Population Monitoring and Radionuclide Decorporation Following a Radiological Or Nuclear Incident

166

NCRP REPORT No. 166

POPULATION MONITORING AND RADIONUCLIDE DECORPORATION FOLLOWING A RADIOLOGICAL OR NUCLEAR INCIDENT

POPULATION MONITORING AND RADIONUCLIDE DECORPORATION FOLLOWING A RADIOLOGICAL OR NUCLEAR INCIDENT

National Council on Radiation Protection and Measurements

NCRP REPORT No. 166

Population Monitoring and Radionuclide Decorporation Following a Radiological or Nuclear Incident

Recommendations of the NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS

April 6, 2010

National Council on Radiation Protection and Measurements 7910 Woodmont Avenue, Suite 400 / Bethesda, MD 20814-3095

LEGAL NOTICE This Report was prepared by the National Council on Radiation Protection and Measurements (NCRP). The Council strives to provide accurate, complete and useful information in its publications. However, neither NCRP, the members of NCRP, other persons contributing to or assisting in the preparation of this Report, nor any person acting on the behalf of any of these parties: (a) makes any warranty or representation, express or implied, with respect to the accuracy, completeness or usefulness of the information contained in this Report, or that the use of any information, method or process disclosed in this Report may not infringe on privately owned rights; or (b) assumes any liability with respect to the use of, or for damages resulting from the use of any information, method or process disclosed in this Report, under the Civil Rights Act of 1964, Section 701 et seq. as amended 42 U.S.C. Section 2000e et seq. (Title VII) or any other statutory or common law theory governing liability.

Disclaimer Any mention of commercial products within NCRP publications is for information only; it does not imply recommendation or endorsement by NCRP.

Library of Congress Cataloging-in-Publication Data Population monitoring and radionuclide decorporation following a radiological or nuclear incident / recommendations of the National Council on Radiation Protection and Measurements. p. ; cm. -- (NCRP report ; no. 166) "April 6, 2010." Includes bibliographical references and index. ISBN 978-0-9823843-7-4 (alk. paper) 1. Radiation--Physiological effect. 2. Radiation--Safety measures. I. National Council on Radiation Protection and Measurements. II. Series: NCRP report ; no. 166. [DNLM: 1. Radiation Effects--Practice Guideline. 2. Decontamination--methods-Practice Guideline. 3. Disaster Planning--Practice Guideline. 4. Radiation Monitoring--Practice Guideline. 5. Radioactive Hazard Release--Practice Guideline. 6. Triage--Practice Guideline. WN 600] RC93.P67 2011 616.9'897--dc23 2011014172

Copyright © National Council on Radiation Protection and Measurements 2011 All rights reserved. This publication is protected by copyright. No part of this publication may be reproduced in any form or by any means, including photocopying, or utilized by any information storage and retrieval system without written permission from the copyright owner, except for brief quotation in critical articles or reviews. [For detailed information on the availability of NCRP publications see page 273.]

Preface In the years since the September 11, 2001 terrorism incident, the National Council on Radiation Protection and Measurements (NCRP) has been active in preparing publications that provide guidance in preventing, preparing for, and responding to possible acts of radiological or nuclear terrorism. Major publications on these subjects include Report No. 138, Management of Terrorist Events Involving Radioactive Material and Commentary No. 19, Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism. In addition, NCRP has published three commentaries (Commentary No. 16, Screening of Humans for Security Purposes Using Ionizing Radiation Scanning Systems; No. 17, Pulsed Fast Neutron Analysis System Used in Security Surveillance; and No. 20, Radiation Protection and Measurement Issues Related to Cargo Scanning with Accelerator-Produced High-Energy X Rays) that discuss measurement and health protection aspects of using radiation-based systems for the detection and interdiction of radiological or nuclear materials and terrorism threats. NCRP has also been preparing new reports related to the treatment and long-term medical management of people affected by deliberate or accidental releases of radiological or nuclear materials including Report No. 161, Management of Persons Contaminated with Radionuclides. The present Report provides guidance for emergency responders and medical centers for the development of radiological response plans that include the efficient screening of a population for internally-deposited radionuclides, decontamination procedures, and treatment by decorporation therapy. While this Report is intended to focus on screening a population for internal contamination, screening is only one aspect of monitoring a population in the aftermath of a radiological or nuclear incident. Thus, this Report also broadly discusses external monitoring and decontamination of the affected population. The social and psychological impacts of a radiological or nuclear incident, and the long-term medical monitoring of the exposed population, are also addressed. A focus is placed on rapid methods for determining the contaminating radionuclides and the screening, decontamination and prompt medical management of contaminated persons.

iii

iv / PREFACE This Report was prepared by NCRP Scientific Committee 4-2 on Population Monitoring and Decontamination Following a Radiological or Nuclear Incident. Members of the Committee were: Richard J. Vetter, Chairman Mayo Clinic Rochester, Minnesota Members Steven M. Becker University of Alabama at Birmingham Birmingham, Alabama

Steven H. King Milton S. Hershey Medical Center Pennsylvania State University Hershey, Pennsylvania

Eugene H. Carbaugh Pacific Northwest National Laboratory Richland, Washington

Adela Salame-Alfie New York State Department of Health Troy, New York

James R. Cassata U.S. Navy Bethesda, Maryland

Lin-Shen Casper Sun U.S. Nuclear Regulatory Commission Rockville, Maryland

Scott Davis University of Washington Seattle, Washington

Katherine Uraneck New York City Department of Health and Mental Hygiene New York, New York

Fun H. Fong, Jr. Smyrna, Georgia

George J. Vargo MJW Corporation, Inc. Avondale, Pennsylvania

P. Andrew Karam New York City Department of Health and Mental Hygiene New York, New York NCRP Secretariat Bruce B. Boecker, Staff Consultant Cindy L. O’Brien, Managing Editor David A. Schauer, Executive Director

The Council expresses its appreciation to the Committee members for the time and effort devoted to the preparation of this Report. NCRP also expresses appreciation to the Centers for Disease Control and Prevention and the U.S. Navy for providing funding for preparation of the Report. Thomas S. Tenforde President

Contents Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 1. Executive Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 2. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 2.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .7 2.2 Purpose of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . . .13 2.3 Target Audiences of this Report . . . . . . . . . . . . . . . . . . . .14 2.4 Scope of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14 3. Background Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . .17 3.1 Internal Deposition of Radionuclides . . . . . . . . . . . . . . . .17 3.1.1 Inhalation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .18 3.1.2 Ingestion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .20 3.1.3 Absorption from Skin Contamination . . . . . . . . .21 3.1.4 Absorption Through Wounds . . . . . . . . . . . . . . . .21 3.2 External Contamination . . . . . . . . . . . . . . . . . . . . . . . . . .22 3.3 Effects of Weather . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .22 3.4 Complications Due to the Presence of Multiple Agents or Serious Injuries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .23 3.5 Radiological Triage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .26 3.6 Proximity to the Incident . . . . . . . . . . . . . . . . . . . . . . . . . .26 3.7 Previous Experience with Internal Contamination . . . . .27 3.8 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .28 4. Settings in Which Persons May Become Contaminated with Radioactive Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .29 4.1 Radiological Dispersal Device . . . . . . . . . . . . . . . . . . . . . .29 4.1.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .30 4.1.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .30 4.2 Aerosol Dispersal into a Public Area . . . . . . . . . . . . . . . .31 4.2.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .31 4.2.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .31 4.3 Contamination of Food or Water Supplies . . . . . . . . . . . .32 4.3.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .32 4.3.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .33 4.4 Improvised Nuclear Device . . . . . . . . . . . . . . . . . . . . . . . .33 4.4.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . .33 4.4.2 Nature of Contamination . . . . . . . . . . . . . . . . . . .35

v

vi / CONTENTS 4.5

Nuclear Reactor Incident . . . . . . . . . . . . . . . . . . . . . . . . . 4.5.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . . 4.5.2 Nature of Contamination . . . . . . . . . . . . . . . . . . . Large-Scale Fires and Incidents . . . . . . . . . . . . . . . . . . . 4.6.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . . 4.6.2 Nature of Contamination . . . . . . . . . . . . . . . . . . . Sealed Radioactive Source Incidents . . . . . . . . . . . . . . . . 4.7.1 Incident Characteristics . . . . . . . . . . . . . . . . . . . . 4.7.2 Nature of Contamination . . . . . . . . . . . . . . . . . . . Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

35 36 37 37 37 38 38 39 39 40

5. Coordination with the Incident Command System . . . . . 5.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2 Incident Command System . . . . . . . . . . . . . . . . . . . . . . . 5.3 Hospital Incident Command System . . . . . . . . . . . . . . . . 5.4 Coordination Between Incident Command System and Hospital Incident Command System . . . . . . . . . . . . . . . . 5.5 Communicating Information from the Scene to the Hospitals and from the Hospitals to the Scene . . . . . . . .

41 41 43 46

6. Radiological Triage and Screening Guidance . . . . . . . . . . 6.1 General Guidance for Emergency Responders . . . . . . . . 6.1.1 Selecting an Appropriate Radiation Survey Instrument . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.1.2 Presurvey Radiation Survey Instrument Checks 6.1.3 Surveying for Radioactive Contamination . . . . . 6.2 Radiological Triage and Screening Procedures . . . . . . . . 6.3 Initial Screening at Scene . . . . . . . . . . . . . . . . . . . . . . . . 6.4 Initial Screening at Hospital . . . . . . . . . . . . . . . . . . . . . . 6.5 Mass Screening Following the Emergency Phase . . . . . . 6.6 Biodosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

53 53

4.6

4.7

4.8

48 48

54 55 56 56 57 60 61 61

7. Clinical Decision Guide: Concept and Use . . . . . . . . . . . . . 65 7.1 Clinical Decision Guide Concept . . . . . . . . . . . . . . . . . . . 65 7.2 Clinical Use of the Clinical Decision Guide . . . . . . . . . . . 68 7.2.1 Decision-Making Process . . . . . . . . . . . . . . . . . . . 68 7.2.2 Use of the CDG Tables . . . . . . . . . . . . . . . . . . . . . 69 7.2.3 Use of a Single-Void Urine Sample Collected During the First 24 h . . . . . . . . . . . . . . . . . . . . . . 70 7.2.4 Using the CDG with an Intake of Multiple Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 7.2.5 Determining an Intake for Times More than 24 h in the Past . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 7.2.6 Special Considerations for Uranium CDGs . . . . 73 7.3 Americium-241: Clinical Decision Guide Fact Sheet . . . 78

CONTENTS

7.4 7.5 7.6 7.7 7.8 7.9 7.10 7.11 7.12

7.13

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Cesium-137: Clinical Decision Guide Fact Sheet . . . . . . .79 Cobalt-60: Clinical Decision Guide Fact Sheet . . . . . . . . .80 Iodine-131: Clinical Decision Guide Fact Sheet . . . . . . . .82 Iridium-192: Clinical Decision Guide Fact Sheet . . . . . . .84 Plutonium-238: Clinical Decision Guide Fact Sheet . . . .85 Plutonium-239: Clinical Decision Guide Fact Sheet . . . .86 Radium-226: Clinical Decision Guide Fact Sheet . . . . . .88 Strontium-90: Clinical Decision Guide Fact Sheet . . . . .89 Uranium: Clinical Decision Guide and Nephrotoxicity Fact Sheet . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .90 7.12.1 Treatment Based on Radiological Properties of Uranium (CDG) . . . . . . . . . . . . . . . . . . . . . . . . . . .90 7.12.2 Treatment Based on Nephrotoxic Properties of Uranium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .91 Clinical Decision Guide Technical Details . . . . . . . . . . . .93 7.13.1 Noniodine Radionuclides . . . . . . . . . . . . . . . . . . . .93 7.13.2 Iodine Radionuclides . . . . . . . . . . . . . . . . . . . . . . .94 7.13.3 Uranium Mass and Activity Calculations . . . . . .94

8. Rapid Determination of Internal Contamination . . . . . . .96 8.1 Rapid Identification of Radionuclide(s) Involved . . . . . . .97 8.2 Screening for External Contamination . . . . . . . . . . . . . . .98 8.3 Direct (in vivo) Screening for Internal Contamination . .99 8.3.1 Detection of Internal Contamination by Direct Measurement . . . . . . . . . . . . . . . . . . . . . . . . . . . . .99 8.3.1.1 GM Survey Meter . . . . . . . . . . . . . . . . .101 8.3.1.2 Whole-Body and Lung Counters . . . . .101 8.3.2 Hospital Equipment for the Detection and Quantitation of Radionuclides . . . . . . . . . . . . . .103 8.3.2.1 GM Survey Meter . . . . . . . . . . . . . . . . .103 8.3.2.2 Nuclear Medicine Thyroid-Uptake Probe . . . . . . . . . . . . . . . . . . . . . . . . . . .104 8.3.2.3 Portal Monitors. . . . . . . . . . . . . . . . . . .105 8.3.2.4 Nuclear Medicine Gamma Camera . . .106 8.3.2.5 Pulse-Height Analyzer and Radionuclide Windows . . . . . . . . . . . . . . . . . . . . . . . .107 8.3.3 Measuring Internal Contamination with Hospital Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .110 8.3.3.1 Using GM Survey Meters to Assess Internal Contamination with Certain Gamma-Emitting Radionuclides . . . . .110 8.3.3.2 Using a Gamma Camera or ThyroidUptake Probe to Assess Internal Contamination with Certain GammaEmitting Radionuclides . . . . . . . . . . . .110

viii / CONTENTS

8.4

8.5

8.3.4 Wound Monitoring . . . . . . . . . . . . . . . . . . . . . . . Indirect (in vitro) Determination of Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.1 Nasal Swabs . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.2 Urine Samples . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.3 Blood Samples . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.4 Fecal Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . 8.4.5 Analysis of Indirect Bioassay Samples . . . . . . . Rapid Screening of Persons to Identify Radionuclide Intakes that Exceed the Clinical Decision Guide . . . . .

