1 1.5 × 10−1 2.0 × 10−1 3.0 × 10−1 5.0 × 10−1 7.0 × 10−1 9.0 × 10−1 1.0 × 100 1.2 × 100 2.0 × 100 3.0 × 100 4.0 × 100 5.0 × 100 6.0 × 100 7.0 × 100 8.0 × 100 9.0 × 100 1.0 × 101 1.2 × 101 1.4 × 101 1.5 × 101 1.6 × 101 1.8 × 101 2.0 × 101 3.0 × 101 5.0 × 101 7.5 × 101 1.0 × 102 1.3 × 102 1.5 × 102 1.8 × 102 2.0 × 102 a Not available
E/Φ [pSv cm2] RLAT LLAT 16.4 20.1 21.2 25.5 25.6 30.3 33.4 38.6 46.8 53.2 58.3 66.6 69.1 79.6 74.5 86.0 85.8 99.8 129 153 171 195 198 224 217 244 232 261 244 274 253 285 261 294 268 302 278 315 286 324 290 328 293 331 299 335 305 338 324 naa 358 naa 397 naa 433 naa 467 naa 501 naa 542 naa
AP 59.8 80.2 99.0 133 188 231 267 282 310 383 432 458 474 483 490 494 497 499 499 496 494 491 486 480 458 437 429 429 432 438 445
PA 41.3 52.2 61.5 77.1 103 124 144 154 175 247 308 345 366 380 391 399 406 412 422 429 431 433 435 436 437 444 459 477 495 514 535
6-21
ROT 34.8 45.4 54.8 71.6 99.4 123 144 154 173 234 283 315 335 348 358 366 373 378 385 390 391 393 394 395 395 404 422 443 465 489 517
ISO 27.2 35.2 42.4 54.7 75.0 92.8 108 116 130 178 220 250 272 282 290 297 303 309 322 333 338 342 345 343 naa naa naa naa naa naa naa
H*(10)/Φ [pSv cm2] 88.0 132 170 233 322 375 400 416 425 420 412 408 405 400 405 409 420 440 480 520 540 555 570 600 515 400 300 285 260 245 250 260
6.5 Conversion coefficients for electrons 6.5.1 Occupational exposure Unshielded whole body irradiation by monoenergetic electrons does not represent a practical situation in occupational exposures and evaluated absorbed doses for electron beams are still sparse. However, irradiation of the skin, the lens of the eye and other superficial organs are of concern in radiological protection for electron energies below 10 MeV because the electron range is small, varying from 50 µm to about 5 cm for electron energies from 60 keV to 10 MeV. Table 6.8 shows some conversion coefficients for organ absorbed doses determined with the MCNP-4 code for the MIRD-type phantoms Adam and Eva, for monoenergetic electrons in the energy range of 100 keV to 10 MeV, incident in the AP geometry Landolt-Börnstein New Series VIII/4
6-22
6 External dosimetry
[Ref. p. 6-42
(Schultz and Zoetelief, data from [96ICR]). ICRU Report 43 contains dose distributions in anthropomorphic phantoms resulting from irradiation by electrons of energies between 5 and 46 MeV [88ICR]. Various workers performed Monte Carlo calculations with different codes (EGS4 [85Nel], MCNP-4 [91Bri], MCNP-BO code [94Gua1, 94Gua2], PTB-BG code [86Gro] etc.) enabling them to derive fluence-to-dose-equivalent conversion coefficients for parallel electron beams of energies between 60 keV and 10 MeV. The conversion coefficients for H’(0.07,α), H’(3,α), Hp,slab(0.07,α), Hp,slab(3,α) and Hp,slab(10,α) were determined with the ICRU sphere or the 4-element ICRU tissue slab phantom, respectively. A compilation of data can be found in [96Cha], in ICRP 74 and ICRU 57 [96ICR, 98ICR]. By appropriately averaging of these data, reference fluence-to-dose-equivalent conversion coefficients were derived as a function of energy for normally incident electrons. Table 6.9 shows these data for depths of 0.07 mm and 3 mm. Table 6.8. Organ absorbed dose per unit fluence DT/Φ and effective dose per unit fluence E/Φ for monoenergetic electrons incident in the AP geometry on an adult anthropomorphic computational model (Schultz and Zoetelief, data from [96ICR]). Energy [MeV]
0.1
0.4
0.6
Skin Testes Bone marrow Stomach Breast Liver Thyroid Effective dose
8
98
171 0 0 0
0.1
1
1.5
1.0 1.5 2.0 2 DT/Ф and E/Ф [pGy cm ] 164 158 153 1 14 37 1 5 11 0 14 43 75
2.7
5.9
0 11
4.0
10.0
150 214 28 3 200 0 121 44
165 345 52 184 325 97 297 131
Table 6.9. Reference conversion coefficients from fluence to directional dose equivalent for monoenergetic electrons and normal incidence Energy H´(0.07,0°)/Φ Energy H´(0.07,0°)/Φ H´(3,0°)/Φ H´(3,0°)/Φ [MeV] [MeV] [nSv cm2] [nSv cm2] [nSv cm2] [nSv cm2] 1.00 0.312 0.301 0.07 0.221 1.25 0.296 0.486 0.08 1.056 1.50 0.287 0.524 0.09 1.527 1.75 0.282 0.512 0.10 1.661 2.00 0.279 0.481 0.1125 1.627 2.50 0.278 0.417 0.125 1.513 3.00 0.276 0.373 0.15 1.229 3.50 0.274 0.351 0.20 0.834 4.00 0.272 0.334 0.30 0.542 5.00 0.271 0.317 0.40 0.455 6.00 0.271 0.309 0.50 0.403 7.00 0.271 0.306 0.60 0.366 8.00 0.271 0.305 0.70 0.344 0.000 10.0 0.275 0.303 0.80 0.329 0.045
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
6-23
6.5.2 Environmental exposure Due to the short ranges of electrons emitted by radionuclides, electrons contribute only to the dose to skin. Skin dose coefficients for a series of monoenergetic electron sources were calculated by Eckerman and Ryman [93Eck1] using the code DOSFACTER of Kocher [88DOE]. The results are shown in Figs. 6.13 and 6.14 for submersion in contaminated air and for exposure to contaminated soil respectively. These data can be then convoluted to the spectra of the various radionuclides, using the energy and intensity of beta and electron emissions of radionuclides to obtain radionuclide specific conversion coefficients. 100
10-1 Contaminated air
Contaminated soil
10-2 -1 -3 h(skin) [pSv s per Bq m ]
-1 -3 h(skin) [pSv s per Bq m ]
10-1 10-2 10-3 10-4
10-3 10-4 10-5 10-6 10-7 Infinite volume Surface
10-8
10-5 10-9 10-6 10-2
10-1
100
101
Electron Energy [MeV]
Fig. 6.13. Electron skin dose coefficient for submersion in air; [93Eck1].
10-10 10-2
10-1
100
Electron Energy [MeV]
Fig. 6.14. Electron skin dose coefficient for exposure to contaminated soil on the surface and in the volume; [93Eck1].
6.6 Doses from external exposure of radionuclides in the environment Only photons, including bremsstrahlung, and electrons emitted by the radionuclides are sufficiently penetrating to traverse the overlying tissues of the body and contribute to the dose to tissues and organs of the body. The energy spectra of emitted radiation are either discrete, as in the case of photons, or continuous, as in the case of beta particles and bremsstrahlung. The dose coefficient H TS for tissue T for any exposure mode S can be expressed as H TS =
∞
∑ [∑ y j ( E i ) H TS, j ( Ei ) + ∫ 0
j =e ,γ
y j ( E ) H TS, j ( E )dE ]
i
where y j ( Ei ) is the yield of radiations of type j and discrete energy Ei and y j (E ) denotes the yield of radiations per nuclear transformation with continuous energy between E and E + dE. The other summation is over all electron and photon radiations. The contribution of the radiations to the dose in tissue or organ T is defined by the quantity H TS which is estimated by means of Monte Carlo calculations and is given as a function of energy for tissue and organ T for each exposure mode [93Eck1]. Landolt-Börnstein New Series VIII/4
6-24
6 External dosimetry
[Ref. p. 6-42
By using the dose conversion coefficients for monoenergetic sources of photon and electron radiation and by scaling them to the emissions of the radionuclides of interest, dose coefficients from radionuclides in the environment can be derived. The following tables contain data from the American Federal Guidance Report No. 12, based on Monte Carlo radiation transport calculations and data obtained from Eckerman [02Eck]. Tables 6.10 and 6.11 give the skin dose and effective dose coefficients for several radionuclides for exposure to contaminated ground surface to a depth of 5 cm and for air submersion respectively. The nuclear decay data used are from Eckerman et. al. [93Eck2] and are based on the ICRP Publication 38 [83ICR] on radionuclide transformations. Table 6.10. Effective dose and skin dose coefficients for exposure depth of 5 cm ; [93Eck1] and [02Eck]. Effective dose Radionuclide Skin dose Radionuclide [Sv/(Bq s m−2)] [Sv/(Bq s m−2)] He-3 0.00 0.00 Ti-45 Be-7 Sc-46 1.04·10−18 8.55·10−19 Be-10 Ca-47 4.21·10−21 3.02·10−20 C-11 2.23·10−17 1.77·10−17 Sc-47 N-13 2.34·10−17 V-47 1.77·10−17 C-14 1.21·10−22 Cr-48 5.50·10−23 O-15 2.72·10−17 Sc-48 1.78·10−17 F-18 V-48 2.16·10−17 1.77·10−17 Ne-19 Ca-49 3.14·10−17 1.79·10−17 −17 −17 Na-22 4.48·10 3.76·10 Cr-49 Na-24 8.02·10−17 Sc-49 6.80·10−17 Al-26 5.52·10−17 V-49 4.54·10−17 Cr-51 Al-28 5.00·10−17 3.01·10−17 Mg-28 Mn-51 2.74·10−17 2.31·10−17 P-30 Fe-52 4.10·10−17 1.81·10−17 Si-31 Mn-52m 3.59·10−18 7.88·10−20 P-32 Mn-52 5.18·10−18 9.05·10−20 Si-32 Mn-53 2.67·10−22 1.43·10−22 Mn-54 P-33 4.11·10−22 2.36·10−22 Co-55 S-35 1.30·10−22 6.08·10−23 Fe-55 Cl-36 1.76·10−19 9.72·10−21 Co-56 Ar-37 0.00 0.00 Mn-56 Cl-38 5.04·10−17 2.53·10−17 Ni-56 K-38 7.71·10−17 5.38·10−17 Co-57 Ar-39 2.29·10−20 3.37·10−21 Ni-57 Cl-39 3.63·10−17 2.46·10−17 Co-58m K-40 6.11·10−18 2.70·10−18 Co-58 Ar-41 2.75·10−17 2.19·10−17 Fe-59 Ca-41 0.00 0.00 K-42 Ni-59 2.52·10−17 5.05·10−18 Co-60m K-43 2.08·10−17 1.68·10−17 −17 −17 Co-60 Sc-43 2.38·10 1.90·10 −17 −17 Cu-60 K-44 6.38·10 3.81·10 −18 −18 Fe-60 Sc-44m 5.84·10 4.79·10 −17 −17 Co-61 Sc-44 4.76·10 3.68·10 −18 −18 Cu-61 Ti-44 1.90·10 1.47·10 −22 −22 Co-62m Ca-45 4.30·10 2.50·10 −17 −17 Cu-62 K-45 4.66·10 3.14·10
to contaminated ground surface to a Skin dose [Sv/(Bq s m−2)] 1.94·10−17 4.09·10−17 2.27·10−17 2.13·10−18 2.80·10−17 8.80·10−18 6.79·10−17 5.92·10−17 6.47·10−17 2.55·10−17 7.49·10−18 0.00 6.64·10−19 3.05·10−17 1.56·10−17 6.24·10−17 7.01·10−17 0.00 1.72·10−17 4.33·10−17 0.00 7.09·10−17 4.24·10−17 3.53·10−17 2.21·10−18 3.85·10−17 4.58·10−23 2.02·10−17 2.39·10−17 0.00 9.16·10−20 5.02·10−17 8.68·10−17 1.08·10−22 3.26·10−18 1.86·10−17 6.56·10−17 3.77·10−17
Effective dose [Sv/(Bq s m−2)] 1.51·10−17 3.45·10−17 1.81·10−17 1.74·10−18 1.74·10−17 7.19·10−18 5.73·10−17 4.99·10−17 5.08·10−17 1.80·10−17 1.47·10−19 0.00 5.43·10−19 1.74·10−17 1.26·10−17 4.15·10−17 5.92·10−17 0.00 1.44·10−17 3.44·10−17 0.00 6.03·10−17 2.88·10−17 2.94·10−17 1.81·10−18 3.26·10−17 9.55·10−24 1.68·10−17 2.03·10−17 0.00 6.82·10−20 4.27·10−17 6.62·10−17 4.60·10−23 1.17·10−18 1.43·10−17 4.60·10−17 1.77·10−17 Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Zn-62 Ni-63 Zn-63 Cu-64 Ga-65 Ni-65 Zn-65 Cu-66 Ga-66 Ge-66 Ni-66 Cu-67 Ga-67 Ge-67 Ga-68 Ge-68 As-69 Ge-69 Zn-69m Zn-69 As-70 Ga-70 Se-70 As-71 Ge-71 Zn-71m As-72 Ga-72 Zn-72 As-73 Ga-73 Se-73m Se-73 As-74 Br-74m Br-74 Kr-74 Br-75 Ge-75 Se-75 As-76 Br-76 Kr-76 As-77 Br-77 Ge-77 Kr-77 Se-77m Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 9.03·10−18 0.00 3.27·10−17 4.00·10−18 3.21·10−17 1.62·10−17 1.18·10−17 1.40·10−17 6.18·10−17 1.42·10−17 3.11·10−22 2.24·10−18 3.01·10−18 4.62·10−17 2.65·10−17 4.59·10−22 3.76·10−17 1.87·10−17 8.79·10−18 4.90·10−19 9.17·10−17 4.65·10−18 2.34·10−17 1.19·10−17 4.65·10−22 3.56·10−17 4.93·10−17 5.69·10−17 2.81·10−18 6.27·10−20 8.13·10−18 6.34·10−18 2.42·10−17 1.68·10−17 1.00·10−16 1.00·10−16 3.16·10−17 2.93·10−17 2.11·10−18 7.90·10−18 2.15·10−17 6.09·10−17 8.86·10−18 2.64·10−19 6.59·10−18 2.75·10−17 2.62·10−17 1.67·10−18
Effective dose [Sv/(Bq s m−2)] 7.41·10−18 0.00 1.92·10−17 3.28·10−18 2.02·10−17 9.42·10−18 9.96·10−18 1.67·10−18 4.13·10−17 1.15·10−17 1.73·10−22 1.82·10−18 2.45·10−18 2.44·10−17 1.66·10−17 6.94·10−24 1.78·10−17 1.50·10−17 7.21·10−18 1.28·10−20 7.02·10−17 2.08·10−19 1.70·10−17 9.70·10−18 7.02·10−24 2.69·10−17 3.11·10−17 4.58·10−17 2.29·10−18 4.03·10−20 5.31·10−18 4.22·10−18 1.84·10−17 1.31·10−17 6.92·10−17 7.53·10−17 2.00·10−17 2.10·10−17 6.11·10−19 6.46·10−18 7.64·10−18 4.46·10−17 7.23·10−18 1.54·10−19 5.42·10−18 1.87·10−17 1.74·10−17 1.36·10−18
Radionuclide As-78 Ge-78 Kr-79 Rb-79 Se-79 Br-80m Br-80 Rb-80 Sr-80 Kr-81m Kr-81 Rb-81m Rb-81 Se-81m Se-81 Sr-81 Br-82 Rb-82m Rb-82 Sr-82 Br-83 Kr-83m Rb-83 Se-83 Sr-83 Br-84 Rb-84 Kr-85m Kr-85 Sr-85m Sr-85 Rb-86 Y-86m Y-86 Zr-86 Kr-87 Rb-87 Sr-87m Y-87 Kr-88 Nb-88 Rb-88 Y-88 Zr-88 Nb-89b Nb-89a Rb-89 Sr-89
6-25 Skin dose [Sv/(Bq s m−2)] 4.38·10−17 5.94·10−18 5.27·10−18 3.60·10−17 1.61·10−22 1.20·10−19 2.55·10−18 5.75·10−17 1.48·10−20 2.61·10−18 1.24·10−19 7.66·10−20 1.33·10−17 2.40·10−19 4.15·10−18 4.04·10−17 5.43·10−17 5.99·10−17 4.23·10−17 1.45·10−20 6.77·10−19 3.47·10−21 1.05·10−17 5.23·10−17 1.69·10−17 5.04·10−17 1.91·10−17 3.37·10−18 1.67·10−19 4.51·10−18 1.06·10−17 7.19·10−18 4.53·10−18 7.38·10−17 5.59·10−18 3.33·10−17 8.93·10−22 6.73·10−18 9.47·10−18 3.99·10−17 1.00·10−16 4.41·10−17 5.27·10−17 8.32·10−18 4.33·10−17 4.89·10−17 5.26·10−17 3.69·10−18
Effective dose [Sv/(Bq s m−2)] 2.17·10−17 4.77·10−18 4.33·10−18 2.35·10−17 7.61·10−23 4.75·10−20 1.39·10−18 2.23·10−17 5.33·10−22 2.12·10−18 9.48·10−20 5.28·10−20 1.06·10−17 1.91·10−19 2.29·10−19 2.40·10−17 4.55·10−17 5.00·10−17 1.93·10−17 5.24·10−22 1.43·10−19 1.35·10−22 8.61·10−18 4.15·10−17 1.36·10−17 2.99·10−17 1.58·10−17 2.56·10−18 4.40·10−20 3.68·10−18 8.73·10−18 1.71·10−18 3.69·10−18 6.12·10−17 4.52·10−18 1.36·10−17 5.55·10−22 5.51·10−18 7.77·10−18 3.25·10−17 7.11·10−17 1.13·10−17 4.53·10−17 6.80·10−18 2.39·10−17 3.35·10−17 3.52·10−17 6.67·10−20
6-26 Radionuclide Zr-89 Mo-90 Nb-90 Sr-90 Y-90m Y-90 Sr-91 Y-91m Y-91 Sr-92 Y-92 Mo-93m Mo-93 Nb-93m Tc-93m Tc-93 Y-93 Zr-93 Nb-94 Ru-94 Tc-94m Tc-94 Y-94 Nb-95m Nb-95 Tc-95m Tc-95 Y-95 Zr-95 Nb-96 Tc-96m Tc-96 Nb-97m Nb-97 Ru-97 Tc-97m Tc-97 Zr-97 Nb-98 Tc-98 Mo-99 Rh-99m Rh-99 Tc-99m Tc-99 Pd-100 Rh-100 Mo-101
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.41·10−17 1.71·10−17 8.42·10−17 1.41·10−20 1.31·10−17 9.85·10−18 1.98·10−17 1.12·10−17 4.07·10−18 2.69·10−17 2.53·10−17 4.52·10−17 2.99·10−20 5.26·10−21 1.39·10−17 2.87·10−17 1.64·10−17 0.00 3.26·10−17 1.09·10−17 4.63·10−17 5.51·10−17 4.70·10−17 1.28·10−18 1.59·10−17 1.38·10−17 1.63·10−17 3.90·10−17 1.54·10−17 5.11·10−17 9.38·10−19 5.16·10−17 1.52·10−17 1.57·10−17 4.73·10−18 3.51·10−20 3.44·10−20 9.34·10−18 5.85·10−17 2.95·10−17 4.45·10−18 1.40·10−17 1.22·10−17 2.38·10−18 7.94·10−22 1.71·10−18 5.49·10−17 3.16·10−17
Effective dose [Sv/(Bq s m−2)] 2.00·10−17 1.38·10−17 7.06·10−17 2.72·10−21 1.07·10−17 1.74·10−19 1.21·10−17 9.19·10−18 1.32·10−19 2.28·10−17 4.71·10−18 3.83·10−17 2.23·10−21 3.94·10−22 1.20·10−17 2.46·10−17 1.78·10−18 0.00 2.72·10−17 9.01·10−18 3.19·10−17 4.59·10−17 1.95·10−17 1.03·10−18 1.32·10−17 1.14·10−17 1.36·10−17 1.52·10−17 1.28·10−17 4.26·10−17 7.71·10−19 4.30·10−17 1.26·10−17 1.14·10−17 3.83·10−18 7.18·10−21 3.02·10−21 3.17·10−18 4.18·10−17 2.45·10−17 2.58·10−18 1.16·10−17 9.99·10−18 1.95·10−18 4.94·10−22 1.29·10−18 4.64·10−17 2.33·10−17
Radionuclide Pd-101 Rh-101m Rh-101 Tc-101 Ag-102 Rh-102m Rh-102 Ag-103 Pd-103 Rh-103m Ru-103 Ag-104m Ag-104 Cd-104 Tc-104 Ag-105 Rh-105 Ru-105 Ag-106m Ag-106 Rh-106m Rh-106 Ru-106 Cd-107 Rh-107 Pd-107 Ag-108m Ag-108 Ag-109m Cd-109 In-109 Pd-109 Ag-110m Ag-110 In-110b In-110a Sn-110 Ag-111 In-111 Sn-111 Ag-112 In-112 Cd-113m Cd-113 In-113m Sn-113 In-114m In-114
[Ref. p. 6-42 Skin dose [Sv/(Bq s m−2)] 6.60·10−18 6.22·10−18 5.09·10−18 9.09·10−18 7.72·10−17 1.08·10−17 4.42·10−17 1.72·10−17 5.50·10−20 6.22·10−21 9.87·10−18 2.98·10−17 5.52·10−17 4.80·10−18 6.22·10−17 1.07·10−17 1.65·10−18 1.76·10−17 5.77·10−17 1.93·10−17 6.01·10−17 2.36·10−17 0.00 2.78·10−19 8.30·10−18 0.00 3.38·10−17 4.49·10−18 8.32·10−20 1.46·10−19 1.35·10−17 9.38·10−19 5.64·10−17 1.52·10−17 6.27·10−17 3.85·10−17 6.03·10−18 1.36·10−18 7.98·10−18 1.17·10−17 3.22·10−17 6.56·10−18 1.72·10−20 7.14·10−22 5.34·10−18 1.81·10−19 1.81·10−18 5.58·10−20
Effective dose [Sv/(Bq s m−2)] 5.39·10−18 5.05·10−18 4.12·10−18 5.78·10−18 5.73·10−17 8.28·10−18 3.67·10−17 1.29·10−17 8.91·10−21 9.09·10−22 8.12·10−18 2.01·10−17 4.59·10−17 3.90·10−18 3.40·10−17 8.72·10−18 1.34·10−18 1.36·10−17 4.83·10−17 1.23·10−17 5.01·10−17 3.91·10−18 0.00 1.62·10−19 5.40·10−18 0.00 2.79·10−17 3.77·10−19 4.61·10−20 5.67·10−20 1.12·10−17 7.58·10−20 4.73·10−17 7.96·10−19 5.23·10−17 2.67·10−17 4.87·10−18 4.71·10−19 6.45·10−18 8.58·10−18 1.16·10−17 4.56·10−18 2.54·10−21 4.45·10−22 4.36·10−18 9.77·10−20 1.46·10−18 4.68·10−20
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Ag-115 Cd-115m Cd-115 In-115m In-115 Sb-115 In-116m Sb-116m Sb-116 Te-116 Cd-117m Cd-117 In-117m In-117 Sb-117 Sn-117m Sb-118m In-119m In-119 Sb-119 Sn-119m I-120m I-120 Sb-120b Sb-120a Xe-120 I-121 Sn-121m Sn-121 Te-121m Te-121 Xe-121 I-122 Sb-122 Xe-122 I-123 Sn-123m Sn-123 Te-123m Te-123 Xe-123 I-124 Sb-124n Sb-124m Sb-124 Cs-125 I-125 Sb-125 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.62·10−17 4.44·10−18 5.53·10−18 3.28·10−18 3.09·10−21 2.04·10−17 4.99·10−17 6.38·10−17 4.72·10−17 9.18·10−19 4.07·10−17 2.48·10−17 4.30·10−18 1.45·10−17 3.33·10−18 2.83·10−18 5.17·10−17 1.21·10−17 2.00·10−17 1.08·10−19 5.18·10−20 1.24·10−16 7.52·10−17 4.94·10−17 1.16·10−17 8.29·10−18 8.62·10−18 2.61·10−20 1.16·10−21 4.22·10−18 1.18·10−17 4.22·10−17 3.36·10−17 1.29·10−17 1.07·10−18 3.04·10−18 4.54·10−18 2.99·10−18 2.70·10−18 9.89·10−20 1.35·10−17 2.38·10−17 3.50·10−22 7.65·10−18 3.89·10−17 1.70·10−17 2.24·10−19 8.82·10−18
Effective dose [Sv/(Bq s m−2)] 1.21·10−17 4.48·10−19 4.03·10−18 2.65·10−18 1.52·10−21 1.56·10−17 4.21·10−17 5.33·10−17 3.67·10−17 6.64·10−19 3.45·10−17 1.86·10−17 1.50·10−18 1.18·10−17 2.67·10−18 2.26·10−18 4.35·10−17 3.38·10−19 1.33·10−17 2.41·10−20 1.13·10−20 9.05·10−17 4.65·10−17 4.15·10−17 7.69·10−18 6.72·10−18 6.84·10−18 7.64·10−21 7.58·10−22 3.41·10−18 9.68·10−18 3.04·10−17 1.66·10−17 7.70·10−18 8.05·10−19 2.41·10−18 2.23·10−18 1.69·10−19 2.16·10−18 2.49·10−20 1.04·10−17 1.86·10−17 7.38·10−23 6.10·10−18 3.10·10−17 1.15·10−17 6.36·10−20 7.21·10−18
Radionuclide Sn-125 Te-125m Xe-125 Ba-126 Cs-126 I-126 Sb-126m Sb-126 Sn-126 Cs-127 Sb-127 Sn-127 Te-127m Te-127 Xe-127 Ba-128 Cs-128 I-128 Sb-128b Sb-128a Sn-128 Cs-129 I-129 Sb-129 Te-129m Te-129 Xe-129m Cs-130 I-130 Sb-130 Ba-131m Ba-131 Cs-131 I-131 La-131 Sb-131 Te-131m Te-131 Xe-131m Cs-132 I-132m I-132 La-132 Te-132 Ba-133m Ba-133 I-133 Te-133m
6-27 Skin dose [Sv/(Bq s m−2)] 1.47·10−17 1.94·10−19 5.09·10−18 3.06·10−18 4.43·10−17 9.69·10−18 3.68·10−17 6.02·10−17 7.67·10−19 8.44·10−18 1.50·10−17 4.28·10−17 6.64·10−20 1.81·10−19 5.34·10−18 1.23·10−18 2.92·10−17 8.51·10−18 6.64·10−17 5.08·10−17 1.30·10−17 5.42·10−18 1.48·10−19 3.15·10−17 1.96·10−18 3.95·10−18 4.32·10−19 1.44·10−17 4.52·10−17 7.32·10−17 1.16·10−18 9.05·10−18 1.31·10−19 8.03·10−18 1.49·10−17 4.15·10−17 2.95·10−17 1.42·10−17 1.58·10−19 1.44·10−17 6.88·10−18 4.99·10−17 4.69·10−17 4.43·10−18 1.13·10−18 7.71·10−18 1.42·10−17 5.26·10−17
Effective dose [Sv/(Bq s m−2)] 5.49·10−18 5.81·10−20 4.07·10−18 2.44·10−18 1.92·10−17 7.70·10−18 2.69·10−17 4.90·10−17 5.82·10−19 6.85·10−18 1.19·10−17 3.26·10−17 2.06·10−20 8.80·10−20 4.28·10−18 9.36·10−19 1.57·10−17 1.57·10−18 5.35·10−17 3.45·10−17 1.06·10−17 4.35·10−18 5.11·10−20 2.47·10−17 5.44·10−19 1.00·10−18 2.33·10−19 8.81·10−18 3.70·10−17 5.61·10−17 9.03·10−19 7.35·10−18 4.20·10−20 6.56·10−18 1.11·10−17 3.18·10−17 2.43·10−17 7.14·10−18 8.13·10−20 1.19·10−17 5.40·10−18 3.94·10−17 3.41·10−17 3.54·10−18 8.70·10−19 6.18·10−18 1.05·10−17 3.96·10−17
6-28 Radionuclide Te-133 Xe-133m Xe-133 Ce-134 Cs-134m Cs-134 I-134 La-134 Te-134 Ba-135m Ce-135 Cs-135m Cs-135 I-135 La-135 Xe-135m Xe-135 Cs-136 Nd-136 Pr-136 Ba-137m Ce-137m Ce-137 Cs-137 La-137 Pr-137 Cs-138 La-138 Nd-138 Pr-138m Pr-138 Xe-138 Ba-139 Ce-139 Nd-139m Nd-139 Pr-139 Ba-140 La-140 Ba-141 Ce-141 La-141 Nd-141m Nd-141 Pm-141 Sm-141m Sm-141 Ba-142
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.67·10−17 5.84·10−19 5.59·10−19 1.72·10−19 3.57·10−19 3.24·10−17 5.80·10−17 2.34·10−17 1.84·10−17 9.93·10−19 3.74·10−17 3.30·10−17 2.87·10−22 3.27·10−17 3.76·10−19 9.07·10−18 5.60·10−18 4.42·10−17 5.38·10−18 5.22·10−17 1.27·10−17 8.22·10−19 3.51·10−19 9.23·10−20 1.52·10−19 1.16·10−17 6.15·10−17 2.45·10−17 5.01·10−19 5.20·10−17 3.32·10−17 2.81·10−17 9.72·10−18 2.78·10−18 3.20·10−17 9.73·10−18 2.32·10−18 4.23·10−18 4.89·10−17 2.62·10−17 1.39·10−18 1.09·10−17 1.61·10−17 1.18·10−18 2.29·10−17 4.46·10−17 3.77·10−17 2.29·10−17
Effective dose [Sv/(Bq s m−2)] 1.61·10−17 4.26·10−19 4.00·10−19 6.73·10−20 2.64·10−19 2.69·10−17 4.52·10−17 1.21·10−17 1.50·10−17 7.60·10−19 3.07·10−17 2.74·10−17 1.55·10−22 2.68·10−17 2.37·10−19 7.37·10−18 4.23·10−18 3.70·10−17 4.25·10−18 3.60·10−17 1.03·10−17 6.18·10−19 2.14·10−19 3.62·10−21 5.59·10−20 8.37·10−18 4.02·10−17 2.09·10−17 3.31·10−19 4.23·10−17 1.43·10−17 1.90·10−17 8.29·10−19 2.19·10−18 2.64·10−17 6.72·10−18 1.74·10−18 3.07·10−18 3.93·10−17 1.45·10−17 1.12·10−18 9.09·10−19 1.31·10−17 8.98·10−19 1.28·10−17 3.37·10−17 2.41·10−17 1.78·10−17
Radionuclide La-142 Pm-142 Pr-142m Pr-142 Sm-142 Ce-143 La-143 Pm-143 Pr-143 Ce-144 Pm-144 Pr-144m Pr-144 Eu-145 Gd-145 Pm-145 Pr-145 Sm-145 Eu-146 Gd-146 Pm-146 Sm-146 Eu-147 Gd-147 Nd-147 Pm-147 Pr-147 Sm-147 Tb-147 Eu-148 Gd-148 Pm-148m Pm-148 Eu-149 Gd-149 Nd-149 Pm-149 Tb-149 Eu-150b Eu-150a Pm-150 Tb-150 Gd-151 Nd-151 Pm-151 Sm-151 Tb-151 Eu-152m Eu-152
