Disposal of radioactive wastes, Volume 15 (Studies in Environmental Science)

DISPOSAL OF RADIOACTIVE WASTES

Studies in Environmental Science Volume 1

Atmospheric Pollution 1978 edited by M. M. Benarie

Volume 2

Air Pollution Reference Measurement Methods and Systems edited by T. Schneider, H. W. de Koning and L. J. Brasser

Volume 3

Biogeochemical Cycling of Mineral-Forming Elements edited by P. A. Trudinger and D. J. Swaine

Volume 4

Potential Industrial Carcinogens and Mutagens by L. Fishbein

Volume 5

Industrial Waste Management by S. E. Jmgensen

Volume 6

Trade and Environment: A Theoretical Enquiry by H. Siebert, J. Eichberger, R. Gronych and R. Pethig

Volume 7

Field Worker Exposure during Pesticide Application edited by W. F. Tordoir and E. A. H. van Heemstra-Lequin

Volume 8

Atmospheric Pollution 1980 edited by M. M. Benarie

Volume 9

Energetics and Technology of Biological Elimination of Wastes edited by G. Milazzo

Volume 10

Bioengineering, Thermal Physiology and Comfort edited by K. Cena and J. A. Clark

Volume 11

Atmospheric Chemistry. Fundamental Aspects by E. Meszaros

Volume 12

Water Supply and Health edited by H. van Lelyveld and B. C. J. Zoeteman

Volume 13

Man Under Vibration. Suffering and Protection edited by G. Bianchi, K. V. Frolov and A. Oledzki

Volume 14

Principles of Environmental Science and Technology by S. E. Jmgensen and I. Johnsen

Volume 15

Disposal of Radioactive Wastes by Z. Dlouhf

Studies in Environmental Science 15

DISPOSAL OF RADIOACTIVE WASTES ZDENEK DLOUHY Nuclear Research Institute, Re5

Contributors : Frantiiek Cejnar Vdclav KOuiim Eduard Maldiek Otakar Vojtcch

ELSEVIER SCIENTIFIC PUBLISHING COMPANY AMSTERDAM OXFORD NEW YORK

1982

Published in co-edition with SNTL, Publishers of Technical Literature, Prague Distribution of this book is being handled by the following publishers for the USA and Canada Elsevier/North-Holland, Inc. 52 Vanderbilt Avenue New York, New York 10017

for the East European Countries, China, Northern Korea, Cuba, Vietnam and Mongolia SNTL, Publishers of Technical Literature, Prague for all remaining areas Elsevier Scientific Publishing Company 1, Molenwerf P.O.Box 21 1 1000 AE Amsterdam, The Netherlands Library of Congress Cataloging in Publication Data Dlouhf Zdengk, 1932 Disposal of radioactive wastes. (Studies in environmental science; Translated from the Czech. Bibliography: p. Includes index. 1. Radioactive waste disposal. TD898.D5913 621.48’38 ISBN 0-444-99724-5 ISBN 0-444-41696-X(Series)

v. 15)

I. Title. 11. Series. 8 1-9826 AACR2

(8 1982 Zdenek Dlouhf Translation 0 1982: LuboS OndriVek and Jana Ondr5Ekovk

All rights mewed. No part of this publication may be reproduced, stored in a retrieval system, or transmitted in any form or by any means, electronic, mechanical, photocopying, recording, or otherwise, without the prior written permission of the copyright owner Printed in Czechoslovakia

Contents

Introduction

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 Radioactive wastes

. . . . . . . . . . . . . . . . . . . . . . . .

1 .'1 Origination of radioactive wastes . . . . . . 1.2 Philosophy of radioactive waste management

. . . . . . . . . . . . . . . . . .

. . . . . . . . . . . . . . . . . . 1.3 Safety and protection in nuclear power production . . . . . . . . . . . . . . . . 1.4 Radioactive wastes and the environment . . . . . . 1.4.1 Radioactive wastes from nuclear power plants 1.4.2 Wastes from reprocessing spent fuel . . . . . 1.4.3 Wastes from other processes . . . . . . . .