112 113 113 114 115 116 116 117

9. Medical Management of Internally-Contaminated Persons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 121 9.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 121 9.2 General Clinical Guidance for the Treatment of Internal Contamination . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 122 9.3 Making Requests for Radiological Countermeasures . . 124 9.4 Medical Management Decisions for PotentiallyContaminated Persons . . . . . . . . . . . . . . . . . . . . . . . . . . 125 9.4.1 Triage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 125 9.4.2 Prioritizing Children and Pregnant Women . . . 126 9.4.3 Medical Decisions During the Early Phases of a Radiation Incident . . . . . . . . . . . . . . . . . . . . . . . 127 9.4.4 Choice of Decorporation Therapy for InternallyDeposited Radionuclides . . . . . . . . . . . . . . . . . . 128 9.4.5 Self-Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . 129 9.4.6 Contaminated Wound Management . . . . . . . . . 129 9.4.7 Using DTPA on Radionuclide-Contaminated Wounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 130 9.5 Medical Management of an Americium-241 Intake . . . 131 9.5.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131 9.5.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131 9.5.3 Medical Follow-Up After Treatment . . . . . . . . . 134 9.6 Medical Management of a Cesium-137 Intake . . . . . . . 134 9.6.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 134 9.6.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 135 9.6.3 Medical Follow-Up After Treatment . . . . . . . . . 135 9.7 Medical Management of a Cobalt-60 Intake . . . . . . . . . 135 9.7.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 135 9.7.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138 9.7.3 Medical Follow-Up After Treatment . . . . . . . . . 138 9.8 Medical Management of an Iodine-131 Intake . . . . . . . 138 9.8.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138 9.8.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 140 9.8.3 Medical Follow-Up After Treatment . . . . . . . . . 140

CONTENTS

9.9

9.10

9.11

9.12

9.13

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Medical Management of an Iridium-192 Intake . . . . . .140 9.9.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .140 9.9.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .145 9.9.3 Medical Follow-Up After Treatment . . . . . . . . . .145 Medical Management of a Plutonium-238 or -239 Intake . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .145 9.10.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .145 9.10.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .146 9.10.3 Medical Follow-Up After Treatment . . . . . . . . . .146 Medical Management of a Radium-226 Intake . . . . . . .146 9.11.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .146 9.11.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .147 9.11.3 Medical Follow-Up After Treatment . . . . . . . . . .147 Medical Management of a Strontium-90 Intake . . . . . .147 9.12.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .147 9.12.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .147 9.12.3 Medical Follow-Up After Treatment . . . . . . . . . .150 Medical Management of a Uranium-235 or -238 Intake 150 9.13.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .150 9.13.2 Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .153 9.13.3 Medical Follow-Up After Treatment . . . . . . . . .153

10. Social, Psychological and Communication Issues Associated with Screening and Monitoring a Population . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .155 10.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .155 10.2 General Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .156 10.3 Program and Center Names . . . . . . . . . . . . . . . . . . . . . .156 10.4 Enlisting the Public as a Partner . . . . . . . . . . . . . . . . . .156 10.5 Communicating with Members of the Public . . . . . . . . .157 10.6 Proactive Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . .157 10.7 Information Hotlines . . . . . . . . . . . . . . . . . . . . . . . . . . . .158 10.8 Special Services, Approaches and Materials for Children . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .159 10.9 Persons with Reproductive and Fertility Concerns . . . .160 10.10 Other Special Populations . . . . . . . . . . . . . . . . . . . . . . . .161 10.11 Addressing Staff Concerns and Information Needs . . . .161 10.12 Staff Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .162 10.13 Training Exercises . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .162 10.14 Understanding and Addressing Responder Concerns and Information Needs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .163 11. Long-Term Follow-Up of Individuals . . . . . . . . . . . . . . . . .165 11.1 Identification of the Population to be Followed . . . . . . .165 11.2 Classification of Persons to be Monitored or Followed . .168 11.3 Follow-Up Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . .171

x / CONTENTS 12. Scalability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 172 13. Assessment of Current Capacity in the United States to Perform Population Screening, Decontamination and Monitoring for Internal Contamination . . . . . . . . . . . . . . 180 13.1 Availability of Equipment and Resources . . . . . . . . . . . 181 13.2 Laboratory Capabilities . . . . . . . . . . . . . . . . . . . . . . . . . 182 13.3 Training Needs on Use of Equipment . . . . . . . . . . . . . . 183 13.4 Radiation Volunteers to Support Population Screening 185 13.5 Biodosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 186 13.6 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 187 14. Conclusions and Recommendations . . . . . . . . . . . . . . . . . 188 14.1 Recommendations for Planning . . . . . . . . . . . . . . . . . . . 188 14.2 Recommendations Related to Screening and Treatment of a Population for Internal Contamination . . . . . . . . . 189 14.3 Recommendations for Additional Work . . . . . . . . . . . . . 190 Appendix A. Radiological Properties of Radionuclides Considered in this Report . . . . . . . . . . . . . . . . . . . . . . . . . . 192 Appendix B. How to Perform a Radiation Survey for Contamination: Instructions for Workers . . . . . . . . . . . . B.1 Screening Survey . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.2 Complete Whole-Body Survey . . . . . . . . . . . . . . . . . . . . B.3 Most Common Mistakes Made During the Survey . . . .

202 202 203 204

Appendix C. How to Distinguish Between Alpha, Beta and Gamma Radiation Using a Geiger-Muller Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 205 C.1 Determining the Presence of an Alpha-Emitting Radionuclide Using Only a Geiger-Muller Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 206 C.2 Determining the Presence of Strontium-90 (or other pure beta emitters) Using a Pancake Geiger-Muller Survey Meter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 206 Appendix D. Survey and Registry Forms . . . . . . . . . . . . . . . . 208 D.1 Contamination Survey Sheet . . . . . . . . . . . . . . . . . . . . . 208 D.2 Registry Form . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 211 D.3 ATSDR Rapid Response Registry Survey Form ATSDR Rapid Response Registry Survey Form . . . . . . . . . . . . . 214 Appendix E. How to Perform Decontamination at Home . . 216

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Appendix F. Using Geiger-Muller Survey Meters to Assess Internal Contamination for Selected Gamma-Emitting Radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .218 Appendix G. Collection and Preparation of Biological Samples for Radioanalysis . . . . . . . . . . . . . . . . . . . . . . . . . .228 G.1 Urine Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .228 G.2 Main Collection Issues . . . . . . . . . . . . . . . . . . . . . . . . . . .228 Appendix H. Shipping of Biological Samples . . . . . . . . . . . . .231 H.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .231 H.2 Regulatory Information: Brief Summary . . . . . . . . . . . .231 H.3 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .232 H.4 General Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples . . . . . . . . . . . . . . . . . . . . . . .232 H.5 Specific Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples . . . . . . . . . . . . . . . . . . . . . . .234 H.6 Website Links for Some Commercial Shippers . . . . . . .234 Appendix I. Population Screening and Monitoring Implications of Two Urban Contamination Incidents . .235 I.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .235 I.2 Overview of Two Cases . . . . . . . . . . . . . . . . . . . . . . . . . .236 I.2.1 Goiânia, Brazil . . . . . . . . . . . . . . . . . . . . . . . . . . .236 I.2.2 London, United Kingdom . . . . . . . . . . . . . . . . . .236 Appendix J. Pregnancy Categories for Drug Use . . . . . . . . . .238 Appendix K. Emergency Phone Numbers for Government Officials to Request Assistance Following a Radiological or Nuclear Incident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .239 Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .240 Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . .247 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .248 The NCRP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .264 NCRP Publications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .273

1. Executive Summary This Report is the second of two reports by the National Council on Radiation Protection and Measurements (NCRP) that focus on measurement of radionuclides deposited internally in a population exposed in a radiological or nuclear incident. The first report, NCRP Report No. 161, entitled Management of Persons Contaminated with Radionuclides (NCRP, 2008a), is an update and expansion of NCRP Report No. 65, Management of Persons Accidentally Contaminated with Radionuclides (NCRP, 1980) that provides detailed guidance for many radionuclides in a much broader range of exposure scenarios. The present Report focuses on screening a population exposed to one or more radionuclides that may be involved in a radiological or nuclear incident. Screening as used in this Report means rapid assessment and measurement for external or internal contamination. Screening is a singular activity intended to enable intervention and management of persons who may have been or who were exposed to radioactive contamination from a radiological dispersal device (RDD) or a nuclear incident. Thus, a group of people who may be contaminated are first screened for external contamination (a single survey). If contaminated externally, they are decontaminated prior to screening for internal contamination (a single measurement). The emphasis is on rapid screening that quickly identifies those patients who may need medical treatment to decorporate internally-deposited radionuclides. This Report uses the term monitoring in a broad sense, which is either the monitoring of a population for internal contamination by taking a number of measurements or collecting a number of bioassay samples over a period of time or the medical monitoring of a population for subsequent medical effects depending on the context. External monitoring for radioactive contamination is not addressed in detail in this Report but is discussed in much greater detail in other referenced literature. The radionuclides in this Report were selected based on information published by various regulatory and advisory organizations including the Centers for Disease Control and Prevention (CDC, 2009) and the U.S. Department of Health and Human Services (DHHS, 2009). This Report addresses screening a population for internal contamination and the possible use of decorporation 1

2 / 1. EXECUTIVE SUMMARY therapy for patients who exceed a level of internally-deposited radionuclide called the Clinical Decision Guide (CDG), a concept developed in NCRP Report No. 161 (NCRP, 2008a). This Report provides information intended for use by emergency responders and planners and public-health officials for development of emergency response plans that include screening a population for internally-deposited radionuclides. However, this Report is not intended to be a procedure manual that can be incorporated directly into an existing radiological response plan. This Report should be used to evaluate and upgrade, to the extent feasible, the capability to screen small, medium and large populations for the presence of internally-deposited radionuclides and to assist in decisions about the possible medical treatment of patients who contain levels of radionuclides in excess of the CDG. This Report also discusses broadly the monitoring of a population for external contamination and decontamination as necessary prior to screening for internal contamination and refers to other literature that treats these subjects more completely. Background information is provided, including various settings in which members of the public might incur internal depositions of various radionuclides. These sections may be of particular interest to those who are involved in development of emergency response plans. Initial screening of individuals at the scene and hospital and mass screening are described, and previous experience with internal radionuclide contamination of members of the public is summarized. Incident command is described briefly to provide the reader with an appreciation for the importance of incident response coordination between the community and the hospital(s) that will be receiving potentially-contaminated patients. It also emphasizes the importance of timely communications during an incident to optimize medical care and treatment of patients to decorporate internally-deposited radionuclides. Efficient and frequent communications facilitate the ability of hospitals to tailor their response to the number of patients who need treatment of acute injuries and screening of patients for internal contamination. The process of sorting victims from a radiological incident based on their risk of having significant radiation exposure or contamination is referred to as radiological triage. The outcome of radiological triage depends partly on whether people receive direct irradiation or contamination from the plume or the contaminated environment. For purposes of this Report, it is assumed that most people who are impacted by the incident or who perceive that they may be contaminated will wish to be screened for contamination.

1. EXECUTIVE SUMMARY

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The CDG can be used by physicians to consider whether the level of radionuclide intake by a patient warrants use of decorporation therapy. This concept was first published in NCRP Report No. 161 (NCRP, 2008a) to provide guidance on when physicians may want to consider the use of decorporation agents to reduce the radiation dose that a patient might receive from an internally-deposited radionuclide. Hospital equipment discussed in this Report that may be used to determine whether the CDG has been exceeded includes nuclear medicine cameras and uptake probes, portal monitors, multichannel analyzers, and portable survey meters for direct screening. Procedures discussed for indirect screening include use of nasal swabs and the collection and analysis of urine and fecal samples. Physicians will be particularly interested in Section 9, which provides guidance on the medical management of patients who have been identified as containing one or more internally-deposited radionuclides. Section 9 provides information on the use of the CDG in making treatment decisions and general guidance for treating patients. It also provides information on how patient care providers can request equipment, supplies and pharmaceuticals from the Strategic National Stockpile (SNS) (CDC, 2008). Medical management guidance is restricted to the radionuclides of interest in this Report and includes consideration of treatment with decorporation agents and over-the-counter products. In addition to the many technical and logistical issues associated with screening patients for internal contamination; social, psychological and communications issues are important. This Report provides guidance on organizing and conducting a screening program in a manner that is sensitive to these issues. It also offers guidance on practical needs such as information hotlines, setting up population screening centers, and considering the needs of special populations including children and pregnant women. Exposure of a population to radioactive materials from an RDD or an improvised nuclear device (IND) may increase the risk of deterministic or stochastic effects from external exposure and internally-deposited radionuclides. This Report describes a long-term follow-up program to monitor (i.e., to follow the health of) this population. Guidance includes the immediate identification of patients who should be included in long-term follow-up and a survey form that includes appropriate identification information for long-term follow-up. Screening activities can be readily undertaken for a few patients but can become complicated or impractical when large numbers of patients present themselves. This Report discusses scalability of emergency response plans and recommends that communities and

4 / 1. EXECUTIVE SUMMARY hospitals consider dividing their responses into three classes based on the numbers of people involved: small (e.g., 1 to 10 people); medium (e.g., tens of people), and large (e.g., over 100 people). Also included in this Report is an assessment of the current capacity within the nation to perform rapid screening of a population exposed to radioactive materials from a radiological or nuclear incident. Surveys of manufacturers and state radiation-control program directors suggest that most communities that responded to the survey have the necessary resources to provide rapid screening of a limited number of patients to determine whether they are internally contaminated. NCRP (2008a) identified major areas of research and development needed to address the management of contaminated persons. These areas are decontamination facilities to remove external contamination, instrumentation and modeling for assessment of internal contamination, bioassay facilities and automation, biomarkers and devices for biodosimetry, software for rapid estimates of organ equivalent dose and effective dose, decorporation agents for use postexposure, medical follow-up of exposed populations, and educational programs in health physics. In addition, this report identified the need for research to develop a more complete understanding of the capabilities of states, communities and hospitals to respond to a radiological incident and to screen populations of different sizes and in different locations for internal contamination. While most communities probably could provide appropriate screening and treatment of a limited number of internally-contaminated patients, more complete information is needed to fully understand capabilities to respond to incidents that create a need to screen large numbers of patients. Several appendices to this Report provide both practical and more detailed information. While the purpose of this Report is to address internal contamination, accurate screening for internal contamination cannot be performed with survey instruments if the patient has external contamination. The appendices contain practical advice on how to perform radiation surveys, how to distinguish between alpha, beta and gamma radiation using a Geiger-Mueller (GM) survey meter equipped with a pancake probe, and how to perform decontamination at home. Survey and registry forms are provided to enter patients into a registry for recordkeeping and for long-term follow-up. GM count rates from CDC (2009) are provided to determine whether a patient has internal contamination in excess of the appropriate CDG values. Procedures are provided for preparation of urine samples and shipping biological samples for analysis. The appendices contain a summary of the 210Po poisoning of a single individual in United Kingdom and contrasts that with the

1. EXECUTIVE SUMMARY

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large-scale 137Cs contamination incident in Goiânia, Brazil. The purpose for these case studies is to compare the procedures used to identify the radionuclide and the individuals who were exposed and to describe monitoring for external contamination and screening for internal contamination. Emergency phone numbers are provided for requesting government assistance following a radiological or nuclear incident. Finally, various U.S. Food and Drug Administration (FDA) drug categories for pregnant women are defined. This Report makes several recommendations and emphasizes a number of points that are highlighted within the various sections. For preplanning purposes, emergency planners should be aware that radiological and nuclear incidents have the potential to cause widespread contamination of people and the consequences will place enormous stress upon emergency response and healthcare organizations. In the development of an emergency response plan, emergency planners should recruit volunteers such as health physicists, radiation-safety officers, and medical personnel who could serve as subject matter experts or perform specific tasks to support planning and response to a radiological incident. Due to their regular duties, nuclear medicine staff may assist but will not be able to support fully the needs of the emergency department for quick screening of patients during a mass casualty incident. Therefore, other hospital staff should be trained and made available to assist with these activities. The planning process should address capabilities for rapid detection and identification of radionuclides. This is particularly challenging for pure alpha- and beta-emitting radionuclides. Screening a population for internal contamination requires planning and practice. In particular, the use of nuclear medicine cameras and thyroid probes to screen patients following exposure to radionuclide contamination from a radiological or nuclear incident requires advance planning, training and rehearsal. It cannot be implemented ad hoc during an emergency. Local municipalities and public-health agencies should develop procedures for requesting, receiving and distributing SNS assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. Planning for population screening should incorporate provisions for establishing stakeholder advisory boards or other mechanisms for making the public a partner and should take into account psychosocial, behavioral and communication issues likely to arise during the screening of a population. Regarding the management of contaminated patients in all but the most extreme cases, standard precautions provide adequate