[Ref. p. 6-42 Skin dose [Sv/(Bq s m−2)] 6.11·10−17 3.83·10−17 0.00 8.94·10−18 1.67·10−18 6.97·10−18 1.94·10−17 6.17·10−18 5.06·10−19 3.33·10−19 3.23·10−17 1.03·10−19 1.58·10−17 2.91·10−17 5.08·10−17 2.44·10−19 5.41·10−18 5.42·10−19 5.12·10−17 3.82·10−18 1.56·10−17 0.00 9.63·10−18 2.71·10−17 2.76·10−18 3.30·10−22 2.39·10−17 0.00 3.94·10−17 4.47·10−17 0.00 4.17·10−17 1.81·10−17 9.06·10−19 8.14·10−18 9.74·10−18 1.15·10−18 3.38·10−17 3.09·10−17 1.61·10−18 3.58·10−17 4.10·10−17 8.52·10−19 2.29·10−17 6.95·10−18 2.46·10−23 1.79·10−17 9.77·10−18 2.31·10−17
Effective dose [Sv/(Bq s m−2)] 4.59·10−17 1.53·10−17 0.00 1.12·10−18 1.23·10−18 4.50·10−18 1.92·10−18 5.04·10−18 1.28·10−20 2.61·10−19 2.67·10−17 5.18·10−20 8.17·10−19 2.44·10−17 3.80·10−17 1.21·10−19 3.12·10−19 2.85·10−19 4.27·10−17 2.93·10−18 1.28·10−17 0.00 7.90·10−18 2.25·10−17 2.08·10−18 1.96·10−22 1.44·10−17 0.00 2.71·10−17 3.71·10−17 0.00 3.45·10−17 9.94·10−18 6.59·10−19 6.59·10−18 6.34·10−18 2.02·10−19 2.72·10−17 2.54·10−17 7.76·10−19 2.45·10−17 2.87·10−17 6.08·10−19 1.55·10−17 5.30·10−18 3.62·10−24 1.46·10−17 4.95·10−18 1.93·10−17 Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Gd-152 Gd-153 Sm-153 Tb-153 Eu-154 Tb-154 Dy-155 Eu-155 Ho-155 Sm-155 Tb-155 Eu-156 Sm-156 Tb-156m Tb-156n Tb-156 Dy-157 Eu-157 Ho-157 Tb-157 Eu-158 Tb-158 Dy-159 Gd-159 Ho-159 Tb-160 Er-161 Ho-161 Tb-161 Ho-162m Ho-162 Tm-162 Yb-162 Ho-164m Ho-164 Dy-165 Er-165 Dy-166 Ho-166m Ho-166 Tm-166 Yb-166 Ho-167 Tm-167 Yb-167 Er-169 Lu-169 Yb-169 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 0.00 1.33·10−18 9.29·10−19 4.03·10−18 2.58·10−17 4.51·10−17 1.14·10−17 9.06·10−19 9.53·10−18 4.93·10−18 2.16·10−18 2.93·10−17 2.28·10−18 2.51·10−19 3.96·10−20 3.64·10−17 7.05·10−18 6.10·10−18 9.49·10−18 2.23·10−20 3.11·10−17 1.59·10−17 4.09·10−19 1.38·10−18 6.51·10−18 2.29·10−17 1.83·10−17 5.95·10−19 3.53·10−19 1.11·10−17 2.98·10−18 3.86·10−17 2.18·10−18 4.32·10−19 5.05·10−19 2.26·10−18 3.62·10−19 4.94·10−19 3.59·10−17 5.66·10−18 3.69·10−17 9.43·10−19 7.69·10−18 2.44·10−18 4.11·10−18 8.53·10−22 2.03·10−17 5.02·10−18
Effective dose [Sv/(Bq s m−2)] 0.00 9.54·10−19 6.08·10−19 3.20·10−18 2.10·10−17 3.88·10−17 9.39·10−18 7.06·10−19 6.25·10−18 1.48·10−18 1.65·10−18 2.25·10−17 1.80·10−18 1.55·10−19 2.63·10−20 3.05·10−17 5.68·10−18 4.02·10−18 7.58·10−18 1.22·10−20 1.81·10−17 1.32·10−17 2.35·10−19 7.53·10−19 5.19·10−18 1.91·10−17 1.52·10−17 3.67·10−19 2.25·10−19 9.19·10−18 2.34·10−18 2.96·10−17 1.69·10−18 2.61·10−19 1.92·10−19 4.16·10−19 2.20·10−19 3.49·10−19 2.98·10−17 4.97·10−19 3.11·10−17 6.35·10−19 6.14·10−18 1.89·10−18 3.18·10−18 5.43·10−22 1.70·10−17 3.89·10−18
Radionuclide Hf-170 Lu-170 Tm-170 Er-171 Lu-171 Tm-171 Er-172 Hf-172 Lu-172 Ta-172 Tm-172 Hf-173 Lu-173 Ta-173 Tm-173 Lu-174m Lu-174 Ta-174 Hf-175 Ta-175 Tm-175 Yb-175 Lu-176m Lu-176 Ta-176 W-176 Hf-177m Lu-177m Lu-177 Re-177 Ta-177 W-177 Yb-177 Hf-178m Lu-178m Lu-178 Re-178 Ta-178b Ta-178a W-178 Yb-178 Hf-179m Lu-179 Ta-179 W-179 Hf-180m Os-180 Re-180
6-29 Skin dose [Sv/(Bq s m−2)] 1.06·10−17 4.79·10−17 6.24·10−19 8.55·10−18 1.36·10−17 7.02·10−21 1.07·10−17 1.40·10−18 3.78·10−17 3.56·10−17 1.29·10−17 7.59·10−18 1.94·10−18 1.45·10−17 8.59·10−18 7.57·10−19 2.06·10−18 1.54·10−17 7.25·10−18 1.80·10−17 2.39·10−17 8.07·10−19 1.90·10−18 9.98·10−18 4.20·10−17 2.47·10−18 4.56·10−17 1.99·10−17 6.68·10−19 1.47·10−17 8.91·10−19 1.76·10−17 5.60·10−18 4.85·10−17 2.42·10−17 9.38·10−18 3.01·10−17 2.03·10−17 1.70·10−18 1.52·10−19 7.64·10−19 1.81·10−17 2.64·10−18 3.56·10−19 6.13·10−19 2.05·10−17 8.23·10−19 2.44·10−17
Effective dose [Sv/(Bq s m−2)] 8.57·10−18 4.10·10−17 6.50·10−20 6.21·10−18 1.12·10−17 4.75·10−21 8.79·10−18 1.00·10−18 3.16·10−17 2.60·10−17 8.07·10−18 6.13·10−18 1.46·10−18 9.38·10−18 6.67·10−18 5.42·10−19 1.65·10−18 1.02·10−17 5.85·10−18 1.51·10−17 1.81·10−17 6.57·10−19 1.78·10−19 8.11·10−18 3.56·10−17 1.89·10−18 3.71·10−17 1.61·10−17 5.39·10−19 9.90·10−18 6.60·10−19 1.44·10−17 3.14·10−18 3.97·10−17 1.84·10−17 2.42·10−18 2.01·10−17 1.64·10−17 1.35·10−18 1.06·10−19 6.00·10−19 1.47·10−17 5.48·10−19 2.45·10−19 4.06·10−19 1.67·10−17 6.04·10−19 1.98·10−17
6-30 Radionuclide Ta-180m Ta-180 Hf-181 Os-181 Re-181 W-181 Hf-182m Hf-182 Ir-182 Os-182 Re-182b Re-182a Ta-182m Ta-182 Hf-183 Ta-183 Hf-184 Ir-184 Re-184m Re-184 Ta-184 Ir-185 Os-185 Ta-185 W-185 Ir-186a Ir-186b Pt-186 Re-186m Re-186 Ta-186 Ir-187 Re-187 W-187 Ir-188 Pt-188 Re-188m Re-188 W-188 Ir-189 Os-189m Pt-189 Re-189 Ir-190n Ir-190m Ir-190 Os-190m Ir-191m
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 5.75·10−19 1.11·10−17 1.13·10−17 2.43·10−17 1.53·10−17 4.61·10−19 1.89·10−17 4.91·10−18 3.97·10−17 8.53·10−18 3.70·10−17 2.30·10−17 4.49·10−18 2.57·10−17 1.70·10−17 5.44·10−18 5.34·10−18 4.03·10−17 7.48·10−18 1.78·10−17 3.44·10−17 1.12·10−17 1.46·10−17 8.03·10−18 2.34·10−21 3.28·10−17 2.08·10−17 1.50·10−17 1.69·10−19 1.12·10−18 4.13·10−17 6.96·10−18 0.00 1.03·10−17 3.07·10−17 3.57·10−18 1.04·10−18 7.88·10−18 3.88·10−20 1.19·10−18 5.34·10−22 6.00·10−18 1.89·10−18 3.20·10−17 5.86·10−22 2.95·10−17 3.32·10−17 1.11·10−18
Effective dose [Sv/(Bq s m−2)] 4.00·10−19 8.97·10−18 9.29·10−18 2.02·10−17 1.26·10−17 3.24·10−19 1.55·10−17 4.00·10−18 2.29·10−17 6.93·10−18 3.09·10−17 1.93·10−17 3.59·10−18 2.16·10−17 1.27·10−17 4.35·10−18 3.90·10−18 3.20·10−17 6.13·10−18 1.48·10−17 2.75·10−17 9.44·10−18 1.20·10−17 2.89·10−18 1.66·10−21 2.74·10−17 1.61·10−17 1.24·10−17 1.15·10−19 2.92·10−19 2.67·10−17 5.69·10−18 0.00 8.05·10−18 2.62·10−17 2.85·10−18 7.83·10−19 1.05·10−18 3.13·10−20 9.09·10−19 9.45·10−24 4.87·10−18 1.10·10−18 2.63·10−17 1.10·10−23 2.42·10−17 2.73·10−17 8.75·10−19
Radionuclide Os-191m Os-191 Pt-191 Ir-192m Ir-192 Au-193 Hg-193m Hg-193 Os-193 Pt-193m Pt-193 Au-194 Hg-194 Ir-194m Ir-194 Os-194 Tl-194m Tl-194 Au-195m Au-195 Hg-195m Hg-195 Ir-195m Ir-195 Pb-195m Pt-195m Tl-195 Hg-197m Hg-197 Pt-197m Pt-197 Tl-197 Au-198m Au-198 Pb-198 Tl-198m Tl-198 Au-199 Hg-199m Pb-199 Pt-199 Tl-199 Au-200m Au-200 Bi-200 Pb-200 Pt-200 Tl-200
[Ref. p. 6-42 Skin dose [Sv/(Bq s m−2)] 9.26·10−20 1.18·10−18 5.46·10−18 3.16·10−18 1.72·10−17 2.67·10−18 2.09·10−17 3.47·10−18 2.30·10−18 1.42·10−19 1.66·10−21 2.11·10−17 2.32·10−21 4.90·10−17 9.47·10−18 9.62·10−21 5.06·10−17 1.59·10−17 4.03·10−18 1.11·10−18 4.05·10−18 3.69·10−18 8.42·10−18 1.49·10−18 3.37·10−17 9.94·10−19 2.52·10−17 1.58·10−18 9.19·10−19 1.41·10−18 4.18·10−19 7.93·10−18 1.10·10−17 9.07·10−18 8.60·10−18 2.46·10−17 3.96·10−17 1.67·10−18 3.40·10−18 2.94·10−17 7.09·10−18 4.65·10−18 4.34·10−17 1.23·10−17 5.03·10−17 3.64·10−18 1.02·10−18 2.64·10−17
Effective dose [Sv/(Bq s m−2)] 6.74·10−20 9.21·10−19 4.38·10−18 2.58·10−18 1.41·10−17 2.11·10−18 1.73·10−17 2.72·10−18 1.17·10−18 1.06·10−19 3.38·10−23 1.77·10−17 5.71·10−23 4.03·10−17 1.68·10−18 4.92·10−21 3.96·10−17 1.30·10−17 3.27·10−18 8.45·10−19 3.29·10−18 3.00·10−18 6.67·10−18 6.48·10−19 2.70·10−17 7.59·10−19 2.11·10−17 1.27·10−18 7.00·10−19 1.11·10−18 3.01·10−19 6.49·10−18 8.91·10−18 7.01·10−18 7.00·10−18 2.02·10−17 3.36·10−17 1.35·10−18 2.74·10−18 2.47·10−17 3.50·10−18 3.76·10−18 3.58·10−17 4.77·10−18 4.07·10−17 2.93·10−18 7.90·10−19 2.21·10−17
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Au-201 Bi-201 Pb-201 Tl-201 Bi-202 Pb-202m Pb-202 Tl-202 Bi-203 Hg-203 Pb-203 Po-203 Tl-204 Bi-205 Pb-205 Po-205 Bi-206 Tl-206 At-207 Bi-207 Po-207 Tl-207 Tl-208 Pb-209 Tl-209 Bi-210m Bi-210 Pb-210 Po-210 At-211 Bi-211 Pb-211 Po-211 Bi-212 Pb-212 Po-212 Bi-213 Po-213 Bi-214 Pb-214 Po-214 At-215 Po-215 At-216 Po-216 At-217 At-218 Po-218 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.62·10−18 2.76·10−17 1.53·10−17 1.37·10−18 5.56·10−17 4.23·10−17 2.07·10−21 9.40·10−18 4.74·10−17 4.92·10−18 6.04·10−18 3.37·10−17 1.92·10−19 3.34·10−17 2.24·10−21 3.18·10−17 6.65·10−17 3.03·10−18 2.65·10−17 3.21·10−17 2.70·10−17 2.48·10−18 6.74·10−17 3.80·10−20 4.51·10−17 5.32·10−18 1.20·10−18 2.22·10−20 1.77·10−22 5.73·10−19 9.71·10−19 3.08·10−18 1.63·10−19 7.85·10−18 2.88·10−18 0.00 4.53·10−18 0.00 3.53·10−17 5.23·10−18 1.73·10−21 4.07·10−21 3.71·10−21 2.25·10−20 3.51·10−22 6.40·10−21 4.28·10−20 1.89·10−22
Effective dose [Sv/(Bq s m−2)] 9.30·10−19 2.25·10−17 1.26·10−17 1.07·10−18 4.61·10−17 3.52·10−17 3.93·10−23 7.67·10−18 4.01·10−17 4.01·10−18 4.89·10−18 2.77·10−17 1.77·10−20 2.83·10−17 4.39·10−23 2.66·10−17 5.57·10−17 5.59·10−20 2.22·10−17 2.61·10−17 2.25·10−17 8.28·10−20 5.58·10−17 3.19·10−21 3.43·10−17 4.34·10−18 2.43·10−20 1.05·10−20 1.47·10−22 4.51·10−19 7.93·10−19 9.07·10−19 1.35·10−19 3.23·10−18 2.34·10−18 0.00 2.32·10−18 0.00 2.57·10−17 4.18·10−18 1.44·10−21 3.33·10−21 3.05·10−21 1.78·10−20 2.92·10−22 5.26·10−21 2.50·10−20 1.57·10−22
Radionuclide Rn-218 Fr-219 Rn-219 Fr-220 Rn-220 Fr-221 Fr-222 Ra-222 Rn-222 Ac-223 Fr-223 Ra-223 Ac-224 Ra-224 Ac-225 Ra-225 Ac-226 Ra-226 Th-226 Ac-227 Pa-227 Ra-227 Th-227 Ac-228 Pa-228 Ra-228 Th-228 Th-229 Pa-230 Th-230 U-230 Pa-231 Th-231 U-231 Np-232 Pa-232 Th-232 U-232 Np-233 Pa-233 U-233 Np-234 Pa-234m Pa-234 Pu-234 Th-234 U-234 Np-235
6-31 Skin dose [Sv/(Bq s m−2)] 1.59·10−20 7.25·10−20 1.17·10−18 2.00·10−19 8.12·10−21 6.22·10−19 6.16·10−18 1.93·10−19 8.40·10−21 8.91·10−20 1.69·10−18 2.50·10−18 3.64·10−18 2.03·10−19 2.86·10−19 9.80·10−20 3.00·10−18 1.32·10−19 1.46·10−19 2.45·10−21 3.16·10−19 4.34·10−18 2.09·10−18 2.15·10−17 2.28·10−17 0.00 3.81·10−20 1.49·10−18 1.30·10−17 8.32·10−21 2.38·10−20 7.69·10−19 2.15·10−19 1.13·10−18 2.44·10−17 1.92·10−17 5.04·10−21 8.66·10−21 1.49·10−18 4.02·10−18 8.26·10−21 2.85·10−17 8.02·10−18 3.93·10−17 1.09·10−18 1.24·10−19 5.61·10−21 3.02·10−20
Effective dose [Sv/(Bq s m−2)] 1.31·10−20 5.94·10−20 9.56·10−19 1.61·10−19 6.68·10−21 5.06·10−19 1.08·10−19 1.58·10−19 6.91·10−21 7.01·10−20 6.90·10−19 2.03·10−18 2.95·10−18 1.66·10−19 2.28·10−19 4.58·10−20 2.07·10−18 1.07·10−19 1.17·10−19 1.80·10−21 2.47·10−19 2.62·10−18 1.69·10−18 1.65·10−17 1.90·10−17 0.00 2.85·10−20 1.18·10−18 1.09·10−17 4.62·10−21 1.56·10−20 6.02·10−19 1.41·10−19 8.79·10−19 2.02·10−17 1.60·10−17 2.05·10−21 3.38·10−21 1.20·10−18 3.27·10−18 4.81·10−21 2.41·10−17 3.33·10−19 3.24·10−17 8.73·10−19 9.49·10−20 1.55·10−21 1.23·10−20
6-32
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 3.27·10−21 8.76·10−19 4.22·10−18 1.55·10−18 8.92·10−18 2.28·10−17 1.50·10−17 1.92·10−17 4.40·10−21 4.38·10−21 2.46·10−18 1.88·10−18 6.61·10−18 4.11·10−21 3.34·10−21 4.80·10−23 6.95·10−18 7.31·10−19 1.86·10−17 3.91·10−21 0.00 7.78·10−18 2.29·10−18 1.64·10−18 4.03·10−21 3.93·10−21 5.66·10−22 9.07·10−21 1.41·10−18 1.27·10−23 1.01·10−17 1.01·10−19 4.63·10−21 6.14·10−20 1.89·10−18 2.08·10−18 2.83·10−20
Effective dose [Sv/(Bq s m−2)] 2.92·10−22 4.94·10−19 3.42·10−18 1.25·10−18 7.02·10−18 1.74·10−17 1.15·10−17 1.59·10−17 7.06·10−22 4.43·10−22 1.96·10−18 1.52·10−18 5.41·10−18 4.69·10−22 3.35·10−22 1.86·10−23 5.68·10−18 3.39·10−19 1.52·10−17 4.45·10−22 0.00 6.46·10−18 1.84·10−18 1.32·10−18 6.60·10−22 5.05·10−22 3.04·10−22 5.56·10−21 1.14·10−18 1.45·10−24 8.04·10−18 4.54·10−20 9.12·10−22 2.36·10−20 1.52·10−18 1.69·10−18 9.02·10−21
Table 6.11. Skin dose and effective dose coefficients for air submersion; [93Eck1]. Effective Dose Radionuclide Skin Dose Radionuclide Skin Dose [Sv/(Bq s m−2)] [Sv/(Bq s m−2)] [Sv/(Bq s m−2)] He-3 0.00 0.00 C-14 2.43·10−16 −15 −15 Be-7 2.74·10 2.19·10 O-15 1.04·10−13 −14 −16 Be-10 1.29·10 1.38·10 F-18 6.94·10−14 −14 −14 C-11 7.91·10 Ne-19 4.56·10 1.21·10−13 −14 −14 N-13 8.68·10 Na-22 4.57·10 1.33·10−13
Effective Dose [Sv/(Bq s m−2)] 2.60·10−18 4.59·10−14 4.56·10−14 4.62·10−14 1.02·10−13
Radionuclide Pu-235 U-235 Np-236a Np-236b Pu-236 U-236 Am-237 Np-237 Pu-237 U-237 Am-238 Cm-238 Np-238 Pu-238 U-238 Am-239 Np-239 Pu-239 U-239 Am-240 Cm-240 Np-240m Np-240 Pu-240 U-240 Am-241 Cm-241 Pu-241 Am-242m Am-242 Cm-242 Pu-242 Am-243 Cm-243 Pu-243 Am-244m Am-244 Cf-244 Cm-244
Skin dose [Sv/(Bq s m−2)] 1.52·10−18 3.01·10−18 2.12·10−18 8.38·10−19 6.04·10−21 4.46·10−21 7.21·10−18 4.03·10−19 7.74·10−19 2.37·10−18 1.79·10−17 1.26·10−18 1.19·10−17 5.03·10−21 3.54·10−21 4.20·10−18 3.17·10−18 2.87·10−21 2.10·10−18 2.07·10−17 5.87·10−21 1.16·10−17 2.69·10−17 4.83·10−21 2.64·10−20 2.89·10−19 9.94·10−18 2.95·10−23 2.05·10−20 2.69·10−19 5.42·10−21 4.03·10−21 7.59·10−19 2.44·10−18 3.91·10−19 2.61·10−18 1.65·10−17 6.02·10−21 4.89·10−21
Effective dose [Sv/(Bq s m−2)] 1.22·10−18 2.45·10−18 1.69·10−18 6.72·10−19 8.63·10−22 8.36·10−22 5.88·10−18 2.96·10−19 6.12·10−19 1.88·10−18 1.49·10−17 1.01·10−18 9.41·10−18 5.78·10−22 4.20·10−22 3.39·10−18 2.57·10−18 1.01·10−21 6.37·10−19 1.72·10−17 6.01·10−22 5.76·10−18 2.21·10−17 5.66·10−22 5.79·10−21 1.85·10−19 8.12·10−18 2.20·10−23 6.37·10−21 1.88·10−19 6.38·10−22 4.94·10−22 5.78·10−19 1.98·10−18 3.01·10−19 4.80·10−20 1.35·10−17 6.80·10−22 4.79·10−22
[Ref. p. 6-42
Radionuclide Pu-244 Am-245 Bk-245 Cm-245 Pu-245 Am-246m Am-246 Bk-246 Cf-246 Cm-246 Pu-246 Bk-247 Cm-247 Cf-248 Cm-248 Bk-249 Cf-249 Cm-249 Bk-250 Cf-250 Cm-250 Es-250 Cf-251 Es-251 Cf-252 Fm-252 Cf-253 Es-253 Fm-253 Cf-254 Es-254m Es-254 Fm-254 Fm-255 Fm-257 Md-257 Md-258
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Na-24 Al-26 Al-28 Mg-28 P-30 Si-31 P-32 Si-32 P-33 S-35 Cl-36 Ar-37 Cl-38 K-38 Ar-39 Cl-39 K-40 Ar-41 Ca-41 K-42 K-43 Sc-43 K-44 Sc-44m Sc-44 Ti-44 Ca-45 K-45 Ti-45 Sc-46 Ca-47 Sc-47 V-47 Cr-48 Sc-48 V-48 Ca-49 Cr-49 Sc-49 V-49 Cr-51 Mn-51 Fe-52 Mn-52m Mn-52 Mn-53 Mn-54 Co-55 Fe-55 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.75·10−13 1.81·10−13 1.88·10−13 8.33·10−14 1.56·10−13 3.78·10−14 4.49·10−14 8.27·10−16 1.38·10−15 2.92·10−16 1.47·10−14 0.00 1.94·10−13 2.66·10−13 1.07·10−14 1.36·10−13 4.20·10−14 1.01·10−13 0.00 1.15·10−13 7.11·10−14 7.91·10−14 2.35·10−13 1.72·10−14 1.58·10−13 6.79·10−15 1.46·10−15 1.74·10−13 7.07·10−14 1.17·10−13 8.02·10−14 1.28·10−14 1.08·10−13 2.40·10−14 2.01·10−13 1.72·10−13 2.46·10−13 9.65·10−14 5.43·10−14 0.00 1.75·10−15 1.18·10−13 5.17·10−14 2.13·10−13 1.99·10−13 0.00 4.67·10−14 1.39·10−13 0.00
Effective Dose [Sv/(Bq s m−2)] 2.08·10−13 1.28·10−13 8.87·10−14 6.38·10−14 4.68·10−14 4.83·10−16 5.36·10−16 8.68·10−18 1.45·10−17 3.11·10−18 1.66·10−16 0.00 7.58·10−14 1.56·10−13 1.15·10−16 6.90·10−14 7.92·10−15 6.14·10−14 0.00 1.48·10−14 4.35·10−14 4.88·10−14 1.14·10−13 1.24·10−14 9.87·10−14 4.70·10−15 1.53·10−17 9.20·10−14 3.89·10−14 9.36·10−14 5.06·10−14 4.67·10−15 4.49·10−14 1.87·10−14 1.57·10−13 1.36·10−13 1.66·10−13 4.68·10−14 7.16·10−16 0.00 1.38·10−15 4.51·10−14 3.27·10−14 1.13·10−13 1.62·10−13 0.00 3.83·10−14 9.16·10−14 0.00
Radionuclide Co-56 Mn-56 Ni-56 Co-57 Ni-57 Co-58m Co-58 Fe-59 Ni-59 Co-60m Co-60 Cu-60 Fe-60 Co-61 Cu-61 Co-62m Cu-62 Zn-62 Ni-63 Zn-63 Cu-64 Ga-65 Ni-65 Zn-65 Cu-66 Ga-66 Ge-66 Ni-66 Cu-67 Ga-67 Ge-67 Ga-68 Ge-68 As-69 Ge-69 Zn-69m Zn-69 As-70 Ga-70 Se-70 As-71 Ge-71 Zn-71m As-72 Ga-72 Zn-72 As-73 Ga-73
6-33 Skin Dose [Sv/(Bq s m−2)] 2.13·10−13 1.51·10−13 9.61·10−14 6.63·10−15 1.17·10−13 3.05·10−19 5.58·10−14 7.13·10−14 0.00 3.46·10−16 1.45·10−13 2.82·10−13 1.64·10−16 3.24·10−14 6.50·10−14 2.25·10−13 1.44·10−13 2.52·10−14 0.00 1.23·10−13 1.64·10−14 1.19·10−13 7.18·10−14 3.29·10−14 7.69·10−14 2.11·10−13 4.26·10−14 1.01·10−15 1.18·10−14 8.50·10−15 1.68·10−13 1.01·10−13 6.62·10−18 1.43·10−13 5.96·10−14 2.44·10−14 1.81·10−14 2.89·10−13 4.17·10−14 8.36·10−14 3.78·10−14 6.71·10−18 1.21·10−13 1.70·10−13 1.86·10−13 1.00·10−14 2.78·10−16 4.37·10−14
Effective Dose [Sv/(Bq s m−2)] 1.73·10−13 8.16·10−14 7.82·10−14 4.97·10−15 9.12·10−14 6.06·10−20 4.44·10−14 5.62·10−14 0.00 2.00·10−16 1.19·10−13 1.87·10−13 1.79·10−18 3.74·10−15 3.72·10−14 1.30·10−13 4.60·10−14 1.92·10−14 0.00 5.00·10−14 8.50·10−15 5.28·10−14 2.67·10−14 2.72·10−14 4.89·10−15 1.23·10−13 3.00·10−14 1.06·10−17 4.90·10−15 6.49·10−15 6.45·10−14 4.29·10−14 1.01·10−19 4.61·10−14 3.99·10−14 1.84·10−14 1.99·10−16 1.92·10−13 8.40·10−16 4.40·10−14 2.53·10−14 1.02·10−19 6.99·10−14 8.26·10−14 1.31·10−13 6.17·10−15 1.55·10−16 1.39·10−14
6-34 Radionuclide Se-73m Se-73 As-74 Br-74m Br-74 Kr-74 Br-75 Ge-75 Se-75 As-76 Br-76 Kr-76 As-77 Br-77 Ge-77 Kr-77 Se-77m As-78 Ge-78 Kr-79 Rb-79 Se-79 Br-80m Br-80 Rb-80 Sr-80 Kr-81m Kr-81 Rb-81m Rb-81 Se-81m Se-81 Sr-81 Br-82 Rb-82m Rb-82 Sr-82 Br-83 Kr-83m Rb-83 Se-83 Sr-83 Br-84 Rb-84 Kr-85m Kr-85 Sr-85m Sr-85
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.39·10−14 8.31·10−14 5.80·10−14 3.31·10−13 3.40·10−13 1.16·10−13 1.01·10−13 2.71·10−14 2.16·10−14 9.61·10−14 1.97·10−13 2.37·10−14 1.20·10−14 1.77·10−14 1.02·10−13 9.74·10−14 6.99·10−15 1.65·10−13 2.75·10−14 1.50·10−14 1.28·10−13 3.71·10−16 7.13·10−16 2.02·10−14 2.11·10−13 1.44·10−16 9.42·10−15 4.04·10−16 4.01·10−16 4.46·10−14 1.40·10−15 3.94·10−14 1.44·10−13 1.54·10−13 1.68·10−13 1.58·10−13 1.42·10−16 1.85·10−14 3.56·10−17 2.77·10−14 1.69·10−13 5.20·10−14 1.88·10−13 5.71·10−14 2.24·10−14 1.32·10−14 1.23·10−14 2.83·10−14
Effective Dose [Sv/(Bq s m−2)] 1.09·10−14 4.78·10−14 3.40·10−14 1.96·10−13 2.26·10−13 5.20·10−14 5.43·10−14 1.78·10−15 1.68·10−14 2.06·10−14 1.26·10−13 1.86·10−14 5.09·10−16 1.40·10−14 4.98·10−14 4.51·10−14 3.63·10−15 6.03·10−14 1.23·10−14 1.12·10−14 6.08·10−14 3.94·10−18 2.37·10−16 3.73·10−15 5.77·10−14 5.00·10−18 5.56·10−15 2.44·10−16 1.63·10−16 2.73·10−14 5.48·10−16 8.69·10−16 6.24·10−14 1.21·10−13 1.34·10−13 5.01·10−14 4.92·10−18 5.34·10−16 1.20·10−18 2.21·10−14 1.14·10−13 3.60·10−14 9.02·10−14 4.18·10−14 6.87·10−15 2.40·10−16 9.48·10−15 2.24·10−14
Radionuclide Rb-86 Y-86m Y-86 Zr-86 Kr-87 Rb-87 Sr-87m Y-87 Kr-88 Nb-88 Rb-88 Y-88 Zr-88 Nb-89b Nb-89a Rb-89 Sr-89 Zr-89 Mo-90 Nb-90 Sr-90 Y-90m Y-90 Sr-91 Y-91m Y-91 Sr-92 Y-92 Mo-93m Mo-93 Nb-93m Tc-93m Tc-93 Y-93 Zr-93 Nb-94 Ru-94 Tc-94m Tc-94 Y-94 Nb-95m Nb-95 Tc-95m Tc-95 Y-95 Zr-95 Nb-96 Tc-96m
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 4.85·10−14 1.28·10−14 2.17·10−13 1.56·10−14 1.37·10−13 3.15·10−15 2.15·10−14 2.51·10−14 1.35·10−13 3.12·10−13 1.83·10−13 1.54·10−13 2.26·10−14 1.56·10−13 1.63·10−13 1.87·10−13 3.69·10−14 7.07·10−14 5.52·10−14 2.66·10−13 9.20·10−15 3.75·10−14 6.24·10−14 8.14·10−14 3.11·10−14 3.85·10−14 8.56·10−14 1.14·10−13 1.32·10−13 2.43·10−16 4.28·10−17 4.62·10−14 8.30·10−14 8.50·10−14 0.00 9.52·10−14 2.95·10−14 1.55·10−13 1.51·10−13 1.80·10−13 1.12·10−14 4.30·10−14 3.76·10−14 4.42·10−14 1.59·10−13 4.50·10−14 1.52·10−13 2.68·10−15
Effective Dose [Sv/(Bq s m−2)] 4.94·10−15 9.59·10−15 1.69·10−13 1.17·10−14 3.97·10−14 3.30·10−17 1.41·10−14 1.99·10−14 9.71·10−14 1.89·10−13 3.33·10−14 1.30·10−13 1.73·10−14 6.62·10−14 8.65·10−14 1.01·10−13 4.37·10−16 5.31·10−14 3.64·10−14 2.05·10−13 9.83·10−17 2.77·10−14 7.92·10−16 3.27·10−14 2.37·10−14 6.22·10−16 6.41·10−14 1.32·10−14 1.06·10−13 1.73·10−17 3.05·10−18 3.53·10−14 6.96·10−14 5.28·10−15 0.00 7.20·10−14 2.36·10−14 8.64·10−14 1.22·10−13 5.39·10−14 2.74·10−15 3.49·10−14 2.99·10−14 3.58·10−14 4.66·10−14 3.36·10−14 1.14·10−13 2.09·10−15
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Tc-96 Nb-97m Nb-97 Ru-97 Tc-97m Tc-97 Zr-97 Nb-98 Tc-98 Mo-99 Rh-99m Rh-99 Tc-99m Tc-99 Pd-100 Rh-100 Mo-101 Pd-101 Rh-101m Rh-101 Tc-101 Ag-102 Rh-102m Rh-102 Ag-103 Pd-103 Rh-103m Ru-103 Ag-104m Ag-104 Cd-104 Tc-104 Ag-105 Rh-105 Ru-105 Ag-106m Ag-106 Rh-106m Rh-106 Ru-106 Cd-107 Pd-107 Rh-107 Ag-108m Ag-108 Ag-109m Cd-109 In-109 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.40·10−13 4.16·10−14 6.51·10−14 1.32·10−14 5.55·10−16 2.71·10−16 5.55·10−14 1.96·10−13 8.53·10−14 3.14·10−14 3.94·10−14 3.42·10−14 7.14·10−15 2.74·10−15 6.11·10−15 1.63·10−13 1.14·10−13 1.94·10−14 1.71·10−14 1.49·10−14 4.77·10−14 2.45·10−13 3.68·10−14 1.19·10−13 5.84·10−14 3.90·10−16 4.49·10−17 2.77·10−14 1.00·10−13 1.56·10−13 1.38·10−14 2.25·10−13 2.90·10−14 1.07·10−14 6.73·10−14 1.58·10−13 7.27·10−14 1.81·10−13 1.09·10−13 0.00 1.50·10−15 0.00 4.42·10−14 9.05·10−14 4.00·10−14 5.59·10−16 9.95·10−16 3.91·10−14
Effective Dose [Sv/(Bq s m−2)] 1.14·10−13 3.31·10−14 2.99·10−14 9.91·10−15 3.72·10−17 2.26·10−17 8.90·10−15 1.14·10−13 6.41·10−14 6.99·10−15 3.06·10−14 2.63·10−14 5.25·10−15 2.87·10−17 3.98·10−15 1.33·10−13 6.48·10−14 1.42·10−14 1.29·10−14 1.09·10−14 1.50·10−14 1.57·10−13 2.15·10−14 9.68·10−14 3.43·10−14 5.32·10−17 6.02·10−18 2.08·10−14 5.48·10−14 1.23·10−13 1.04·10−14 9.61·10−14 2.26·10−14 3.47·10−15 3.56·10−14 1.29·10−13 3.18·10−14 1.35·10−13 1.06·10−14 0.00 5.11·10−16 0.00 1.41·10−14 7.24·10−14 1.25·10−15 1.59·10−16 2.28·10−16 2.98·10−14
Radionuclide Pd-109 Ag-110m Ag-110 In-110b In-110a Sn-110 Ag-111 In-111 Sn-111 Ag-112 In-112 Cd-113m Cd-113 In-113m Sn-113 In-114m In-114 Ag-115 Cd-115m Cd-115 In-115m In-115 Sb-115 In-116m Sb-116m Sb-116 Te-116 Cd-117m Cd-117 In-117m In-117 Sb-117 Sn-117m Sb-118m In-119m In-119 Sb-119 Sn-119m I-120m I-120 Sb-120b Sb-120a Xe-120 I-121 Sn-121m Sn-121 Te-121m Te-121