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

8 10

11 15 17 25 26 26 28

2 Characteristics of radioactive wastes . . . . . . . . . . . . . . . . . 30 2.1 Wastes from mining and processing radioactive raw materials . . . . . . . . . . . 31 2.2 Wastes from nuclear power plants . . . . . . . . . . . . . . . . . . . . . 2.3 Wastes from spent fuel reprocessing . . . . . . . . . . . . . . . . . . . . 2.4 Wastes from research centres and from the production and use of radionuclides

. . . . . . .

36 47 58

. . . . . . . . . . . . .

65

3.1 Processing liquid wastes from radioactive raw material mining and treatment . . . . . 3.2 Processing solid wastes from radioactive raw material mining and processing . . . . . 3.3 Processing liquid wastes from nuclear power plants, research centres and from application of radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3.1 Filtration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3.2 Chemical treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3.3 Ion exchange . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3.4 Evaporation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3.5 Comparison of liquid waste processing methods . . . . . . . . . . . . . . . 3.3.6 Some other methods of liquid waste processing . . . . . . . . . . . . . . . 3.3.7 Processing liquid wastes from nuclear power plants . . . . . . . . . . . . . 3.4 Processing solid radioactive wastes from nuclear power plants and research fac and application of radionuclides . . . . . . . . . . . . . . . . . . . . . . . . 3.4.1 Fragmentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4.2 Compacting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4.3 Incineration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4.4 Other methods of solid wastes treatment . . . . . . . . . . . . . . . . . . 3.4.5 Processing solid wastes from nuclear power plants . . . . . . . . . . . . . 3.4.6 Final waste treatment . . . . . . . . . . . . . . . . . . . . . . . . . .

65 69

3 Processing liquid and solid radioactive wastes

72 73 75 78 80 83 83 86 91 91 92 93 96 97 97

4 Processing wastes from spent fuel reprocessing 4.1 4.2 4.3 4.4

. . . . . . . . . . . . .

High-level waste disposal in deep geological formations . . . . Processing intermediate-level wastes from spent fuel reprocessing Processing wastes containing alpha emitters . . . . . . . . . Radioactive wastes from fluoride reprocessing . . . . . . . .

5 Processing gaseous radioactive wastes

98

. . . . . . . . . . 102 . . . . . . . . . . 104 . . . . . . . . . . 105 . . . . . . . . . . 106

. . . . . . . . . . . . . . . . 107

5.1 Fundamentals of gaseous radioactive waste processing . . . . . . . . . . . . . 5.2 Separation methods for processing radioactive gaseous wastes from nuclear facilities 5.2.1 Gaseous fission products . . . . . . . . . . . . . . . . . . . . . . . . . 5.2.2 Iodine radionuclides . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.2.3 Radioactive aerosols . . . . . . . . . . . . . . . . . . . . . . . . . . .

6 Fixation of radioactive concentrates . . . . . . . . . . . . . . . 6.1 Cementation of radioactive wastes . . . . . . . . . . . . . . . . . . . . . . .

6.2 Incorporation into bitumen . . . . . . . . . . . . . . . . . . . . . . . . . . 6.2.1 Categories and properties of bitumen . . . . . . . . . . . . . . . . . . 6.2.2 Composition of bitumen block . . . . . . . . . . . . . . . . . . . . . . 6.2.3 Process of bituminization . . . . . . . . . . . . . . . . . . . . . . . . . 6.2.4 Technological installations . . . . . . . . . . . . . . . . . . . . . . . . 6.2.5 Evaluation of bituminization . . . . . . . . . . . . . . . . . . . . . . .

7 Solidification of high-level radioactive wastes

.

.

130 140 140 142 145 146 153

155

155 157 157 163 166 167 168 . 170 170 . 170 171 171 172 173 . 174 174 175 176 176 177

. . . . . . . . . . . . . .