6 / 1. EXECUTIVE SUMMARY protection to healthcare workers to prevent secondary radioactive contamination. Patients who have suffered life-threatening injuries should be given medical care immediately, without regard to contamination. The presence of external contamination should almost never delay urgent medical care. Emergency personnel should be aware that patients with the most serious injuries are also likely to be the most contaminated both externally and internally. The presence of internal contamination is rarely life-threatening to a patient. Therefore, treatment of conventional injuries that may be immediately life threatening should take precedence over decorporation of internal contamination. Rapid identification of the radionuclide(s) involved in a radiological or nuclear incident is essential to the selection of appropriate methods for assessment of internal contamination and subsequent treatment decisions. The CDG may be used by physicians as a basis for medical treatment of individuals who have internal radionuclide deposition. The CDG is not intended to instruct physicians on a specific course of action such as administration of decorporation agents. Rather, CDGs are intended as a tool to be used to help a physician determine when radiation exposure may have clinical significance. NCRP recommends a bioassay action-level or benchmark equal to one-half of the CDG value for identifying those individuals for whom continued bioassays should be considered for the purpose of correlating internal contamination with long-term biological effects. Community or hospital emergency response plans should include resource needs to scale the response to the size of the incident. The following recommendations are made in this Report to improve capability of the nation to screen populations for internal contamination following exposure to a radiological incident. Additional surveys of local communities and states should be conducted to assess more accurately their capability to screen populations for internal contamination. Consideration should be given to utilizing strategically located clinical laboratories to analyze bioassay samples for radionuclide identification and concentrations. A limiting factor here is the small number of radiochemists in the United States. Additional research should be conducted to define radiological instrument or bioassay trigger levels that correspond to the CDG for a wider variety of radionuclides such as those identified in NCRP Report No. 161 (NCRP, 2008a), which may be produced by the explosion of an IND.

2. Introduction 2.1 Overview Hospitals, first-responder groups, local and state health departments, and nongovernmental organizations have extensive experience responding to a variety of disasters such as major storms and airplane crashes. Experience with some other kinds of threats is far more limited (e.g., no large-scale radiological or nuclear accidents or terrorism incidents have occurred in the United States). In recognition of the need for an all-hazards approach, planners now include response to radiological incidents in their emergency plans. Many hospitals include in their emergency plans a component for responding to an incident that could result in a few patients who are externally contaminated with radionuclides (e.g., an industrial accident involving one or more radionuclides). Until recent years, these plans anticipated small numbers of patients and focused mostly on external contamination. Currently, emergency response planning has increasingly included possible incidents that could result in tens, hundreds, or even larger numbers of patients with both external and internal radionuclide contamination (HSC, 2009). This necessitates development of a flexible, scalable approach to emergency response, one that will help manage the variety, intensity and duration of an emergency (JCAHO, 2009). Thus, for radiation emergency response, the plan should accommodate the variety of radiation sources that could be involved in the emergency, the intensity and magnitude of an incident, and the duration of the incident. Furthermore, for some scenarios hospitals should not assume that conditions during an incident will allow local agencies to support the hospital’s response. Therefore, hospital emergency plans should be flexible (e.g., using their own security staff to control traffic near the entrance to the emergency department if local law enforcement is occupied at the scene of the incident). NCRP Report No. 161 (NCRP, 2008a) presents detailed information on a broad range of exposure settings, radionuclides, and numbers of persons that might be involved. The focus of this Report is much narrower and is directed primarily to actions associated with a population of people that might be involved in a large-scale radiological incident. Therefore, the list of radionuclides of most interest (Table 2.1) is a small subset of the list in Report No. 161 (NCRP, 7

8 / 2. INTRODUCTION 2008a) and is similar to lists being used by other organizations including CDC (2009) and DHHS (2009). The radionuclides 131I, 235 U and 239Pu are also listed relative to a possible nuclear incident. In the event of an incident involving an IND, the number of detectable radionuclides will significantly exceed those considered in this Report and the radioactive fallout will reach across many jurisdictions, potentially involving multiple states (HSC, 2009). Population screening following a nuclear detonation will be conducted primarily to detect and remove external contamination. Considerably more work is needed to develop a robust, rapid screening program to identify internal contamination of survivors of a nuclear detonation and to make medical treatment decisions in the face of unparalleled medical demands. However, in such an incident, decorporation decisions could be based on internal contamination by 131I and 137 Cs, radionuclides that are included in Table 2.1. One of the first tasks for emergency management following a radiation exposure incident or environmental release of radioactive material is to recognize that radionuclides were involved and to identify the radionuclide(s). The U.S. Environmental Protection Agency (EPA) Radiological Emergency Response Team based in the EPA Office of Radiation and Indoor Air and regional offices, provides advice and responds to emergencies involving releases of radioactive materials by providing environmental measurement and guidance activities (EPA, 2010), but communities and states should be aware that they are responsible for initial assessments because the Radiological Emergency Response Team will not be able to respond immediately. Community or state agencies should also request assistance from the U.S. Department of Energy (DOE) Federal Radiological Monitoring and Assessment Center (FRMAC). FRMAC coordinates federal radiological monitoring and assessment activities with those of local and state agencies (DOE, 2009a). Within several hours of notification, DOE will deploy a team of health-physics specialists from the Radiological Assistance Program and will make a determination of whether to send Phase I of a Consequence Management Response Team consisting of technical and management personnel who should reach the site of the emergency within 6 to 10 h. The Radiological Assistance Program and Consequence Management Response Team Phase I response will be supported by the Consequence Management Home Team consisting of technical and scientific personnel from DOE laboratories (NNSA, 2009). A complete FRMAC for the radiological emergency should be fully operational within 24 to 36 h and will operate under the parameters of the Incident Command System (ICS) (Section 5) to

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TABLE 2.1—Characteristics of radionuclides considered in this Report.a Half-lifeb

Emissions

60

5.27 y

β, γ

90

28.8 y

β

Medical treatment, radioisotopic thermal generators

131 d

8.02 d

β, γ

Medical diagnosis and treatment

137

30.2 y

γ

192

73.8 d

β, γ

226

1,600 y

α, β, γ

Self-luminous products, cancer therapy (no longer used, legacy products remain)

235

7.04 × 108 y

α, β, γ

Nuclear reactor fuel, nuclear weapons

238

4.47 × 109 y

α, β, γ

Anti-tank weapons, radiation shielding

238

87.7 y

α, n

239

24,100 y

α

241

432 y

α, γ

Radionuclide

Co

Sr d

I

Cs d

Ir Ra d

Ud Ud Pu

Pu d Am

Usesc

Industrial radiography, stereotactic radiosurgery, commercial irradiators

Radiography, well logging, soil density gauges, self-shielded irradiators Industrial radiography, brachytherapy

Neutron generators, calibration sources, radioisotopic thermal generators Nuclear weapons Soil moisture, hydrocarbon content, smoke detectors

aAdditional data on the radiological characteristics of the radionuclides listed in this table and their progeny are given in Appendix A. b ICRP Publication 107 (ICRP, 2008). cSee NCRP Report No. 161 (NCRP, 2008a) for typical source activities. d Likely presence of radioactive decay progeny of importance as well as the indicated radionuclide.

10 / 2. INTRODUCTION develop a monitoring and screening plan. The Interagency Modeling and Atmospheric Assessment Center will provide plume model predictions which will be helpful in determining where people were relative to the plume when they became contaminated. But once again, communities should prepare to provide immediate response to the emergency including emergency medical treatment and radiological response. Because federal support may not arrive for hours, local emergency management will be responsible to provide appropriate triage, screening, and subsequent external decontamination of individuals as needed. Local emergency management and health professionals can call the CDC Operations Center for immediate telephone advice on emergency response and emergency patient care. After external decontamination of each casualty, screening for internal contamination should be completed. However, emergency patients should be stabilized before external decontamination is attempted beyond clothing removal. Other literature sources provide guidance and methodology on external decontamination of patients (e.g., CDC, 2007; CRCPD, 2006; NCRP, 2001; 2005; 2008a). Ideally, first responders would survey medically-stable patients for external contamination and decontaminate them as necessary before release or transportation to a hospital. However, in a large-scale incident, some people are likely to self-evacuate and report directly to an emergency department or community reception center. Therefore, medical personnel should confirm that external decontamination is complete prior to screening or treatment of the patient for internal contamination. Emergency response personnel should be aware that the most seriously injured are also the most likely to be externally and internally contaminated due to their proximity to the incident (DHHS, 2009). After confirmation that external decontamination is complete, an assessment of potential internal contamination should be made (Section 8). The goal of screening the exposed population is to determine whether internal contamination is high enough to justify appropriate medical treatment to expedite biological removal of the internal contamination (decorporation therapy) thereby reducing the subsequent irradiation of critical tissues (Section 9). Clinical decisions on decorporation therapy are guided by a CDG for each radionuclide. The CDG concept is discussed in Section 7. The CDG represents a level of internal radionuclide deposition at, or above, which a physician may wish to consider medical treatment to enhance removal of the radionuclide from the body. Members of the public may be identified for bioassay, medical treatment, biodosimetry, dose reconstruction, and entry into long-

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term follow-up programs and registries (Section 11). The purpose of screening the population is to reduce radiation exposure to people by making timely and adequate measurements of potentiallycontaminated individuals so they can receive prompt and adequate external decontamination, medical treatment of injuries, and decorporation therapy. The current capacity to implement these activities at the local and state levels is limited (Section 12). However, the local community should be prepared to monitor for external contamination, decontaminate, screen for internal contamination, administer pharmaceutical countermeasures (decorporation), and enter patients into a registry during the first 24 to 48 h or more after which federal assistance should become available. Section 13 provides an assessment of the current capacity within the nation to perform rapid screening of a population exposed to radioactive materials from an RDD, IND, or similar mass casualty incident. Surveys of manufacturers and state radiation-control program directors suggest that most communities have the necessary resources to provide rapid screening of a limited number of patients to determine whether they are internally contaminated. Resources considered include those discussed in Section 8. Accurate screening for internal contamination will be difficult in patients with external contamination. This Report contains several appendices that provide both practical and more detailed information on screening. Appendix A provides additional information on the radiological properties of radionuclides covered in this Report. Appendix B explains how to perform screening and wholebody surveys for external contamination and Appendix C provides information on how to distinguish between alpha, beta and gamma radiations using a GM survey meter. Related survey and registry forms are given in Appendix D. Appendix E contains instructions for people who perform self-decontamination of external radionuclide contamination at home. Appendix F provides important information on how GM survey instruments can be used to assess internal depositions of certain gamma-emitting radionuclides. Detailed instructions for the collection of bioassay samples for in vitro bioassay analyses are given in Appendix G and instructions for possible shipments of these samples to analytical facilities are given in Appendix H. Appendix I provides a summary of the 210Po poisoning of a single individual in London, United Kingdom and contrasts that with the large-scale 137Cs contamination incident in Goiânia, Brazil. The purpose of Appendix I is to compare the procedures used to identify the radionuclide and the individuals who were exposed and to describe screening and monitoring of individuals. Appendix J defines FDA drug categories for pregnant women

12 / 2. INTRODUCTION and Appendix K contains emergency phone numbers for contacting government officials to request assistance with a radiological or nuclear incident. NCRP has a long history of providing recommendations on response to radiological emergencies and has issued several publications that offer specific advice on response to a radiological emergency. The first of these reports, NCRP Report No. 29, Exposure to Radiation in an Emergency (NCRP, 1962), provided general advice and was superseded by NCRP Report No. 42, Radiological Factors Affecting Decision-Making in a Nuclear Attack (NCRP, 1974), which provided guidance for use in response to a large-scale nuclear disaster involving an intense and uncontrolled exposure of a population. NCRP Report No. 65, Management of Persons Accidentally Contaminated with Radionuclides (NCRP, 1980) was a comprehensive report directed primarily to occupational contamination incidents occurring in nuclear facilities. Many data from, and ideas on, treatment of contaminated patients were collected in this report and its recommendations were directed toward physicians responsible for managing these patients. Report No. 65 provided treatment information for many chemical elements and described pertinent information on numerous radionuclides. NCRP Commentary No. 10, Advising the Public About Radiation Emergencies: A Document for Public Comment (NCRP, 1994) focused on communication issues. NCRP Report No. 138, Management of Terrorist Events Involving Radioactive Material (NCRP, 2001), dealt with releases of radioactive materials as the result of a deliberate act of terrorism and provided recommendations on medical management of contaminated individuals, psychosocial issues, communication with members of the public and media, and training of responders. Commentary No. 19, Key Elements of Preparing Emergency Responders for Nuclear and Radiological Terrorism (NCRP, 2005) focused on detection and personal protection equipment requirements for emergency responders, decontamination equipment and medical supplies, and radiationprotection training of emergency responders. NCRP Report No. 156, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and Procedures for Their Assessment, Dosimetry and Treatment (NCRP, 2006) provides advice on treatment of individuals who may have been contaminated internally through wounds. The most recent report, NCRP Report No. 161, Management of Persons Contaminated with Radionuclides (NCRP, 2008a) is an update of NCRP Report No. 65 that provides recommendations for a broader range of radionuclides and exposure scenarios and for onsite and prehospital actions when responding to contamination incidents and the management of contaminated patients at the hospital.