6-35 Skin Dose [Sv/(Bq s m−2)] 2.15·10−14 1.57·10−13 8.22·10−14 1.71·10−13 1.29·10−13 1.66·10−14 2.19·10−14 2.29·10−14 4.22·10−14 1.33·10−13 2.88·10−14 8.48·10−15 2.41·10−15 2.18·10−14 8.20·10−16 1.05·10−14 2.95·10−15 1.11·10−13 3.99·10−14 2.97·10−14 1.81·10−14 6.18·10−15 6.52·10−14 1.58·10−13 1.82·10−13 1.50·10−13 3.37·10−15 1.29·10−13 8.79·10−14 3.17·10−14 5.16·10−14 1.03·10−14 1.25·10−14 1.46·10−13 7.11·10−14 8.20·10−14 7.09·10−16 3.42·10−16 3.86·10−13 2.55·10−13 1.39·10−13 4.46·10−14 2.40·10−14 2.72·10−14 1.07·10−15 3.71·10−15 1.23·10−14 3.18·10−14
Effective Dose [Sv/(Bq s m−2)] 4.20·10−16 1.27·10−13 2.46·10−15 1.39·10−13 7.15·10−14 1.25·10−14 1.38·10−15 1.68·10−14 2.30·10−14 3.23·10−14 1.19·10−14 9.06·10−17 2.53·10−17 1.12·10−14 3.15·10−16 3.89·10−15 1.59·10−16 3.46·10−14 1.48·10−15 1.05·10−14 6.86·10−15 6.55·10−17 4.02·10−14 1.18·10−13 1.45·10−13 1.02·10−13 1.98·10−15 9.89·10−14 5.14·10−14 4.07·10−15 3.06·10−14 7.15·10−15 6.11·10−15 1.19·10−13 1.26·10−15 3.53·10−14 1.50·10−16 7.04·10−17 2.49·10−13 1.31·10−13 1.14·10−13 2.00·10−14 1.79·10−14 1.78·10−14 5.24·10−17 3.90·10−17 8.99·10−15 2.50·10−14
6-36 Radionuclide Xe-121 I-122 Sb-122 Xe-122 I-123 Sn-123m Sn-123 Te-123m Te-123 Xe-123 I-124 Sb-124n Sb-124m Sb-124 Cs-125 I-125 Sb-125 Sn-125 Te-125m Xe-125 Ba-126 Cs-126 I-126 Sb-126m Sb-126 Sn-126 Cs-127 Sb-127 Sn-127 Te-127m Te-127 Xe-127 Ba-128 Cs-128 I-128 Sb-128b Sb-128a Sn-128 Cs-129 I-129 Sb-129 Te-129m Te-129 Xe-129m Cs-130 I-130 Sb-130 Ba-131m
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.40·10−13 1.25·10−13 6.03·10−14 3.36·10−15 9.40·10−15 3.58·10−14 3.28·10−14 8.48·10−15 6.32·10−16 4.52·10−14 7.39·10−14 2.33·10−18 2.46·10−14 1.26·10−13 5.97·10−14 1.39·10−15 2.65·10−14 7.13·10−14 1.94·10−15 1.50·10−14 9.26·10−15 1.62·10−13 3.37·10−14 1.24·10−13 1.73·10−13 6.65·10−15 2.38·10−14 5.58·10−14 1.41·10−13 8.49·10−16 1.14·10−14 1.57·10−14 3.85·10−15 1.07·10−13 5.38·10−14 1.99·10−13 1.73·10−13 4.50·10−14 1.52·10−14 1.10·10−15 1.05·10−13 1.49·10−14 3.57·10−14 8.29·10−15 5.48·10−14 1.36·10−13 2.29·10−13 3.94·10−15
Effective Dose [Sv/(Bq s m−2)] 8.62·10−14 4.31·10−14 2.02·10−14 2.19·10−15 6.49·10−15 6.14·10−15 6.98·10−16 5.81·10−15 1.51·10−16 2.82·10−14 5.04·10−14 4.67·10−19 1.58·10−14 8.62·10−14 3.01·10−14 3.73·10−16 1.87·10−14 1.54·10−14 3.35·10−16 1.08·10−14 6.41·10−15 4.96·10−14 2.01·10−14 7.01·10−14 1.28·10−13 1.84·10−15 1.78·10−14 3.12·10−14 9.03·10−14 1.12·10−16 3.34·10−16 1.12·10−14 2.54·10−15 4.06·10−14 4.33·10−15 1.41·10−13 9.08·10−14 2.77·10−14 1.13·10−14 2.81·10−16 6.71·10−14 1.56·10−15 2.86·10−15 9.14·10−16 2.30·10−14 9.67·10−14 1.50·10−13 2.64·10−15
Radionuclide Ba-131 Cs-131 I-131 La-131 Sb-131 Te-131m Te-131 Xe-131m Cs-132 I-132m I-132 La-132 Te-132 Ba-133m Ba-133 I-133 Te-133m Te-133 Xe-133m Xe-133 Ce-134 Cs-134m Cs-134 I-134 La-134 Te-134 Ba-135m Ce-135 Cs-135m Cs-135 I-135 La-135 Xe-135m Xe-135 Cs-136 Nd-136 Pr-136 Ba-137m Ce-137m Ce-137 Cs-137 La-137 Pr-137 Cs-138 La-138 Nd-138 Pr-138m Pr-138
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 2.55·10−14 7.84·10−16 2.98·10−14 4.87·10−14 1.40·10−13 8.85·10−14 6.89·10−14 4.82·10−15 3.92·10−14 2.22·10−14 1.58·10−13 1.49·10−13 1.39·10−14 1.36·10−14 2.19·10−14 5.83·10−14 1.74·10−13 1.06·10−13 1.04·10−14 4.97·10−15 9.60·10−16 2.88·10−15 9.45·10−14 1.87·10−13 8.88·10−14 6.35·10−14 1.30·10−14 1.10·10−13 9.10·10−14 9.06·10−16 1.11·10−13 1.49·10−15 2.97·10−14 3.12·10−14 1.25·10−13 1.71·10−14 1.69·10−13 3.73·10−14 1.20·10−14 1.45·10−15 8.63·10−15 8.68·10−16 4.01·10−14 2.17·10−13 7.09·10−14 1.92·10−15 1.52·10−13 1.25·10−13
Effective Dose [Sv/(Bq s m−2)] 1.92·10−14 2.38·10−16 1.69·10−14 2.91·10−14 8.84·10−14 6.55·10−14 1.92·10−14 3.49·10−16 3.11·10−14 1.42·10−14 1.05·10−13 9.41·10−14 9.32·10−15 2.44·10−15 1.62·10−14 2.76·10−14 1.07·10−13 4.34·10−14 1.28·10−15 1.33·10−15 3.52·10−16 7.95·10−16 7.06·10−14 1.22·10−13 3.15·10−14 3.94·10−14 2.16·10−15 7.93·10−14 7.25·10−14 9.50·10−18 7.54·10−14 7.75·10−16 1.90·10−14 1.10·10−14 9.94·10−14 1.15·10−14 9.72·10−14 2.69·10−14 1.83·10−15 7.30·10−16 9.28·10−17 3.00·10−16 2.20·10−14 1.15·10−13 5.84·10−14 1.07·10−15 1.13·10−13 3.72·10−14
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Xe-138 Ba-139 Ce-139 Nd-139m Nd-139 Pr-139 Ba-140 La-140 Ba-141 Ce-141 La-141 Nd-141m Nd-141 Pm-141 Sm-141m Sm-141 Ba-142 La-142 Pm-142 Pr-142m Pr-142 Sm-142 Ce-143 La-143 Pm-143 Pr-143 Ce-144 Pm-144 Pr-144m Pr-144 Eu-145 Gd-145 Pm-145 Pr-145 Sm-145 Eu-146 Gd-146 Pm-146 Sm-146 Eu-147 Gd-147 Nd-147 Pm-147 Pr-147 Sm-147 Tb-147 Eu-148 Gd-148 Pm-148m Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.07·10−13 6.16·10−14 8.94·10−15 9.17·10−14 3.50·10−14 8.75·10−15 2.52·10−14 1.66·10−13 1.07·10−13 1.02·10−14 6.58·10−14 4.67·10−14 4.24·10−15 8.42·10−14 1.39·10−13 1.27·10−13 8.37·10−14 2.16·10−13 1.44·10−13 0.00 5.67·10−14 6.44·10−15 3.96·10−14 9.64·10−14 1.72·10−14 1.76·10−14 2.93·10−15 8.71·10−14 5.08·10−16 8.43·10−14 8.33·10−14 1.66·10−13 1.22·10−15 4.44·10−14 2.64·10−15 1.43·10−13 1.33·10−14 4.64·10−14 0.00 2.77·10−14 7.67·10−14 1.95·10−14 8.11·10−16 9.75·10−14 0.00 1.27·10−13 1.22·10−13 0.00 1.18·10−13
Effective Dose [Sv/(Bq s m−2)] 5.48·10−14 2.54·10−15 5.97·10−15 7.12·10−14 1.77·10−14 4.75·10−15 8.07·10−15 1.11·10−13 3.92·10−14 3.10·10−15 2.88·10−15 3.45·10−14 2.59·10−15 3.39·10−14 9.07·10−14 6.44·10−14 4.84·10−14 1.37·10−13 4.01·10−14 0.00 3.50·10−15 3.43·10−15 1.21·10−14 5.78·10−15 1.35·10−14 1.94·10−16 7.63·10−16 6.95·10−14 2.20·10−16 2.65·10−15 6.78·10−14 1.09·10−13 5.49·10−16 1.12·10−15 1.26·10−15 1.15·10−13 8.61·10−15 3.34·10−14 0.00 2.14·10−14 5.98·10−14 5.72·10−15 8.67·10−18 3.90·10−14 0.00 7.29·10−14 9.83·10−14 0.00 9.01·10−14
Radionuclide Pm-148 Eu-149 Gd-149 Nd-149 Pm-149 Tb-149 Eu-150b Eu-150a Pm-150 Tb-150 Gd-151 Nd-151 Pm-151 Sm-151 Tb-151 Eu-152m Eu-152 Gd-152 Gd-153 Sm-153 Tb-153 Eu-154 Tb-154 Dy-155 Eu-155 Ho-155 Sm-155 Tb-155 Eu-156 Sm-156 Tb-156m Tb-156n Tb-156 Dy-157 Eu-157 Ho-157 Tb-157 Eu-158 Tb-158 Dy-159 Gd-159 Ho-159 Tb-160 Er-161 Ho-161 Tb-161 Ho-162m Ho-162
6-37 Skin Dose [Sv/(Bq s m−2)] 7.97·10−14 3.09·10−15 2.42·10−14 4.99·10−14 2.19·10−14 1.02·10−13 8.50·10−14 2.05·10−14 1.34·10−13 1.31·10−13 3.25·10−15 9.12·10−14 3.32·10−14 1.90·10−19 5.07·10−14 4.85·10−14 6.90·10−14 0.00 5.00·10−15 1.45·10−14 1.23·10−14 8.29·10−14 1.38·10−13 3.27·10−14 3.39·10−15 3.46·10−14 4.01·10−14 7.29·10−15 9.98·10−14 1.46·10−14 1.11·10−15 3.56·10−16 1.04·10−13 1.94·10−14 3.57·10−14 2.90·10−14 1.06·10−16 1.21·10−13 4.70·10−14 1.89·10−15 1.91·10−14 1.98·10−14 7.34·10−14 5.23·10−14 2.59·10−15 7.69·10−15 3.22·10−14 1.01·10−14
Effective Dose [Sv/(Bq s m−2)] 2.76·10−14 1.95·10−15 1.75·10−14 1.68·10−14 7.08·10−16 7.51·10−14 6.64·10−14 2.22·10−15 6.77·10−14 7.75·10−14 1.88·10−15 4.21·10−14 1.40·10−14 2.46·10−20 3.87·10−14 1.36·10−14 5.28·10−14 0.00 3.11·10−15 2.04·10−15 8.86·10−15 5.75·10−14 1.14·10−13 2.56·10−14 2.14·10−15 1.65·10−14 4.43·10−15 4.84·10−15 6.38·10−14 4.93·10−15 6.24·10−16 9.73·10−17 8.34·10−14 1.48·10−14 1.09·10−14 2.04·10−14 5.34·10−17 5.00·10−14 3.58·10−14 9.93·10−16 2.16·10−15 1.43·10−14 5.19·10−14 4.11·10−14 1.40·10−15 8.93·10−16 2.54·10−14 6.70·10−15
6-38 Radionuclide Tm-162 Yb-162 Ho-164m Ho-164 Dy-165 Er-165 Dy-166 Ho-166m Ho-166 Tm-166 Yb-166 Ho-167 Tm-167 Yb-167 Er-169 Lu-169 Yb-169 Hf-170 Lu-170 Tm-170 Er-171 Lu-171 Tm-171 Er-172 Hf-172 Lu-172 Ta-172 Tm-172 Hf-173 Lu-173 Ta-173 Tm-173 Lu-174m Lu-174 Ta-174 Hf-175 Ta-175 Tm-175 Yb-175 Lu-176m Lu-176 Ta-176 W-176 Hf-177m Lu-177m Lu-177 Re-177 Ta-177
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.24·10−13 6.99·10−15 1.93·10−15 8.33·10−15 2.82·10−14 1.61·10−15 5.79·10−15 9.90·10−14 4.46·10−14 1.08·10−13 3.88·10−15 2.95·10−14 1.17·10−14 1.38·10−14 2.83·10−15 5.90·10−14 1.73·10−14 3.00·10−14 1.46·10−13 1.81·10−14 4.22·10−14 3.80·10−14 3.17·10−17 3.21·10−14 5.46·10−15 1.07·10−13 1.16·10−13 5.76·10−14 2.23·10−14 6.45·10−15 5.08·10−14 3.89·10−14 2.89·10−15 6.53·10−15 5.36·10−14 2.15·10−14 5.32·10−14 9.10·10−14 6.93·10−15 2.72·10−14 3.74·10−14 1.25·10−13 8.74·10−15 1.39·10−13 5.89·10−14 7.13·10−15 5.17·10−14 3.36·10−15
Effective Dose [Sv/(Bq s m−2)] 8.50·10−14 4.92·10−15 1.06·10−15 8.03·10−16 1.35·10−15 8.96·10−16 1.21·10−15 7.84·10−14 1.72·10−15 8.78·10−14 2.35·10−15 1.59·10−14 5.39·10−15 9.48·10−15 2.97·10−17 4.75·10−14 1.13·10−14 2.29·10−14 1.21·10−13 3.67·10−16 1.64·10−14 3.00·10−14 1.77·10−17 2.29·10−14 3.40·10−15 8.64·10−14 7.10·10−14 2.30·10−14 1.66·10−14 4.42·10−15 2.55·10−14 1.72·10−14 1.84·10−15 4.94·10−15 2.75·10−14 1.54·10−14 4.24·10−14 4.81·10−14 1.75·10−15 7.65·10−16 2.11·10−14 1.03·10−13 5.98·10−15 9.67·10−14 4.24·10−14 1.50·10−15 2.76·10−14 2.15·10−15
Radionuclide W-177 Yb-177 Hf-178m Lu-178m Lu-178 Re-178 Ta-178b Ta-178a W-178 Yb-178 Hf-179m Lu-179 Ta-179 W-179 Hf-180m Os-180 Re-180 Ta-180m Ta-180 Hf-181 Os-181 Re-181 W-181 Hf-182m Hf-182 Ir-182 Os-182 Re-182b Re-182a Ta-182m Ta-182 Hf-183 Ta-183 Hf-184 Ir-184 Re-184m Re-184 Ta-184 Ir-185 Os-185 Ta-185 W-185 Ir-186a Ir-186b Pt-186 Re-186m Re-186 Ta-186
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 5.11·10−14 3.60·10−14 1.36·10−13 9.06·10−14 5.68·10−14 1.04·10−13 5.87·10−14 5.65·10−15 6.09·10−16 1.07·10−14 5.26·10−14 2.99·10−14 1.45·10−15 2.58·10−15 5.82·10−14 3.19·10−15 7.11·10−14 3.67·10−15 3.26·10−14 3.62·10−14 7.03·10−14 4.76·10−14 1.84·10−15 5.82·10−14 1.46·10−14 1.35·10−13 2.46·10−14 1.08·10−13 6.71·10−14 1.93·10−14 7.85·10−14 6.83·10−14 2.62·10−14 3.12·10−14 1.21·10−13 2.19·10−14 5.00·10−14 1.16·10−13 3.52·10−14 4.01·10−14 5.20·10−14 4.52·10−15 9.55·10−14 6.41·10−14 4.10·10−14 7.24·10−16 2.03·10−14 1.49·10−13
Effective Dose [Sv/(Bq s m−2)] 3.91·10−14 8.82·10−15 1.03·10−13 4.80·10−14 7.12·10−15 5.73·10−14 4.32·10−14 4.12·10−15 3.83·10−16 1.62·10−15 3.84·10−14 1.66·10−15 9.00·10−16 1.50·10−15 4.33·10−14 1.96·10−15 5.33·10−14 1.43·10−15 2.35·10−14 2.42·10−14 5.52·10−14 3.37·10−14 1.16·10−15 4.08·10−14 1.03·10−14 6.07·10−14 1.83·10−14 8.49·10−14 5.39·10−14 9.94·10−15 5.99·10−14 3.39·10−14 1.19·10−14 1.04·10−14 8.75·10−14 1.67·10−14 3.99·10−14 7.25·10−14 2.74·10−14 3.18·10−14 8.23·10−15 4.97·10−17 7.51·10−14 4.33·10−14 3.27·10−14 4.14·10−16 9.97·10−16 7.02·10−14
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Ir-187 Re-187 W-187 Ir-188 Pt-188 Re-188m Re-188 W-188 Ir-189 Os-189m Pt-189 Re-189 Ir-190n Ir-190m Ir-190 Os-190m Ir-191m Os-191m Os-191 Pt-191 Ir-192m Ir-192 Au-193 Hg-193m Hg-193 Os-193 Pt-193m Pt-193 Au-194 Hg-194 Ir-194m Ir-194 Os-194 Tl-194m Tl-194 Au-195m Au-195 Hg-195m Hg-195 Ir-195m Ir-195 Pb-195m Pt-195m Tl-195 Hg-197m Hg-197 Pt-197m Pt-197 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.03·10−14 0.00 4.09·10−14 9.18·10−14 1.18·10−14 3.91·10−15 5.35·10−14 2.91·10−15 4.14·10−15 7.16·10−18 1.82·10−14 2.15·10−14 8.89·10−14 7.52·10−18 8.24·10−14 9.12·10−14 4.07·10−15 3.67·10−16 4.35·10−15 1.71·10−14 8.81·10−15 5.53·10−14 9.16·10−15 6.21·10−14 1.26·10−14 2.44·10−14 3.07·10−15 2.07·10−17 6.19·10−14 2.65·10−17 1.34·10−13 5.85·10−14 5.22·10−17 1.47·10−13 4.41·10−14 1.35·10−14 4.12·10−15 1.38·10−14 1.11·10−14 3.53·10−14 2.19·10−14 9.97·10−14 5.92·10−15 7.52·10−14 1.02·10−14 3.35·10−15 1.86·10−14 1.06·10−14
Effective Dose [Sv/(Bq s m−2)] 1.54·10−14 0.00 2.13·10−14 7.52·10−14 7.90·10−15 2.56·10−15 3.13·10−15 1.10·10−16 2.77·10−15 1.24·10−19 1.34·10−14 3.08·10−15 6.81·10−14 1.38·10−19 6.32·10−14 7.03·10−14 2.62·10−15 2.31·10−16 2.78·10−15 1.21·10−14 6.84·10−15 3.61·10−14 6.03·10−15 4.69·10−14 7.70·10−15 3.29·10−15 3.76·10−16 4.07·10−19 4.94·10−14 6.23·10−19 1.04·10−13 4.73·10−15 2.17·10−17 1.03·10−13 3.41·10−14 8.52·10−15 2.73·10−15 8.78·10−15 8.38·10−15 1.78·10−14 2.17·10−15 7.12·10−14 2.44·10−15 5.94·10−14 3.62·10−15 2.26·10−15 3.25·10−15 9.73·10−16
Radionuclide Tl-197 Au-198m Au-198 Pb-198 Tl-198m Tl-198 Au-199 Hg-199m Pb-199 Pt-199 Tl-199 Au-200m Au-200 Bi-200 Pb-200 Pt-200 Tl-200 Au-201 Bi-201 Pb-201 Tl-201 Bi-202 Pb-202m Pb-202 Tl-202 Bi-203 Hg-203 Pb-203 Po-203 Tl-204 Bi-205 Pb-205 Po-205 Bi-206 Tl-206 At-207 Bi-207 Po-207 Tl-207 Tl-208 Pb-209 Tl-209 Bi-210m Bi-210 Pb-210 Po-210 At-211 Bi-211
6-39 Skin Dose [Sv/(Bq s m−2)] 2.43·10−14 3.75·10−14 4.08·10−14 2.66·10−14 7.40·10−14 1.16·10−13 8.23·10−15 2.71·10−14 8.55·10−14 4.38·10−14 1.49·10−14 1.27·10−13 6.36·10−14 1.43·10−13 1.31·10−14 1.13·10−14 7.50·10−14 2.78·10−14 8.99·10−14 4.43·10−14 4.89·10−15 1.57·10−13 1.17·10−13 2.72·10−17 2.63·10−14 1.39·10−13 1.56·10−14 1.87·10−14 1.00·10−13 1.24·10−14 9.70·10−14 2.92·10−17 9.10·10−14 1.90·10−13 3.36·10−14 7.76·10−14 9.31·10−14 7.67·10−14 3.06·10−14 2.34·10−13 9.35·10−15 1.59·10−13 1.63·10−14 2.30·10−14 1.28·10−16 4.81·10−19 1.96·10−15 3.07·10−15
Effective Dose [Sv/(Bq s m−2)] 1.78·10−14 2.39·10−14 1.81·10−14 1.86·10−14 5.26·10−14 9.47·10−14 3.67·10−15 7.63·10−15 6.83·10−14 9.32·10−15 1.02·10−14 9.32·10−14 1.32·10−14 1.08·10−13 8.17·10−15 2.33·10−15 5.98·10−14 2.62·10−15 6.08·10−14 3.35·10−14 3.25·10−15 1.24·10−13 9.29·10−14 4.96·10−19 2.00·10−14 1.13·10−13 1.04·10−14 1.30·10−14 7.59·10−14 1.71·10−16 7.98·10−14 5.45·10−19 7.29·10−14 1.51·10−13 3.95·10−16 6.09·10−14 7.04·10−14 6.08·10−14 4.53·10−16 1.69·10−13 1.00·10−16 9.65·10−14 1.12·10−14 2.58·10−16 4.48·10−17 3.89·10−19 1.37·10−15 2.04·10−15
6-40 Radionuclide Pb-211 Po-211 Bi-212 Pb-212 Po-212 Bi-213 Po-213 Bi-214 Pb-214 Po-214 At-215 Po-215 At-216 Po-216 At-217 At-218 Po-218 Rn-218 Fr-219 Rn-219 Fr-220 Rn-220 Fr-221 Fr-222 Ra-222 Rn-222 Ac-223 Fr-223 Ra-223 Ac-224 Ra-224 Ac-225 Ra-225 Ac-226 Ra-226 Th-226 Ac-227 Pa-227 Ra-227 Th-227 Ac-228 Pa-228 Ra-228 Th-228 Th-229 Pa-230 Th-230 U-230 Pa-231
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 3.06·10−14 4.47·10−16 4.05·10−14 1.35·10−14 0.00 3.39·10−14 0.00 1.28·10−13 2.77·10−14 4.71·10−18 1.12·10−17 1.01·10−17 8.03·10−17 9.57·10−19 1.86·10−17 2.12·10−16 7.56·10−19 4.30·10−17 2.04·10−16 3.38·10−15 8.53·10−16 2.20·10−17 2.02·10−15 4.76·10−14 5.51·10−16 2.28·10−17 3.05·10−16 2.30·10−14 8.87·10−15 1.08·10−14 6.35·10−16 9.40·10−16 3.01·10−15 2.15·10−14 4.79·10−16 6.37·10−16 1.10·10−17 1.08·10−15 3.19·10−14 6.50·10−15 7.88·10−14 6.56·10−14 0.00 1.50·10−16 5.41·10−15 3.73·10−14 4.51·10−17 1.07·10−16 2.44·10−15
Effective Dose [Sv/(Bq s m−2)] 2.59·10−15 3.56·10−16 8.95·10−15 6.24·10−15 0.00 6.16·10−15 0.00 7.25·10−14 1.09·10−14 3.81·10−18 8.51·10−18 7.79·10−18 5.38·10−17 7.75·10−19 1.37·10−17 9.71·10−17 4.21·10−19 3.40·10−17 1.53·10−16 2.46·10−15 4.40·10−16 1.72·10−17 1.32·10−15 5.79·10−16 4.03·10−16 1.77·10−17 1.87·10−16 2.20·10−15 5.47·10−15 8.01·10−15 4.29·10−16 6.37·10−16 2.40·10−16 5.57·10−15 2.84·10−16 3.21·10−16 5.12·10−18 7.38·10−16 7.01·10−15 4.43·10−15 4.49·10−14 5.16·10−14 0.00 8.10·10−17 3.36·10−15 2.91·10−14 1.48·10−17 4.56·10−17 1.57·10−15
Radionuclide Th-231 U-231 Np-232 Pa-232 Th-232 U-232 Np-233 Pa-233 U-233 Np-234 Pa-234m Pa-234 Pu-234 Th-234 U-234 Np-235 Pu-235 U-235 Np-236a Np-236b Pu-236 U-236 Am-237 Np-237 Pu-237 U-237 Am-238 Cm-238 Np-238 Pu-238 U-238 Am-239 Np-239 Pu-239 U-239 Am-240 Cm-240 Np-240m Np-240 Pu-240 U-240 Am-241 Cm-241 Pu-241 Am-242m Am-242 Cm-242 Pu-242
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 2.52·10−15 3.82·10−15 6.94·10−14 5.57·10−14 3.44·10−17 5.92·10−17 4.78·10−15 1.66·10−14 4.57·10−17 8.41·10−14 5.48·10−14 1.24·10−13 3.46·10−15 7.50·10−16 4.25·10−17 1.82·10−16 4.78·10−15 8.64·10−15 9.17·10−15 5.76·10−15 4.83·10−17 3.57·10−17 2.14·10−14 1.54·10−15 2.54·10−15 9.97·10−15 5.09·10−14 3.94·10−15 4.31·10−14 4.09·10−17 2.91·10−17 1.56·10−14 1.60·10−14 1.86·10−17 2.61·10−14 5.79·10−14 4.68·10−17 5.93·10−14 9.15·10−14 3.92·10−17 3.12·10−15 1.28·10−15 3.14·10−14 1.17·10−19 1.36·10−16 8.20·10−15 4.29·10−17 3.27·10−17
Effective Dose [Sv/(Bq s m−2)] 4.58·10−16 2.56·10−15 5.38·10−14 4.26·10−14 7.24·10−18 1.17·10−17 3.39·10−15 8.55·10−15 1.42·10−17 6.83·10−14 1.21·10−15 8.72·10−14 2.49·10−15 2.94·10−16 6.11·10−18 4.19·10−17 3.45·10−15 6.46·10−15 4.74·10−15 1.92·10−15 4.68·10−18 3.86·10−18 1.55·10−14 8.87·10−16 1.76·10−15 5.29·10−15 4.04·10−14 2.85·10−15 2.56·10−14 3.50·10−18 2.50·10−18 9.26·10−15 6.95·10−15 3.48·10−18 2.13·10−15 4.67·10−14 4.17·10−18 1.55·10−14 5.88·10−14 3.42·10−18 5.87·10−17 6.74·10−16 2.11·10−14 6.33·10−20 2.49·10−17 6.09·10−16 4.02·10−18 2.90·10−18
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Am-243 Cm-243 Pu-243 Am-244m Am-244 Cf-244 Cm-244 Pu-244 Am-245 Bk-245 Cm-245 Pu-245 Am-246m Am-246 Bk-246 Cf-246 Cm-246 Pu-246 Bk-247 Cm-247 Cf-248 Cm-248
Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.75·10−15 9.79·10−15 8.15·10−15 3.11·10−14 5.25·10−14 4.65·10−17 3.91·10−17 2.69·10−17 1.62·10−14 1.58·10−14 5.36·10−15 4.00·10−14 8.56·10−14 6.42·10−14 5.31·10−14 3.35·10−17 3.49·10−17 8.82·10−15 7.43·10−15 1.79·10−14 3.17·10−17 2.67·10−17
Effective Dose [Sv/(Bq s m−2)] 1.85·10−15 5.30·10−15 9.61·10−16 3.63·10−16 3.59·10−14 4.74·10−18 3.40·10−18 2.08·10−18 1.45·10−15 9.26·10−15 3.49·10−15 1.86·10−14 4.74·10−14 3.06·10−14 4.27·10−14 3.92·10−18 3.10·10−18 5.35·10−15 4.20·10−15 1.38·10−14 3.25·10−18 2.35·10−18
Radionuclide Bk-249 Cf-249 Cm-249 Bk-250 Cf-250 Cm-250 Es-250 Cf-251 Es-251 Cf-252 Fm-252 Cf-253 Es-253 Fm-253 Cf-254 Es-254m Es-254 Fm-254 Fm-255 Fm-257 Md-257 Md-258
6-41 Skin Dose [Sv/(Bq s m−2)] 4.07·10−17 1.91·10−14 1.59·10−14 6.43·10−14 3.02·10−17 0.00 2.21·10−14 1.12·10−14 5.35·10−15 3.08·10−17 2.95·10−17 1.66·10−15 4.55·10−17 4.55·10−15 9.83·10−20 3.76·10−14 5.65·10−16 3.43·10−17 3.95·10−16 7.18·10−15 6.20·10−15 1.82·10−16
Effective Dose [Sv/(Bq s m−2)] 4.68·10−19 1.45·10−14 1.02·10−15 4.12·10−14 3.09·10−18 0.00 1.76·10−14 5.01·10−15 3.65·10−15 3.63·10−18 3.45·10−18 1.75·10−17 1.60·10−17 3.12·10−15 1.01·10−20 2.11·10−14 1.57·10−16 4.76·10−18 8.82·10−17 4.15·10−15 4.52·10−15 3.89·10−17
6-42
6 External dosimetry
6.7 References 68Dre 69Sny 69Spi 70Sto 75Emm 75ICR 77ICR 78Sny 79Lic 80Cri 82Kra 83ICR 83Rou 85ICR 85Nel 85Sai 86Gro 88DOE 88ICR 88Zan 89Vei 90Hol 90Sai 91Bri
Drexler, G.: Proceedings of the Symposium on Microdosimetry, Ispra (Italy), 13-15 November 1967. European Communities, Brussels. Report No. EUR 3747 d-f-e, 1968, p. 433. Snyder, W.S., Ford, M.R., Warner, G.G., Fisher jr., H.L.: Medical Internal Radiation Dose Committee (MIRD) Pamphlet No. 5, Supplement No. 3: J. Nucl. Med. 10 (1969). Spiers, F.W.: Delayed effects of bone-seeking radionuclides, Mays, C.W., Jee, W.S.S., Lloyd, R.D., Stover, B.J., Dougherty, J.H., Taylor, G. (eds.), Salt Lake City: University of Utah Press, 1969, p. 95. Storm, E., Israel, H.I.: Nucl. Data Tables A 7 (1970) 565. Emmett, M.B.: Oak Ridge National Laboratory Report No ORNL4972. Oak Ridge, TN, 1975. International Commission on Radiological Protection: ICRP Publication 23. Oxford, UK: Pergamon Press, 1975. International Commission on Radiological Protection: ICRP Publication 26. Oxford, UK: Pergamon Press, 1977. Snyder, W.S., Ford, M.R., Warner, G.G., Fisher jr., H.L.: Medical Internal Radiation Dose Committee (MIRD) Pamphlet No. 5, New York: The Society of Nuclear Medicine, 1978. Lichtenstein, H., Cohen, M.O., Steinber, H.A., Trubetzkoys, E.S., Beer, M.: Computer code manual of the electric power research institutes (MAGI) EPRI-CCM-8. Palo Alto, California: Mathematical application group Inc., 1979. Cristy, M.: Oak Ridge National Laboratory Report No. ORNL/NUREG/TM-367. Oak Ridge, TN, 1980. Kramer, R., Zankl, M., Williams, G., Drexler, G.: GSF-Report S-885. Neuherberg, Germany: GSF-National Research Center for Environment and Health, 1982. International Commission on Radiological Protection: ICRP Publication 38. Oxford, UK: Pergamon Press, 1983. Roussin, R.W., Knight, J.R., Hubbell, J.H., Howerton, R.J.: Report No. ORNL-RSIC-46 (ENDF-335), Oak Ridge, TN: Radiation Shielding Information Center, Oak Ridge National Laboratory, 1983. International Commission on Radiation Units and Measurements: ICRU Report 39. Bethesda, MD: ICRU Publications, 1985. Nelson, W.R., Hirayama, H., Rogers, D.W.O.: SLAC-265-UC-32. Stanford, CA: Stanford Linear Accelerator Center, 1985. Saito, K., Moriuchi, S.: Radiat. Prot. Dosim. 12 (1985) 21. Grosswendt, B, Roos, M.: Medizin Physik (1986) 265. Department of Energy DOE/EH-0070 DOE, Washington DC: Department of Energy, 1988. International Commission on Radiation Units and Measurements: ICRU Report 43. Bethesda, MD: ICRU Publications, 1988. Zankl, M., Veit, R., Williams, G., Schneider, K., Fendel, H., Petoussi, N., Drexler, G.: Radiat. Environ. Biophys. 27 (1988) 153. Veit, R., Zankl, M., Petoussi, N., Mannweiler, E., Williams, G., Drexler, G.: GSF-Report 3/89. Neuherberg, Germany: GSF - National Research Center for Environment and Health, 1989. Hollnagel, R.A.: Radiat Prot. Dosim. 30 (1990) 149. Saito, K., Petoussi, N., Zankl, M., Veit, R., Jacob, P., Drexler, G.: GSF-Report 2/90. Neuherberg, Germany: GSF-National Research Center for Environment and Health, 1990. Briesmeister, J.F.: Los Alamos National Laboratory Report LA-12625-M. Los Alamos, New Mexico, 1977.