8.1 Possible application of fission products and transuranium elements . . . . . . 8.2 Separation of long-lived radionuclides and separation of valuable radionuclides . 8.2.1 Separation and isolation methods . . . . . . . . . . . . . . . . . . . . 8.2.2 Separation of elements . . . . . . . . . . . . . . . . . . . . . . . . .

107 11 1 111 120 125

. 130

. . . . . . . . . . . . .

Properties of fixation products . . . . . . . . . . . . . . . . . . . . . . . . . Solidification process requirements . . . . . . . . . . . . . . . . . . . . . . . Final form of fixation products . . . . . . . . . . . . . . . . . . . . . . . . . Fixation technologies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.4.1 Denitration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.4.2 Concentration and calcination . . . . . . . . . . . . . . . . . . . . . . 7.5 High-temperature process . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.6 Main solidification processes for high-level wastes . . . . . . . . . . . . . . . 7.6.1 HARVEST process . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.6.2 Continuous French solidification process . . . . . . . . . . . . . . . . . 7.6.3 LOTES low-temperature process . . . . . . . . . . . . . . . . . . . . . 7.6.4 PAMELA process . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.6.5 Aluminothermal process . . . . . . . . . . . . . . . . . . . . . . . . . 7.6.6 SANDIA process . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.7 Methods of monitoring product properties and testing fixation process . . . . . . 7.7.1 Microstructural analysis . . . . . . . . . . . . . . . . . . . . . . . . . 7.7.2 Macrostructural properties . . . . . . . . . . . . . . . . . . . . . . . . 7.7.3 Chemical composition . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.7.4 Chemical resistance . . . . . . . . . . . . . . . . . . . . . . . . . . . 7.7.5 Physical properties . . . . . . . . . . . . . . . . . . . . . . . . . . .

7.1 7.2 7.3 7.4

8 Use of radioactive wastes as raw material

.

180

. . . 184 . . . 186 . 188 . 189

8.2.2.1 8.2.2.2 8.2.2.3 8.2.2.4 8.2.2.5 8.2.2.6 8.2.2.7 8.2.2.8

Strontium . . . . . . . . . . . . . . . . . . . . . . . . . . . . Promethium and cerium . . . . . . . . . . . . . . . . . . . . . . Cesium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Zirconium and niobium . . . . . . . . . . . . . . . . . . . . . . Ruthenium . . . . . . . . . . . . . . . . . . . . . . . . . . . Technetium, rhodium and palladium . . . . . . . . . . . . . . . Krypton and xenon . . . . . . . . . . . . . . . . . . . . . . . . Americium and curium . . . . . . . . . . . . . . . . . . . . . .

9 Radioactive waste disposal

.

....................

189 190 191 192 192 192 193 193 195

9.1 Selection of site for radioactive waste disposal . . . . . . . . . . . . . . . . . . 199 9.1.1 Criteria for determining the suitability of radioactive waste disposal sites . . . . 199 9.1.2 Data indispensable for site evaluation . . . . . . . . . . . . . . . . . . . 203 9.1.3 Site evaluation process . . . . . . . . . . . . . . . . . . . . . . . . . . 204 9.1.4 Use of radioactive tracers for site evaluation . . . . . . . . . . . . . . . . 206 9.2 Land burial . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 208 9.2.1 Basic units of radioactive waste disposal site . . . . . . . . . . . . . . . . 208 9.2.2 Operation of storage site . . . . . . . . . . . . . . . . . . . . . . . . . 213 9.2.3 Properties of stored wastes . . . . . . . . . . . . . . . . . . . . . . . . 215 9.2.4 Waterproofing of storage areas . . . . . . . . . . . . . . . . . . . . . . 218 9.2.5 Storage of waste with high-level radioactivity in containers . . . . . . . . . . 219 9.3 Depositing radioactive wastes into geological formations . . . . . . . . . . . . . 220 9.3.1 Storing wastes in salt deposits . . . . . . . . . . . . . . . . . . . . . . . 220 9.3.2 Storage of gaseous radioactive wastes from reprocessed spent fuel . . . . . . . 224 9.4 Storage of radioactive wastes in seas and m a n s . . . . . . . . . . . . . . . . . 225 9.5 Waste effluents released into the environment . . . . . . . . . . . . . . . . . . 227 9.5.1 Radioactive wastes released into the atmosphere . . . . . . . . . . . . . . . 227 9.5.2 Waste discharge into surface waters (rivers. seas and oceans) . . . . . . . . . 235