2.2 PURPOSE OF THIS REPORT

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CDC is the lead federal agency charged with assisting local, tribal and state agencies with population monitoring and decontamination (CDC, 2005a). CDC hosted a Population Monitoring Roundtable in January 2005 that convened representatives from various federal agencies, local and state public-health agencies, clinicians, private public-health organizations, and health professional organizations to assist CDC in evaluating the best methods and techniques for performing internal population screening for radiation, inventorying current U.S. capability to perform this screening, and developing guidance that local and state publichealth agencies can use in preparing to respond to a radiological or nuclear incident (CDC, 2005b). Roundtable participants recognized the complexity of determining internal contamination in a population that had been contaminated externally. They did not identify specific methodologies for determining internal contamination but suggested that external contamination might be a good indicator for internal contamination, and that scalability should be considered. They also recommended that brief histories be elicited for exposed or potentially-exposed individuals, to include basic demographic information and a description of the circumstances regarding their exposures. This information could be used subsequently if resources are available to conduct follow-up for late health effects. They also identified challenges in treating patients for internal contamination including identifying those who need treatment, addressing scalability for the number of patients who are identified as containing internally-deposited radionuclides, and specific agebased guidelines for treatment. 2.2 Purpose of this Report The purpose of this Report is to provide recommendations for local, state and federal responding communities regarding internal radionuclide depositions, decorporation therapy, equipment and laboratory needs. This Report is intended to provide general advice on screening members of the public for internal contamination following an emergency during which people may have been contaminated with radionuclides. Any early radiation injury will most likely be due to external radiation exposure from large particles that fall quickly to the ground and are not respirable. Internal contamination is unlikely to pose a short-term danger to the population (NA/IOM, 2009). Nevertheless, it is important to be aware that following the explosion of an RDD or IND, patients with the most serious medical injuries are likely to have been close to the explosion and are likely to be internally contaminated. Also, in the first

14 / 2. INTRODUCTION 10 to 15 min, anyone within 500 m of the explosion is at risk of inhaling a large amount of activity (Harper et al., 2007). This Report also provides advice on treatment methods to reduce internal radionuclide depositions in individuals who exceed a CDG of radioactive material. Finally, this Report provides recommendations for registering contaminated individuals in a registry for longterm follow-up of internally-contaminated patients. 2.3 Target Audiences of this Report This Report should be useful to a wide range of preparedness and response organizations, emergency management jurisdictions, and policy makers at the local, state and federal levels involved in emergency preparedness and response. However, this Report is aimed primarily at emergency responders including hazardous materials teams, hospitals, and health departments (NCRP, 2005) who will be involved in screening for external contamination followed by decontamination as necessary, treatment of physical injuries, screening for internal contamination, and providing decorporation therapy, if necessary, in patients who may have been exposed to radionuclides during an incident. This Report is based largely on recommendations in other publications, particularly NCRP Report No. 161 (NCRP, 2008a) and is intended as a guide to those who would participate in the screening of small to large numbers of patients who may contain internal radionuclides from exposure to radioactive contamination following a radiological or nuclear incident. This Report is presented in sections that address separate issues including screening the exposed population, treatment of those who have quantities of an internally-deposited radionuclide that warrant removal efforts, and long-term follow-up of the exposed population. Thus, emergency responders may find the sections on screening a population to be most useful whereas physicians, nurses, and healthcare providers may find most value in the section on treatment of patients. However, the entire Report will be useful to those who are responsible for emergency planning. 2.4 Scope of this Report This Report addresses screening of a population, small or large, that may have been exposed to radioactive material from a radiological incident. This population includes emergency responders and anyone potentially exposed to radioactive materials from the incident. It does not address screening of workers exposed to radionuclides as a result of other types of employment, members of the public exposed to patients who have been administered radioactive

2.4 SCOPE OF THIS REPORT

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drugs or sources, or any other situation addressed by a licensed program that uses radioactive material. The radionuclides specifically covered in this Report (Table 2.1) have been limited to those that CDC (2009) has highlighted plus 226 Ra. NCRP (2008a) discusses numerous additional radionuclides and provides more detailed internal dose estimates. Bioassay methods in use today were developed and designed for the adult worker (e.g., whole-body counter size and metabolic models). Unlike adult workers, young people and members of the public who have special needs require unique methods, models and protocols to screen for the presence of radionuclides and to estimate radiation dose. This Report discusses and provides advice on the following issues related to screening and treatment of members of the public of any age who are internally contaminated: • recommended methods, protocols and practices that enable determination of internal contamination: - direct bioassay (in vivo); and - indirect bioassay (in vitro such as urine, feces, or nasal swabs). • elimination of external contamination prior to direct and indirect bioassays; • social, psychological and communication issues related to population monitoring and screening; • radionuclide-specific baseline values measured in the U.S. population, or if none, whether a national baseline data gathering program should be recommended; • age-specific internal contamination trigger levels (CDG values) that would indicate that decorporation therapy should be considered [e.g., countermeasures such as Prussian blue or diethylenetriamine pentaacetic acid (DTPA)]; • recommendations for the receipt, physical security, distribution, and administration of SNS assets; • recommendations for maintaining medical and pharmaceutical supplies and assets locally; • age-specific internal contamination trigger levels that indicate the need for a person to be entered in a long-term follow-up program or registry, and guidance for the following: - administration of a long-term health surveillance program; - content of the personal history needed to plan for subsequent long-term follow-up (i.e., specific questions to ask); and - type of follow-up testing needed.

16 / 2. INTRODUCTION • recommendations against the use of self-administered decorporation therapy (e.g., laxative, aluminum and magnesium antacid, etc.); • discussion of scalability of the above recommendations based on the size and magnitude of the incident or potentiallyexposed population; and • discussion of the current capacity in the United States to perform these functions, what is needed to augment these capacities, the training recommendations and to whom this training should be provided.

3. Background Information Radioactive materials are used in many types of settings, and people may become contaminated or may experience an intake of radionuclides in nearly all of these settings. Knowledge of the setting in which a contamination incident has occurred can help determine many relevant factors surrounding the incident, including the type of radiation involved (alpha, beta or gamma), the likelihood of an intake, the type of intake (inhalation, ingestion, absorption through the skin, or through wounds), the chemistry and biokinetics of the radionuclides, and the number of people likely to be involved. However, while large numbers of people in a densely populated area near the detonation of an RDD or IND might become contaminated and require external decontamination, few if any will be internally contaminated to a level that requires medical treatment (DHHS, 2009). After intake, radionuclides are distributed within the body according to the physical and chemical properties of the materials in which they are incorporated or the elemental chemical properties of the radionuclides after dissolution in body fluids. Some radionuclides are preferentially deposited in a particular organ or tissue and impart a higher radiation dose to it than to the remainder of the body, whereas others are distributed relatively uniformly in the body, imparting similar doses to most tissues. Until they are excreted from the body or diminished by radioactive decay, internally-deposited radionuclides will irradiate body organs and tissues. Radionuclides have different radiological properties [e.g., halflife, type(s) of radiation emitted, and specific activity]. The chemical form of a radionuclide, which influences its distribution within the body, and the radiological properties determine the radiation doses received by various organs and tissues. Detailed discussions of dosimetric aspects of the radionuclides covered in this Report can be found in NCRP Report No. 161 (NCRP, 2008a). 3.1 Internal Deposition of Radionuclides Skin provides a barrier that keeps most potentially-harmful contaminants outside the body. Exceptions include tritiated water and unbound radioiodine. However, contaminants can enter the body if the skin is breached or if they are inhaled or ingested (Figure 3.1). 17

18 / 3. BACKGROUND INFORMATION

Fig. 3.1. Generic biokinetic diagram showing routes of entry, metabolic pathways, and possible bioassay samples for internally-deposited radionuclides (adapted from ICRP, 1997).

3.1.1

Inhalation

The inhalation of radionuclides can lead to health risks to the persons exposed. Particles of certain sizes may penetrate deeply into the lungs, come into contact with body fluids, dissolve and enter the systemic circulation. Less soluble particles may remain in the lungs for months or years chronically irradiating the lung tissue and providing a longer-term opportunity for dissolution of the particles and absorption of their contents into the blood. Increased knowledge of anatomy and physiology of the respiratory tract and of the behavior and biological effects of inhaled radioactive particles is reflected in the detailed models published by the International Commission on Radiological Protection (ICRP) in Publication 66 (ICRP, 1994a) and Supporting Guidance 3 (ICRP, 2002a). These models make it possible to calculate radiation doses to various regions of the respiratory tract from the inhalation of airborne particles for a wide range of exposure conditions. NCRP Report No. 125 (NCRP, 1997) also provides a summary of scientific information and mathematical models that describe the deposition and clearance of various inhaled radionuclides from studies of exposed laboratory animals and people.

3.1 INTERNAL DEPOSITION OF RADIONUCLIDES

/ 19

Factors that influence the radiation dose pattern received from inhaled radionuclides (ICRP, 1994a) include: • Particle size: The aerodynamic and thermodynamic characteristics of inhaled radioactive particles determine whether particles are of an inhalable size, whether they are deposited or exhaled, and if deposited, where in the respiratory tract (ICRP, 1994a). Large (i.e., >10 μm) and fine (10 pediatric

Regional

Tier 5

Mass Casualties

>100 patients contaminated externally

Numerous blast injuries

>500

>20 adult >10 pediatric

National

Tier 6

TABLE 12.2—Example of mass casualty incident response tiers for emergency department response to injuries from an explosive radiological device.a

178 / 12. SCALABILITY

Not implemented

Limited

Expanded

Full

9Sztajnkrycer,

9

Full/unified command

11 – 20 patients contaminated internally

M. (2009). Personal communication (Mayo Clinic, Rochester, Minnesota).

Modeled after Mayo Clinic Level 1 Trauma Center Mass Casualty Incident Response.9 Triage category “Immediate”; life threatening injury. c Assumes that most patients were decontaminated before arrival at emergency department. d Hospital Incident Command System (Section 5.3).

b

a

HICSd

1 – 10 patients contaminated internally Unified command

>20 patients contaminated internally

12. SCALABILITY

/ 179

13. Assessment of Current Capacity in the United States to Perform Population Screening, Decontamination and Monitoring for Internal Contamination Section 8 described an approach to screen a potentially-contaminated population for internal contamination using radiation survey meters and existing nuclear medicine equipment. This section presents a snapshot of past and current capabilities in the United States to use existing nuclear medicine equipment to perform assessments of internal radionuclide contamination of patients during a mass casualty incident. The information presented in Section 13.1 was compiled from three sources: • a survey of gamma cameras conducted by Phillips Medical Systems in 2000;10,11 • a survey of state radiation-control programs conducted by CRCPD to determine the availability of equipment in hospitals that could be used to screen or monitor patients for internally-deposited radionuclides following exposure during an RDD or IND incident;12 and • a quarterly survey of critical assets available at New York State hospitals, conducted by the New York State Department of Health that included questions on availability of radiation-detection equipment.12 10Garrard,

J.L. (2000). Personal communication (Phillips Medical Systems, Andover, Massachusetts). 11Wood, C.M. (2007). Personal communication (Centers for Disease Control and Prevention, Atlanta). 12Salame-Alfie, A. (2008). Personal communication (New York State Department of Health, Troy, New York).

180

13.1 AVAILABILITY OF EQUIPMENT AND RESOURCES

/ 181

13.1 Availability of Equipment and Resources Even though the type and number of gamma cameras available at hospitals differ at various locations across the country mainly due to advances in technology and replacement of old equipment, gamma cameras are a resource that should be factored into community and hospital radiological emergency response plans. There is large variability in the types of gamma cameras available; the survey in 2000 of gamma cameras available at hospitals included 21 listed manufacturers and over 580 models installed. There were many different factors specific to individual cameras such as the thickness of the crystal (ranging from three-eighths to one inch), types of collimators (low, medium and high energy, with low energy being used for most of the routine diagnostic studies), and energy windows. Due to this variability, it was recommended that hospitals consider contacting the manufacturers to request specific modifications to the computer programs for their gamma camera to include a preset on the camera to allow for the detection of the gammaemitting radionuclides that may be found after an RDD or IND attack. Furthermore, it was also recommended that hospitals be requested to develop specific calibration factors for their cameras to minimize the need for the operators to manually perform those changes, thus allowing them to quickly convert their cameras to this nontraditional use. Results from a survey of RCPs13 showed that the percentage of hospitals with gamma cameras among the responding states range from 5 to 100 % and the availability of mobile cameras range from 0 to 46 %. The availability of thyroid counters at hospitals ranges from 6 to 75 %, though some radiation-control program responders indicated that hospitals could use the gamma cameras to measure thyroid uptakes of radioiodine. All nuclear medicine departments are required by regulation to possess calibrated radiation detection and measuring instruments, but it is difficult to assess the availability of GM survey meters outside of emergency departments. To obtain information on availability of survey meters at hospitals, it is necessary to survey the hospitals. Results from a survey of critical assets available in New York State hospitals provided an indication that such equipment is readily available at 82 % of the hospitals.13 The survey also provided data on the percent of hospitals that have pocket dosimeters (58 %) and the percent of hospitals that have portable radiation detectors 13Salame-Alfie,

A. (2008). Personal communication (New York State Department of Health, Troy, New York).

182 / 13. ASSESSMENT OF CURRENT CAPACITY available in the emergency department (11 %). Other information that was gathered from the New York State hospital survey included the percentage of hospitals that have a decontamination team (76 %), and, of those, 9 % have a separate decontamination plan for pediatric and special-needs patients. The majority of hospitals do not have portable portal monitors. Though many hospitals (~50 %) have waste-monitoring stations, those monitors are fixed and located at the waste-handling area. It may be possible to use them during an emergency if contaminated victims are directed to enter the hospital via the waste-handling area or if the portal monitor can be moved to the emergency department. About 68 % of the RCPs that responded to the survey have purchased portable portal monitors or have made arrangements to use portal monitors available at nuclear power plants or national laboratories.14 Most states that have nuclear power plants have access to whole-body counters (mostly located at the nuclear power plants). Few states have mobile decontamination facilities or equipment. Most RCPs reported that there is a National Guard civil support team in their state, but that the RCPs do not share resources with the civil support teams. Most state programs have access to spectrum analyzers within their program. Only a few state programs keep a list of physicists, medical physicists, dosimetrists, and radiation-safety officers that may be available to assist during a radiological emergency, though some have identified volunteers within these groups and have a program setup to qualify them as volunteers. 13.2 Laboratory Capabilities A critical component of a bioassay program is the ability to analyze the biological samples collected. Most hospitals are able to collect urine and blood samples during routine diagnosis. Traditionally, bioassays to assess internal radionuclide depositions in occupational settings are conducted on 24 h urine samples. It has been recognized that this protocol would not work during a radiological mass casualty incident due to the number of samples and the difficulty of obtaining and analyzing such a large number of samples. To address this issue, CDC laboratories have been working on a new methodology to analyze “spot” urine samples that involve a small amount of urine. As discussed in Section 8, CDC is also in the process of obtaining Clinical Laboratory Improvement 14Salame-Alfie,

A. (2008). Personal communication (New York State Department of Health, Troy, New York).

13.3 TRAINING NEEDS ON USE OF EQUIPMENT

/ 183

Amendments approval for a gross-alpha/gross-beta method using liquid scintillation and is developing a method to use gamma spectroscopy to conduct quick screening of urine samples to expedite analysis of those that require detailed isotopic analysis. General information on the collection and shipping information is provided in Appendices F and G. Most hospitals are traditionally able to collect and analyze blood samples for complete blood cell counts with differentials. The results of these analyses will provide doctors with an initial indication of high radiation exposure from an external source. CDC is also working on new protocols for blood sample analyses.15 As discussed in Section 8, the use of fecal samples to determine internal contamination during a radiological incident with mass casualties is not practical. These samples have often been collected to assess occupational exposures. The logistics of sample collection and complex analytical processes render then unsuitable for mass casualty applications. In addition, only a few laboratories have analytical capabilities to analyze them and mass casualty scenarios would overwhelm their capacities. While there are a number of laboratories that can analyze environmental samples for activity, very few laboratories in the country have the capability and capacity to conduct bioassays. Results of the CRCPD survey16 indicated that ~72 % of the existing state laboratories are able to handle radiological environmental samples in-house, but very few could do radiological clinical analyses. Some states have identified commercial laboratories that can handle such samples, but few have an existing contract/agreement to perform these analyses. 13.3 Training Needs on Use of Equipment To use the equipment described above, hospital staff need to be trained in its proper use. Every hospital that has a nuclear medicine department has at least one staff person trained in the use of the gamma camera and survey meters. During a mass casualty incident, the nuclear medicine staff will not be able to support the needs for quick screening of patients and they will need other hospital staff available to assist with these activities. Hospitals should have staff trained in areas other than their traditional roles, such as in the use 15Jones, R. (2008). Personal communication (Centers for Disease Control and Prevention, Atlanta, Georgia). 16Salame-Alfie, A. (2008). Personal communication (New York State Health Department, Troy, New York).