Landolt-Börnstein New Series VIII/4
6 External dosimetry 91Gro 91ICR 91Pet 92Hal 92ICR1 92ICR2 92Zan 93Eck1 93Eck2 93Gro 93ICR 93Yam 94Gro 94Gua1 94Gua2 94Hir 94ICR 94Yam 94Zub 95ISO 95Sai 95Til 96Cha 96Dim 96ICR 97Zan 98Cla 98ICR 98Sai 99Sta 99Zan 00Pel
6-43
Grosswendt, B.: Radiat Prot. Dosim. 35 (1991) 221. International Commission on Radiological Protection: ICRP Publication 60. Oxford, UK: Pergamon Press, 1991. Petoussi, N., Jacob, P., Zankl, M., Saito, K.: Radiat. Prot. Dosim. 37 (1991) 31. Halbleib, J.A., Kensek R.P., Mehlhorn, T.A., Valdez, G.D., Seltzer, S.M., Berger, M.J.: Report No SAND-1634 / UC-405. Albuquerque, New Mexico and Livermore, California: SANDIA National Laboratories, 1992. International Commission on Radiation Units and Measurements: ICRU Report 47. Bethesda, MD: ICRU Publications, 1992. International Commission on Radiation Units and Measurements: ICRU Report 48. Bethesda, MD: ICRU Publications, 1992. Zankl, M., Petoussi, N., Drexler, G.: Health Phys. 62 (1992) 395. Eckerman, K.F., Ryman, J.C.: Federal Guidance Report No. 12. Oak Ridge, TN: Oak Ridge National Laboratory, 1993. Eckerman, K.F., Westfall, R.J., Ryman, J.C., Cristy, M.: Oak Ridge National Laboratory Report No ORNL/TM-12350. Oak Ridge, TN: Oak Ridge National Laboratory, 1993. Grosswendt, B.: PTB-Report Dos-22, Braunschweig: Physikalisch-Technische Bundesanstalt, 1993. International Commission on Radiation Units and Measurements: ICRU Report 51. Bethesda, MD: ICRU Publications, 1993. Yamaguchi, Y.: Radioisotopes 42 (1993) 35. Grosswendt, B.: Radiat Prot. Dosim. 54 (1994) 85. Guaraldi , R., Padoani, F.: ENEA Report RT/ERG/94/17, 1994. Guaraldi , R., Padoani, F.: ENEA Report RT/ERG/94/21, 1994. Hirayama, H.: Radiat. Prot. Dosim. 51 (1994) 107. International Commission on Radiation Units and Measurements: ICRU Report 53. Bethesda, MD: ICRU Publications, 1994. Yamaguchi, Y.: Radiat. Prot. Dosim. 55 (1994) 123. Zubal, I.G., Harrell, C.R., Smith, E.O., Rattner, Z., Gindi, G., Hoffer, P.B.: Med. Phys. 21 (1994) 299. International Organization for Standardization: ISO FDIS 4037-3, 1995. Saito, K., Jacob, P.: Radiat. Prot. Dosim. 58 (1) (1995) 29. Till, E., Zankl, M., Drexler G.: GSF-Report 27/95. Neuherberg, Germany: GSF-National Research Center for Environment and Health, 1995. Chartier, J.-L., Grosswendt, B., Gualdrini, G.F., Hirayama, H., Ma, C.-M., Padoani, F., Petoussi, N., Seltzer, S.M., Terrisol, M.: Radiat. Prot. Dosim. 63 (1996) 7. Dimbylow, P.J.: Proc. Voxel phantom development 6-7 July 1996, Dimbylow, P.J. (ed.), Chilton, UK: National Radiological Protection Board, 1996, p. 1. International Commission on Radiological Protection: ICRP Publication 74. Oxford, UK: Pergamon Press, 1996. Zankl, M., Drexler, G., Petoussi-Henss, N., Saito, K.: GSF-Report 8/97. Neuherberg, Germany: GSF - National Research Center for Environment and Health, 1997. Clark, M.J., Chartier, J.-L., Siebert, B.R.L., Zankl, M.: Radiat. Prot. Dosim. 78 (1998) 91. International Commission on Radiation Units and Measurements: ICRU Report 57. Bethesda, MD: ICRU Publications, 1998. Saito, K., Petoussi-Henss, N., Zankl, M.: Health Phys. 74 (6) (1998) 698. Stabin, M.G., Tagesson, M., Thomas, S.R., Ljungberg, M., Strand, S.E.: Appl. Radiat. Isot. 50 (1999) 73. Zankl, M.: Health Phys. 76 (1999) 162. Pellicioni, M.: Radiat. Prot. Dosim. 88 (2002) 279.
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6 External dosimetry Xu, X.G, Chao, T. C., Bozkurt A.: Health Phys. 78 (2000) 476. Zankl, M., Wittmann, A.: Radiat. Environ. Biophys. 40 (2001) 153. Eckerman, K.F. Private communication, 2002. Petoussi-Henss, N., Zankl, M., Fill, U., Regulla, D.: Phys. Med. Biol. 47 (2002) 89. Roesler, S., Heinrich, W., Schraube, H.: Radiat. Prot. Dosim. 98 (2002) 367.
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7 Internal dosimetry of radionuclides
This Chapter reviews the behaviour of radionuclides in the body. It summarises the biokinetic and dosimetric models that have been developed by the International Commission on Radiological Protection (ICRP) for assessing radiation doses, and hence risks, from intakes by different routes, including inhalation and ingestion. These models have been widely accepted around the world for use in radiological protection. They have been incorporated in the European and International Basic Safety Standards as well as in many national regulations and guidance notes around the world. Future developments in this area are also examined. Finally, methods that can be used to assess intakes of radionuclides by direct and indirect monitoring procedures and requirements for dose assessment are summarised.
7.1 Introduction People may be exposed to radionuclides in a number of ways. They may be taken into the body as a result of occupational exposure or uptake from the environment. They are used extensively in medical diagnosis and treatment as well as in biomedical research. People may also be exposed externally by submersion in a radioactive cloud. For occupational exposure, the main route of intake is by inhalation, although a fraction of any material deposited in the respiratory system will be transferred to the throat and swallowed, giving the opportunity for absorption in the gastrointestinal tract. Intakes by direct ingestion may occur and some radionuclides may be absorbed through the intact skin. Damage to the intact skin by cuts or other wounds can also result in the entry of radionuclides into the body. For members of the public, the main route of intake of radionuclides will be by ingestion in food and drinking water although intakes by inhalation may also occur, in particular in the case of accidental releases into the environment. For medical applications the method of administration will depend upon the specific nature of the diagnostic investigation or treatment. Knowledge of the behaviour of radionuclides in the body is important for assessing radiation doses resulting from intakes or superficial contamination. For occupational and public exposure the calculation of radiation doses provides the basis for controlling exposures to within accepted limits, for assessing the consequences of the presence of radionuclides in the working or natural environment or determining the need for treatment in the case of accidental intakes. In medical situations radiation doses are needed for optimising diagnostic and treatment schedules. In the case of administration of radionuclides for clinical research, for example on the behaviour of radiopharmaceuticals, assessment of radiation doses is needed for estimating risks for ethical considerations. The radiation dose received by a tissue as a result of the intake of a radionuclide will depend upon a number of factors. These include: the route of intake, the physico-chemical form, its biokinetic behaviour and pathways in the body, organ(s) of accumulation, rate of removal (by physical decay, biological turnover and excretion) and the quality of the emitted radiation (α, β, γ). Biological variation (age, sex, dietary habits etc) will also influence behaviour in the body. Thus the determination of the radiation dose to tissues and the assessment of the possible biological effects resulting from the intake of a particular radionuclide requires a knowledge of all the pertinent physical, physiological, biokinetic and chemical data. Landolt-Börnstein New Series VIII/4
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This Chapter reviews the biokinetic behaviour of radionuclides in the body and illustrates the key features through examples. It summarises the models that have been developed by the International Commission on Radiological Protection (ICRP) for assessing doses and hence risks from intakes of radionuclides and examines some future developments in this area. These models have been widely accepted around the world for use in radiological protection. They provide the basis for dose coefficients (doses per unit intakes, Sv Bq−1) for assessing radiation doses from intakes by inhalation and ingestion and have been incorporated in the European and International Basic Safety Standards [96E1, 96I2] as well as in many national regulations and guidance notes around the world. The use of these dose coefficients for assessing the risks from intakes of radionuclides is illustrated. A recent development by ICRP has been the issue of a report giving dose coefficients for the embryo and foetus following intakes of radionuclides by the mother before or during pregnancy [01I1]. A further report is presently being prepared that will give doses to the newborn child from radionuclides consumed in mothers’ milk. The development of the document is summarised. A new dosimetric model for the human alimentary tract is also being prepared. The conceptual basis for this model is reviewed. Finally, methods that can be used to assess intakes of radionuclides by monitoring procedures and the requirements for dose assessment are described in principal; more detailed information is given in Sections 10.3.2 and 10.3.3 of Chapter 10.
7.2 Biokinetics of radionuclides in the body The principal routes by which radionuclides may enter and move around the body and which must be considered in internal dosimetry are summarised in Figure 7.1. Radionuclides passing through the gastrointestinal tract, or deposited in the air passages of the lungs, in a wound or on the outer layer of skin will irradiate these tissues. Soluble forms of radionuclide(s) that are transportable can readily enter the bloodstream and their subsequent fate depends upon their chemical characteristics. If poorly transportable they will only slowly enter the bloodstream or the lymphatic system. Any insoluble particles entering the systemic circulation will be taken up by the reticuloendothelial cells of the liver, spleen and red bone marrow. Here they may remain for up to the life-span of the individual. To facilitate calculation of doses to tissues following intakes of radionuclides, the ICRP has developed a number of generalised biokinetic models to describe their movement and behaviour in the body. Specific models were given by ICRP for adult workers in Publication 30 [79I1, 80I1, 80I2, 88I2] to describe the behaviour of radionuclides in the main organs of intake − the lungs and gastrointestinal tract as well as the skin. For radionuclides that have entered the blood and systemic circulation, activity subsequently deposited in tissues was generally assumed to be uniformly distributed throughout them and therefore the radiation dose depends solely on the organ mass and both the physical half-life and the biological half-time of the radionuclide (see Chapters 3 and 4). A specific model was needed for the skeleton, however, because of the morphology of skeletal bone and the heterogeneous distribution of deposited activity. Biokinetic models were also given in the various parts and supplements of Publication 30 to describe the behaviour of radionuclides in the body after their entry into the blood. More recently, ICRP has provided age-dependent biokinetic models for selected radionuclides in Publications 56, 67, 69, 71 and 72 [89I1, 93I1, 95I1, 95I2, 96I1] and has given dose coefficients (Sv Bq−1) for six ages: 3-month-old infants, 1-, 5-, 10- and 15-year old children and adults. The requirement for age-dependent models and dose coefficients became apparent in the aftermath of the Chernobyl accident when it was realised that, whilst some countries had developed such models there were no models that were generally accepted around the world. Such models are essential for assessing doses to the public from intakes of radionuclides in foods and drinking water, for making comparisons with dose limits and for informing decisions on the acceptability for consumption of foods that may be marketed in many countries. For the development of “age-dependent” models there was a need to include anatomical and physiological information, such as age dependent mass and turnover rate of the skeleton. These, socalled physiologically based models provide a framework in which both human and animal data on the behaviour of radionuclides in the body can be integrated and allow a more realistic approach to the calculation of doses to individuals of different ages. They also have the important advantage that they can take into account excretion and are therefore more appropriate for the interpretation of bioassay data. Landolt-Börnstein New Series VIII/4
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Extrinsic removal Inhalation Skin Lymph nodes
Exhalation
Ingestion
Respiratory tract
Direct absorption
Liver
Gastrointestinal tract
Transfer compartment
Sweat
Wound
Subcutaneous tissue
Other organs
Kidney
Urinary bladder Skin Urine
Faeces
Fig. 7.1. Summary of the main routes in intake, transfers and excretion of radionuclides in the body; [97I2].
In addition to the development of age-dependent biokinetic models a new human respiratory tract model (HRTM) was issued in Publication 66 [94I2]. This model has been applied in all the recent calculations of dose coefficients for workers and the public issued by ICRP (see compilations in 94I1, 96I1, 99I1). Table 7.1 summarises the recent ICRP publications giving revised biokinetic models and dose coefficients. A further development has been the issue of a report giving dose coefficients for the embryo and foetus following intakes of radionuclides by the mother before or during pregnancy. ICRP Publication 88 gives biokinetic models for 31 elements and also dose coefficients for selected radionuclides [01I1]. Presently being prepared by ICRP is a further report that will give doses to the newborn child from radionuclides consumed in mothers’ milk. The development of the document is summarised in Section 7.2.8. A new dosimetric model for the human alimentary tract is also being developed that will also be age-dependent. The conceptual basis for this new model is summarised in Section 7.2.2.
7.2.1 Inhalation A model for describing the deposition and clearance of inhaled radionuclides in adults who are occupationally exposed was first given in Publication 30 of ICRP [79I1]. This lung model separated the respiratory system into three distinct regions, the naso-pharynx (NP), the tracheo-bronchial region (TB) and the pulmonary region (P). It gave information on the deposition of inhaled radionuclides in these regions as a function of the activity median aerodynamic diameter (AMAD) of inhaled particulates and on the rate of clearance of the material from the respiratory system in terms of three default clearance Classes. These had clearance times from the pulmonary part of the respiratory system of days (Class D), weeks (Class W) and years (Class Y). The default particle size was taken to have an AMAD of 1 µm. Information was also given on the transfer of the three Classes of material to lymphatic tissue associated with the respiratory system. A main feature of this lung model is that it calculated only the average dose to the lungs (TB and P regions). Although the model was developed for adults it has been used for younger ages but without any changes to parameter values, other than organ mass.
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Table 7.1 Summary of ICRP reports on dose coefficients for workers and members of the public from intakes of radionuclides. ICRP Application Intake Contents Publication No. (year) 56 (1989) Publica Inhalation and Age-dependent systemic models, and tissue dose ingestion coefficients for selected radioisotopes, for H, C, Sr, Zr, Nb, Ru, I, Cs, Ce, Pu, Am and Np. Issued before ICRP Publication 60 [91I1], and hence giving dose equivalents using the tissue weighting factors from ICRP Publication 26 [77I1]. It was also issued before ICRP Publication 66 [94I2] and hence used the ICRP Publication 30 lung model [79I1]. The dose coefficients given in ICRP Publication 56 were superseded by those in ICRP Publications 67 and 71, which used the tissue weighting factors from Publication 60. 67 (1993) Publica Ingestion Age-dependent systemic models, and tissue dose coefficients for selected radioisotopes, for S, Co, Ni, Zn, Mo, Tc, Ag, Te, Ba, Pb, Po and Ra. Updated systemic models are given for Sr, Pu, Am and Np. 68 (1994) Workers Inhalation and Effective dose coefficients for workers, for about 800 ingestion radionuclides: selected radioisotopes of the 91 elements covered in ICRP Publication 30, Parts 1-4 [79I1, 80I1 and I2, 88I2]. The inhalation dose coefficients for workers exposed to 226Ra given in ICRP Publication 68 were revised in Annexe B of ICRP Publication 72. Applies the Human Respiratory Tract Model, HRTM [94I2]. 69 (1995) Publica Ingestion Age-dependent systemic models, and tissue dose coefficients for selected radioisotopes, for Fe, Sb, Se, Th and U. 71 (1995) Publica Inhalation Tissue dose coefficients for selected radioisotopes of elements covered in ICRP Publications 56, 67 and 69, plus Ca and Cm for which age-dependent systemic models are given. Applies the HRTM [94I2]. 72 (1996) Publica Inhalation and Effective dose coefficients for members of the public for ingestion radioisotopes of the 31 elements covered in ICRP Publications 56, 67, 69, and 71, plus radioisotopes of the further 60 elements covered in ICRP Publications 30 and 68. Applies the HRTM [94I2]. CD-ROM Publica Inhalation and A database of equivalent doses to individual tissues (1998) and ingestion corresponding to the effective dose coefficients in ICRP workers Publications 68 and 72. Inhalation dose coefficient for 10 particle sizes. 88 (2001) Embryo Inhalation and Dose coefficients for the offspring for intakes by the mother and foetus ingestion by (worker or public) before or during pregnancy of the mother radionuclides of the 31 elements covered in Publications 68 and 72. CD-ROM2 Embryo Inhalation and Database of dose coefficients extending information on (2002) and foetus ingestion by radionuclides in Publication 88. the mother a Age-dependent dose coefficients (3 months, 1-, 5-, 10-, and 15-years and adult)
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The Human Respiratory Tract Model (HRTM) described in ICRP Publication 66 [94I2] was developed to replace the lung model given in ICRP Publication 30. It takes into account extensive data on the behaviour of inhaled materials that had become available since the ICRP Publication 30 model was developed. As in the earlier model, deposition and clearance are treated separately. The scope of the model was extended to apply explicitly to all members of the population, giving reference values for 3-month-old infants, 1-, 5-, 10- and 15-y-old children and male and female adults. The main features of the model are summarised below. In the new model, the respiratory tract is represented by five regions (Fig. 7.2). The extrathoracic (ET) airways are divided into ET1, the anterior nasal passage and ET2, which consists of the posterior nasal and oral passages, the pharynx and larynx. The thoracic regions are bronchial (BB: trachea, generation 0 and bronchi, airway generations 1-8), bronchiolar (bb: airway generations 9-15), and alveolar-interstitial (AI: the gas exchange region). Lymphatic tissue is associated with the extrathoracic and thoracic airways (LNET and LNTH, respectively). Reference values of dimensions and scaling factors for subjects of different ages are specified in the model. A main feature of the HRTM, compared with the Publication 30 model, is the calculation of doses to these specific tissues in the five regions and allowance for their differences in radiosensitivity.
Posterior Nasal Passage ü ý þ
Pharynx
ET 1 Extrathoracic
Nasal Part Oral Part
ET 2
Larynx
BB Trachea Thoracic Bronchial Main Bronchi Bronchi
Bronchioles
bb
Bronchiolar Alveolar Interstitial
Al
bb Bronchioles Terminal Bronchioles
Al
Respiratory Bronchioles Alveolar Duct + Alveoli
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Fig. 7.2. Respiratory tract regions defined in the Human Respiratory Tract Model; [94I2].