10 Transport of radioactive wastes

. . . . . . . . . . . . . . . . . .

10.1 Transport of radioactive wastes on installation site . 10.2 Organization of waste collection . . . . . . . . . . 10.3 Transport of radioactive materials . . . . . . . . .

238

. . . . . . . . . . . . . . . 238 .............. ..............

11 Economic problems of radioactive wastes disposal

239 241

. . . . . . . . . .

246

References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

259

Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

261

7

Introduction

The lack of conventional power resources and the growing demands on electric power have given nuclear power great importance. At the end of 1970, the installed capacity of nuclear power plants in the world was 15 GW(e); by the middle of 1978, 215 nuclear power plants with a total output of 102.5 GW(e) had been commissioned in 21 countries throughout the world. By the end of 1976, 140 nuclear power plants were operating outside the socialist countries; in that year they produced a total of 1386 EJ (385 TWh) of electric power. In the socialist countries, the USSR, GDR and Bulgaria operate nuclear power plants and in Czechoslovakia the first nuclear power plant, the A-1 at JaslovskC Bohunice, was commissioned on December 25th, 1972, and the second at the end of 1978 in the same locality. Nuclear power plants operating in different parts of the world now have a sum of experience corresponding to more than 1700 years of operation and in some countries nuclear power plants account for more than 20 % of total power production. Data indicating future trends of development are not unified. Sober forecasts for 1980 indicate 120 GW(e) as being the most probable world production, a growth trend which is predicted to continue well into the future. Long-term prognoses consider an output of 250-300 GW(e) for 1990 as being realistic and forecasters believe that the total output will further increase, to reach a world nuclear power production of 500- 1000 GW(e) by 2000. By that time nuclear power plants will be turning out 55 % of total world power production, mainly to the detriment of power produced from fossil fuels. As concerns the countries of the Council of Mutual Economic Assistance, it is estimated that it will prove necessary to install in these countries a total nuclear power capacity of 150 GW(e) by 2000. Czechoslovakia will contribute to the development programme by the construction of WWER power plants in the first stage, followed by fast neutron reactor power plants with a minimum unit output of 1000 MW(e). Under the light-water reactor power plants development programme, the construction and commissioning of nuclear power plants with a total installed capacity of the order of 10 000 MW(e) is envisaged by 1990.

8

These forecasts proceed from the assumed need of electric power on the one: hand and the size of world fuel reserves and the technical level of power installations on the other. For obvious reasons nuclear power production has become one of the fastest developing industries in the history of mankind. The growing number of nuclear power plants and stringent environmental laws have brought to the fore the problem of the safe disposal of radioactive wastes which accompany the whole fuel cycle from uranium mining and fuel processing to power production and the reprocessing of spent fuel. The amounts and nature of these wastes are extremely varied and depend on many factors, as do their processing and disposal.