184 / 13. ASSESSMENT OF CURRENT CAPACITY of survey meters, portal monitors, contamination control, sample labeling, conducting screening surveys, decontamination, etc. Based on the activities that are expected to take place at the hospital in response to a radiological mass casualty incident, it is recommended that hospital staff be trained in the following areas: • radiation safety and the effects of radiation on patients; • effects of radiation and radioactive contamination on a developing fetus; • how to determine whether a victim is contaminated; • collection and measurement of nasal swabs; • how to survey wounds for evidence of radioactive shrapnel or for radionuclide contamination; • wound cleaning and treatment if radionuclide contamination is detected or suspected; • how to determine the severity of radiation injury, the potential for radiation injury to worsen, the impact of nonradiation injuries on radiation injury, and other risks that may be involved; • current strategies for surveying large numbers of people (including walk-through whole-body monitors and rapid personnel surveys using hand-held instruments); • decontamination of large numbers of people (mass decontamination); • performing radiological surveys (including identifying alpha, beta or gamma radiation): - interpretation of the survey meter readings; - how to perform a contamination survey; and - contamination-control practices: assessing the need for contamination-control measures, and contamination-control of the victim. • contamination-control among medical and emergency response personnel; • use of personal protective equipment; • examples of contamination-control measures: - ambulance and treatment area contamination control; - contamination-control actions in the emergency department; - working with contaminated patients; - instructions for leaving a controlled area: patient and responder; and - decontamination (self, patient and equipment). • record-keeping; • patient records;

13.4 RADIATION VOLUNTEERS

/ 185

• workers (emergency responders); and • proper labeling of laboratory samples (includes using specific containers/documentation with the samples). 13.4 Radiation Volunteers to Support Population Screening The Radiation Studies Branch of CDC supports and has been working with several state radiation-control programs to establish a Radiation Volunteers Corps, similar to the Medical Volunteers Corps. The Florida Bureau of Radiation Control has established a radiation volunteer program and the information presented here is based on their program.17 The initial steps in the development of this program included the determination of: • • • •

intended duties of the corps; professions that may qualify with limited training; whether a mechanism for volunteering already exists; infrastructure needed for staging a population screening or community reception center; • providing initial training; and • supporting annual training, infrastructure, and logistics. The radiation response volunteers thus identified will not be considered emergency responders but will help fill the gap identified in the National Response Framework, Radiation Response Annex section dealing with population screening. These individuals are expected to be able to help in response to the need to monitor large numbers of people at the community reception centers, population screening centers, entrance to Red Cross shelters, or other locations identified in the emergency response plan for this activity. These individuals are also expected to “staff-up” between 12 h after the incident and until such time as federal assets can be mobilized. The volunteers are individuals trained in identification of contamination and implementation of decontamination procedures as part of their current job duties and may have some of the following additional qualifications: • have knowledge and are experienced in answering questions regarding health risk from radiation; • are able to collect or use epidemiological information; 17Gilley,

D.B. (2009). Personal communication (State of Florida Department of Health, Bureau of Radiation Control, Tallahassee).

186 / 13. ASSESSMENT OF CURRENT CAPACITY • have training and experience in disaster mental health; and • can provide “reach-back” supervision directly with the radiation-control program through established communication channels. The Florida program determined that there was already an existing mechanism for volunteering through the Medical Reserve Corps, which is a specialized component of the Citizen Corps. This group has medical and public-health professionals ready to serve their communities in times of need and more information on this resource can be found on the Medical Reserve Corps website (MRC, 2011). Other states may want to consider establishing a similar mechanism for a volunteer corps. 13.5 Biodosimetry The proceedings of the BioDose 2008 International Symposium (Simon et al., 2010) are an excellent resource on the current status and future direction of biodosimetry for high-level radiation exposures. In these proceedings, Swartz et al. (2010) provide a critical assessment of biodosimetry methods for large-scale incidents. They focus on the use of biodosimetry to: • identify individuals who did not receive a significant radiation exposure; • classify the exposed persons into different treatment categories as needed; and • guide both short- and long-term medical treatments. Knowledge of the strengths and weaknesses of various biodosimetry techniques related to these three stages will be critical to their effective use in managing a mass-casualty event. One new approach for early-response assessment of radiation exposure that shows promise is measurement of plasma protein serum amyloid A as a complementary approach to conventional biodosimetry for early assessment of radiation exposures and, when coupled with peripheral blood cell counts, provides early diagnostic information for the effective management of radiation casualty incidents (Ossetrova et al., 2010). However, current national resources for biodosimetry are limited and need to be enhanced to provide timely and adequate dose assessments (Blakely et al., 2005). Physicians who are considering these biodosimetry methods should contact REAC/TS (Appendix K) for assistance. High throughput biodosimetry will be needed if large numbers of people are exposed to an IND and possibly an RDD. This is well illustrated by the 1987 radiation incident in Goiânia, Brazil

13.6 CONCLUSIONS

/ 187

(Appendix I) where ~112,000 people reported for screening, of whom 46 required treatment with Prussian blue (IAEA, 1988). Identifying those patients who do not need medical intervention will be equally crucial to reduce demand on limited resources and to reassure those patients. New, fully-automated approaches to biodosimetry have the potential for much higher throughputs (Garty et al., 2010). The World Health Organization has formulated the general scope and concept of a global biodosimetry laboratory network for radiation emergencies (BioDoseNet) and is developing the technical details of the network (Blakely et al., 2009). 13.6 Conclusions The capability to use nuclear medicine equipment available at hospitals and clinics exists in every state. Many operators, however, are not trained in how to change the settings to allow them to use the gamma cameras to measure internal contamination resulting from a radiological incident. It is unknown at this time which hospitals have developed protocols to allow them to use the equipment during a mass casualty radiological incident. Therefore, if a hospital is planning to use this equipment during a mass casualty incident, it is in their best interest to request modifications to the gamma camera’s computer from their vendors, and to request that they develop specific calibration factors to minimize the need for the operators to manually perform those changes. Hospitals also need to develop hospital-specific protocols for handling contaminated patients to minimize contamination of equipment and other areas of the hospital. Staff using the equipment should be trained in their use during an emergency, and this should be made an integral part of the plans developed by the hospitals to quickly assess contaminated patients. Additionally, using the equipment in this fashion should be practiced during hospital emergency exercises. Finally, additional research should be conducted to determine whether concerns of emergency personnel to report to a radiological or nuclear incident will impact the capacity of agencies to respond adequately. Hospitals should be aware of other statewide resources available that may be used to supplement their resources. This could be accomplished by contacting their state radiation-control program. A list of contacts for these radiation-control programs is available from CRCPD (2010).

14. Conclusions and Recommendations This Report has discussed screening a population for internal contamination following exposure to radionuclides that are considered most likely to be involved in a radiological or nuclear incident. The Report contains many recommendations that focus on procedures for screening a population to determine whether individuals have radionuclide intakes that exceed the CDG. The CDG was developed in NCRP Report No. 161 (NCRP, 2008a) as an operational quantity that physicians could use as a basis for medical treatment of individuals having one or more internally-deposited radionuclides. Many communities and hospitals are likely to be capable of screening a few tens of exposed individuals, but little information exists to determine the capability of screening large numbers of people who may have been exposed to the radionuclide(s) involved in a radiological or nuclear incident. 14.1 Recommendations for Planning The following recommendations are highlighted in the Report and are intended to help local communities plan and prepare to screen a population for internal contamination following a radiological or nuclear incident. The information provided in this Report should be used to evaluate and to upgrade, to the extent feasible, the capability of a local community to screen small, medium and large populations for the presence of internally-deposited radionuclides and to treat members of the population medically to reduce radiation dose to patients who contain high levels of one or more radionuclides based on comparisons of these internal depositions with CDG values in this Report or other guidelines. • In the development of a response plan for a radiological or nuclear incident, community leadership should identify personnel such as health physicists, radiation-safety officers, or medical personnel who could serve as subject matter experts or perform specific tasks to support response to a radiological or nuclear incident. • Screening a population for internal contamination requires planning and practice. In particular, the use of nuclear 188

14.2 RECOMMENDATIONS RELATED TO SCREENING

/ 189

medicine cameras to screen patients following exposure to radionuclide contamination from an RDD or IND incident requires advance planning, training and rehearsal. It cannot be implemented ad hoc during an emergency. • Local municipalities and public-health agencies must develop procedures for requesting, receiving and distributing SNS assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. Lists of possible decorporation agents for radionuclides beyond those covered in this Report are given in Sections 3 and 12 of NCRP Report No. 161 (NCRP, 2008a). 14.2 Recommendations Related to Screening and Treatment of a Population for Internal Contamination The following points are emphasized to help local communities screen (and treat, if necessary) members of a population who may have been internally contaminated during a radiological or nuclear incident such as an RDD or IND: • Patients who have suffered life-threatening injuries should be given medical care immediately, without regard to contamination. • People with the most serious injuries are also likely to be the most contaminated both externally and internally. • The presence of internal contamination is rarely life-threatening and decorporation should not take precedence over treatment of conventional injuries that may be acutely life threatening. • The CDG may be used by physicians as a basis for considering medical treatment of individuals who have an internal radionuclide deposition. • The CDG is not intended to instruct physicians on a specific course of action such as administration of decorporation agents. Rather, the CDGs are intended as a tool to be used to help a physician determine when radiation exposure may have clinical significance. • REAC/TS should be used as a resource for information and assistance when considering the treatment of internallycontaminated individuals (REAC/TS, 2010).

190 / 14. CONCLUSIONS AND RECOMMENDATIONS • Rapid identification of the radionuclide(s) involved in a radiological or nuclear incident is essential to the selection of appropriate methods for assessment of internal contamination and subsequent treatment decisions. • The rapid detection and identification of pure alpha- or betaemitting radionuclides (e.g., 210Po, 90Sr/90Y, respectively) are challenging and should be addressed in the planning process. • In all but the most extreme cases, standard precautions provide adequate protection to healthcare workers to prevent secondary contamination. The presence of external contamination should almost never delay urgent medical care. • The use of nuclear medicine cameras to screen patients following exposure to radionuclide contamination from an RDD or IND incident requires advance planning, training and rehearsal; it cannot be implemented ad hoc in an emergency. • Local municipalities and public-health agencies must develop procedures for requesting, receiving and distributing SNS assets. This planning should include providing the diagnostic and medical management guidelines to healthcare providers for the use of decorporation agents for internally-deposited radionuclides. • Planning for population screening should incorporate provisions for establishing stakeholder advisory boards or other mechanisms for making the public a partner. • NCRP recommends a bioassay action-level or benchmark equal to one-half of the CDG value for identifying those individuals for whom continued bioassays should be considered for the purpose of correlating internal contamination with long-term biological effects. • Scalability should be considered in each hospital or community radiation-response plan. 14.3 Recommendations for Additional Work The following recommendations should be implemented by appropriate federal agencies to improve capability of the nation to screen populations for internal contamination following exposure to a radiological or nuclear incident: • Additional surveys of local communities and states should be conducted to determine more accurately their capability to screen populations for internal contamination. • Consideration should be given to utilizing strategicallylocated radiochemistry laboratories to analyze bioassay samples for radionuclide identification and concentrations. A

14.3 RECOMMENDATIONS FOR ADDITIONAL WORK

/ 191

limiting factor here is the low number of available radiochemists in the United States. • Additional research should be conducted to define radiological instrument or bioassay trigger levels that correspond to the CDG for a wider variety of radionuclides such as those identified in NCRP Report No. 161 (NCRP, 2008a) that may be produced by the explosion of an RDD or IND.

Appendix A Radiological Properties of Radionuclides Considered in this Report Tables A.1a and A.1b are a compilation of the radionuclides considered in this Report and their relevant radiological data in SI and previous units, respectively, as given in NCRP Report No. 161 (NCRP, 2008a) and drawn in large part from ICRP Publication 107 (ICRP, 2008). These data include half-life, decay mode, energy, and specific activity of the primary radiation. Also given are air kerma rate constants and electron constants. The air kerma rate constants are radiation doses in air at a distance of 1 m from an unshielded point source of unit activity [1 Bq (1 μCi)]. The electron constants are radiation doses at a depth of 70 μm from a 1 cm2 source of unit activity [1 Bq (1 μCi)] on the skin surface. For those radionuclides that typically exist as equilibrium mixtures with their progeny (e.g., 137Cs/137mBa and 90Sr/90Y), the air kerma rate and electron constants are provided for both the parent and progeny radionuclides; progeny radionuclides are indicated with an asterisk. The total radiation dose from an equilibrium mixture of a parent radionuclide and its progeny is the sum of the doses from both the parent and progeny. Tables A.1a and A.1b provide dose rate information from radionuclide sources external to the body (at 1 m and on the skin surface).

192

Co

8.02 d

11.8 d 30.2 y 2.55 m 73.8 d

I

*131mXe

Cs

131

137

*137mBa

192

226Ra

See 238U chain

64.1 h

*90Y

Ir

28.8 y

5.27 y

Half-life

Sr

90

60

Radionuclidea

4.6 × 106 3.1 × 106 3.2 × 103 2.0 × 1010

β 2.28 γ 0.364, β 0.807 γ 0.0298 β 1.176 γ 0.662 γ 0.316, β 1.38

β– β–

IT β– IT β– EC

3.4 × 105

2.0 × 107

5.1 × 103

β 0.546

β–

4.1 × 104

Specific Activity (GBq g–1)

γ 1.17, 1.33 β 2.51

Energy of Prominent Radiations (MeV)

β–

Decay Modeb

3.2 × 10–17

2.3 × 10–17

6.1 × 10–23

3.3 × 10–18

1.5 × 10–17

1.7 × 10–21

—

8.5 × 10–17

Air-Kerma Rate Constant [Gy s–1 (Bq m–2)–1]

5.5 × 10–10

6.9 × 10–11

4.7 × 10–10

6.5 × 10–10

4.8 × 10–10

6.7 × 10–10

5.0 × 10–10

3.1 × 10–10

Electron Constante [rad h–1 (µCi m–2)–1]

TABLE A.1a—Radiological properties of radionuclides in this Report given in SI units (NCRP, 2008a).