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7.2.1.1 Deposition The amount of an aerosol inhaled depends upon breathing parameters and these are influenced by age, body size and level of physical exertion. Deposition of inhaled particles in the HRTM is calculated for each region of the respiratory system, with account taken of both inhalation and exhalation. This is done as a function of particle size, breathing parameters and/or work load, and is assumed to be independent of chemical form. Four standard levels of activity are defined in the HRTM ranging from sleep through to heavy exercise, and the different proportions of time spent at these reference levels are specified for representative individuals at the six standard ages [94I2]. The fraction of inhaled aerosol deposited in the various regions of the lung depends in turn upon the particle size, typically taken as a log-normal distribution. Age dependent default deposition parameters are given for a range of particle sizes from 0.6 nm activity median thermodynamic diameter (AMTD) to 20 µm activity median aerodynamic diameter (AMAD). Previously, a 1 µm AMAD was taken as the default particle size for occupational exposure, but ICRP Publication 66 now recommends 5 µm AMAD as being more typical of the workplace. For members of the public the default is taken as 1 µm AMAD. Table 7.2 compares regional deposition in the respiratory system for the models of ICRP Publications 30 and 66. For the old and new defaults for workers, total deposition is about 30 % higher in the new model (82 % c.f. 63 %), with the extrathoracic region dominating, although a large fraction of this is in the ET1 region and thus unavailable for systemic uptake. Conversely, deposition in the deep lung (bronchiolar and alveolar-interstitial regions) is a factor of four higher in the ICRP 30 model (25 % c.f. 6.4 %). Table 7.2 Comparison of regional deposition for ICRP 30 Lung and ICRP 66 Respiratory Tract Models. Publication 30 model Publication 66 model Adult Adult mem- Worker ber of public 1 µma 1 µma 1 µma 5 µma Region [%] Region [%] [%] [%] Nasal passage (NP) 30 Extrathoracic (ET1) 15 17 34 Extrathoracic (ET2) 19 21 40 Total (34) (38) (74) Trachea and bronchial tree (TB)
8
Bronchial (BB) Bronchiolar (bb) Total
1.3 2.0 (3.3)
1.2 1.7 (2.9)
1.8 1.1 (2.9)
Pulmonary region (P)
25
Alveolar-interstitial (AI)
11
11
5.3
Total
63
Total
48
52
82
a Activity Median Aerodynamic Diameter (AMAD) The variation in deposition parameters between individuals, depending upon age, gender and habits, is an important difference from the Publication 30 model [79I1] for which particle size was the only factor that influenced deposition (see 1 µm entries of Table 7.2). Previously, this distinction was not needed, since the Publication 30 model was intended only for reference adults who were occupationally exposed. In contrast, the ICRP Publication 66 model [94I2] has been designed for application to all members of the population. In the new model, deposition, but not clearance, is strongly influenced by age.
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7.2.1.2 Clearance Subsequent to deposition, material is cleared from the respiratory tract. For material deposited in the anterior nose ET1, clearance is affected extrinsically by such means as nose-blowing or sneezing. The ET1 deposit is cleared directly from the body and makes no subsequent contribution to gut or systemic tissue doses. Removal from all other regions is treated as two competing processes: particle transport (by mucociliary clearance to the throat or translocation to lymph nodes) and absorption to blood. It is assumed that these clearance processes compete independently with each other and have no age or gender dependence. Transport processes include mechanical transport to the gut by mucociliary action and removal by macrophages to the lymph nodes. Particle transport rates are taken to be fixed for all materials and a single compartment model describes clearance by this mechanism (Fig. 7.3). Absorption rates, however, are determined by solubility of inhaled materials and default parameters are recommended for Fast (Type F), Moderate (Type M) and Slow (Type S) absorption. This corresponds roughly to the ICRP Publication 30 classification scheme and chemical forms previously assigned to Class D, W or Y are now provisionally treated as Type F, M or S, respectively. The correspondence between the two schemes is not exact, e.g. the D, W or Y classification refers to whether total pulmonary lung clearance (by absorption to blood or clearance to the throat and then through the gut) is of the order of days, weeks or years, whereas Type F, M or S refers only to the absorption component. The mechanical clearance of the deposited activity is not dependent on the chemical form. The main clearance components for the two models, in the form of approximate biological half-times, are summarised in Table 7.3. Qualitatively, residence times in the lung are reduced, quite drastically, for Type F compared to Class D and elevated for Types M and S relative to Classes W and Y. Anterior nasal Nasooropharynx/ larynx
Extrathoracic LNET
LNTH
ET2′
ETseq
0.01
Bronchi
Bronchioles
ET1
0.001
0.01
BB2
bb2
GI tract
BB1 2
0.03 bbseq
100
Environment
10
0.03 BBseq
1
bb1 0.0001 0.001 0.02
Alveolar interstitial
0.00002
AI3 AI2 AI1
Thoracic
Fig. 7.3 Compartment model representing time-dependent particle transport from each respiratory tract region. Rates shown alongside arrows are reference values in units of d−1. It is assumed that (i) the AI deposit is divided between AI1, AI2 and AI3 in the ratio 0.3:0.6:0.1; (ii) the fraction of the deposit in BB and bb that is cleared slowly (BB2 and bb2) is 50 % for particles of physical size 2.5 µm, and the fraction retained in the airway wall (BBseq and bbseq) is 0.7 % at all sizes; (iii) 0.05 % of material deposited in region ET2 is retained in its wall (ETseq) and the rest in compartment ET2′ which clears rapidly to the GI tract. The model as shown above would describe the retention and clearance of a completely insoluble material. However, there is in general simultaneous absorption to body fluids of material from all the compartments except ET1; [94I1].
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Table 7.3. Comparison of approximate clearance half-times for Publication 30 Lung and Publication 66 Respiratory Tract Model ICRP 30 ICRP 66
Class D W Y
Pulmonary clearance T½ 0.5 d (100 %) 50 d (60 %) 500 d (60 %)
Type F M S
Dominant absorption T½ 10 min (10 %) 140 d (90 %) 7000 d (99.9 %)
By considering the relative rates of the two independent clearance processes (mechanical and absorption) in the HRTM and the amount absorbed in the gut after clearance to the throat it is possible to calculate the fraction of material ultimately transferred to the blood and systemic circulation, both directly from the lungs and indirectly via the gut. Because particle transport rates are fixed for all lung Types, the proportion of material escalating to the gut increases as the classification changes from Type F to Type S. It is interesting to compare the amounts transferred to the circulation for inhalation of the different Types. This is illustrated in Fig. 7.4, ignoring the effect of radioactive decay. For Type F material, such as soluble (transportable) forms of radioisotopes of caesium and iodine (e.g. 137Cs, 131I) there is rapid translocation to the blood with about 25 % of the intakes being taken up by within a day. In contrast for Type S materials, such as 239PuO2, transfer to the blood is much slower with about 0.15 % transferred after 1000 days. Examples of the lung clearance Types adopted for various chemical forms of a selection of radionuclides are given in Table 7.4. Although the HRTM model provides these default clearance Types there is also provision for including material specific absorption parameters when information is available. ICRP has issued a guidance document on this application of the HRTM [02I1].
Fraction of inhaled activity transferred to blood
1 Type F 10-1
Type M
10-2 Type S 10-3
10-4 10-2
10-1
1 10 10 2 Time after intake [d ]
10 3
10 4
Fig. 7.4. Cumulative fraction of inhaled activity absorbed into blood directly from the respiratory tract as a function of time after intake for each default absorption Type (in the absence of radioactive decay), for a reference worker; [02I1].
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Table 7.4 Examples of lung clearance Types adopted by ICRP for workers; [94I1]. Radionuclide Chemical form Inhalation Type Tritium Tritiated water SR2a Cobalt Unspecified compounds, Mb Oxides, hydroxide, halides and nitrate S Strontium Unspecified compounds Fb Strontium titanate S Zirconium Unspecified compounds F Oxides, hydroxide, halides and nitrate M Zirconium carbide S Niobium Unspecified compounds M Oxides and hydroxide S Ruthenium Unspecified compounds F Halides Mb Oxides and hydroxides S Iodine All compounds Fb Vapour Vb Caesium All compounds Fb Cerium Unspecified compounds Mb Oxides, hydroxides and fluorides S Polonium Unspecified compounds F Oxides, hydroxides and nitrate M Radium All compounds M Thorium Unspecified compounds M Oxide and hydroxides S Uranium Most hexavalent compounds, e.g. UFO6, F UO2F2 and UO2(NO3)2 Less soluble compounds, e.g. UO3, UF4, Mb UCI4 and most other hexavalent compounds Highly soluble compounds, e.g. UO2, U3O8 S Plutonium Unspecified compounds Mb Insoluble oxides Sb Americium All compounds Mb Sc Trace contaminant Curium All compounds Mb a Excretion and retention functions for inhalation of 3H2O given in Figure 7.15. b Excretion and retention functions for inhalation of 5 µm AMAD aerosols given in Figures 7.16-7.25. c Trace contaminant formed from 241Pu in matrices of nuclear fuel in insoluble (Type S) forms (Fig. 7.25). 7.2.1.3 Gases and vapours For radionuclides inhaled in particulate form, it is assumed that entry into and deposition in the respiratory tract is governed by the size distribution of the aerosol particles [94I2]. The situation is different for gases and vapours, for which the radionuclide has a specific behaviour at its site of entry into the respiratory tract, depending on the chemistry of the compound. Almost all inhaled molecules contact airway surfaces, but usually return to the air unless they dissolve in, or react with, the surface lining. The fraction of an inhaled gas or vapour that is deposited in each region thus depends on its solubility and reactivity. Generally, however, the regional deposition of a gas or vapour cannot be predicted on a mechanistic basis, from knowledge of its physical and chemical properties, but has to be obtained from an in vivo experimental study. Landolt-Börnstein New Series VIII/4
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The HRTM assigns gases and vapours to three classes: • Class SR-1 (soluble or reactive). Deposition may occur throughout the respiratory tract. Retention in respiratory tract tissues and uptake to the systemic circulation may be less than 100 % of the inhaled activity, although this is the default assumption; e.g. tritium gas and tritiated methane, carbon monoxide. • Class SR-2 (highly soluble or reactive). Total deposition occurs in the extrathoracic airways (ET2). Subsequent retention in the respiratory tract and absorption to body fluids are determined by the chemical properties of the specific gas or vapour; e.g. tritiated water, organically bound tritium and carbon dioxide. • Class SR-0 (insoluble and non-reactive). Negligible deposition in the respiratory tract. External irradiation from submersion in the cloud of gas, and internal irradiation from gas within the respiratory tract. e.g. from all radioisotopes of argon, krypton and xenon (except 37Ar, 94I1). Subsequent retention in the respiratory tract and absorption to body fluids are determined by the chemical properties of the specific gas or vapour. ICRP Publications 68 and 71 as well as the guidance document [94I1, 95I2, 02I1] give information on the assignment of gases and vapours to these three classes, and for selected Class SR-1 compounds information on fractional deposition and subsequent clearance. As an alternative to any of the three default Types defined in ICRP Publication 66, very fast uptake to body fluids (Type V) may be recommended. Although consideration has to be given to the total respiratory tract deposition, regional deposition does not need to be assessed for such materials, since, for the purposes of dose calculation, they can be treated as if they were injected directly into body fluids. Examples are tritiated water and tritiated methane, methyl iodide and methane. 7.2.1.4 Dosimetry In the ICRP Publication 30 lung model, doses to the respiratory system were averaged over 1 kg of lung tissue and the energy of charged particle emissions in the TB, P and respiratory lymph node regions was assumed to be completely absorbed within the lung, i.e. the absorbed fractions were unity for charged particles. In the HRTM model, doses are calculated to several specific regions of the lung and account is taken of variations in radiosensitivity. Absorbed fractions are energy dependent and prescribed functions are given for all source and target combinations and particle types. The target cells identified for the assessment of doses are: basal cells of the epithelium in both extrathoracic regions; basal and secretory cells in the bronchial epithelium; Clara cells (a type of secretory cell) in the bronchiolar epithelium; and endothelial cells, such as those of capillary walls and type II epithelial cells, in the alveolar-interstitial (AI) region. The overall dose to the lung is then taken to be a weighted sum of the doses to the following regions: bronchial, bronchiolar, pulmonary and lymphatic with weighting factors of 0.333, 0.333, 0.333 and 0.001, (the sum is 1), respectively. These weights are known as regional apportionment factors to distinguish them from the tissue weighting factors used in the calculation of effective dose (Chapter 4). They represent the contribution from each region towards the total radiation detriment associated with irradiation of the lung.
7.2.2 Ingestion The model of the gastrointestinal tract (GI) presently used by ICRP to describe the behaviour of ingested radionuclides is that given in ICRP Publication 30 [79I1]. Radionuclides contaminating food or drink, or cleared from the lung by mucociliary action are swallowed, pass down the oesophagus and enter the gastrointestinal (GI) tract, which is treated as four compartments (Fig. 7.5). Absorption is usually described by f1 values which give fractional absorption into the systemic circulation (e.g. f1 = 1.0, absorption = 100 %, f1 = 0.01, absorption = 1 %). The transport of material through the GI tract is described in terms of movement through the four regions. Landolt-Börnstein New Series VIII/4
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Stomach (ST): Its contents are acidic, and little absorption takes place other than for very soluble radionuclides such as caesium or iodine for which absorption from the stomach is assumed to be complete (f1 = 1.0). All other radionuclides are assumed to be absorbed in the small intestine. The residence time for food in the stomach varies from minutes to hours depending on many factors − the amount and composition of food, exercise and emotions. In the dosimetric model the mean residence time is taken to be 1 h. Small Intestine (SI): The principle site of absorption. The contents are alkaline, so that elements which hydrolyse such as rare earths and actinides are not normally readily absorbed. The mean residence time is assumed to be 4 hours. Recommended f1 values are given in ICRP publications for specific radionuclides (e.g. 226Ra: f1 = 0.2, 144Ce: f1 = 0.0005, 239PuO2: f1 = 0.00001). Upper Large Intestine (ULI): Water is absorbed here from the semi-liquid contents. The mean residence time is taken to be 13 hours. Lower Large Intestine (LLI): This region acts as a store for food residues and often forms the critical organ for long-lived non-transportable ingested radionuclides. The mean residence time is taken to be 24 hours.
6WRPDFK 67
6PDOOLQWHVWLQH 6,
%RG\IOXLGV
8SSHU/DUJH ,QWHVWLQH
/RZHU/DUJH ,QWHVWLQH
Fig. 7.5. Biokinetic model for the gastrointestinal tract (based upon 79I1).
Table 7.5 gives the transit time and mass of the contents for the different regions of the GI tract that are assumed for dose calculations. Typical f1 values are given in Table 7.6 for some important elements.
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7 Internal dosimetry of radionuclides
Table 7.5 Regional masses and residence times in the gastrointestinal tract; [79I1]. Portion of GI tract that is the Mass of contents [g] critical tissue Stomach (ST) 250 Small Intestine (SI) 400 Upper Large Intestine (ULI) 220 Lower Large Intestine (LLI) 135 a Transfer rate
[Ref. p. 7-68
ICRP Publication 30 dosimetric model for the Mean residence time [days]
Ka days−1
1/24 2/24 13/24 24/24
24 6 1.8 1
Table 7.6 Examples of f1 values adopted by ICRP for workers; [94I1]. Radionuclide Chemical form f1 Tritium Tritiated water 1.0a Organically bound tritium 1.0a Cobalt Unspecified compounds 0.1a Oxides, hydroxides and inorganic 0.05 compounds Strontium Titanate 0.01 Unspecified compounds 0.3a Zirconium All compounds 0.002 Niobium All compounds 0.010 Ruthenium All compounds 0.05a Iodine All compounds 1.0a Caesium All compounds 1.0a Cerium All compounds 5 × 10−4 a Polonium All compounds 0.10 Radium All compounds 0.20 Thorium Oxide and hydroxides 2.0 × 10−4 Unspecified compounds 5.0 × 10−4 Uranium Unspecified compounds 0.02a Most tetravalent compounds, e.g. UO2, 0.002 U3O8, UF4 Plutonium Nitrate 1 × 10−4 a Insoluble oxides 1 × 10−5 a Unspecified compounds 5 × 10−4 a Americium All compounds 5 × 10−4 a Curium All compounds 5 × 10−4 a Chemical forms of radionuclides for which retention and excretion functions are given in Figures 7.15-7.25. The f1 values recommended for workers are not necessarily appropriate for food and drinking water. Moreover the absorption of radionuclides tends to be greater in the newborn although the results of animal studies suggest that gut absorption decreases as age increases, reaching adult f1 values by about the time of weaning in most cases. An expert group set up by the Nuclear Energy Agency (NEA) within the Organisation for Economic Cooperation and Development (OECD) [88N1] suggested f1 values to be used as average values for the first year of life. The expert group recommended that for fractional absorption values between 0.01 and 0.5 in adults, an increase by a factor of 2 be assumed for the first year of life; but for elements with a fractional absorption in adults of 0.001 or less, a value 10 times that of the adult should be assumed. This general approach has been adopted in the current ICRP documents when more specific data are not available.
Landolt-Börnstein New Series VIII/4
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In the dosimetric model for the GI tract doses are calculated to the radiosensitive mucosal cell layer. For low LET radiation (β particles and γ-rays) this is nominally taken to be one half of the average energy absorbed per gram of the contents and for high LET (α) radiation one two-hundredth. The Publication 30 model for the GI tract has a number of distinct limitations in use and although it has been used for calculating doses to infants and children it does not have specific age-dependent parameter values. For this reason a Task Group of ICRP is developing a new model for the Human Alimentary Tract (HAT). The proposed new model is illustrated in Figure 7.6.
Salivary glands
Oral cavity Pharynx
Teeth
Oeso Oeso phagus 1 phagus 2
Oral mucosa
Slow
Fast
B L O O D
Secretory organs
Stomach
Stomach wall
Hepatic artery
Small intest
SI wall
Liver
Left colon
LC wall
Right colon
RC wall
Sigmoid Rectum
SR wall
B L O O D
Portal vein
Fig. 7.6. Proposed structure for the new human alimentary tract model; [03M1].
The revision of the Publication 30 model was motivated by a number of developments: • The 1990 recommendations of ICRP introduced specific risk estimates and tissue weighting factors, wT for radiation-induced cancer of the oesophagus, stomach and colon, requiring dose estimates for each of these regions [91I1]. The Publication 30 model did not include the oral cavity, or the oesophagus and treated the colon as two regions – upper and lower large intestine (Figure 7.5). The new model for the alimentary tract will comprise the oral cavity, including the mouth, teeth, salivary glands and pharynx, the oesophagus, the stomach, the small intestine, including duodenum, jejunum and ileum, the large intestine, including ascending, transverse and descending colon, rectum and anal canal. • Since the development of the ICRP Publication 30 model, a considerable body of data has become available on the transit of materials through the different regions of the alimentary tract. These data have been obtained using non-invasive, mainly scintigraphic techniques and include studies of differences between solid and liquid phases, age and gender related differences and the effect of disease conditions. These data are being used to set default transit rates for the defined regions of the alimentary tract for the six age groups given in ICRP Publication 56 (Table 7.1) [89I1]. • Information has become available for morphometrical and physiological parameters and on the location of sensitive cells in different regions of the alimentary tract. • More information has become available concerning absorption, retention and transfer from different regions of the alimentary tract. • Extensive age-, gender- and health-dependent information is available.
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Development of the new model has been described [03M1]. The new HAT model will be applicable to children and adults under all circumstances of exposure. It considers the movement of radionuclides throughout the alimentary tract from ingestion to elimination. It takes account of sites of radionuclide absorption and retention in the alimentary tract and routes of secretion of absorbed radionuclides into the alimentary tract. Doses will be calculated for sensitive cells in each region: mouth, oesophagus, stomach, small intestine and colon. The new model is more detailed and morphological than the previous gastro-intestinal tract model [79I1]. The new model is physiologically based. It includes consideration of absorption in regions other than the small intestine when such information is available. The model can also be used for radiopharmaceuticals. The model provides the flexibility needed to calculate dose to the alimentary tract for a wide range of exposure conditions and for specific individuals. A gut transfer factor, equivalent to the f1 value, will take account of absorption from the small intestine and from other regions of the alimentary tract where information is available. An important development in the new model is the calculation of doses to sensitive cells in the different regions of the alimentary tract. The location of sensitive epithelial stem cells in the various regions is considered separately; that is for the mouth, oesophagus, stomach, small intestine and colon. Doses from radionuclides in the gut lumen, retained radionuclides and radionuclides in transit to blood are considered. It is expected that the report will be used as the basis for future dosimetric calculations for both ingested radionuclides and radionuclides passed through the throat and swallowed after inhalation.
7.2.3 Cuts and wounds The presence of cuts, abrasions, burns or other pathological damage to the skin may greatly increase the ability of radioactive materials to reach subcutaneous tissues and thence the blood and systemic circulation. Although much of the material deposited at a wound site may be retained at the site, and can be surgically excised, soluble (transportable) material can be transferred to the blood and hence to other parts of the body. These events occur only as a result of accidents, each event will, therefore, be unique and need to be assessed by occupational health physicists and medical staff. To date, ICRP has not given advice on the interpretation of wound monitoring data following accidents involving radionuclides. The biokinetic models that have been developed for various radionuclides are, however, applicable to the soluble component of any deposit in cuts or wounds that enters the blood circulation. Insoluble material will be slowly translocated to regional lymphatic tissue, where it will gradually dissolve and eventually enter the blood. A variable fraction of insoluble material can be retained at the wound site or in lymphatic tissue for the remainder of the individual's life. If particulate material enters the blood it deposits principally in phagocytic cells in the liver, spleen and bone marrow. The United States National Committee on Radiological Protection and Measurements (NCRP) has established a Committee to review the problem of wound dosimetry. The report it is preparing will contain an extensive compilation of human and experimental data on the behaviour of radionuclides at wound sites. Four default categories for wound retention of deposited material have been proposed as summarised below: • • • •
weakly retained (≤10 % retained at one day, 80 % retained in one day; >50 % at 64 days).
In addition default categories are also being considered for colloids, particles and fragments. In reviewing the experimental data available, various chemical forms of elements/radionuclides are being allocated to these categories on the basis of either the results of studies in experimental animals or their chemical characteristics [03G1]. Once the report is complete it will need to be reviewed by the NCRP and it is possible these categories may change. The ICRP is awaiting publication of the report before deciding on its future work in this area. Landolt-Börnstein New Series VIII/4
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7.2.4 Absorption through intact skin The intact skin provides an effective barrier against the entry of most radioactive materials into the body, exceptions of practical importance being tritium oxide as liquid or vapour, organic carbon compounds and iodine as vapour or in solution. No generalised model has been adopted by ICRP for estimating absorption of radionuclides through the skin although it would be possible to develop specific models. For example, the behaviour of tritiated organic compounds following direct absorption through the skin would be expected to be significantly different from that after inhalation or ingestion. For skin contamination, both the radiation dose to the area of skin contaminated and the dose to the whole body as a result of absorption need to be considered. ICRP [77I1, 91I1] has recommended that for skin contamination doses should be calculated to sensitive cells, assumed to be at a depth of 70 µm (as a reasonable average value). ICRP [79I1] addressed the uptake of tritiated water vapour by assuming the uptake is instantaneously distributed within body water in the same manner as the inhaled water vapour. That is, for airborne HTO vapour, the dose per unit uptake through the intact skin is the same as the dose per unit activity inhaled. For deposited activity doses are to be calculated as an average to each cm2 of skin tissue. This applies to activity distributed over the skin surface or aggregated in particles. No specific models are recommended by ICRP for calculating doses from β particles deposited on the skin (also see Section 8.2).