9

1 Radioactive Wastes

The diversity of radionuclides and materials which they may contaminate is so great that a general and unified definition of radioactive wastes has yet to be made. In approximate terms radioactive waste will, however, include all materials and substances which have come into direct contact with the radioactive medium and contain or may contain radionuclides, and all such materials that escape from the technological process or from active areas, or are removed from such areas because there is no other use for them. Their activity ranges from levels which differ very little from the level of natural radioactivity up to high-level materials with a specific activity of lo2TBq (lo4 Ci/kg) and more generated in the reprocessing of spent fuel. These wastes occur in the solid, liquid or gaseous state. Many authors often use the terms low-, intermediate- and high-level wastes, and these terms do not always help us to determine correctly the nature of the wastes. To remove these difficulties, the International Atomic Energy Agency recommended the Standardization of Radioactive Waste Categories (1) which includes solid, liquid and gaseous wastes. The Standardization is shown in Table 1. The basic criterion for the categorization is the state of the wastes prior to processing and the categories should be applicable to all wastes, ranging from radioisotopes and radioactive raw materials from mining to the reprocessing of spent fuel. Liquid wastes are divided by volume beta and gamma activities into five categories, and it is assumed that these wastes do not contain alpha emitters. Solid wastes are divided into four categories. The first three, with a negligible content of alpha emitters, differ in the irradiation rate on the surface of the container in which the waste is discarded, and the fourth category includes wastes containing alpha emitters and a negligible amount of beta and gamma emitters. The sole criterion for the categorization of solid wastes is the transport of wastes. Gaseous wastes have been divided into three categories by volume activity but it is generally assumed that experience in this field is insufficient and that the categorization lacks a true theoretical basis. In view of the fact that the definitions of the individual categories do not take into consideration the differences in the radiotoxicity of radionuclides and that

10

Table 1 : Categories of Radioactive Wastes Recommended by the International Atomic Energy Agency Liquid wastes ~~

Category

volume activity A (Bq/m3)

1

A 5 104 104 < A 5 107 107 iA 5 109 109 < A S 1014 1014 < A

2 3 4 5

Category

I

I

dose rate $ . (A/kg)

p, -emitters 11

D < D 5 10-7 10-7 < D u -emitters volume activity (Bq/m3) 10-8

Gaseous wastes Category 1 2 3

volume activity A (Bq/rn3)

A



"Cr, 56Mn, 59Fe and 6oCo. Only radionuclides with a half-life of more than 1 day contribute to the total activity. Corrosion products include 6oCo, 59Fe and "Cr, and fission products 90Sr, 95Zr, 95Nb, 99Mo, I 3 l I , 137Cs, I4'Ba and I4'La. Low-level sorbents and concentrates from evaporators have the same composition.

40

The activity of concentrates from evaporators is initially determined by the activity of I 3 l I ; after 2 months of storage 95Zr + 95Nbwill predominate and after a further 6 months the dominant radionuclides will be 6oCo which, together with I3’Cs, will continue to be the main carrier of the activity of the wastes. During storage for 12 months the total specific activity will decrease from the initial 437 GBq/m3 (1.18 x Cijl) to 107 MBq/m3 (0.9 x Ci/l). Table 7: Amount and Character of Solid Wastes and Concentrates Generated per day by WWERtype Nuclear Power Plants Volume (m3/d)

Type of material

440 MW(e) or 500 MW(e)

2 x 440 or 500 MW(e) 1OOO MW(e)

2 x lOOOMW(e)

Solid Wastes paper and textiles hard materials reactor parts aerosol filters filters with activated charcoal

0.041 0.027 0.014 0,110

0.082 0,055 0.022 0.200

0.082 0.055 0.049 0.200

0.200

0.300

0.310

Total

0.392

0.659

0.696

I

Volume (m3/d)

Type of material

Concentrates low-level sorbents high-level sorbcnts concentrates

Total

370-0.37 37 x lo3 610 x lo6

0.055 0.082 0.500

0.10

37 x 104

0.637

1.06

0.14 0.82

0.14 0.20

1.10 1.44

It is estimated that from the WWER plant, gases and aerosols will be discharged from a 100-m stack in daily amounts of 16 TBq (430 Ci) of radioactive gases and 4 GBq (0.1 Ci) of radioactive aerosols, which are roughly 10 % of the permissible limit. Compared with PWR nuclear, power plants in other countries the Czechoslovakian nuclear power plant generates a larger amount of concentrates from evaporators, because concentration is carried out only to a salt content of 200 kg/m3 and a considerable stand-by capacity is retained for disposal of wastes from the removal of operating failures. There is also a difference of several orders of magni-