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

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γ 0.0633, β 0.195 γ 1.00, β 2.27 γ 0.0136, β 2.072

β– β– IT β– α

24.10 d

1.17 m

6.70 h 2.5 × 105 y

*234Th

*234mPa

*234Pa

*234U

γ 0.0163, α 4.78

γ 0.016, α 4.20

α SF

U

4.5 × 109 y

β 0.389

238

γ 0.186, α 4.397

α β–

7.1 × 108 y

Energy of Prominent Radiations (MeV)

25.5 h

U

Half-life

Decay Modeb

*231Th

235

Radionuclide

a

2.3 × 10–1

7.4 × 107

2.5 × 1010

8.6 × 105

1.2 × 10–5

2.0 × 107

8.0 × 10–5

Specific Activity (GBq g–1)

TABLE A.1a—(continued)

2.8 × 10–18

7.4 × 10–17

7.4 × 10–19

2.9 × 10–18

2.0 × 10–18

1.9 × 10–17

1.3 × 10–17

Air-Kerma Rate Constant [Gy s–1 (Bq m–2)–1]

9.3 × 10–13

9.2 × 10–10

6.6 × 10–10

1.1 × 10–10

2.7 × 10–13

2.5 × 10–10

4.7 × 10–11

Electron Constante [rad h–1 (µCi m–2)–1]

194 / APPENDIX A

γ 0.609, β 3.27 α 7.69 γ 0.800, β 4.394 γ 0.108, β 0.0631

β– α α β– β– α

19.9 m 1.6 × 10–4 s 1.30 m

22.2 y

*214Po

*210Tl

*210Pb

*214Bi

β 2.88, α 6.69

α β–

1.5 s

*218At

γ 0.352, β 1.02

β–

26.8 m

*214Pb

β 0.259, α 6.00

α β–

3.10 m

*218Po

α 5.49

α

3.82 d

*222Rn

γ 0.186, α 4.78

α

1,600 y

*226Ra

γ 0.0153, α 4.69

α

7.5 × 104 y

*230Th

2.8 × 103

2.5 × 1010

1.2 × 1016

1.6 × 109

1.3 × 1012

1.2 × 109

1.0 × 1010

5.7 × 106

3.7 × 101

7.6 × 10–1

9.7 × 10–18

1.0 × 10–16

3.1 × 10–21

5.0 × 10–17

—

1.5 × 10–17

—

1.5 × 10–20

5.2 × 10–19

2.5 × 10–18

—

8.4 × 10–10

7.8 × 10–16

6.4 × 10–10

6.8 × 10–13

6.9 × 10–10

4.2 × 10–14

1.6 × 10–14

1.4 × 10–11

1.8 × 10–12

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 195

432 y

241

Am

26 m

*235mU IC 0.000076 γ 0.0600, α 5.49

IT α

γ 0.0136, α 5.176

α

2.4 × 104 y

Pu

239

γ 0.0173, α 5.46, α 5.50

α SF

87.7 y

Pu

238

β 1.53

β–

4.2 m

*206Tl

γ 0.305, β 1.31

8.15 m

*206Hg

β–

138 d

*210Po

α 5.30

5.013 d α

Energy of Prominent Radiations (MeV)

β 1.162

Decay Modeb

β– α

Half-life

*210Bi

Radionuclide

a

9.8 × 10–18

—

1.1 × 109 1.3 × 102

1.1 × 10–18

2.6 × 10–18

1.7 × 10–20

6.2 × 10–18

3.6 × 10–22

1.5 × 10–23

Air-Kerma Rate Constant [Gy s–1 (Bq m–2)–1]

2.3 × 100

6.3 × 102

8.0 × 109

4.1 × 109

1.7 × 105

4.6 × 106

Specific Activity (GBq g–1)

TABLE A.1a—(continued)

7.6 × 10–13

—

1.3 × 10–13

2.5 × 10–13

6.2 × 10–10

6.7 × 10–10

8.5 × 10–17

5.9 × 10–10

Electron Constante [rad h–1 (µCi m–2)–1]

196 / APPENDIX A

b

Names preceded by an asterisk are radioactive progeny that may be present in significant quantities. EC = electron capture IT = isometric transition SF = spontaneous fission c Beta energy is the endpoint energy of the spectrum, IC denotes internal conversion electrons, neutrons accompany spontaneous fission. d Includes the contribution from annihilation photons and, in the case of spontaneous fission, the prompt and delayed photons e In the case of spontaneous fission, includes the contribution of prompt beta decay.

a

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 197

Co

64.1 h 8.02 d

11.8 d 30.2 y 2.55 m 73.8 d

*90Y

I

*131mXe

Cs

Ra

137

*137mBa

Ir

131

192

226

See 238U chain

28.8 y

5.27 y

Half-life

Sr

90

60

Radionuclide

a

1.2 × 105 8.4 × 104 8.6 × 101 5.4 × 108

β 2.28 γ 0.364, β 0.807 γ 0.0298 β 1.18 γ 0.662 γ 0.316, β 1.38

β– β–

IT β– IT β– EC

9.2 × 103

5.4 × 105

1.4 × 102

β 0.546

β–

1.1 × 103

Specific Activity (Ci g–1)

γ 1.17, 1.33 β 2.51

Energy Primary Radiationsc (MeV)

β–

Decay Mode

b

4.2 × 10–7

3.0 × 10–7

8.1 × 10–13

4.4 × 10–8

2.0 × 10–7

2.2 × 10–11

—

1.1 × 10–6

Air-Kerma Rate Constantd [Gy s–1 (Bq cm–2)–1]

7.3 × 100

9.2 × 10–1

6.3 × 100

8.7 × 100

6.4 × 100

8.9 × 100

6.6 × 100

4.1 × 100

Electron Constante [rad h–1(µCi cm–2)–1]

TABLE A.1b—Radiological properties of radionuclides in this Report given in previous units (NCRP, 2008a).

198 / APPENDIX A

γ 0.163, α 4.21 γ 0.0633, β 0.195 γ 1.00, β 2.27 γ 0.0136, β 2.07

α SF β– β– IT

γ 0.0153, α 4.687 γ 0.186, α 4.78 α 5.49

α α α α

1.17 m

6.70 h 2.5 × 105 7.5 × 104 y

1,600 y

3.82 d

*234mPa

*234Pa

*234U

*230Th

*226Ra

*222Rn

γ 0.0163, α 4.78

β–

24.10 d

*234Th

U

4.5 × 109 y

β 0.389

238

γ 0.186, α 4.397

α β–

7.0 × 108 y

25.5 h

U

*231Th

235

1.5 × 105

9.9 × 10–1

2.1 × 10–2

6.2 × 10–3

2.0 × 106

6.9 × 108

2.3 × 104

3.4 × 10–7

5.3 × 105

2.2 × 10–6

2.0 × 10–10

7.0 × 10–9

3.3 × 10–8

3.7 × 10–8

9.9 × 10–7

9.8 × 10–9

3.8 × 10–8

2.7 × 10–8

2.5 × 10–7

1.8 × 10–7

2.1 × 10–4

1.9 × 10–1

2.5 × 10–2

1.2 × 10–2

1.2 × 101

8.7 × 100

1.4 × 100

3.6 × 10–3

3.3 × 100

6.2 × 10–1

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 199

γ 0.609, β 3.272 α 7.69 γ 0.800, β 4.39 γ 0.108, β 0.0631

β– α α β– β– α

19.9 m 1.6 × 10–4 s 1.30 m

22.2 y

*214Po

*210Tl

*210Pb

*214Bi

β 2.88, α 6.69

1.5 s

*218At

α β–

26.8 m

*214Pb

γ 0.352, β 1.023

3.10 m

*218Po β–

Decay Mode

Energy Primary Radiationsc (MeV)

β 0.259, α 6.00

Half-life

b

α β–

Radionuclide

a

7.6 × 101

6.9 × 108

3.2 × 1014

4.4 × 107

3.5 × 1010

3.3 × 107

2.8 × 108

Specific Activity (Ci g–1)

TABLE A.1b—(continued)

1.3 × 10–7

1.4 × 10–6

4.1 × 10–11

6.7 × 10–7

—

2.0 × 10–7

—

Air-Kerma Rate Constantd [Gy s–1 (Bq cm–2)–1]

—

1.1 × 101

1.0 × 10–5

8.5 × 100

9.0 × 10–3

9.2 × 100

5.6 × 10–4

Electron Constante [rad h–1(µCi cm–2)–1]

200 / APPENDIX A

432 y

*241Am

IC 0.000076 γ 0.0600, α 5.49

IT α

1.3 × 10–7

—

3.1 × 107 3.4 × 100

1.5 × 10–8

3.5 × 10–8

2.3 × 10–10

8.2 × 10–8

4.8 × 10–12

2.0 × 10–13

6.2 × 10–2

1.7 × 101

2.2 × 108

1.1 × 108

4.5 × 103

1.2 × 105

1.0 × 10–2

—

1.7 × 10–3

3.3 × 10–3

8.3 × 100

9.0 × 100

1.1 × 10–6

7.8 × 100

b

Names preceded by an asterisk are radioactive progeny that may be present in significant quantities EC = electron capture IT = isometric transition SF = spontaneous fission c Beta energy is the endpoint energy of the spectrum, IC denotes internal conversion electrons, neutrons accompany spontaneous fission. d Includes the contribution from annihilation photons and, in the case of spontaneous fission, the prompt and delayed photons e In the case of spontaneous fission, includes the contribution of prompt beta decay.

a

26 m

*235mU

γ 0.0136, α 5.16

α

2.4 × 104 y

Pu

239

γ 0.0173, α 5.46, α 5.499

α SF

87.7 y

Pu

238

β 1.53

β–

4.2 m

*206Tl

γ 0.305, β 1.31

β–

8.15 m

*206Hg

α 5.30

α

138 d

*210Po

β 1.16

β– α

5.01 d

*210Bi

A. RADIOLOGICAL PROPERTIES OF RADIONUCLIDES

/ 201

Appendix B How to Perform a Radiation Survey for Contamination: Instructions for 18 Workers18 In performing a contamination survey with a hand-held instrument, first ensure the instrument is functioning properly. It is advisable to wrap the meter probe with plastic wrap to protect the probe from contamination (except if you are surveying for alpha contamination). Make sure that the instrument has batteries and that they work. To do this, turn your instrument to battery check. If the batteries are acceptable, turn the dial to a measurement mode and use a check source to verify the instrument is operating properly. B.1 Screening Survey If a large population must be surveyed, it is acceptable to perform only a screening survey of the head, face, shoulders and feet rather than a more detailed survey, since these are the most likely locations to become contaminated. You may also consider using portal monitors. If only performing a screening survey for beta-gamma activity, it is acceptable to hold the survey-meter probe ~3 to 5 cm (~1 to 18Adapted

from the Handbook for Responding to a Radiological Dispersal Device. First Responders Guide—The First 12 Hours (CRCPD, 2006).

202

B.2 COMPLETE WHOLE-BODY SURVEY

/ 203

2 inches) away from the body [instead of ~1 cm (0.5 inch) for alpha activity], and move it twice as fast as the normal ~3 to 5 cm s–1 (1 to 2 inches s–1). (If the probe is moved too quickly, its detection capability may be reduced.) Check with local and state radiation-control personnel to determine the extent of contamination survey required. Return the probe to its holder on the meter when finished. Do not set the probe down on the ground. The probe should be placed in the holder with the sensitive side of the probe facing to the side or facing down so that the next person to use the meter can monitor his/her hands without handling the probe or allowing contamination to fall onto the probe surface. B.2 Complete Whole-Body Survey If feasible, perform a complete, whole-body contamination survey and record the findings on the contamination survey sheet. To begin a body survey, the individual should stand with their legs spread and arms extended. First holding the probe about a ~1 cm (0.5 inch) away from the surface to be surveyed, slowly [~3 to 5 cm s–1 (1 to 2 inches s–1)] move the probe over the head, and proceed to survey the shoulders, arms, and bottoms of the feet. Care must be taken not to permit the detector probe to touch any potentially-contaminated surfaces. It is not necessary to perform the personnel contamination survey in exactly the order listed below, but a consistent procedure should be followed to help prevent accidentally skipping an area of the body. Pause the probe for ~5 s at locations most likely to be contaminated: • top and sides of head, face (pause at mouth and nose for ~5 s; high readings may indicate internal contamination); • front of the neck and shoulders; • down one arm (pausing at elbow), turn arm over; • backside of hands, turn over (pause at palms for ~5 s); • up the other arm (pausing at elbow), turn arm over; • shoe tops and inside ankle area; and • shoe bottoms (pause at sole and heel). As with the screening survey, return the probe to its holder on the meter when finished. Do not set the probe down on the ground. The probe should be placed in the holder with the sensitive side of the probe facing to the side or facing down so that the next person to use the meter can monitor his/her hands without handling the probe or allowing contamination to fall onto the probe surface.

204 / APPENDIX B B.3 Most Common Mistakes Made During the Survey • Holding the probe too far away from the surface (should be ~3 to 5 cm (~1 to 2 inches) away for a screening survey or about 0.5 inch or less for an alpha survey). • Moving the probe too fast [should be ~5 to 10 cm s–1 (~2 to 4 inches s–1) for a beta-gamma survey or ~3 to 5 cm s–1 (~1 to 2 inches s–1) for an alpha survey]. • Contaminating the probe; probe background should be observed and compared to initial background. If within a factor of two, it is acceptable to continue to use probe. Otherwise, check with radiation-control personnel. Wrapping the probe in plastic wrap will help prevent surface contamination but this wrapping should not be used when surveying for alpha contamination.

Appendix C How to Distinguish Between Alpha, Beta and Gamma Radiation Using a Geiger-Muller 19 Survey Meter19 This appendix describes a technique using a pancake GeigerMuller (GM) survey meter that may be employed by emergency responders to make a quick, initial determination of the type of radiation (alpha, beta or gamma) present at the scene. Many studies show that the most likely radionuclide(s) to be used in a radiological or nuclear incident would be either a gamma-ray emitter or a beta-gamma emitter. However, it is possible that the radionuclide may be a pure beta emitter such as 90Sr or an alpha emitter such as 239 Pu. This methodology was developed to assist emergency responders in making an initial determination of the type of radiation present. This determination should be used to make decisions until hazardous materials or radiation-control staff arrives at the site with more sophisticated instrumentation to verify the type of radiation and identify the radionuclide(s). Pancake GM survey meters will respond to beta, gamma, and x radiation. They have very limited response to alpha radiation. NaI(Tl) survey instruments will respond primarily to gamma radiation or x rays. Do not be misled into thinking that radionuclides are not present by the lack of response or low reading from a NaI(Tl) survey meter, since it cannot detect alpha and may respond poorly to beta radiation. 19Adapted

from the Handbook for Responding to a Radiological Dispersal Device. First Responders Guide—The First 12 Hours (CRCPD, 2006).