7.2.5 Systemic behaviour of radionuclides The fraction of an intake of a radionuclide entering the systemic circulation is referred to as the uptake. In Publication 30 [79I1, 80I1, 80I2, 86I1] ICRP reviewed biokinetic data for each element for use in the calculation of limits on internal exposure to radionuclides by workers for intakes by inhalation and ingestion. Element-specific biokinetic models were given for the distribution and retention of radionuclides following their entry into the blood. The ICRP 30 models applied specifically to workers and not to members of the public. More recently, Publications 56, 67, 69 and 71 have revised the biokinetic models for selected elements and these have been applied in the calculation of dose coefficients for both workers and members of the public [89I1, 93I1, 94I1, 95I1, 95I2] (Table 7.1). If a radionuclide that enters the blood is an isotope of an element that is required by the body then it will follow the normal metabolic pathways for that element (e.g. Na, P, K, Ca, Fe). If it has similar chemical properties to an element that is normally present then it will tend to follow the biokinetic pathways of that element, although its rate of movement between the various compartments in the body may be different (e.g. 90Sr and 226Ra behave similarly to Ca, 137Cs and 86Rb similarly to K). For other radionuclides their behaviour in the body will depend upon their affinity for biological ligands and other transport systems in the body and, as a result, the extent of uptake is unpredictable and must be assessed from the available human or animal data (e.g. 95Nb, 106Ru, 239Pu, 241Am). Radionuclides entering the blood may distribute throughout the body (e.g. 3H, 42K, 137Cs); they may selectively deposit in a particular tissue (e.g. 131I in the thyroid; 90Sr in bone) or they may deposit in significant quantities in a number of tissues (e.g. 239Pu, 241Am, 144Ce). Some examples of the behaviour of selected radionuclides are given below. Limited information is also given on methods of treatment for accidental intakes. More information on decorporation of radionuclides is given in Chapter 9. 7.2.5.1 Elements that distribute widely in body tissues Hydrogen Tritium labelled water (HTO), given either orally or by intravenous injection, is rapidly absorbed from the lungs and absorption from the gut is also essentially complete (f1= 1.0). HTO distributes throughout the body water and is subsequently lost from the body with a biological half-time of about 10 days as a result of excretion in the urine, sweat, faeces and via the lungs (i.e. about 7 % of the total body water is lost per day). The addition of HTO to the body water has been a standard method of determining total body water by isotope dilution. For example, following intravenous injection of 1000 kBq of HTO the activity in a urine sample 6 hours later was 20 Bq ml−1. The total body water is then: Landolt-Börnstein New Series VIII/4
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1,000,000 Bq = 50,000 ml (50 litres) 20 Bq/ml The total body water in Reference Man is 42 litres [75I1]. The rate of loss of tritiated water can be increased by increasing the fluid intake (see Chapter 9). Doubling the fluid intake from 2 to about 4 litres per day can reduce the half-time of tritiated water to about 5-6 days. This rate of loss of tritiated water is also found in countries with a hot climate, such as India. In practice, a small fraction of tritium in body tissues becomes incorporated into organic compounds − amino acids, carbohydrates, etc. − and is retained with a longer half-time. For adults, ICRP [89I1] assumes that this fraction is 0.03 (3 %) and is lost with a half-time of 40 days, while the half-time of 10 days applies to the remaining fraction of 0.97 (97 %). For members of the public, ingesting foods containing tritium, absorption of organically-bound forms and their incorporation into body tissues will lead to longer retention of a larger component. ICRP [89I1] assumes that in the adult, the half-times of 10 days and 40 days apply to equal fractions (0.5) of activity entering blood; values the same as those recommended for organically bound forms of 14C. Total body water =
Caesium 137
Cs, together with 90Sr, is a major component of nuclear fission. As a result of nuclear weapons testing Cs has been injected into the atmosphere and weapons fallout has resulted in the contamination of the food chain and man [77U1]. The first observation of the presence of weapons fallout 137Cs in man was reported in 1956 [56M1]. Since then it has been shown to be present in everyone as a result of contamination of the environment by nuclear test explosions, routine releases from nuclear sites and by the accident at Chernobyl in 1986. 137Cs may enter the body either by inhalation or through the foodchain. Absorption from the gut is almost 100 % (f1 = 1.0, see Section 7.2.2). Once inside the body caesium ions (Cs+) behave very similarly to potassium ions (K+) and are rapidly taken up by cells. There is a considerable concentration factor between the cells and plasma. Generally tissue/plasma concen-tration ratios are the same for K+ and Cs+ but there are a number of exceptions and some variations between different species. In particular, muscle accumulates Cs+ more effectively than K+ and is the main site of long term deposits in the body. At equilibrium the muscle accounts for more than 50 % of the total body 137 Cs in man and bone about 8 %. Accumulation of Cs+ by cells is both by diffusion and by the ion pump that normally accumulates K+. There is a continual turnover of Cs+ in body tissues. For the purposes of dosimetry the retention of 137Cs in man is taken to have two components [89I1]. The first accounts for about 10 % of the administered activity and is excreted mainly in the urine with a half-time(T½) of about 2 days. The second component (90 %) has a half-time of about 110 days in males (range about 50 to 150 days). The long half-time mainly reflects the slow turnover rate in muscle tissue. The retention half-time for the long-term component in females is less than in males, with a mean value for adults of about 60 - 65 days. The use of the ICRP value of 110 days is therefore likely to be conservative for adult females. In children the half time is less than in adults. Thus for a 5 year-old the half- time is taken to be 55 days [89I1]. In cases of accidental intakes, Prussian Blue (ferric ferrocyanide) can increase the rate of excretion of 137 Cs from the body (see Chapter 9). If Prussian Blue is ingested it will accumulate any caesium secreted into the gut preventing it being re-absorbed. The half-time of retention can be reduced to about 40 days; the rate of loss is dependent upon the turnover rate of 137Cs in tissues and its loss into the gastrointestinal tract. 137
Ruthenium 106
Ru is also produced in nuclear fission. Its absorption from the gastrointestinal tract is quite low; the value for gut absorption (f1) is taken to be 0.05. The distribution of 106Ru in mice, rats, monkeys and dogs is fairly uniform throughout all tissues after an initial period during which the kidneys contain the highest concentration. The animal data have been used by ICRP [89I1] to define a retention function for Landolt-Börnstein New Series VIII/4
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ruthenium in man. Any ruthenium entering the blood is assumed to be distributed throughout the body with retention components of 35 % (T½ = 8 days); 30 % (T½=35 days) and 20 % (T½=1000 days). The remaining activity entering the blood (15 %) is taken to be promptly excreted. Because of the relatively short retention time in the body of most of the intake and a physical half-life of 368 days, the effective dose from ingestion of 106Ru is dominated by the radiation dose to the large intestine from beta-particles emitted while it passes through the GI tract. 7.2.5.2 Elements that deposit mainly in particular organs or tissues Iodine Radioactive isotopes of iodine are important components of nuclear fission, particularly in the first few days and weeks after a release into the atmosphere. If taken into the body they are accumulated by the thyroid gland, as was demonstrated particularly after the accident at Chernobyl. The most important isotope is 131I which has a physical half-life of 8.04 days. Radioactive isotopes of iodine are also widely used in medicine. They are used to demonstrate changes in thyroid function, to treat hyperthyroidism or to kill tumour cells in the treatment of thyroid cancer. The thyroid gland consists of a bi-lobed body in the neck region. It produces the hormones thyroxine and tri-iodothyronine which are important for regulating the body's metabolic rate. Disorders of the gland can result in either an under- or over-active gland (hypo- or hyperthyroidism). The gland weighs about 20 g in the healthy adult (2 g in the newborn child) and is made up of 20 to 40 million spherical vesicles (follicles) per lobe. Each follicle is surrounded by a single layer of cuboidal epithelial cells (acinar cells) lying upon a basement membrane and in proximity to numerous blood capillaries. The vesicles are filled with a structureless semi-fluid material − the so-called “colloid” which contains the active component of the gland (a protein-storage form of hormone). When the thyroid is quiescent, colloid is abundant and the follicles large (about 300 µm in diameter). When the thyroid is active colloid is scanty and follicles small (about 50 µm in diameter). The gland contains about 10,000 µg of iodide in the average normal adult. For adults in Europe about 225 µg of stable iodide enters the extracellular (iodine) space from the diet each day, absorption occurring across the small intestine within 1-2 hours. About 70 µg of iodide per day is trapped by the thyroid and converted to thyroid hormones while most of the rest is excreted in the urine. The amount of ingested iodide that is taken up by the gland is thus about 30 % [83S1, 87S1]. The fractional absorption varies between different individuals and there are significant differences between countries because of varying levels of stable iodine in the diet. For the purpose of dosimetric modelling ICRP has recommended that the uptake of radioiodine by the gland should be taken to be 30 %. The iodide synthesised into hormone leaves the gland with a half-time of about 80 days (adult) and enters other tissues. From this source most of the iodine (about 80 %) is metabolised back to free iodide with a half-time of about 8 days and re-enters the iodide space, the rest is excreted in the faeces. In adults the total amount of stable iodine excreted is approximately equal to the amount absorbed. Iodide or elemental radioactive iodine (e.g. 131I) may be ingested (f1 = 1.0) or the volatile compounds inhaled. As the gland is small, and it takes up about 30 % of radioiodine entering the blood, the concentration of radioiodine in the gland, and hence the radiation dose, is more than a thousand times that to other tissues. The turnover of stable iodine, and hence radioiodine, is low (T½ = 80 days in adults) and thus short-lived isotopes (e.g. 131I T½ = 8.0 days) will decay mainly in the gland rather than being returned to the blood. In children, although the turnover rate is faster the mass of the gland is smaller and hence for a similar intake of radioiodine the dose can be higher. Milk consumption is the most important pathway for the uptake of radioiodine from the human food chain after a release of radioiodine into the environment and children have a high consumption of milk. As a consequence children are the most sensitive (critical) group following such a release. Various drugs have been used to reduce the uptake of radioiodine into the thyroid gland after an intake. The safest and most effective procedure is to administer a large single oral dose (20-200 mg) of potassium iodide or iodide (see Chapter 9). It is effective within an hour and reduces the subsequent uptake of radioiodine into the gland. The daily intake is suddenly increased from about 225 µg to Landolt-Börnstein New Series VIII/4
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20-200 mg. However, the uptake of iodide into the thyroid remains at about 70 µg d−1 so that only a small fraction of the stable iodine and hence radioiodine present in the iodide space is then transferred to the gland. Thus, if 20 mg of stable iodide is given promptly the fractional uptake of radioactive iodine by the gland can be reduced to less than 0.001 (0.1 %). Since the half-time of iodide in the iodide space is about 10 hours, with rapid uptake by the gland or excretion from the body, the effectiveness of administering large amounts of stable iodide diminishes the greater the delay; after 48 hours, it is of little value at all. It has no effect on any radioiodine that has been taken up by the gland. Administration of stable iodide is the treatment of choice for accidental intakes of radioiodine and is recommended by the World Health Organisation [99W1]. Alkaline earth elements The bone is a highly specialised form of connective tissue. It consists of a soft organic matrix of collagen and ground substance in which is deposited calcium hydroxyapatite (Ca10 (PO4)6 (OH)2). The cells of bone consist of a proliferating population of stem cells which differentiate into: osteocytes − the cells responsible for bone maintenance; osteoblasts − the cells responsible for bone formation; and osteoclasts − the cells responsible for bone removal. There are two types of bone: • Hard cortical bone, which makes up about 80 % of the bone mass and which is penetrated by blood vessels and Haversian systems; and • Trabecular bone or spongy bone which makes up the remaining 20 % of the bone mass. The trabeculae are generally 100-200 µm in diameter and usually do not contain blood vessels. The spaces between the trabeculae are large (up to 1000-2000 µm) and contain the vascular (red) marrow. The surface area of trabecular bone is estimated to be about 4 times that of cortical bone [79I1]. The surfaces of bone are covered with non-mineralised layers of connective tissue. This is the periosteum on the external bone surface and the endosteum on the internal surface. The processes of bone turnover, resulting from the laying down of new bone and removal of old bone, continue throughout life although slowing down with increasing age. Calcium is thus an important component of the skeleton and other alkaline earth elements can be substituted for it in the bone matrix. A number of substitutions are possible in the lattice structure without disturbing the symmetry of the crystal lattice. ICRP has developed age-dependent biokinetic models for the alkaline earth elements (Ca, Ba, Sr and Ra) [93I1]. These models allow for the recycling of radionuclides between the skeleton, blood and soft tissues. They are physiologically based and take account of bone growth and remodelling as a function of age. They can also be applied to the interpretation of bioassay data. These models are essentially dynamic in nature. Material initially deposited on bone surfaces may be buried in bone volume by the formation of new bone or resorbed to red bone marrow. Activity in marrow is either locally or systemically recycled back to bone surfaces, and is also replenished by resorption from bone volume. A fraction will also be excreted from the body. The rate at which some of the processes operate can depend on the type of bone, as well as age, and allowance is made for this by division into cortical and trabecular compartments. This division also accommodates age-dependent dosimetry − in adults most of the active marrow is associated with trabecular bone, whereas for children this is more evenly distributed between both types of bone. 7.2.5.3 Elements that deposit in a number of tissues Plutonium, americium and curium Plutonium and other higher actinides are produced in thermal reactors, although the relative amounts generated depend upon the irradiation time. At low irradiation times almost pure 239Pu\240Pu are produced (subsequently given as 239Pu). As the irradiation time increases other isotopes of plutonium (e.g. 241Pu) Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
7-19
and higher actinides (e.g. 241Am, 242Cm, 244Cm) are produced. 239Pu is the isotope that has been processed in greatest quantity and for which most biological data are available. 238Pu is also an important isotope and is used as a power source in satellites and in cardiac pacemakers. Plutonium metal is highly reactive, oxidising in moist air if present in a finely divided form. The oxide is chemically inert and insoluble, particularly if produced at a high temperature (1000 °C). Am and Cm metals readily react with oxygen to produce oxides which are much more soluble than PuO2. In the processing of nuclear fuel soluble nitrates and other soluble complexes are formed. The behaviour of plutonium in the body depends upon its chemical form. Plutonium dioxide is chemically inert and largely insoluble in biological fluids, particularly if produced at a high temperature. Soluble plutonium compounds (e.g. plutonium nitrate, plutonium citrate) readily hydrolyse at the pH of biological fluids. Following hydrolysis, there is a strong tendency to polymerise, forming a colloidal insoluble compound. Alternatively, soluble compounds can react with naturally occurring complexing agents in body fluids which can readily move around the body. Which of these reactions predominates depends upon the site of entry of plutonium. If into blood, then complex formation is likely; if into a wound or the lung, then colloidal polymers are likely to be formed. The main route of entry of plutonium into the body is by inhalation, although it can also enter through cuts and wounds. In adults very little is absorbed from the gut (ICRP has adopted gut absorption, f1, values of 5 × 10−4 for soluble plutonium compounds and plutonium in foodstuffs, 1 × 10−4 for plutonium nitrate, and 1 × 10−5 for plutonium oxide, 239PuO2) or across the intact skin. For americium and curium ICRP has adopted an f1 of 5 × 10−4 for all compounds [94I1, 96I1]. After inhalation, deposition of plutonium in the lungs is determined by particle size as detailed in Section 7.2.1.1. Subsequent clearance from the lungs depends upon its chemical form. Whatever the chemical form inhaled, a fraction, consisting of any soluble material will be rapidly transported to blood and this is excreted through the kidneys or deposited in tissues (mainly the liver and skeleton). The remaining material, consisting of colloidal polymers or material with a low solubility (e.g. PuO2), is initially retained in the lungs. Material retained in the lungs is largely taken up by scavenger cells (macrophages) in the lungs. These cells may migrate to lymph nodes or reach the muco-ciliary escalator, be swallowed and excreted in the faeces. Alternatively materials in macrophages may gradually dissolve and translocate to blood. In studies with experimental animals retention half-times of the long-term component of retention have varied from 100 to 1000 days or more although soluble compounds are cleared more rapidly. The relative proportions of the long and short retention components depend upon the material initially deposited. For example, in the case of a polydisperse aerosol of high temperature calcined plutonium dioxide deposited in the lungs the amount rapidly moving to blood is normally less than 0.4 % and would be treated as default absorption Type S in the HRTM (Section 7.2.1.2). For a plutonium nitrate aerosol it may be greater than 20 % and would be treated as Type M. Compounds of americium and curium, particularly the oxides, are more soluble than plutonium compounds in the respiratory system and more readily absorbed. They would generally be classified as Type M but any americium or curium trapped in insoluble particles of PuO2 would behave as Type S. The principal sites of deposition of Pu in the body after translocation to the blood are the liver and skeleton. This is also the case for Am and Cm. All three actinides are classified as bone-surface seeking elements. That is, they are assumed to be uniformly distributed on endosteal bone surfaces of cortical and trabecular bone after their deposition in the skeleton. In practice, surface deposits of Pu, Am and Cm have been shown to be progressively buried by the formation of new bone. In addition, activity is lost from bone surfaces during resorption and some transfer to bone marrow takes place, particularly for Pu. In ICRP Publication 67, a dynamic model has been adopted to describe their behaviour in the body and to take account of their movement in bone as well as between tissues and excretion (Fig. 7.7 [93I1]).
Landolt-Börnstein New Series VIII/4
7-20
7 Internal dosimetry of radionuclides
Other soft tissue
Intermediate turnover (ST 1)
Slow turnover (ST 2)
Rapid turnover (ST 0)
Skeleton
Liver Cortical surface
Cortical marrow
Trabecular surface
Trabecular marrow
Cortical volume
Liver 2 Liver 1
Blood Trabecular volume
[Ref. p. 7-68
GI tract contents Kidneys Urine
Urinary bladder contents
Other kidney tissue
Gonads
Faeces
Urinary path
Fig. 7.7. Diagram of the biokinetic model for plutonium and americium; [93I1].
Comparisons of relative skeletal retention in the biokinetic model are complicated by the compound structure of the skeleton and recycling between compartments which gives rise to several retention components. However, calculation shows that after short-term losses are complete it is possible to discern an effective retention half-time in the skeleton of nearly 100 years (Fig. 7.8). The figure also gives the retention of plutonium in the skeleton for 3-month-old infants and 10-year-old children. Because of the faster turnover of bone at younger ages, the initial uptake of plutonium by the skeleton is greater although the rate of loss is also faster. Fig. 7.9 gives comparable data for the liver. Although there are differences in the initial uptake, reflecting the effect of age and deposition in the skeleton, overall liver retention is similar for all age groups. The peak in uptake at 5-10 years reflects the uptake of activity lost to the blood from the skeleton. 60
60
Infant Age 10y Adult
50
Infant Age 10y Adult
Injected Pu in liver [%]
Injected Pu in skeleton [%]
80
40
40 30 20
20
10
0
10
30 20 Years after injection
40
50
Fig. 7.8. Model predictions of the plutonium content of the skeleton as a function of age and time after injection.
0
10
30 20 Years after injection
40
50
Fig. 7.9. Model predictions of the plutonium content of the liver as a function of age and time after injection.
Intravenous injection of the complexing agent diethylenetriaminapentaacetic acid (DTPA) as the calcium zinc salt is the only accepted therapeutic method for removing soluble forms of actinides from the body. It forms chelate complexes with these actinides which can be excreted in the urine and hence Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-21
can effectively clear them from the systemic circulation and some that has recently deposited in bone and other tissues. It is unable to remove intracellular deposits or activity that has been buried in bone and must, therefore, be administered soon after an intake. It can remove some soluble complexes from the lungs. Local injection of DPTA into contaminated wounds can remove more soluble forms of plutonium from the body than the same amount given intravenously, provided the DPTA completely infiltrates the wound site. Further information on DTPA treatment is given in Chapter 9.
7.2.6 Excretion In the biokinetic model described in ICRP Publication 30, no specific information was given on excretion in urine and faeces, although the models were used in Publication 54 [88I1] for interpreting excretion data. In the 1990 recommendations of the ICRP [91I1], however, the urinary bladder and the colon are given explicit tissue weighting factors wT (see Section 7.3.1) and the revised biokinetic models given by ICRP now give specific information on excretion pathways in urine and faeces [93I1]. For assessing doses from systemic activity lost into the faeces, the model for the gastrointestinal tract is used (Section 7.2.2) assuming secretion of radionuclides from the blood into the upper large intestine. A model for the urinary bladder has been adapted for calculating doses to the bladder wall [93I1]. The bladder is taken to be of fixed size containing 15, 25, 65, 75, 85 and 115 ml of urine in the 3-month-old, 1-, 5-, 10-, 15-yearold children and adults, respectively. These volumes represent the average content of the bladder during the time period between voids. The rate at which radionuclides enter the bladder is based on their elimination rates from body tissues and the urine to faecal excretion ratio adopted for the biokinetic data for each element. For some elements, the biokinetic data directly address excretion. The number of voids per day for the 3-month-old and 1-year-old are taken as 20 and 16, respectively. For all other ages, 6 voids per day are assumed.
7.2.7 Embryo and foetus During pregnancy, radionuclides that have entered the mother's body, either before or after conception, can irradiate the developing embryo and foetus. The radiation dose to the offspring will depend upon a number of factors. These include: • the transfer of radionuclide(s) to the developing offspring from maternal blood and from deposits in the tissues of the mother; • the distribution and retention in foetal tissues; • the physical half-life and formation of decay products; • growth of the offspring; and • photon irradiation from radionuclides in the placenta and maternal tissues. Radiation doses will also be received by the newborn child from radionuclides retained at birth. During the foetal period of development, radionuclides can cross the placenta to reach the tissues of the embryo and foetus from the maternal circulation. The processes involved in this transfer may include simple diffusion, facilitated transport, active transport, movement through pores and channels, and pinocytosis [83S1, 87S1]. Most human data on the placental transfer of radionuclides are available from studies with labelled metabolites, radiopharmaceuticals and other radionuclides used in nuclear medicine, although some data are also available for radionuclides in fallout from weapons testing or for radionuclide releases into the environment as a result of nuclear accidents (e.g. 90Sr, 131I, 137Cs). Analysis of autopsy samples has also given information on both naturally occurring and artificially produced radionuclides. Information is additionally available on levels of stable elements in the placenta and foetal tissues that can be compared with those in the adult. The rather limited amount of human data available has made it Landolt-Börnstein New Series VIII/4
7-22
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
essential to use the results of animal studies in the development of dosimetric models for man, although even here information can be very limited for many elements. Chemical analogy is also of value in model development. ICRP has issued Publication 88 [01I1] giving dose coefficients for the embryo and foetus following intakes of radionuclides by the mother. It covers selected radionuclides of the 31 elements covered in Publications 56, 67, 69 and 71 (Table 7.1) and applies to the offspring of both members of the general public and workers. In the development of biokinetic and dosimetric models, two approaches have been used. Where sufficient information is available, element-specific models have been given. This applies, for example, to tritiated water, caesium, iodine and the alkaline earths. When appropriate human data are not available, animal studies have provided the main basis for model development using a generic modelling approach. It has been assumed, in the absence of more specific information, that the dose to all tissues of the embryo, taken to be up to the end of 56 days after conception, (i.e. the end of the second month of gestation), can be approximated by the dose to the maternal uterus. All organs and tissues of the developing embryo thus receive the same dose. The general approach that has been adopted for calculating equivalent doses to the organs and tissues of the developing foetus from experimental studies in animals is to use average concentrations of a radionuclide in the foetus (CF) and mother (CM) obtained shortly after injection. Where possible the value of the CF:CM concentration ratio adopted has been based on results obtained in a number of different animal species. The total activity transferred to and retained in the foetus from 57 days of gestation to birth at 38 weeks (266 days) is calculated for each radionuclide from the CF:CM ratio. This ratio at the time of the intake is assumed to stay constant for the remaining period of the pregnancy. This is taken to be a conservative assumption. Some examples of (CF:CM) concentration ratios adopted are given in Table 7.7. The concentration ratio may depend upon the time of the intake in relation to the start of the pregnancy. Thus for an acute intake of plutonium at any time before pregnancy the CF:CM ratio is taken to be 0.03; this ratio is then maintained at this value through the period of gestation. For an acute intake during the first trimester of pregnancy, however, a ratio of 0.1 is adopted, subsequently increasing to 0.3 for an intake at the end of the second trimester (180 days) and 1.0 for an intake at term (266 days). Again this ratio is assumed to be kept constant over the remaining period of the pregnancy. Table 7.7 Concentration ratios for elements in the foetus and mother (CF:CM) following intakes by the mother before or during pregnancy and corresponding ratios for the placenta (CPl:CM)
Element H in HTO Organic carbon Sulphur Zinc Zirconium Ruthenium Caesiuma Cerium Plutonium Americium
CF:CM Intakes prior to pregnancy 1.6 1.5 1 2 0.2 0.01 1 0.01 0.03 0.01
Intakes during pregnancy 1.6 1.5 2 2 0.2 0.2 1 0.05 0.1/0.3/1b 0.1
CPl:CM 1 1.5 2 1 1 0.1/2c 1 0.1/1c 0.1/5c 0.1/2c
a Half-time of long-term component in mother during pregnancy taken to be 50 days b Intakes in 1st and at the end of the 2nd/3rd trimester (see text) c Intakes before/during pregnancy Dosimetric models were developed by ICRP that allowed for the calculation of doses to the embryo and to foetal tissues from radionuclides deposited either in the tissues of the embryo/foetus, in the placenta or in the mother. To provide data that could be used for assessing a range of possible intake scenarios dose coefficients have been given for acute and chronic intakes by the mother at various times Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-23
both before conception or during pregnancy. Dose coefficients for the offspring following ingestion of radionuclides by the mother are given for a range of f1 values in Publication 88, while dose coefficients for the offspring after inhalation of radionuclides by the mother are given for both 1 and 5 µm AMAD aerosols (the default particle sizes for members of the public and workers, respectively) and appropriate lung absorption Types. For acute exposures, intakes of radionuclides are taken to occur at the time of conception and after 5, 10, 15, 25 and 35 weeks of the pregnancy and at 6 months and 2½ years before conception. For chronic exposures, intakes are taken to occur during the year of pregnancy, starting from conception and for 1 year or for 5 years prior to conception. This range of intake scenarios was selected to allow doses to the offspring to be calculated for any pattern of intake by the mother. Equivalent doses to the date of birth have been given in Publication 88 for the brain, for the most sensitive 8-15 weeks of gestation, and for the tissue receiving the highest dose. The effective dose to birth has also been given using the wT values recommended by ICRP in Publication 60 [91I1]. Whilst these values are not strictly appropriate for exposures in utero, they have been used as no alternative weighting factors are available and the calculation of effective dose provides a useful quantity for comparison with doses to the reference adult. Effective doses (to age 70 years) received after birth are also given, together with the total committed effective dose (before and after birth) received by the offspring. The total committed effective dose to the offspring, eoffspring due to maternal intake of the radionuclide is the sum of the effective dose received during the in utero period, ein utero and the committed effective dose during the subsequent 70 years of post-natal life, epostnatal. That is: eoffspring = ein utero + epostnatal
(1)
where ein utero =
∫
8w 0
h&uterus (t )dt +
∑w ∫ T
T
38 w 8w
h&T (t )dt
(2)
and e postnatal =
∑w ∫ T
T
70 y birth
h&T (t )dt
(3)
where the limits of integration in the first two integrals are in weeks and in the last term is in years. The value h&T is the equivalent dose rate to individual tissues of the offspring during foetal life and after birth. In the case of the embryo the dose to the tissues of the uterus is taken as a surrogate for the dose to the embryo. In conjunction with the preparation of ICRP Publication 88 a CD-ROM has been issued which gives much more detailed information than in the publication [02I2]. In addition to the doses given in Publication 88 it provides equivalent doses to 15 organs and tissues in the foetus as well as an average equivalent dose to the remainder tissues. It also gives doses to the offspring to a number of times after birth (10, 20, 40, 70 years). Additionally, doses have been given for a range of ten inhaled particle sizes. Publication 88 gives only dose coefficients to the offspring and no information is provided on comparative doses to the adult. Such a comparison has, however, been published [02S1]. The main findings are that, in general, doses to the offspring are similar to or less than those to the reference adult. For a few radionuclides the dose to the offspring can exceed that to the adult. This applies to 3H, 14C, 35S and 59Fe, to radioisotopes of I and to radioisotopes of the alkaline earth elements including 45Ca, 89Sr, 90 Sr, 224Ra and 226Ra. For radioisotopes of iodine and the alkaline earth elements, the greatest doses result from intakes during the last trimester of the pregnancy when there is the greatest foetal demand for iodine and calcium. Whilst in most cases the doses to the offspring for the radionuclides covered in Publication 88 exceed those to the reference adult by a factor of about 2 to 3, in the case of some bone-seeking radionuclides the difference can be around a factor of 10 for intakes of short lived isotopes towards the end of the period of pregnancy. Some illustrative dose coefficients for the offspring following inhalation of 137Cs by the mother (as a member of the public) are given in Table 7.8. In this case the doses to the Landolt-Börnstein New Series VIII/4
7-24
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
offspring are highest for intakes early in the pregnancy as they reflect doses received by the mother over the period of the pregnancy. For intakes later in the pregnancy doses are lower as, although some 137Cs will be retained in the newborn child, the half-time of retention is less than in the mother. Table 7.8 Dose coefficients [Sv Bq−1] for the offspring from acute intakes by inhalation of 137Cs (T ½ = 30 y) by the mother, as a member of the public (AMAD = 1 µm, absorption Type F, f1 = 1.0). Reference adult = 4.6 × 10−9 Sv Bq-1 Scenario [weeks]a −130b −26 0c 5 10 15 25 35 a b c d na
Highest organ dose (in utero)d 7.0 × 10−13 6.2 × 10−10 2.5 × 10−9 2.4 × 10−9 2.3 × 10−9 2.2 × 10−9 1.7 × 10−9 6.1 × 10−10
Brain (8-15 weeks) 1.9 × 10−13 1.7 × 10−10 6.7 × 10−10 1.1 × 10−9 9.4 × 10−10 na na na
ein utero 7.0 × 10−13 6.2 × 10−10 2.5 × 10−9 2.4 × 10−9 2.3 × 10−9 2.2 × 10−9 1.7 × 10−9 6.1 × 10−10
epost natal 1.1 × 10−14 9.6 × 10−12 1.5 × 10−11 2.5 × 10−11 4.1 × 10−11 6.7 × 10−11 1.8 × 10−10 4.7 × 10−10
eoffspring 7.1 × 10−13 6.3 × 10−10 2.5 × 10−9 2.4 × 10−9 2.3 × 10−9 2.3 × 10−9 1.9 × 10−9 1.1 × 10−9
Intake at indicated time; negative times are prior to pregnancy −130 weeks = acute intake 2.5 years before conception 0 = acute intake at time of conception For 137Cs all tissues receive the same dose Not applicable
7.2.8 Transfer in maternal milk Models are presently being developed by ICRP for the transfer of radionuclides to mothers’ milk that will allow the calculation of dose coefficients for intakes by the offspring [03H1]. The publication will cover a review of biokinetic data relevant to an assessment of the transfer of radionuclides to breast milk following intake by the mother, the development of models, and the calculation of doses to the newborn child resulting from the transfer of radionuclides to milk after inhalation or ingestion by the mother. It is assumed that lactation lasts for a period up to 6 months after birth and that milk consumption increases to 800 ml d−1 over the first week and then remains constant to the end of lactation. Doses to the infant will be given for a range of intake scenarios. It is proposed that in the publication dose coefficients will be given for acute intakes by the mother at 26 weeks before conception, for intakes during pregnancy at 5, 15, and 35 weeks after conception and for intakes after birth at 1, 10 and 20 weeks of age. In addition doses from protracted exposures throughout pregnancy and lactation will also be included. These dose coefficients should give a sufficient amount of information to understand the implications for doses to the offspring for intakes at various times either before or after birth. Data for additional acute intake times and for chronic exposures as well as for inhalation of a range of particle sizes will be included on a CD-ROM. The dose coefficients for intakes by the 3-month-old infant given in previous publications [96I1] will be used to calculate the doses from the intakes by the suckling infant in milk. Some preliminary information is given in Table 7.9 (from 03H1) which gives the ratios of infant (offspring) dose to adult dose for chronic intake of various radionuclides throughout pregnancy and lactation. The values for lactation include contributions from activity retained in the mother from intakes during pregnancy as well as transfer to milk from intakes by her during lactation. The results of preliminary model calculations showed that intakes during pregnancy contribute an estimated 15 % of activity in milk for 137Cs, 210Po and 241Am, about 10 % for 45Ca and 90Sr, 4 % for 239Pu and negligible amounts for 131I. Doses to the infant from milk consumption are estimated to exceed adult doses in the cases of 45Ca and 131I. Very similar ratios of infant to adult dose are obtained when considering acute maternal intake by ingestion, during early lactation; that is, for maximum transfer to milk. Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-25
Also shown in Table 7.9 are ratios of dose to the offspring and adult from activity transferred to the embryo and foetus during pregnancy. These ratios are based on the dose coefficients for the offspring given in ICRP Publication 88 [01I1]. These offspring doses from in utero exposure include contributions from activity retained in the tissues of the child at birth, ranging from about 90 % of the total “foetus” dose for 239Pu to less than 10 % for 137Cs. Only in the cases of 131I and 210Po do the doses to the offspring from transfer in milk exceed that resulting from in utero transfer. Table 7.9 Comparison of doses following chronic maternal ingestion of radionuclides throughout pregnancy and lactation Ratio of offspring : adult dosea Radionuclide Foetusb Infant in milkc 45 Ca 12 2.7 90 Sr 1.5 0.8 131 I 1.0 2.4 137 Cs 0.4 0.4 210 Po 0.1 0.2 239 Pu 0.04 3ppm)
No effect on SCC susceptibility in 100 h decon, but 500 h decon. Increased on sensitised 304 SS. No susceptibility increase to SCC
304SS, Inconel 600 304SS, Low alloy steel, (SA533B) Inconel 600
PWR (DH: >3ppm) BWR (DO: 0.2ppm)
No susceptibility increase to SCC
304SS, Low alloy steel, (SA533B) Inconel 600 304SS, Inconel 600
BWR (DO: 0.2ppm)
Sensitized 304SS increased susceptibility to SCC in some cases.
BWR (DO: 0.2ppm)
Can-Decon and PNSCitrox increased susceptibility of sensitised 304SS to SCC
304 SS 316 NG SS
BWR (DO: 0.2ppm)
No susceptibility increase to SCC
304SS
BWR (DO: 0.2ppm)
No susceptibility increase to SCC
No acceleration on propagation rate of existing crack
Can-Decon increased 304SS SCC susceptibility. No effect with LOMI
Corrosion
Reporting organisation London 50µm IGA was observed Nuclear Ltd. on sensitised [85EPR] 304 SS. No descrip- Ontario tion Hydro [85EPR1] Pitting on 304 SS and 1.5µm IGA symptom on Inconel 600 No IGA
London Nuclear Ltd. [85EPR2] London Nuclear Ltd. [85EPR3] G.E. [86EPR]
Can-Decon caused 200µm IGA on 304SS and 100µm IGA on Inconel 600. No IGA with LOMI. G.E. 80µm IGA on 304SS and [86EPR1] 60µm IGA on Inconel 600 IGA were observed as the same extent as the above two columns No description
G.E. [86Man]
G.E. [86Man]
Slight IGA G.E. symptom was [86Man] detected in some cases
8-12
8 Decontamination
[Ref. p. 8-34 Existing Recirculation System
Vent UV Modules
Temporary Decon. Equipment
Spray Ring
Chemical Injection Tank Chemical Injection Tank
RPV Pump 20% speed Reactor Recirculation System Pump
CRD Housing Decon Loops: 200 m3/ h /each
Demineralizer
Main Circulation Pump
Heater
Cooler
Fig. 8.3 Fukushima 1 Unit 2 Flow diagram; [00Wil]
8.1.2.2 Fuel assemblies and decontamination Fuel surface crud usually contains a 100 to 1000 times higher inventory of radioactivity than recirculation system oxide layers. A decontamination reagent is easily decomposed by gamma rays and neutrons from the fuel even in reactor lay down periods. In spite of these difficult conditions fuel elements may have to be decontaminated for two reasons. In case the crud deposits cause too high pressure drops to achieve a homogeneous reactor coolant water flow cleaning is required. In addition, the treatment of spent fuel elements depends on a certain degree of cleanliness. The SGHWR (100 MWe, closed in 1991, UK) and Candu reactors often performed the decontamination of fuel assemblies in situ. The LOMI developed for this purpose has self-regenerating ability in the presence of radiolysis. Pressure-tube-type reactors are usually designed for on-power refueling. This requires more time than required for BWRs and PWRs to discharge all fuel from the core. In spite of that, Fugen (165 MW, Japan) a pressure tube type reactor, was decontaminated after discharging the fuel to obtain a higher decontamination factor and to reduce radioactive waste. NPP Paks found a thermal-hydraulic anomaly in the reactor core caused by corrosion product deposits. Consequently, the coolant flow through the fuel assemblies was insufficient resulting in a temperature asymmetry in the reactor core. The fuel assemblies were removed from the core and successfully cleaned applying the CORD UV process. 8.1.2.3 Decontamination of sodium cooled systems These systems may be decontaminated effectively by acid solutions. Within the research and development programme of the CEC an inorganic acid-based process has been evaluated and tested by CEA for the RAPSODIE reactor in France [89Cos]. Decommissioning operations began in 1987. Preliminary cleaning and water rinsing after isolation of the main vessel eliminated all traces of residual sodium. Main contamination nuclides were 137Cs, 63Ni and 54Mn. After steam-cleaning to remove the residual sodium, the specimens were highly rusted. One of the first reagents to provide satisfactory results was a mixture of nitric acid and sulphuric acid at 85 °C. In order to improve the effectiveness of the decontamination, the aggressiveness of the reagent was Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
8 Decontamination
8-13
enhanced by adding cerium (IV) in sulphate form. This reagent is sufficiently oxidising (Redox potential Eh = 1.610 V) iron, chromium and nickel in austenitic steels. Together with an alkaline washing excellent decontamination results were obtained: The estimated initial contamination level of 5500 Bq cm−2 was reduced to less than 10 Bq cm−2. The low residual contamination values allowed to estimate the pipes to be melted down for reutilization and release and an authorisation has been applied for. The dose rates were uniform throughout the facility, ranging from 1 to 15 µGyh−1 8.1.2.3 Gas cooled reactors (WAGR) An important feature of the Windscale AGR dismantling programme was the removal and disposal of the four heat exchangers [89Cro]. Each heat exchanger contains tube banks (or superheater banks), with plain Cr/Mo low alloy steel tubes and evaporator and economiser banks with finned mild steel tubes. Contamination was found to be incorporated in this oxide layer and to consist predominantly of 137Cs with some 134Cs and 60Co. Average values of 1.6 × 103 Bq cm−2 for Cs and 0.2 × 103 Bq cm−2 for Co were measured, giving a total for the superheater of about 2 ×1011 Bq. To remove the activity it was necessary to remove the oxide layer from the tubes. The decontamination was performed by spraying with a 3000-litre mixture of 0.5 molar (3.15 % by weight) nitric acid and 0.0025 molar citric acid at ambient temperature. In total the radiation level was removed by 70 %, activity were removed, the manSv-uptake was reduced remarkably and the targets reached.
8.1.3 Decontamination techniques for segmented parts 8.1.3.1 Chemical decontamination The chemical decontamination of an item removed from a nuclear plant or facility is generally carried out by immersion in a tank containing the chemical reagent [92Com]. The size of the tank depends on the dimensions of the item to be decontaminated. A common size is one, which is 1-2 m square with a depth of 0.5-1 m. Tanks for water rinsing are always installed. In sequential multistep processes the availability of several tanks can be useful in order to reduce the time needed. Chemical decontamination is characterised by the following parameters: • Type and nature of the chemical reagent; • Temperature of the process; • Duration of the process. The effectiveness of decontamination can be improved by increasing the duration of the treatment and the temperature. Optimum results are usually obtained with the solvents at elevated temperatures (up to 120 °C). During the decontamination process, as the concentration of the contaminants in the solution increases, the item being cleaned may become re-contaminated. This problem can be minimised by cleaning the least contaminated items first and by cleaning or replacing the solution if the concentration of contaminants exceeds certain levels. It should be noted that strong corrosive attack of the base metal may not result in high decontamination factors. These can however be achieved without significant corrosion of the base metal. Strong solutions of nitric and phosphoric acid used in the USA Bonus programme resulted in the removal of up to 0.2 mm of the inner wall of pipes, but only an average decontamination factor of approximately 100 was achieved. Results to date with non-aggressive processes indicate that decontamination factors as high as 2000 can be achieved without significant corrosion of the base metal [81Nel].
Landolt-Börnstein New Series VIII/4
8-14
8 Decontamination
[Ref. p. 8-34
Some multistep processes are commonly used for removing highly adhesive contamination layers. In many cases chemical decontamination can be used as a single step in complex processes e.g., before electropolishing, items covered with thick oxide layers are submitted to chemical decontamination in order to soften the oxide. 8.1.3.1.1 Chemical reagents Since the 1950s several chemical reagents have been commonly used for cleaning contaminated items. Lists with more than 100 chemical products can be read in dedicated decontamination handbooks [88Int, 81Nel, 82Com]. Some examples of chemical reagents used are given in Table 8.4. Table 8.4 Some reagents typically used in chemical decontamination of reactor components depending on type of material to be decontaminated. Materials Reagents Aluminium HNO3, Na-EDTA + 2 % detergent, 10 % citric acid, sulphamic acid …. Brass HNO3, 5 % AC Carbon steel Inhibited HCl, inhibited sulphamic acid, EDTA, citric + oxalic acids, APAC, CITROX, Copper Phosphorous, nitric and acetic acids Nickel & Alloys AP, 25 % HNO3 + 25 % HF, AP + AC, AP + CITROX, EDTA Monel 25 % sulphamic acid Stainless steel AP- CITROX, EDTA-CITROX, 30 % HF + 20 % HNO3, 70-80 % H3PO4, 0.4 M Cr SO4 + 0.5 M H2SO4, AP-HNO3, AP-AC, AP-OX, APACE Zircalloy OX + H2O2 + glucosic acid, EDTA, 8 M HNO3, … AP: alkaline permanganate, AC: ammonium citrate, OX: oxalic acid, CITROX: citric + oxalic acid, APACE: AP + AC + EDTA. There is a wider range of solvents to choose from for decommissioning programmes since corrosion of the base metal is of little concern. Certain solvents exhibit a time dependency in the mixing, heating, recirculation and draining cycle that affects both the chemical solution stability and the solubility of contained contamination. Each process under consideration would have to be evaluated for the effect of a loss-of-flow accident and associated cooling of the solvent. Factors considered would include toxic or explosive gas generation, excessive plate-out and excessive corrosion. The selected process must include appropriate emergency procedures, e.g. emergency draining, gas detection, and emergency ventilation. 8.1.3.1.2 Spent decontaminant solutions The selection of the chemical reagent directly affects the features of the secondary wastes arising from the process. It is obvious that continuously renewing the solution increases the decontamination effectiveness [85Pas] but the quantity of spent solution to treat and to dispose of also increase dramatically. In latter years the regeneration of chemicals have become a fundamental step in all chemical decontamination processes. Several conventional chemical processes can be used for regenerating the spent solutions either on their own or in combination and they include: ion exchange, evaporation/distillation or electrodialysis. The problem of limiting the secondary wastes arising from the decontamination process, sometimes leads to the selection of other similar processes like electropolishing or ultrasound with chemicals rather than chemical decontamination. As stated previously, only a detailed cost/benefit analysis can provide the actual criteria for selecting the best option for decontamination. Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
8 Decontamination
8-15
8.1.3.2 Electrochemical decontamination 8.1.3.2.1 In-Tank Electrochemical decontamination can be considered in principle to be a chemical decontamination assisted by an electrical field. Nevertheless the best electropolishing is a process widely used in nonnuclear industrial applications to produce a smooth polished surface on a variety of metals and alloys. It can be considered the opposite of electroplating as metal layers are removed from a surface rather than added as a coating. Usually the object to be electropolished is immersed in a tank of electrolyte and is used as the anode in the electrolytic cell. The passage of electric current results in the anodic dissolution of surface material and, for normal operating conditions, a progressive smoothing of the surface. A progressive dissolution of the surface material occurs within a certain range of voltage and current density [83Int]. If the voltage and current densities are too low, the surface is attacked non-uniformly, causing etching rather than polishing. Similarly voltages that are too high cause severe pitting of the surface [87Pav].
(+)
DC Power supply
( -)
Anode
Cathode Hydrogen Oxygen
Containment Tank
Electrolyte
aa aaaa aa
aa a Fig. 8.4 Electropolishing device Surface to be decontaminated
If the anode is a contaminated material such as metal or alloy, all radioactive contamination on the surface (or entrapped within surface imperfections) can be removed and released into the electrolyte by this surface dissolution process [85Pas]. The process produces a very smooth (0.02-0.03 mm), non-reactive and non-adsorbing surface resistant to recontamination during further operations. Experience has shown that electropolishing is an effective technique for removing both fixed and smearable radionuclide contamination. Moreover it is fast and easily controlled. In general, there are two methods of application for electropolishing. The most common method is immersing the item to be decontaminated in a tank filled with a suitable electrolyte. The second method involves the use of “in-situ” mobile devices that are able to electropolish part of the surface of the item, which, because of size or installation, cannot be electropolished in a tank. Phosphoric acid is normally used as the electrolyte in electropolishing because of its stability, safety and applicability to a variety of alloy systems. Moreover, the non drying nature of phosphoric acid helps minimise airborne contamination, and the good complexing characteristics of phosphoric acid for metal ions is a significant factor in minimising recontamination from the electrolyte. Representative operating conditions for decontamination using phosphoric acid electrolytes are: solution temperatures of 5 to 25 °C, phosphoric acid concentrations of 40 to 85 %, electrode potentials of 8 to 12 V and current densities of 5 to 25 A cm−2. Landolt-Börnstein New Series VIII/4
8-16
8 Decontamination
[Ref. p. 8-34
The direct current power supply converts alternating current in to direct current which generates the current flow between anode and cathode required for the electrochemical reaction. Voltage requirements range from 0 to 24 V with sufficient amperage to provide the required current densities. From experience gained in non-nuclear industrial applications, electropolishing normally uses phosphoric acid as the electrolyte (sulphuric acid is an alternative). However, during the last decade a variety of different electrolytes have been tested and developed with particular reference to the problem of reducing the secondary wastes arising from the process. As a consequence new processes usually called electrodecontamination or electropickling decontamination have been proposed. These consider the use of basic solutions as well as organic acid mixtures. Finally, among the electrodecontamination processes for “in situ” applications, electrobrushing using an electrobrush continuously fed by an electrolyte should be mentioned. Electrolytic decontamination can be used to remove fixed or imbedded contamination on iron-based alloys, including stainless steel, as well as on copper, aluminium, lead, and molybdenum. However the effectiveness of the decontamination can be limited by the presence of foreign materials on the surface of the items to be decontaminated. Materials such as oil, grease, oxide (rust), and paint or other coatings should be removed before decontamination. In decontaminating (mainly for decommissioning purposes) reactor coolant systems, the systems are usually covered by oxide layers that in principle work as a barrier for electropolishing. This problem can be overcome by increasing the electropolishing time. Nevertheless some new processes consider the periodic switching of polarity between cathode and anode, as well as changing voltage and current in order to increase the removal of the surface materials [86Gau]. Generally, at least two (stainless steel) tanks are required for performing electropolishing. One tank contains the electrolyte, electrodes, and parts to be decontaminated (as anode). The other tank holds the water used for rinsing the parts after decontamination. Power supply amperage capacities up to 2700 A are common. The cathode is normally a piece of copper, or stainless steel, positioned in the electrolyte within 30-100 mm from the item to be decontaminated. In addition for special items, the walls of the tank for immersion electropolishing can also serve as the cathode. To control vapours released from the electrolyte during the electropolishing process an extraction hood is located alongside the electropolishing tank. Provision for heating and agitating the electrolyte and rinse tank is also required. Studies on “in-tank” electropolishing became of relevance in the early 1970s in the USA where they were used to decontaminate hot-cells, glove-boxes, and tools contaminated by alpha emitters. Decontamination carried out in conjunction with Rockwell Hanford Operations and United Nuclear Industries in the USA, show that components heavily contaminated with PuO were decontaminated from 1 million dpm per 100 cm2 to background in less than 10 minutes [78All]. Typical decontamination times range from 5 to 30 min, corresponding to the removal of 10 to 50 mm of surface material at a current density of 2-15 A/dm−2. It is usually necessary to remove the anode contacts at least once during a cycle in order to decontaminate the area under the contacts. Since the early 1980s commercial use of electropolishing in decontamination of reactor coolant water systems and components for decommissioning purposes was made at the KRB power station (reactor A) in Germany [83Eic, 01Eic]. Electrochemical decontamination by electropolishing causes a steady increase of dissolved iron in the phosphoric acid. If the content of iron exceeds 100 g dm−3, a precipitation of iron phosphate occurs and this stops the efficiency of the decontamination process. Therefore the acid has to be exchanged or regenerated periodically. The regeneration of phosphoric acid is based upon the reaction of Fe2+ with oxalic acid (see Fig. 8.5). Electrochemical decontamination of steel, however, generates a high percentage of Fe in the phosphoric acid, which cannot be precipitated to iron oxalate. The high content of Fe3+ is reduced to Fe2+ by subsequent pickling. When a high portion of Fe2+ is obtained, the phosphoric acid has to be mixed with an aqueous solution of oxalic acid. The activity (mainly 60Co) is mostly separated from the solution by precipitation together with the iron. The iron oxalate is dried and stored for subsequent processing. The initial concentration of the phosphoric acid can be achieved by an evaporation process [89Sta].
Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
8 Decontamination Oxalic Acid
Decontamination of 1 Mg Steel Material + -
8-17
Iron Oxalate
Precipitation Thermolysis
H 3 PO 4 Chemical, Electrochemical
Vaporization
Waste: 15 kg Iron Oxide
Fig. 8.5 Flow chart of the phosphoric acid regeneration
Several electrolytes were investigated and proposed as alternatives to phosphoric and sulphuric acid. The need for new electrolytes was initially motivated by the incompatibility of phosphoric and sulphuric acids with the existing treatment facilities and by the possibility of generating secondary liquid wastes which are more easier to process. This regeneration process was extensively proved at KRB Gundremmingen in Germany where more than 200 m3 of phosphoric acid (concentration 20-40 %) have been regenerated. The iron oxalate can be converted to iron oxide by pyrolytic decomposition. With this method, KRB-A-Reactor materials like pipes, pumps, and housings were decontaminated. Decontamination Factors of more than 100000 were reached. 8.1.3.2.2 “In-situ” In the case of “in-situ” electrochemical decontamination, the surface of the item to be decontaminated is flooded with electrolyte through a gap between the cathode of the device and the item's surface. The inside of tanks, housings and other relatively open vessels, without internal components, can be decontaminated without removal using an expandable “bladder” with a conductive surface that serves both as the cathode and as a displacement device to minimise the electrolyte volume. Several devices have been developed in the USA for application of “in-situ” electropolishing [82Gar1]: The internal cathode device consists of a perforated, tubular, copper or stainless steel cathode section with insulator-spacers at each end and has the provision for pumping the electrolyte and feeding power through the insulator at one end. The perforated tubular section permits flow of the electrolyte to the pipe surface being decontaminated, thus accomplishing the electropolishing action. An improved device with four module heads was designed and used in Germany for the Obrigheim power plant [84KWU]. The pump stream device consists of a perforated, disc-shaped, copper or stainless steel cathode facing the surface to be decontaminated, with an insulated handle for flow of electrolyte and supply of power. The electrolyte flows out of the end of the device in a stream and impinges on the surface being decontaminated. 8.1.3.2.3 Electrobrushing Electrobrushing is an electrodecontamination process for selected areas. The item to be decontaminated is used as the anode, while an electrobrush serves as the cathode. The brush itself is a cellulose sponge wetted by a continuous feed of an electrolyte, such as 5 % sulphuric acid solution inhibited with 1 g dm−3 ethyl quinolinlium. Decontamination is carried out by scrubbing at a current of 15 to 40 A at 15 to 20 V, and decontamination factors of around 30 are reported at a rate of 0.6 m2h−1 [81Nel].
Landolt-Börnstein New Series VIII/4
8-18
8 Decontamination
[Ref. p. 8-34
The disadvantages of this process include the production of large volumes of aqueous radioactive waste and excessive attack of the surface by the electrolyte. In addition, if the decontamination is performed manually rather than remotely, the radiation exposure to operators may be high. Single electropolishing and brushing processes can be combined. Subsequently, in decontaminating cast steel components at the KRB Gundremmingen power station in Germany, it was found that brushing prior to electropolishing resulted in a 35 % reduction of the time required for the galvanic decontamination process and the reduction of dissolved iron in the electrolyte was also found to be a benefit of prebrushing [89Sta]. 8.1.3.3 Jetting decontamination techniques The impingement of either a liquid or a solid media (or a liquid solid mixture called slurry) can be successfully used for decontamination. Extensive use of jetting methods to clean surfaces and items has been made in many industries and for varied applications. As a result of this several decontamination methods have been studied, developed and as a result of this several contaminated materials. Generally, jetting decontamination processes have a high flexibility and can therefore be applied to both large surfaces e.g. floors and walls, and relatively small-contaminated items and systems. Contaminated glove-box internals and several types of tools have also been cleaned using jetting processes. The problem of amount and characteristic of secondary wastes is one of the main concerns for jetting processes. The amount of waste can be strongly reduced by recirculating, treating and rinsing the impinging jetting media. Particular care must be applied to processes using abrasives. A variety of nozzles and lance configurations can be used for high-pressure water cleaning depending on the configuration of the item to be decontaminated. A straight jet can be used on the end of a long handled lance to reduce worker exposures for decontamination of accessible tank interiors, walls or floors. Self-propelling mole nozzles on a flexible high-pressure lance or hose can be used to decontaminate the inside of tubing or pipes. Because high pressure water cleaning is very effective for the removal of smearable surface contamination, water lances have been successfully used to decontaminate pump internals, valves, cavity walls, spent fuel pool racks, reactor vessel walls and heads, fuel handling equipment, feedwater spargers, floor drains, sumps, interior surfaces of pipes and storage tanks. Although decontamination factors of up to several hundred are commonly achieved, the normal factor for most applications is up to 50. Decontamination factors of 2 to 50 with water as the agent and of 40 with the addition of a proper cleaning agent were achieved at a particular site [79Rem]. 8.1.3.3.1 Abrasive jetting Abrasive jetting is a very effective decontamination method in which an abrasive medium is propelled by a jet of air (dry blasting) or water (wet blasting) against the surface to be cleaned [82Gar, 81Rem]. Typical abrasives are sand, alumina and metals, metal oxides and sawdust. Sand is the most common abrasive because it is inexpensive and a good scouring agent. Abrasive cleaning can be wet or dry. Abrasive particles are impelled at high velocity against the surface to be cleaned by air, water or a mixture of the two, as in the following air abrasive blasting, water abrasive blasting, air slurry blasting. Alternatively the particles may be carried in a viscous matrix and rubbed against the surface (liquid honing or abrasive slurry cleaning) or the abrasive may be in the form of stones, which can be used to grind or hone the surface to be cleaned. For effective cleaning either highpressure air at about 1 MPa or water at pressures similar to those used in hydrolaser systems are used, depending on the application. The abrasive can also be attached to a flexible backing to form a type of sandpaper or emery cloth or it can be forced against the surface by centrifugal action.
Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
8 Decontamination
8-19
The different types of abrasive can be grouped into three general classifications depending on hardness: Hard abrasive materials: For decontamination, hard abrasive materials (harder than the material to be removed) are commonly used [64Ame]. In addition, there are three special sub-classifications: Cleanable abrasives, soluble abrasives and system compatible abrasives. The last one being important only for the decontamination of items required for reuse. A cleanable abrasive simplifies the problem of secondary wastes by markedly reducing the quantity of contaminated material to be disposed of after decontamination. One type of cleanable abrasive that can be used on non-stainless steel surfaces is steel shot. To facilitate cleaning the particles should be smooth; unfortunately smooth particles also reduce the abrasive action. A soluble abrasive is a material, which can be used in solid form in an air or water jet and can later be removed from the system by dissolving it in a liquid and flushing it out of the system. Boron oxide, which dissolves in water to form boric acid, is such a material. There are a number of methods of application of abrasive jetting, which depend on the carrier fluid, flow velocity and kind of abrasive used. Abrasive blasting is often carried out in special cabinets [91Bru]. A specific case of abrasive blasting is dry ice blasting which uses ice pellets (2-3 mm in diameter) produced by CO2 flashing at 40 °C. The use of dry ice pellets as an abrasive media strongly reduces the volume of secondary wastes produced [70Ayr]. The vacuum blasting method is a modification of air abrasive blasting for “in-situ” use where the discharge nozzle is surrounded by a concentric hood. An air exhaust line is attached to the hood and the blast air, debris and spent abrasive are drawn out through the exhaust line. The debris and abrasive are separated and the abrasive is reused. Smaller, hand-held units are also available. Water abrasive blasting has been used successfully to decontaminate a wide variety of contaminated components. Methods of remote application have been developed for “in-situ” cleaning to supplement the more common manual application methods. Decontamination factors of 200 to 300 are commonly achieved. 8.1.3.3.2 Freon jetting Systems have been developed to remove loose contamination from surfaces and equipment using commercial freon (trichlorotrifluoroethane) cleaning solvents [88Int]. Freon has a low viscosity and surface tension, which allows it to penetrate into cracks and crevices and remove contamination, including that associated with grease, oil, etc. Freon is non-flammable and chemically inert and can therefore be used to clean many types, of equipment without damaging delicate components. Most radioactive contaminants are insoluble in freon and can be removed by filtration or distillation, allowing recycling of the freon. Freon decontamination is carried out by directing a high pressure (15 MPa) jet of the liquid onto the surface to be cleaned. The decontamination is usually carried out inside a glove box [81McV], but experimental units have been developed for “in-situ” cleaning. The freon liquid and particles of contaminant are collected in a sump; the liquid is then filtered to remove the contaminants, cleaned and recycled. The freon is distilled as required to remove any radioactive material, which has dissolved in the liquid. The use of this decontamination technique is often limited due to legislative and regulatory restrictions in the industrial use of freon and freon-compounds due to their potential effect on the environment. 8.1.3.4 Ultrasonic decontamination Ultrasound consists of longitudinal mechanical waves and has been used over a long period of time for cleaning dirty surfaces in non-nuclear industry. In particular, ultrasonic cleaning has been used with good success for removing oil, grease, dirt and scale from a variety of items of various sizes and configurations. As a result of this ultrasound was used as one of the first methods for surface decontamination purposes in the nuclear industry [81McV]. Landolt-Börnstein New Series VIII/4
8-20
8 Decontamination
[Ref. p. 8-34
The process is particularly appropriate for decontaminating items with complex surfaces where other decontamination techniques are not suitable. It has been successfully used to clean dirt from holes, cracks and crevices in parts made of metal, glass, and a variety of plastics. Its most common application is for decontamination of tools and items that are evenly contaminated and it is carried out by immersion in an ultrasonic tank containing water (or water with chemical additives) [82Gar]. New applications have been developed replacing water with more aggressive chemicals in order to combine the cleaning effect of both the chemical and ultrasound. 8.1.3.4.1 Ultrasonic technique The two main components of ultrasonic cleaning systems are the ultrasonic generator and the transducer (or vibrator). The ultrasonic generator converts normal 50-60 Hz power to a high frequency supply usually in the range of 18 to 25 kHz. The high frequency alternating current is then fed to a transducer to produce vibration in the liquid. The transducer is generally made of piezoelectric material (i.e. material that will elongate or contract depending upon the polarity). As a consequence of the vibration, compressive and rarefacted phases are present in the solution. In the rarefaction phase, cavities are generated (mainly originating from nucleation points) and during the compressive phase these collapse causing a phenomenon called “cavitation”. When an item is immersed in the solution, the collapse of cavities causes scrubbing on its surface and hence produces a cleaning action. It is important to note that the presence of many nucleation points causes cavitation mainly on the surface of the items. The cavitation occurs even if the surfaces are inhomogeneous and complicated or located in inaccessible zones. Calculations indicate that during cavitation localised peak pressures as high as 70 MPa can be reached. These conditions produce a strong cleaning action on any surface upon which they act [82Gar]. Two factors play a fundamental role in the action of ultrasound: the cavitation threshold and the scrubbing factor. The cavitation threshold is the pressure difference inside the fluid, which allows the cavitation phenomenon to take place, and is directly correlated with the ultrasonic power applied to the solution. 8.1.3.4.2 Ultrasound in conjunction with chemicals It is well known that in order to increase the effectiveness of the process in terms of scrubbing effects, or decontamination factors, an appropriate liquid should be selected. In ultrasonic cleaning, specific chemical agents are commonly added and the ultrasonic cleaning combines the effect of cavitation of a liquid at the surface to be cleaned with the chemical action of the liquid. The physical and chemical properties of the liquid are important. Decontamination factors as low as 2 have been obtained by using pure water at room temperature. The addition of a cleaning or wetting agent in the amount of 2 to 5 % by weight, with an increase in temperature to about 80 °C, can greatly increase the cleaning ability of water. The addition of small amounts of citric acid, or other chemicals, can also enhance the cleaning ability. Reports on the effectiveness of ultrasonic decontamination from the 40 or more nuclear plants where it has been used are mixed. Some plants use it on a regular basis with good success, obtaining decontamination factors in the range of 5 to 100 [80Man]. Other plants report little success, and some have stopped using ultrasonic cleaning entirely [82Gar]. KWU-Service used ultrasonic decontamination to clean primary recirc pumps at several power plants (Biblis A-B, Neckarwestheim, Unterweser, Borselle and Atucha). This equipment has also been used in more recent power plants. At the KWU Centre in Karlstein, Germany, a large amount of scaffolding and small tools were decontaminated up to the release limit [84KWU].
Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
8 Decontamination
8-21
8.1.3.4.3 Decontamination by mobile ultrasonic tools A specially designed ultrasonic hand-held wall cleaner and floor cleaner has been designed at the Argonne National Laboratory in the USA for decontamination of flame-sprayed zinc on hot cell liners [70Ayr]. 8.1.3.5 Decontamination by foams Foam decontamination uses liquid foam generated from an acid or acid mixture, using air, nitrogen or an inert gas. The foam also contains various chemical additives such as inhibitors, foam stabilisers and surfactants [82Com]. Typical acids used are hydrochloric, nitric, hydrofluoric, sulphuric, and phosphoric and organic acids can also be considered. The foam is produced in a foam generator and the density can vary considerably − typically 1 litre of acid solution is used to produce 20 dm3 of foam. Foams have been successfully used in decontamination. The use of foams started more than 40 years ago [60Ayr] and at that time, inhibited hydrochloric acid, together with special surfactants, was used. Decontamination factors in the range of 5 to 50 are obtained with a single foam application using 7-10 % phosphoric acid [82Gar]. 8.1.3.6 Decontamination by gels A gel medium is defined as a semisolid system obtained by flocculating and immobilising particles in a continuous medium. The problem in utilising this method of decontamination is the generation and maintenance of adequate gel systems in combination with decontaminating chemicals. For example, it has been found difficult to develop an adequate gel system using alkaline permanganate solutions [82Gar]. Gels can be made from either organic based or inorganic based systems and contain in the gel formulation decontaminating chemicals which are normally acids such as phosphoric, sulphuric, or nitric [80Des]. High decontamination factors (in excess of 50) have been obtained in the laboratory for mild steel, stainless steel, aluminium, copper and Plexiglas. Large decontamination tests by gel spraying decontamination have been performed at CEA-CEN-Cadarache, France, on 17 Mg of steel from the German ISAR-BWR (consisting of 11 Mg of frame and 6 Mg of pipes) [89Brun]. The chemicals used were sulphuric acid (2 mol dm−3) and hydrofluoric acid (1.6 mol dm−3) and during the decontamination 100 litres of gels were used. 8.1.3.7 Decontamination by pastes Cleaning pastes are widely used for treating metal surfaces, particularly for stainless steel, and can be extremely effective for decontamination. They consist of a filler, carrier and use an acid or mixture of acids as the active agent. The concept of decontamination pastes follows the method used for the older paste systems, however new acid systems particularly effective in removing radioactive contaminants have been developed. These pastes, when applied in a thin layer on contaminated surfaces, can provide effective decontamination, together with generating relatively small quantities of waste. 8.1.3.8 Mechanical decontamination techniques Mechanical techniques include many decontamination methods based on the use of mechanical tools or devices to remove the surface contamination [56USS]. Mechanical devices are commonly used for cleaning industrial tubing and piping and can be adapted for the decontamination of similar items in a nuclear installation. Landolt-Börnstein New Series VIII/4
8-22
8 Decontamination
[Ref. p. 8-34
8.1.3.9 Decontamination by strippable coatings Decontamination by strippable coatings consists of the application of a coating over the surface to be decontaminated. This coating is then left on the surface for a set period (from few hours to a few days) and then removed/stripped resulting in removal of the contamination. Strippable coating formulations usually consist of high molecular weight, film forming, synthetic polymers such as polyethylene, polyvinylacetate, polyvinylchloride, acrylics, etc., dispersed as an emulsion in an aqueous base. These coatings usually contain an active agent e.g., an acid or mixture of acids, which attack the contaminants on the surface to which the coatings are applied. The coatings may be applied with a brush, spray system, roller or other similar method. In some cases, it may be necessary to apply two or more coats to ensure that the coating has sufficient strength to be readily removed from the surface without tearing. The coatings are applied in varying thicknesses from 0.5 to 2 mm [79TMI]. Usually the coating is then manually stripped off the surface in sheets, compacted and placed in waste containers. 8.1.3.10 Melting Melting is considered as a decontamination process since it can be used to reduce the specific contamination. The method completely destroys components and is effective only for contaminants that are volatile or more soluble (e.g. plutonium) in the slag than the molten metal. The decontamination efficiency varies widely depending on the radioisotope present. The radionuclides remaining in the molten material are distributed homogeneously and effectively immobilised, thus reducing the possibility of the spread of contamination. The melting should take place in a suitable refinery, which has filters on the gas exhaust system to protect the environment [85Pfl]. Melting is extensively used in Germany. From 1984 to 1989 more than 2000 Mg of low-level contaminated scrap (3 MeV) liberating electrons in the chamber wall. In practice, neither the idealized conditions of charged particle equilibrium nor that of the Bragg-Gray principle can be fully realized and thus a variety of corrections must be applied in order to make absolute dosimetric measurements possible. For radiation protection measurements, however, this complex task can be avoided when chambers are used that are calibrated under conditions similar to those in which the instrument will be applied and in terms of the desired dosimetric quantity (e.g., exposure, absorbed dose, ambient dose equivalent, or the corresponding dose rates). As an example, Fig. 10.4 shows a portable survey instrument with an air-equivalent chamber that can be used for dose and dose rate measurements over a wide range of photon energies.
Fig. 10.4 Dose and dose rate survey meter with an air- equivalent ionization chamber (volume 600 cm3) for the measurement of X- and γ-rays in the energy range between 6 keV and 3 MeV. Using an additional plastic shielding, even photons up to an energy of 7.5 MeV can be measured. The probe is detachable for remote measurements. (Courtesy Step Sensortechnik). Landolt-Börnstein New Series VIII/4
Ref. p. 10-25]
10 Measuring techniques
- - - -
Charging electrode
+ + + + + + + +
- - - -
10-7
C Insulator
Fig. 10.5. Schematic representation of a condenser-type ionization chamber.