205

206 / APPENDIX C C.1 Determining the Presence of an Alpha-Emitting Radionuclide Using Only a Geiger-Muller Survey Meter Although alpha emitters may not appear to be as hazardous as gamma-ray emitters, they are very harmful when inhaled or ingested. Therefore, it is important to check for the presence of alpha-emitting radionuclides. Because the instruments normally available to emergency responders will not readily respond to alpha particles, it is important to use appropriate respiratory protection when monitoring for radionuclides. Procedure: • Take readings at approximately 8 cm (3 inches) and 1 cm (0.5 inch) (as close as possible without touching) above the potentially-contaminated surface with the window facing down. • If the instrument reading increases by more than a factor of three at the 1 cm (0.5 inch) measurement [as compared to the 8 cm (3 inch) measurement], suspect alpha contamination (such as 239Pu). • Next, place a sheet of paper on the surface and take a reading with the window side down directly on top of the paper. The alpha radiation will not penetrate the paper, and the window-down reading should significantly decrease to near background level. If the window-down measurement taken over the paper does not significantly decrease, the radionuclide is likely not an alpha emitter. (Note that some alpha emitters, such as 241Am, also emit a low-energy gamma which will not be stopped by a sheet of paper.) C.2 Determining the Presence of Strontium-90 (or other pure beta emitters) Using a Pancake Geiger-Muller Survey Meter Strontium-90 is a pure beta emitter, and most NaI(Tl) instruments respond poorly to energetic beta emitters. However, 90Sr beta radiation can be easily detected and measured with a survey meter connected to an end-window, side-window, or pancake GM probe (preferred). For the purposes of this Report, suspect the presence of 90 Sr if a pancake GM survey meter reads high, such as between 1,000 and 10,000 cpm (20 to 200 times background) but there is no corresponding increase in readings using a NaI(Tl) survey meter (still reads near background). When 90Sr is shielded by certain materials, the beta radiation cannot be detected. However, the interaction of the beta radiation

C.2 DETERMINING THE PRESENCE OF STRONTIUM-90

/ 207

with the shielding materials can produce x rays, which can be detected by GM, NaI(Tl), and other types of gamma-ray identification survey meters. Procedure: • Take a measurement with the window side of the pancake probe (mesh covered side) facing down at ~15 cm (~6 inches) above the area being surveyed where the survey meter reads between 500 to 1,500 cpm. Then take another measurement with the window side facing up (away from the ground) at the same height. • Compare the two measurements: - If only 90Sr (or another pure beta-emitter or a very weak gamma emitter such as 241Am) is present, the window-up reading will be near background (depending on the model of the GM pancake probe, background should be in the range of 25 to 75 cpm), and the window-down reading should be 10 or more times greater than the window-up reading. This is because the beta emissions are not able to penetrate the back side of the GM pancake probe. - If an energetic gamma emitter is present (e.g., 137Cs, 192Ir, 60Co), the window-down reading at 15 cm (6 inches) will be approximately twice the window-up reading. • Take another measurement with the window side of the pancake probe facing down at ~1 m (~3 feet) from the area being surveyed where the meter reads between 500 to 1,500 cpm. Then take another measurement with the window-up at the same height. Compare the two measurements. If a gammaemitting radionuclide is present, both readings will be approximately the same.

208

20

Middle initial:

Drivers license #:

Phone:

Last name:

Reproduced with minor changes from the Handbook for Responding to a Radiological Dispersal Device (CRCPD, 2006).

Parent or guardian (if child):

Location at time of incident:

Date/time:

Address:

Date of birth:

First name:

D.1 Contamination Survey Sheet20

Survey and Registry Forms

Appendix D

Survey instrument and detector type:

1,000 cpm

Back

>10,000 cpm

Mark contamination locations and survey reading on the diagrams below (circle if readings are in cpm mR h–1 μR h–1)

D.1 CONTAMINATION SURVEY SHEET

/ 209

Person sent to medical facility:

Yes

Yes

Nasal area reading of 100,000 cpm (0.5 mR h–1): No

Valuables returned:

Yes

Clothing and valuable bag number:

Person sent to decontamination area:

Monitored by:

Comments:

Yes

No

No

(if yes, refer to medical facility)

No

210 / APPENDIX D

Last

Yes

Possibly pregnant:

No

No

Sex:

Other (specify)

If yes, estimate term:

Place of birth:

First

Emergency services Yes

Public

Witness to the incident:

Member of:

Email address:

Telephone:

Current permanent full address

Social security #:

Nationality:

Date of birth:

Full name:

Date:

D.2 Registry Form

M

M.I. F

D.2 REGISTRY FORM

/ 211

1 μSv h–1

Personal survey instrument

Instrument type:

No

212 / APPENDIX D

Remarks:

Organization:

Telephone:

Date:

Full name: Time:

D.2 REGISTRY FORM

/ 213

214 / APPENDIX D D.3 ATSDR Rapid Response Registry Survey Form

D.3 ATSDR RAPID RESPONSE REGISTRY SURVEY FORM

/ 215

Appendix E How to Perform Decontamination at 21 Home21 You may have been exposed to low levels of radioactive particles. The particles may have settled on your hair, skin and clothing as dust. You are not in immediate danger from these small radioactive particles. However, you should go home or to a facility designated by the emergency operations center to decontaminate. Removal of outer clothing should reduce your contamination by up to 90 %. In order to help protect your health and safety as well as others, please follow these directions. Because radiation cannot be seen, smelled, felt or tasted, people at the site of an incident will not immediately know if you have been exposed to radioactive materials. You can take the following steps to limit your contamination: • Get out of the immediate area quickly. Go directly home, inside the nearest safe building, or to an area to which you are directed by law-enforcement or health officials. Do not go to a hospital unless you have a medical condition that requires treatment. • If radioactive material is on your clothes, removing them will reduce the external contamination and decrease the risk of internal contamination. Prompt removal of outer clothing will also reduce the length of time that you are exposed to radiation. When removing the clothing, be careful of any clothing that has to be pulled over the head. Try to 21Adapted

from the Handbook for Responding to a Radiological Dispersal Device. First Responders Guide—The First 12 Hours (CRCPD, 2006).

216

E. HOW TO PERFORM DECONTAMINATION AT HOME

•

•

•

•

/ 217

either cut the article off or prevent the outer layer from coming in contact with the nose and mouth area. You may also hold your breath while carefully pulling the article over the head. Removal of clothes should be done in a garage or outside area if available, where the ground can be washed with a hose. If an outside area is not available, the removal of clothing should take place in a room where the floor can be easily cleaned, such as the tub or shower areas. (Disposable sweeping cloths are good for decontaminating smooth floor surfaces.) Clothing should be rolled up with the contaminated side “in” to minimize cross contamination. If possible, place the clothing in a plastic bag (double-bagging is best to reduce the chances of a rupture), and leave it in an out-of-the-way area, such as the corner of a room or garage. Keep people away from it to reduce their exposure to radiation. You may be asked to bring this bag for follow-up readings or for disposal at a later time. Keep cuts and abrasions covered when handling contaminated items to avoid getting radioactive material in the wound. Shower and wash all of the exposed parts of your body and hair using lots of soap and lukewarm water to remove contamination. This process is called decontamination. Simple washing will remove most of the radioactive particles. Do not use abrasive cleaners, or scrub too hard. Do not use hair conditioners. If you are going to a monitoring location, it is best to change clothes and shower before being monitored. Contact your local or state radiation-control program for additional guidance.

Appendix F Using Geiger-Muller Survey Meters to Assess Internal Contamination for Selected GammaEmitting Radionuclides Section 8.5 describes a method for rapid screening of patients to determine whether internal contamination exceeds the CDG. Tables F.1, F.2, and F.3 are based on those developed by Hurtado (2006) and Juneja (2011), and the data currently are available from CDC (2009). The values in these tables are based on the response of a typical thinwindow (“pancake”) detector [Ludlum 44-9® (Atlanta Nuclear, Rockland, Massachusetts) pancake GM probe]. For each particular radionuclide, inhalation exposures are based on a 1 μm AMAD particle size and Type-M lung absorption rate in the case of 60Co and 192Ir and Type-F lung clearance rate in the case of 137Cs. The distance for checking the patient was fixed at 6 and 30 cm directly in front of the sternum (AP) or directly in the middle of the back (PA). The numerical values indicated on each table below are the net counts per minute that would indicate an intake at the CDG level, defined as the intake of radionuclide that would result in an effective dose of 250 mSv for an adult or 50 mSv for a child. Results given for a child are based on calculations for a 10 y old child (Section 8.5). If the indicated count rate exceeds the tabulated value, it should be assumed that the CDG has been met or exceeded. Consult CDC (2009) to determine whether tables have been updated or whether tables for additional radionuclides, ages or distances have been added. This screening method assumes that the patient is free of external contamination. If a reading from the chest (AP) indicates that a patient has an intake in excess of the CDG, it would be prudent to measure from the back (PA) position to confirm that the indicated activity is from an intake rather than external contamination. 218

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 219

TABLE F.1a—Adult male count rate (counts per minute) corresponding to 1 CDG for inhaled 60Co (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

12,900

9,200

4,800

3,100

2

11,300

8,500

4,600

2,900

4

10,700

8,100

4,500

2,800

6

10,600

8,000

4,400

2,800

8

10,600

7,800

4,300

2,700

10

10,500

7,700

4,200

2,700

12

10,400

7,600

4,100

2,600

14

10,200

7,500

4,000

2,500

16

10,000

7,400

3,900

2,500

18

9,800

7,300

3,800

2,400

20

9,600

7,200

3,700

2,400

24

9,100

7,000

3,400

2,300

48

6700

6,100

2,200

1,800

72

5,500

5,600

1,700

1,500

220 / APPENDIX F TABLE F.1b—Adult female count rate (counts per minute) corresponding to 1 CDG for inhaled 60Co (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

15,500

9,600

5,400

3,200

2

14,500

9,100

5,100

3,000

4

13,500

8,600

4,900

2,900

6

13,000

8,300

4,800

2,800

8

12,800

8,100

4,700

2,700

10

12,500

7,900

4,500

2,600

12

12,000

7,700

4,400

2,500

14

11,800

7,600

4,200

2,400

16

11,500

7,400

4,100

2,400

18

11,200

7,300

4,000

2,300

20

11,100

7,100

3,800

2,300

24

10,500

6,900

3,500

2,100

48

8,200

5,900

2,300

1,700

72

7,300

5,400

1,800

1,400

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 221

TABLE F.1c—Child count rate (counts per minute) corresponding to 1 CDG for inhaled 60Co (Type M, 1 μm AMAD) using a GM survey meter (calculated from data in CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

2,500

870

1,600

520

2

2,300

820

1,500

490

4

2,200

790

1,400

460

6

2,100

770

1,300

440

8

2,100

750

1,300

430

10

2,000

730

1,300

420

12

2,000

700

1,200

400

14

1,900

680

1,200

390

16

1,800

660

1,200

380

18

1,800

640

1,200

370

20

1,700

610

1,100

360

24

1,600

570

1,100

340

48

1,300

380

950

270

72

1,200

290

880

230

222 / APPENDIX F TABLE F.2a—Adult male count rate (counts per minute) corresponding to 1 CDG for inhaled 137Cs (Type F, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

6,300

3,800

2,300

1,500

2

5,700

3,700

2,200

1,400

4

5,000

3,500

2,000

1,400

6

4,700

3,400

1,900

1,300

8

4,400

3,300

1,800

1,300

10

4,100

3,200

1,700

1,300

12

3,900

3,100

1,600

1,200

14

3,700

3,000

1,600

1,200

16

3,500

3,000

1,500

1,200

18

3,400

2,900

1,500

1,200

20

3,300

2,900

1,400

1,100

24

3,100

2,800

1,300

1,100

48

2,700

2,600

1,100

1,000

72

2,500

2,500

1,000

1,000

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

/ 223

TABLE F.2b—Adult female count rate (counts per minute) corresponding to 1 CDG for inhaled 137Cs (Type F, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

10,000

5,000

3,200

1,700

2

9,600

4,900

3,000

1,700

4

8,300

4,400

2,800

1,600

6

7,200

3,900

2,600

1,500

8

6,300

3,400

2,400

1,400

10

5,500

3,100

2,200

1,300

12

4,900

2,800

2,100

1,300

14

4,300

2,600

2,000

1,200

16

3,900

2,400

1,900

1,200

18

3,600

2,300

1,800

1,200

20

3,300

2,200

1,700

1,100

24

2,800

2,000

1,600

1,100

48

1,800

1,600

1,300

1,000

72

1,600

1,500

1,200

900

224 / APPENDIX F TABLE F.2c—Child count rate (counts per minute) corresponding to 1 CDG for inhaled 137Cs (Type F, 1 μm AMAD) using a GM survey meter (calculated from data in CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

2,300

730

1,200

400

2

2,200

700

1,100

390

4

1,900

640

1,000

360

6

1,700

590

900

340

8

1,500

550

800

330

10

1,300

520

720

310

12

1,100

490

660

300

14

1,000

460

610

290

16

910

440

570

280

18

830

420

530

280

20

770

410

510

270

24

660

380

460

260

48

430

300

380

230

72

370

270

360

220

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

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TABLE F.3a—Adult male count rate (counts per minute) corresponding to 1 CDG for inhaled 192Ir (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

14,400

9,500

6,000

3,300

2

12,500

8,700

5,600

3,000

4

12,000

8,200

5,500

2,900

6

12,000

8,000

5,400

2,800

8

12,000

7,900

5,300

2,800

10

12,000

7,800

5,200

2,800

12

11,600

7,700

5,100

2,700

14

11,500

7,600

5,000

2,700

16

11,300

7,500

4,900

2,600

18

11,100

7,500

4,700

2,600

20

10,800

7,400

4,500

2,500

24

10,300

7,200

4,200

2,400

48

7,600

6,500

2,700

2,000

72

6,100

6,000

2,000

1,700

226 / APPENDIX F TABLE F.3b—Adult female count rate (counts per minute) corresponding to 1 CDG for inhaled 192Ir (Type M, 1 μm AMAD) using a GM survey meter (CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm

30 cm

AP

PA

AP

PA

1

17,800

10,200

6,600

3,400

2

16,300

9,500

6,200

3,200

4

15,000

9,000

5,900

3,000

6

14,500

8,500

5,700

2,800

8

14,000

8,200

5,600

2,700

10

13,500

8,000

5,400

2,600

12

13,200

7,800

5,300

2,600

14

12,800

7,700

5,100

2,500

16

12,500

7,500

5,000

2,400

18

12,200

7,400

4,800

2,400

20

11,900

7,300

4,600

2,300

24

11,300

7,200

4,300

2,200

48

9,200

6,400

2,800

1,800

72

8,200

6,000

2,100

1,600

F. USING GEIGER-MULLER SURVEY INSTRUMENTS

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TABLE F.3c—Child count rate (counts per minute) corresponding to 1 CDG for inhaled 192Ir (Type M, 1 μm AMAD) using a GM survey meter (calculated from data in CDC, 2009). Distance from Sternum or Middle of Spine Time Post Incident (h)

6 cm AP

30 cm PA

AP

PA

1

2,900

1,100

1,700

560

2

2,700

1,000

1,500

530

4

2,500

960

1,400

490

6

2,400

940

1,400

460

8

2,300

910

1,300

440

10

2,200

890

1,300

430

12

2,100

470

1,300

420

14

2,100

840

1,200

410

16

2,000

810

1,200

400

18

2,000

780

1,200

390

20

1,900

760

1,200

380

24

1,800

700

1,100

370

48

1,500

460

1,000

300

72

1,300

350

980

260

Appendix G Collection and Preparation of Biological Samples for Radioanalysis G.1 Urine Samples Urine analysis is particularly useful for those radionuclides which do not emit types of radiation that are sufficiently penetrating to be detected outside the body. Analysis of the amount of the radionuclide in urine can be done by measuring activity (such as for cesium) or by chemical analysis (such as for uranium). By use of a biokinetic model, the radionuclide concentration in urine and time of intake are used to estimate the body burden of the radionuclide and its distribution among different organs that may follow internal deposition via different routes of exposure. Figure 3.1 is a generic biokinetic diagram for internally-deposited radionuclides including the various routes by which these radionuclides can enter the body and eventually be excreted. G.2 Main Collection Issues • Discard first void: Void is the term used to mean a single relief of the bladder such that a person empties all the available urine from their body. Generally, the first void after a potential intake should be discarded. It should not be collected as part of the sample to be measured for two primary reasons. First, typically there has not been enough elapsed time for the radioactive material to pass from where it enters the body, into the blood, into the kidney, and into the bladder. Second, if it does enter a partially filled bladder then the radioactive material will be diluted and the sample may result in the underestimation of the intake. 228

G.2 MAIN COLLECTION ISSUES

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• Minimum time to wait before collecting sample: There exists some finite amount of time between inhalation, ingestion, absorption or injection of a radionuclide and it appearing in urine. The time it takes is variable and depends on many factors including method of entry, biokinetics of the individual, radionuclide involved, chemical form of the radionuclide, and physical form of the radionuclide. Generally, waiting 4 h from the time of probable entry to the start of collecting a 24 h sample is sufficient to obtain the measurement. • Period of collection: Ideally, urine should be collected over a 24 h period. This kind of sample is called a “24 h urine sample.” Collecting a sample for a shorter time or from a single spot sample may result in a less accurate determination of intake activity. NCRP Report No. 87 (NCRP, 1987) states the following: “Although various kinetic models relate urinary excretion to the amount of a radionuclide in the body, the actual data on which most of these models are based exhibit considerable variability among samples from a single individual or from different individuals (Pochin, 1968; Snyder et al., 1972). Thus, a single spot sample taken from a person usually has a high degree of physiologically related uncertainty and does not provide a reliable estimate of body burden.” With that stated, sometimes it is only possible to collect a spot sample due to circumstances. In these cases, the first void should especially be discarded, if possible. When it is only possible to get a spot sample of urine, a correction should be made to scale the sample result up to an approximate 24 h sample for evaluation. More detailed discussion on this scaling procedure is provided in Section 7.2.3 of this Report and Section 10.3.1.2 of NCRP Report No. 161 (NCRP, 2008a). • The volume of urine collected from an individual over a 24 h period will vary somewhat depending on the person and situation. Therefore, the measurement of activity in the sample is normalized to remove the volume dependency. The 24 h reference urine volumes given in ICRP Publication 89 are 1.6 L for men and 1.2 L for women (ICRP, 2002b). Sometimes a 24 h urine sample is incorrectly interpreted to mean a sample collected 24 h after the potential intake. This distinction should be made clear when communicating with people who are unfamiliar with the concept of a 24 h

230 / APPENDIX G urine sample. • Cross-contamination: The concern for cross-contamination is that radioactive material external to the body may end up in the urine sample during collection of the urine. When a low level of contaminate is to be measured, cross-contamination can significantly affect the accuracy of the measurement (Sun et al., 1993). To prevent cross-contamination it is recommended that at the beginning of the 24 h collection period and before collecting the void, the person giving the sample should shower and put on clean clothes. Also, the container used to hold the urine should be kept closed until needed. • Containers: Containers for collecting a 24 h urine sample should hold 2 L or more since reference 24 h urinary excretion is 1.6 L for men and 1 L for women. Commercially obtained containers are readily available for this purpose and have openings that are wide enough to make collection easy for both men and women. In a catastrophic incident an empty plastic 2 L soda bottle could be used. If possible, it should be rinsed and dried. Because the opening is too small to be practically used it may be necessary to urinate into a wider container, such as a measuring cup, and then transfer the urine to the larger 2 L soda bottle. • Documentation: Tracking the urine sample with the proper documentation is critical to maintaining the chain of custody and to the proper determination of the body burden. • Personal information: Include, for example, name, age, sex, address, telephone number, email address, and social security number (or at least the last four digits of the social security number if possible). • Technical information: Include, for example, date/time of probable intake, start date, time and duration of the urine collection, number of voids, if first void was discarded. • Other considerations: Clean containers must be used for collection and storage. • All biological samples: Subject to deterioration by bacteriologic action that may interfere with subsequent analysis. Prompt analysis following collection is preferred. When samples are kept longer then a day, they should be refrigerated, acidified to minimize precipitation, or have preservatives added to inhibit bacterial growth. Appropriate handling techniques should be used to avoid exposure to possible diseases like hepatitis.

Appendix H Shipping of Biological Samples H.1 Introduction It may be necessary to send a biological sample for testing and analysis. Possible types of biological samples could include nasal swabs, sputum, urine, or feces. Hospital or clinical laboratory personnel most likely are aware of shipping regulations and have materials on hand to ship specimens to the analytical laboratory that will be performing the analysis. Additionally, the analytical laboratory should be consulted for shipping instructions. The following details are provided for background information and for those instances when neither the hospital or clinical laboratory nor the analytical laboratory is able to provide instructions. However, readers should be aware that the shipping requirements provided here may have changed subsequent to publication of this Report. The commercial carrier should be contacted any time there are questions about shipping requirements. There is a distinction between shipping clinical biological samples and hazardous biological samples. This appendix is written to assist in the preparation and shipping of clinical samples for routine testing and not those biological samples that may contain infectious diseases. Sometimes clinical biological samples are referred to as diagnostic specimens or exempt biological samples. Sources for detailed information on shipping biological samples include the U.S. Department of Transportation (DOT, 2009), Federal Express® (Memphis, Tennessee) (FedEx, 2010), Glode and Gillum (2010), the International Air Transport Authority (IATA, 2010), and USPS (2009). H.2 Regulatory Information: Brief Summary DOT regulates transport within U.S. borders. Title 49 of the Code of Federal Regulations (49 CFR) has specific requirements 231

232 / APPENDIX H for packaging and labeling hazardous goods. In addition, DOT requires that those who ship or receive hazardous goods receive both awareness training (about the reasons for regulation) and job function (about how to properly prepare a shipment). The United Nations has a committee of experts (International Civil Aviation Organization), which has harmonized international shipping regulations; United Nations’ “certified” packaging has been tested and certified to meet criteria for safe transport of hazardous goods. The International Air Transport Authority (IATA), an aviation trade organization, absorbs the International Civil Aviation Organization technical instructions into IATA regulations. Noncompliance with regulations may result in large fines, and in some instances (e.g., concealing cultures in personal baggage on an airline), personal civil and criminal liability may result. H.3 Definitions These definitions are according to DOT regulations 49 CFR Part 173.134 (DOT, 2009). Similar definitions are in IATA (2010). • Diagnostic/exempt specimen (e.g., urine samples being shipped for routine testing not related to diagnosis of an infectious disease are exempt from shipping requirements for infectious disease specimens): Any human or animal material being shipped not known or suspected of containing a pathogen. This includes secreta, excreta, blood, blood components, tissue, and body fluids. For example, blood samples sent for a cancer screen are considered diagnostic specimens but blood samples sent to confirm HIV seroconversion are classified as infectious substances because the presence of a pathogen is suspected. Diagnostic specimens must be shipped with the triple packaging explained below. • Biological product: Material prepared and manufactured in accordance with certain regulations of the U.S. Department of Agriculture or DHHS (e.g., vaccine preparations). • Infectious substance: In hazardous goods regulations, “infectious substance” includes any toxin (classified as 6.1) or infectious agents (classified as 6.2) which affect humans or animals. This includes infectious agents listed in 42 CFR Part 72.3 of the regulations of DHHS and any other agent that causes or may cause severe, disabling or fatal disease. H.4 General Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples • All shipments must include four basic requirements:

H.4 GENERAL SHIPPING GUIDELINES

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- watertight primary containers - watertight secondary containers - absorbent material - sturdy outer packaging • Examples of suitable watertight primary containers include the following: - plastic canister - glass/plastic vial - sealed plastic bag - glass/plastic jar Note: Use sealed plastic bags for tissue and solid samples only. • Examples of suitable watertight secondary containers include the following: - screw-cap can - sealed Styrofoam® (Dow Chemical, Midland, Michigan) container [minimum of 2.5 cm (1 inch) thick] - sealed plastic bag - plastic container • Examples of suitable absorbent materials include the following: - super absorbent packer - cellulose wadding - cotton balls - paper towels Note: Absorbent materials must be placed between the primary and secondary container. The quantity should be sufficient to absorb all liquid in the shipment. • Examples of sturdy outer packing include the following: - corrugated fiberboard - wood - rigid cooler - rigid plastic container Note: The packaging must be of good quality (typically capable of passing a drop test of 1.2 m) and must be marked on the outside with a label indicating it is a clinical biological sample (such as the words exempt human specimen or diagnostic specimens). Corrugated cardboard is the usual choice. Styrofoam® boxes, plastic bags, or paper envelopes are unacceptable outer containers for shipping biological materials. Exact packing requirements and labeling are specific to the shipper chosen.

234 / APPENDIX H H.5 Specific Shipping Guidelines for Clinical, Diagnostic and Exempt Biological Samples Due to the highly specific nature of shipping regulations, specific details for shipping patient specimens are not provided here. The clinical or hospital laboratory or emergency department where biological specimens are collected should follow their standard protocol when shipping these specimens to the analytical laboratory that will be performing the radioactivity analysis. It is highly recommended that the analytical laboratory and commercial carrier be contacted for any special packaging and shipping instructions. H.6 Website Links for Some Commercial Shippers • DHL® (Germany): http://www.dhl-usa.com; 1-800-225-5345 • Federal Express® (Memphis, Tennessee): http://www.fedex. com; 1-800-Go-FedEx. Additional assistance for noninfectious shipment inquiries is available from FedEx® Packaging Design and Development Department at 1-800-633-7019 • United Parcel Service® (Atlanta, Georgia): http://www.ups. com; 1-800-554-9964 • World Courier® (Stamford, Connecticut): http://www.worldcourier.com; 1-800-221-6600

Appendix I Population Screening and Monitoring Implications of Two Urban Contamination Incidents

I.1 Background Two contamination incidents provide some objective evidence on the level of response that may follow release of radioactive contamination in an urban environment. The two incidents are the 1987 widespread release of 137Cs in urban Goiânia, Brazil, and the November 2006 210Po incident in London. In 1987, accidental dispersal of 137Cs resulted from the actions of unauthorized salvagers breaking open an abandoned teletherapy machine in Goiânia. In the London case, the focus was a very limited intentional administration to a single individual, but it was not discovered for some time after the poisoning and contamination spread to a number of individuals that required the emergency response to be scaled up. In Goiânia, the initial contamination was not recognized or suspected by the involved individuals and ultimately became widespread in an urban area. The follow-up actions to both incidents might be comparable to those likely to follow the use of an RDD or IND. Comparison of the scenarios includes the nature of the radionuclides and exposure scenario, the extent of contamination spread, the resulting population monitoring and bioassays, the number of confirmed intakes and potential doses, and the ameliorative actions undertaken.

235

236 / APPENDIX I I.2 Overview of Two Cases I.2.1

Goiânia, Brazil

The Goiânia incident involved the unrecognized breaching by unauthorized salvagers of a 50.9 TBq (1,375 Ci) 137Cs source from an abandoned teletherapy machine (IAEA, 1988). Cesium-137 emits energetic beta particles and relatively-high-energy gamma rays, making the radionuclide quite easy to detect using portable survey instruments (GM detectors) and gamma scintillator [NaI(Tl)] detector systems. In addition to the hazard posed by the highly penetrating gamma-ray radiation and the energetic beta particles, the soluble cesium chloride salt form was readily dispersible and subject to intake by ingestion and inhalation. Once intake occurred, the biokinetics of 137Cs resulted in a rapid distribution throughout the body causing essentially a total body distribution and irradiation from internally-deposited radioactive material. Within 2 d, one of the salvagers sought medical attention for what was thought initially to be an allergic reaction to bad food, but was in actuality a manifestation of the ARS. It was several additional patients and almost two weeks before radiation exposure was recognized as a likely agent. During this time, contamination was spread over a large urban area. Ultimately, four patients died from their exposure and an additional 249 others showed external contamination with 137Cs. A total of 46 persons underwent medical therapy using Prussian blue for decorporation of 137Cs. Approximately 112,000 members of the public were screened for contamination by personal survey. I.2.2

London, United Kingdom

The radionuclide poisoning case in London provides an example of how the apparent internal contamination of a single person may require subsequent scaling of response to monitor a number of people who may have come in contact with the radionuclide. This case demonstrates response to a radiological incident that is discovered days after the initial exposure. This case also demonstrates the impact on effective screening of a population that is facilitated by good communication, effective and timely press releases, and excellent efforts to monitor and report back to members of the public. Polonium-210 was used as the agent in an assassination on November 1, 2006 in London, United Kingdom (Harrison et al., 2007). As of this writing, the crime is still under investigation and many of the details are not publicly available. The nature of 210Po is unusual in that it is the only alpha-emitting radionuclide that is distributed broadly when taken into the body in a relatively soluble form. This

I.2 OVERVIEW OF TWO CASES

/ 237

distribution pattern included alpha radiation exposure of the bone marrow that resulted in the symptoms of the ARS. It was just a matter of days before the victim died (November 23, 2006) that radiation poisoning with 210Po was diagnosed. The diagnosis was made following recognition that a large-volume urine sample appeared to be emitting radiation. This was detected by a passing hospital radiation technologist using a GM survey meter. The low-yield gamma rays emitted by 210Po were of sufficient quantity to elevate the background reading. Once the radionuclide was suspected, surveys quickly revealed its presence, not only in the patient, but also on objects with which he had contact. Although only one person was the target of the assassination, contamination was found in various locations, including offices, restaurants, coffee bars, night clubs, soccer stadium, airplanes, cars, and three hotels. Literally, thousands of people were potentially exposed, including 460 overseas visitors from 52 countries outside the United Kingdom. Monitoring and bioassay of those inside the United Kingdom consisted of announcing that concerned citizens should contact the National Health Service, which then determined potentials for exposure and need for follow-up based on a triage script from the Health Protection Agency. The Health Protection Agency referred overseas visitors to their respective national health authorities for consideration of follow-up. Where indicated, follow-up consisted of a 24 h urine sample radiochemically analyzed for 210Po. The action levels established were based on committed effective dose levels of 40 y of age if the projected dose to thyroid is ≥5 Gy, to adults 18 to 40 y of age if the projected dose is ≥0.1 Gy, and to pregnant or lactating women or persons