Other types of portable ion chambers are based on the charge integration principle illustrated in Fig. 10.5. In this case, the chamber electrodes are connected in parallel with a capacitor, which is initially charged up to establish an electric field in the chamber. When the chamber is exposed to radiation, the ionization caused in the gas-filled chamber is collected by the electrodes and the charge stored in the capacitor is reduced. The resulting drop in chamber voltage can be used as a measure of the total integrated ionization charge. In the case of a self-reading ‘pocket dosimeter’ employed for personal radiation protection measurements, a simple rugged quartz-fiber is mounted inside the ionization volume which is deflected when the device is charged. The position of the fiber, which varies as the charge on the capacitor is reduced due to ionizing radiation, is observed through a small built-in microscope that has a scale in the eyepiece (Fig. 10.6). The position against the scale can be calibrated in terms of the total radiation recorded by the pocket dosimeter. Over a longer period of time, however, leakage currents across the insulator surface can not be avoided and this limits the accuracy of the dosimeter. Progress in integrated circuit technology resulted in the development of direct ion storage (DIS) dosimeters, which combine a gas-filled ionization chamber with a non-volatile electronic charge storage element [96Kah, 00Kno, 99Wer]. A schematic representation of a DIS dosimeter is shown in Fig. 10.7. The charge is initially placed on the floating gate of a standard EEPROM (electrically erasable and programmable read-only memory) cell by injecting electrons by a tunneling process through the silicon oxide. Electrons are trapped at this gate for many years since at normal operating temperatures they have no conductive discharge path when the silicon dioxide formation is made of high-purity material (cf. Section 10.1.4). When the chamber is exposed to ionizing radiation, the ions produced in the fill gas are collected by the charged floating gate, which results in a reduction of the charge stored. Assessment of this quantity can be performed without changing the charge distribution by measuring the channel conductivity of the field effect transistor (FET), which forms the basis of the EEPROM, by means of an electronic readout unit. The passive electronic dosimeter thus makes it possible to instantly and nondestructively readout the accumulated dose whenever required. The first commercial DIS personal dosimeter consist of a series of separate dosimeter elements housed inside a small hermetically sealed container. Three elements are used for the measurement of the personal dose equivalent Hp(10) in the range from 1 µSv to 40 Sv and two elements for the determination of Hp(0.07) in the range between 10 µSv and 40 Sv [99Wer] (the definition of Hp(d) is given in Section 4.5.3.4). The dose-rate linearity is flat up to 40 Sv/h thus guaranteeing accurate dose assessment in accident situations. Field lens
Eye lens
Scale
a
Fiber
Insulator
Ionization chamber
Electrode
Bellows
Charging pin
Fig. 10.6. Pen-size direct-reading ion chamber dosimeter for personal radiation protection measurements in a γ- or X-ray environment. (a) Simplified schematic representation. (b) Pocket dosimeters for different dose and energy ranges. The nominal voltage required to ‘set the dosimeter to zero’ is adjusted with the charging unit (potentiometer). A test source is used to check the correct function and calibration of the personal dosimeters. (Courtesy Thermo Eberline ESM). Landolt-Börnstein New Series VIII/4
10-8
10 Measuring techniques
Silicon oxide
Source
Fill gas
Channel
Floating gate
Drain
Silicon
[Ref. p. 10-25
Tunneling process
Fig. 10.7. Schematic representation of a DIS dosimeter consisting of combination of a gas-filled ionization chamber and an EEPROM.
10.1.2.3 Proportional counters A serious limitation of ionization chambers is that they are not sensitive enough to detect individual ionizing particles and thus can not be used for particle counting. This limitation can be overcome by operating an ionization detector in the proportional region to take advantage of the gas amplification process described in Section 10.1.2.1. However, proportional counters are not simply ionization chambers operated at high voltages (predominantly) in the pulse mode, but are specially constructed devices designed to optimize the gas amplification effect [00Kno]. The most important difference is that proportional counters always contain a thin anode wire to create a high electric field. Important geometrical factors are – among others – the uniformity, smoothness, and centricity of the thin anode wire with a diameter of between 5 µm and 50 µm. Specific demands on the gases are: low working voltage, high gas amplification, good proportionality, and high rate capability. These conditions are met by using mixtures of a noble gas and a polyatomic organic gas, such as 90 % argon and 10 % methane (P10 gas) or 96 % helium and 4 % isobutane. The organic additives, denoted as ‘quenchers’, improve the stability and performance of the counter by absorbing secondary ultraviolet photons − that are emitted from excited gas atoms − in a mode that does not lead to further ionization and thus avoids a transitory electrical breakdown. In practice, gain factors between 102 and 106 can be achieved. For particle detection and counting, proportional counters are operated in the pulse mode. Fig. 10.8 shows the equivalent circuit of a proportional counter, which replaces the circuit plotted in Fig. 10.1 for a current-type chamber. As mentioned above, the electrical field increases steeply near the central wire of a ionization detector and thus the amplification process initiated by an ionizing event in the sensitive volume results in an electron ‘avalanche’ in a region extending only a fraction of a millimeter from the anode surface. However, at the moment when the electrons are collected at the anode wire (within about 1 µs), the positive ions are still so close to the center wire that there is almost no change of the electric voltage. The output signal – a voltage pulse V(t) – is thus predominantly determined by the slower drift of the positive ions outward towards the cathode. Most of this process develops while the ions are still relative close to the wire and thus a sharply defined fast-rising electrical pulse can be observed. The subsequent decrease of the voltage pulse depends on the relative time constant of the external load circuit, which is given by the product of the resistance R and the equivalent capacitance C of both the detector and the measuring circuit (usually a preamplifier). When the capacitance of the circuit is fixed, the height of the voltage pulse is directly proportional to the charge generated within the detector and thus to the amount of energy the incident particle deposited in the gas [00Kno]. Based on this feature, a discrimination between particles depositing different amounts of energy in the gas volume – such as α- and β-particles – can be achieved. This can be realized, for example, by means of two separate read-out channels with different discriminator levels (cf. Fig. 10.8) in order to simultaneously detect either α-particles (high level) or both α- and β-particles (low level). Alternatively, the proportional counter can be connected to a multichannel analyzer, which records and stores pulses according to their height and thus allows the direct discrimination between different particles. If the channel number is related to the energy loss of the incident particles in the cavity by means of a suitable calibration procedure, a proportional counter can also be used for particle identification (particle spectroscopy).
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R
V (t)
Voltage V ( t )
Counter
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VD
Time t
Fig. 10.8. Equivalent circuit of a counter operated in the pulse mode (cf. Fig. 10.1). VD is the discriminator level. Only pulses with a pulse height exceeding this level are counted.
In practice, it is not possible to record all types and energies of radiation with sufficient efficiency by one detector only and thus different detector designs are used for α-, β-, and γ-radiation. For the detection of α- and low-energy β-particles, it is of particular importance to use counters with thin window foils (such as metalized plastic) or windowless entrance cathodes made of a metal wire grid to reduce absorption or energy-loss of the particles transversing the entrance material as far as achievable. Since such systems can be sealed only heavily, stationary counters are normally operated as gas flow units to avoid gradual contaminations and loss of fill gas. To monitor for contaminations on floors and surfaces of objects in laboratories or on hands, shoes, and clothing of personnel working with radioactive material, large-area proportional counters are used such as those shown in Fig. 10.9. They consist of a cathode container filled with gas, which encloses either a meander-shaped anode wire or multiple anode wires. Proportional counters (PC) made of a tissue equivalent (TE) plastic wall (most often A-150 plastic) and filled with TE gas mixtures (propane or methane gases mixed with carbon dioxide and nitrogen) are standard instruments in microdosimetry [83ICR]. An important feature of TEPCs, which are based on the Bragg-Gray cavity theory described in Section 10.1.2.2, is that the pressure of the filling gas can be adjusted so that a charged particle crossing the cavity deposits an identical amount of energy as a charged particle crossing a real tissue volume of microscopic dimensions [95Wak]. Multichannel pulse-height measurements with a low-pressure TEPC in radiation fields with an intensity low enough to allow the detection of single events thus give the distribution of the energy deposited by individual primary charged particles in a microscopically small tissue volume. Since the height of the recorded pulses strongly depends on the ionization density along the tracks of the charged particles − which varies considerably between different types of charged particles such as electrons, protons and heavier ions − a low pressure TEPC not only acts as recorder of deposited energy, but also as a spectrometer able to distinguish charged particles with a different linear energy transfer (LET) and thus provides an estimate on radiation quality [95Wak, 02Wak].
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Fig. 10.9. Large-area proportional counters for contamination detection. a: Sealed handheld monitor for detection of surface contaminations with α-, β- and γ-isotopes. (Courtesy Thermo Eberline ESM). b: Continuous gas-flow contamination monitor for detection of α- and β-contaminations of hands, shoes, and clothing. (Courtesy Berthold Technologies).
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Low pressure TEPCs have been an important laboratory tool in experimental microdosimetry for many years. The recent availability of microelectronics and progress in digital electronics enabled the development of portable TEPCs for area monitoring and thus the application of this microdosimetric method in practical radiation protection [89Men, 95Sch, 00Sha]. Particular advantages of these devices are the possibility (1) to separate dose fractions due to photons and neutrons in mixed n-γ fields, which release weakly respectively densely ionizing secondaries in the TE wall of the counter, and (2) to directly measure the operational quantity ambient dose equivalent [89Die]. 10.1.2.4 Geiger-Müller counters Although the design and operation of Geiger-Müller (GM) counters are in many respects similar to those of proportional counters, there are three important differences: Firstly, GM counters are operated at substantially higher tube voltages (cf. Fig. 10.2) so that a particle entering the gas volume triggers an electron avalanche extending along the entire length of the anode wire. As a consequence, the size of the output voltage pulses is more or less independent on the number of original ion pairs that initiate the gas amplification process and thus on the type and energy of particle entering the sensitive gas volume. The voltage between the electrodes that is required to sustain an avalanche ionization can be determined by exposing the counter to a constant source of radiation and observing the counting rate as a function of the applied voltage. Secondly, the UV absorbing quench gas used in proportional counters is omitted in GM tubes since ultraviolet photons emitted from excited gas atoms are essential to the process of propagating the discharge throughout the tube. Instead, other ‘quenchers’ such as gaseous halogens (Cl or Br) or organic substances (ethyl alcohol or ethyl formate) are added with a typical concentration of 5-10 % to the primary fill gas to prevent repeated or continuous gas discharge through the mechanism of charge transfer collisions. A detailed description of the underlying complex mechanism can be found in [86Att, 00Kno]. Thirdly, immediately after a discharge, a dense cloud of positive ions exists near the central wire and reduces the electric field in the counter to a great extent. This space charge not only terminates the discharge of the tube but also prevents that a further avalanche can be generated before the positive ions have moved − at least part of the distance − towards the cathode. The time between the detection of the initial pulse and the time at which a succeeding pulse can be counted because its amplitude exceeds the discriminator level is denoted as resolving time. Typical values are between 100 to 300 µs. In contrast, the resolving time of a proportional counter is less than a few microseconds. During the resolving time, the GM counter is ‘dead’ and any particles entering the tube during that time are lost [94ICR1]. In practice, gain factors between 106 and 1010 can be realized with a GM counter. The resulting voltage pulses have a height between 1 and 10 V and can easily be detected with simple electronic circuits − often completely without external amplification. GM tubes are thus simple, rugged, and relatively inexpensive particle-counting instruments. As mentioned above, however, they suffer from extremely long resolving times and are thus seldom used when accurate measurements are required at count rates greater than a few hundred counts per second. In many cases, GM counters are provided with removable covers on the entrance window in order to differentiate between penetrating (γ- and high-energy β-particles) and low-penetrating (α- and low-energy β-particles) radiation by measuring the difference between the count rates with and without the cover in place. As in the case of proportional counters, the entrance window must be sufficiently thin to permit passage of α-particles. For the detection of γ-particles, on the other hand, the thickness of the entrance window or of the cover should approximate the maximum range of the secondary electrons produced in the window or cover to increase detection sensitivity.
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10.1.3 Scintillation detectors The major limitation of gas-filled counters, namely the low detection efficiency for X- or γ-rays, can be overcome by the use of solid or liquid detector materials, which have a much higher density than gases. Scintillation materials are frequently utilized for photon or neutron detection. In these materials a small fraction of the energy deposited by charged particles will be emitted as visible or ultraviolet light on a time scale of nanoseconds to milliseconds, whereby the intensity of the light flash is proportional to the energy deposited in the scintillator. As shown in Fig. 10.10, a scintillation detector consists basically of scintillator material that is optically coupled to a photomultiplier tube to convert the light photons released in the scintillator into an electrical pulse which can then be amplified and analyzed electronically. Scintillator
γ − particle Analyzer
a
Semitransparent photocathode
Light shield, reflector Photomultiplier
gas envelope
HV
Shield e−
Amplifier
Pre − amplifier
Light photon
Dynodes
Anode
b
Fig. 10.10. Schematic representation of (a) a pulse-mode scintillation detector and (b) a linear focussed photomultiplier tube showing the cascade effect due to amplification of electrons from the photocathode by increasing secondary emission when the electrons strike the dynodes.
Scintillators fall into two major categories: inorganic and organic materials, the choice of which depends strongly on the type of measurement to be performed. Physical properties of a few representative scintillator materials are given in Table 10.3; a more comprehensive list can be found in [00Kno, 94Leo]. Inorganic scintillators are crystals of alkali halides (such as NaI, CsI) or oxides (such as Bi4Ge3O12, ‘BGO’) grown at high temperatures. In these materials, scintillation is a property of the electronic band structure of the crystals: When an ionizing particle enters the scintillator, it can raise electrons from the valence into the conduction band. The electrons and holes formed by this excitation process recombine and emit a photon. In the pure scintillator material, however, de-excitation is an inefficient process due to self-absorption. Therefore, small amounts of an ‘activator’ (e.g., thallium in the case of NaI) are added. These impurities create energy states within the forbidden band gap of the scintillator over which electrons can alternatively de-excite from the conduction band into the valence band. A more detailed description of the scintillation process can be found in [00Kno, 95Tso]. Since energy spacing between activator energy states is less than that between the conduction and valence bands of the pure solid, the emitted photons do not have enough energy to raise other electrons from the valence band to the conduction band and thus cannot be effectively reabsorbed by the scintillator. Moreover, the change in energy of the emitted photons results in a shift of the wavelength of maximum emission from the ultraviolet into the visible region, where the sensitivity of most photomultiplier tubes is maximal. Inorganic scintillators tend to contain elements with a high atomic number and have a relatively high density (cf. Table 10.3). Consequently, the photoelectric effect is the main interaction mechanism for X- or γ-rays in the energy range between 10 keV to 1 MeV, making inorganic scintillators favorable for particle identification by means of spectroscopic measurements (see below). They also have a high light output, but are hampered by a relatively slow response. Organic scintillators, on the other hand, are aromatic hydrocarbon compounds which contain benzenoid rings. They are broadly classed into three types: crystalline, liquid, and plastic, all of which utilize the ionization produced by charged particles to generate optical photons, usually in the blue to green wavelength regions. Examples of pure organic crystals are anthracene (C14H10) and trans-stilbene (C14H12). Plastic scintillators are non-fluid solutions consisting of fluorescent organic compounds disLandolt-Börnstein New Series VIII/4
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solved in a solidified polymer matrix or fluid solutions with similar fluorescent organic compounds [02Sai]. In contrast to inorganic scintillators, the fluorescence process in organic scintillators is an inherent molecular property which is characterized by the excited states of the individual molecules. Therefore, fluorescence can be observed independent of the physical state of the material. Organic scintillators are generally faster in their response than inorganics and are more suitable for β-particle spectroscopy and especially for fast neutron detection due to the high hydrogen fraction in their composition (cf. Section 10.1.7). Moreover, some organic scintillator materials (such as BC 501 / NE 213, cf. Tab. 10.3) offer the possibility to discriminate between photons and neutrons due to differences in scintillator response. In these materials, electrons released by γ-quanta cause scintillations at a rate faster than that due to photons created by neutrons. Table 10.3. Physical properties of a few representative scintillator materials (data from [02Sai]). Material Type Density RefracDecay Light Wave- Main application tion outputa constantb lengthc index [g/cm3] [%] [ns] [nm] Inorganic scintillators NaI(Tl) Crystal 3.67 1.85 100 250 415 γ, X-rays CsI(Tl) Crystal 4.51 1.79 45 1005 550 γ, heavy particles CsI(Na) Crystal 4.51 1.84 85 630 420 γ, heavy particles BGO Crystal 7.13 2.15 20 300 480 γ, X-rays LSO Crystal 7.40 1.82 63 40 420 γ, X-rays Organic scintillators Anthracene Crystal 1.25 1.62 100 30 447 General purpose Trans-stilbene Crystal 1.16 1.63 50 4.5 410 γ, fast n BC 400 / NE 102 Plastic 1.03 1.58 65 2.4 423 General purpose BC 422 / NE 111 Plastic 1.03 1.58 55 1.6 370 Ultra-fast timing BC 501 / NE 213 Liquid 0.87 1.51 78 3.2 425 Fast n with γ discrimination BC 509 / NE 226 Liquid 1.61 1.38 20 3.1 435 γ, insensitive to n a
Given relative to NaI(Tl) for inorganic scintillators and relative to anthracene for organic scintillators. The light output of anthracene is 40-50 % of NaI(Tl). b Main component. c Maximum emission.
As mentioned above, photomultiplier tubes (PMT) have two different functions: conversion of ultraviolet and visible photons emitted by the scintillator into an electrical signal and signal amplification. Fig. 10.10 shows the essential parts of a PMT, which are mounted inside an evacuated glass envelope, namely a photocathode, typically 10 to 12 electrodes denoted as dynodes, and an anode. The photons from the scintillator strike the photocathode − usually made of a semiconductor material formed from antimony plus one or more alkali metals − and release photoelectrons with an efficiency of about 10-30 % [94Leo]. These photoelectrons are attracted to the first dynode, which is at a higher potential than the cathode, so that the electrons strike the dynode with a sufficiently high kinetic energy to eject three to four secondary electrons from the surface. Since each dynode has a more positive voltage than the preceding one, this amplification process is repeated with each successive dynode, so that a multiplication factor of 106 or more can be obtained with a twelve stage PMT. The average gain of the dynode chain is independent of how many electrons are simultaneously ejected from the photocathode. As a consequence, the size of the electrical output signal at the anode is proportional to the number of electrons leaving the photocathode. To achieve a good performance, it is important to match the emission spectrum of the scintillator to the quantum efficiency of the photocathode material. The current measured at the anode of the PMT is fed into an RC circuit as shown in Fig. 10.8 to produce an electrical voltage pulse.
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The previous discussion reveals, that the output of a scintillator-PMT combination is proportional to the amount of energy deposited by an incident particle in the scintillator and can thus be used for particle spectroscopy. To this end, the voltage pulse goes into an amplifier and is then fed into a multichannel analyzer, which records and stores pulses according to their voltage amplitude into different channels. By using γ-ray sources of known energy, the channel numbers can be related to energy. As an example, Fig. 10.11 shows the pulse-height spectrum of 60Co recorded from a large NaI(Tl) scintillator. 100
100 Photopeaks Photopeaks Backscatter
60
Compton region
40
Relative count rate [% ]
Relative count rate [% ]
80
80 Compton edges
Backscatter 3 2
1 3
20
0
0 0
200
a
400
600 800 1000 1200 1400 1600 Energy [ keV ]
0
b
200
400
600 800 1000 1200 1400 1600 Energy [ keV ]
Fig. 10.11. Pulse-height spectra of 60Co measured with (a) a large NaI(Tl) scintillation detector and (b) a highpurity germanium detector. The two characteristic photopeaks are at energies of 1.17 and 1.33 MeV. In (b) a single (1) and a double escape peak (2) related to the incident 1.33-MeV-photons is apparent, as is an annihilation peak at 511 keV (3) due to pair production interactions in surrounding materials. Note that incident 1.17-MeVphotons do not give rise to escape peaks, since the photon energy is only slightly above the threshold for pair production interactions where the cross section is still very low.
A pulse-height spectrum recorded from a radiation source depends not only on the characteristics of the radiation to be measured but also on the type of scintillator used and the mechanisms by which the incident particles transfer their energy to the detector material. Since the ranges of charged particles are very short in most solid and liquid materials, they deposit their energy almost completely in the detector material giving rise to a well-defined peak in the spectrum at the particle energy. In the case of γ-quanta, energy is deposited to the detector primarily by the photoelectric effect, Compton scattering, and pair production. An incident photon undergoing a photoelectric interaction in the scintillator transfers (nearly) all of its energy to an electron and thus contributes to the ‘photopeak’ in the pulse-height spectrum, which is located at the energy of the incident photon. In Compton scattering, however, only part of the energy is transferred to the detector, via the recoil electrons. The scattered photon may either be absorbed by a photoelectric interaction within the scintillator or may escape from the detector. In the first case, the total energy of the incident photon is absorbed and the event will contribute to the photopeak. In the second case, however, the energy deposited by the recoil electron depends on the scattering angle. The ‘Compton region’ in the spectrum thus ranges from near zero (small-angel scattering) up to a maximum energy (‘Compton edge’) for 180° Compton scattering. If the energy of the incident photon exceeds 1022 keV, pair production can occur. When the positron created by this process comes to rest, it combines with an electron to create a pair of 511 keV annihilation photons. If one or both of these photons escape, the energy deposited in the scintillator is reduced by 511 or 1022 keV, respectively. As a consequence, additional ‘photopeaks’ − denoted as single and double escape peaks − appear in the spectrum at energies of 511 keV and of 1022 keV below the corresponding full-energy photopeaks (cf. Fig. 10.11). Finally, lowenergy peaks may appear in the spectrum resulting from γ-quanta that are scattered in material outside of the scintillator, and enter the detector having lost most of their energy. However, these backscatter peaks Landolt-Börnstein New Series VIII/4
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area usually easy to identify due to their low energies (3000 keV) down into the energy range of in vivo counting (Table 10.17). Table 10.17. Shielding parameters for some selected shielded rooms for in vivo measurement devices Shielding materials
Reference Background Background reduction factor index [cpm/cm3] 0.43 85Sch 70 cm silica sand, 0.3 cm Pb, 0.04 cm Cd, 0.1 cm Cu, 73 0.55 cm plastic 15 cm Fe, 1 cm Pb, 0.2 cm Fe 15 cm Fe, 0.9 cm Pb
100 40 - 57
0.36 0.34
85Sum 61Kie
Partial shielding Some installations comprise more open structures, which eliminate direct paths for radiation between the detector and the laboratory. Examples are the “shadow shield” design used for assessing whole-body radioactivity in which the subject lies on a bed moving under a fixed detector in a central turret. Other arrangements embodying the same principle can be devised to assess the radioactivity of individual organs or regions. In another simple arrangement the detector and the back of a chair or bed holding the subject are shielded for the investigation of radioactive deposits in larger regions. With all partially shielded counters the background response below 200 keV is likely to be much larger than in a shielded room, because they respond to photons scattered by the subject into the detector. For this reason a shielded room is essential for the sensitive assessment of low-energy photon emitters. More information about shielding is given for example in the IAEA “Directory of whole-body radioactive monitoring” [70IAE]. In addition to the shielding of the environmental radiation it is very important that all materials in the detector systems, the mounting facilities and the shielding are selected for low level of intrinsic radioactivity. Also the natural radioactivity in the construction materials of the surrounding building is important for the background. The natural activity concentration of 40K can vary in between 200 Bq/kg and 800 Bq/kg in bricks and 320 Bq/kg and 800 Bq/kg in cement, respectively [92Zik]. Thus, there is a large potential for background reduction by proper selection of the construction materials. Moreover it is important to minimise the amount of material close to the detector system in order to minimise the background component due to Compton scattering in the direct vicinity of the detectors. Last but not least there is a need for air filtration in order to reduce the background component due to airborne radioactive materials. 10.3.2.4.3.2 Active methods There is different kind of active methods for the reduction of the detector background signal, most of them being based on anticoincidence techniques. Firstly this technique was applied for proportional counters in order to reduce the background due to β particles or Compton electrons from the environment. For this purpose the counting volume of the proportional counter was surrounded by guard counters and all coincident absorption events in both the counting volume and one of the guard counters were discriminated. Thus only the events due to photoelectric absorption in the counting volume were processed, this resulting in very good detection features especially for low energy photons such as the plutonium L X-rays. On the other hand the sensitivity of these detectors for photons with higher energy as for example the 59.6 keV γ-rays of 241Am is very low, even when the proportional counters were operated with heavy counting gases such as Xe under high pressure of 2 or 3 bar. However, for in vivo measurement of plutonium via the L X-rays it is essential to measure simultaneously the 241Am activity present, because the L X-ray yield of 241Am is one order of magnitude higher than that of plutonium. The energy resolution of the proportional counters does not allow for discrimination of the 241Am L X-rays Landolt-Börnstein New Series VIII/4
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from the plutonium L X-rays and so the contribution from 241Am must be determined by measurement of 241 Am via the 59.6 keV γ-rays in the same measuring geometry. It was mainly due to this reason that the proportional counters were replaced in most laboratories by the phoswich detectors, which were commercially available since the late sixties.
Photo multiplier Other source of radiation
CsI (Tl)
(4) NaI (Tl) (1)
(2)
(3)
Be. entrance window Source material
Fig. 10.39. Measuring principle of a phoswich detector.
The phoswich detector has been developed by Laurer for the in vivo measurement of low energy photon emitters such as 210Pb, 239Pu and 241Am [68Lau]. The detector consists of a large area NaI(Tl) crystal the thickness of which being just enough to fully absorb the low energy photons (typically 1-3 mm). The NaI(Tl) crystal is backed by a CsI(Tl) crystal for the detection of scattered photons due to Compton effects in the NaI(Tl) crystal (Fig. 10.39). The photomultiplier tubes detect the scintillation light from both crystals. However, because of the different scintillation decay times of the materials, it is possible to assess by pulse shaping techniques whether the scintillation light is originating from the NaI(Tl) crystal, the CsI(Tl) crystal or from both crystals. Thus it is possible to discriminate the Compton scattering events (in both crystals) from the full absorption events, this resulting in a significant reduction of the detector specific Compton continuum of the NaI(Tl) crystal. When comparing the count rates of a phoswich detector in the low energy range (10 - 100 keV) with and without applying the pulse shape discrimination technique a reduction of about one order of magnitude is observed. This figure, however, does not correspond to the actual reduction of the Compton continuum from the NaI(Tl) crystal because most of the scintillations detected by the photomultipliers are due to absorption events in the CsI(Tl) crystal only. Actually the reduction of the NaI(Tl) Compton continuum is less than a factor 2 because the CsI(Tl) anti Compton shield covers less than a 2π space angle. Phoswich detectors provide good detection features for low energy photon emitters if no high-energy photon emitters such as 137Cs are present. Compton continuum, backscatter peaks and characteristic X-ray peaks due to those high-energetic emitters may influence the spectrum in the low energy region significantly, this giving rise for problems in background prediction.
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10.3.2.5 Spectrum evaluation The interpretation of a photon-energy spectrum of body radioactivity will involve initially the identification of radionuclides responsible for its individual features. The next stage, unless the spectrum is dominated by the contributions from a single radionuclide, will generally involve resolution into the constituent components. In a further process, the response attributable to a particular contributor will be translated into an estimate of body or organ radioactivity; this is accomplished by reference to a spectrum representing a known radioactivity of the nuclide measured in the same conditions. Methods of deconvoluting photon-energy spectra of body radioactivity do not differ in principle from those applied in X- and gamma-ray spectrometry generally, except that account must often be taken of the effect on spectral shape of scatter in a large attenuating mass. The process is at its simplest in the estimation of peak areas from semiconductor detectors. The good energy resolution of such instruments allows the effective background response underlying a spectral peak to be reliably deduced from the adjacent continuum. In the case of scintillation counters, the width of spectral peaks makes this approach often inapplicable especially in case of multiple peaks. It will generally be necessary to first subtract an appropriate spectrum of counter background and a more rigorous analytical procedure will generally be required. The activity of the given radionuclide (q) can be expressed as follows q=
where
N t ⋅ y ⋅η
(10.3.2.3)
N is the number of net counts in the full energy peak area, t is the measuring time, y is the yield of γ (or X)- ray, η is the counting efficiency at the given energy for the respective measuring geometry
This is the most dominating evaluation method in gamma spectrometry. Stripping method When deconvolution of spectra from scintillation counters is required, a “stripping” process is sometimes followed. Reference spectra are derived for each nuclide present, each representing the response from known amounts of the nuclide in appropriate measuring geometry and in relevant absorbing media. The reference spectrum containing the peak with the highest energy is selected, and it is normalised to the subject's spectrum on the basis of count rate in an energy region where only that nuclide contributes. Subtraction of the normalised spectrum gives a residue representing the remaining components, which is treated in the same way. The activity of each radionuclide is calculated directly from the fraction of its reference spectrum, which must be subtracted. In principle, the process can be repeated until the residue consists of the response from a single nuclide only. In practice, unacceptable errors are likely to accumulate if the number of stages exceeds two or three, particularly in relation to minor components in a spectrum; moreover, the method will generally be inapplicable when the dominant peaks of different components overlap. Linear regression analysis A more satisfactory procedure in many situations is to adopt a method of linear regression analysis, to derive the proportions of each reference spectrum which, when combined, gives rise to the best fit to the subject's spectrum. Facilities for such analyses are embodied in several commercially available computer programmes for processing γ ray spectra; alternatively, they can be developed locally. Utilizing a much larger portion of the spectrum, instead of the restricted regions successively considered in the stripping process, this method gives improved statistical accuracy in the estimates of the various components; moreover, realistic estimates of this accuracy may be derived in the matrix-inversion procedures. As with other methods of deconvolution, this approach has its limitations. In particular, it demands stability of the spectrometer during the measurement, especially if the nuclides present possess overlapping spectral features. It is also important that the locations of peaks in the subject's spectrum should coincide with those in the relevant reference spectra; where minor drifts occur between the measurements, adjustments can
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often be made prior to the analysis if appropriate routines are available. However, the validity of the analysis depends also on the spectral shapes of the reference standards according with those of the corresponding components in the subject's spectrum. This method is used practically exclusively in scintillation gamma spectrometry, however the procedure can be extended to the deconvolution of any kind of complex distribution of measured data. Such application is unfolding profile scanning data to quantify the measured distribution pattern in terms of activity deposited in different body regions or organs. 10.3.2.6 Measuring geometries In principle, the in vivo measuring systems can be allocated to two different types of systems, namely geometry dependent and geometry independent systems. Here the geometry is defined as the detector configuration in relation to the photon-emitting source in the body. 10.3.2.6.1 Geometry depending systems Static geometry The most common kind of geometry dependent systems are those having detectors that are positioned close to the subject looking to specific organs or tissues. The advantages include high efficiency, better subject positioning and less space requirement for the system. The use of such a static geometry dependent system is extremely important in measuring low energy photons where the efficiency of the detection system needs to be maximized. Isotopes of iodine, and also 99mTc may concentrate in the thyroid gland. The range of photon energies encountered is 27 keV (125I) to several hundred keV. Some HPGe detectors used for assessment of actinides in lungs are of suitable diameter (ca. 50 mm) in relation to the size of the thyroid, and are large enough to provide adequate detection efficiency over most or all of this energy range. Alternatively, a planar germanium detector or thin NaI(Tl) crystal may be adopted for photon energies