41

tude in the amount of sorbent wastes because current projects do not include the regeneration of non-exchange packing material. The major part of the activity will accumulate on sorbents. When radionuclides with a short half-life are neglected, a nuclear power plant with two WWER-440 units will generate an annual volume of 62 m3 of waste sorbents with a total activity of 3.3 TBq (90 Ci), mostly consisting of 13'Cs, 6oCo and "Sr. Therefore, new projects at Czechoslovakian nuclear plants also include the regeneration of sorbents. Another very widespread type are BWR reactor nuclear power plants. The Muhlenberg nuclear power plant in Switzerland has an installed capacity of 306MW(e). Gaseous wastes are discharged from a 125-m stack at an average rate of 100 m3/d. The permissible annual limit is 330 PBq (9 x lo6 Ci) of xenon and 1.9 TBq (50 Ci) of halogens; the actual amounts discharged are 15 PBq (4 x lo5 Ci) of xenon and 22 GBq per year (0.6 Ci) of I3'I. Gaseous wastes are purified using filters with activated charcoal and absolute filters, and liquid wastes (76.7 m3/d) are demineralized, filtered and decanted. Sorbents (0.05 m3/d) are stored in the dry state in tanks. This generates an annual amount of 30m3 of wastes which are incinerated and 5 m3 of wastes which are not. The Tarapur power plant in India has an installed capacity of 380 MW(e). Gaseous wastes are discharged from a 11I-m stack at an average rate of 56 m3/s. The permissible annual limit is 670 PBq (18 x lo6 Ci) of rare gases and 3 TBq (82 Ci) of halogens and aerosols, the actual amount of rare gas effluent discharged being 14 % of the permissible level and that of halogens and aerosols 8 %. Liquid wastes (247 m3/d) are processed by filtration, chemical precipitation, evaporation and demineralization, generating annual amounts of 73 m3 of concentrates and sludges and 10m3 of ion exchangers, Solid wastes are compacted and gaseous wastes are purified by absolute filters and filters with activated charcoal. The Fukushima-1 power plant in Japan has an installed capacity of 480 MW(e). Gaseous wastes are discharged from a 120 m stack and have an annual activity of 180 TBq (4900 Ci). The activity of the wastes is reduced before discharge by absolute filters, filters with activated charcoal and partial decomposition. Liquid wastes consist of water that has leaked through the equipment (49 m3/d), washout water (13 m3/d), chemical water (19 m3/d) and water from active washeries. Cooling waters are treated by filtration and demineralization and the other liquid wastes by a combination of filtration, evaporation and demineralization. These processes generate 100 m3 of sludges, 45 m3 of concentrates and 25 m3 of ion exchangers annually. The concentrates are fixed in cement and solid wastes (0.7 m3/d) are compacted. In the USA a 1000 MW(e) nuclear power plant will have annual activity of 150 TBq (4000 Ci). Gaseous wastes contain 11 TBq (300 Ci) of 85Kr, 96 TBq (2600 Ci) of '33Xe, 1 1 GBq (0.3 Ci) of 131X,40 GBq (1 Ci) of 13'1, 1.5 GBq (0.04 Ci) of radioactive aerosols, 1.5 TBq (40 Ci) of tritium, 350 GBq (9.5 Ci) of

42

14C, 930 GBq (25 Ci) of 41Ar and 41 TBq (1 100 Ci) of other rare gases, Gaseous wastes are purified using filters with activated charcoal and absolute filters and distilled at low temperatures. Liquid wastes (90 m3/d) are processed by filtration, sorption on ion exchangers and evaporation. The concentrate (0.41 m3/d) is solidified with cement and solid wastes (0.32 m3/d) are compacted. In the USSR nuclear power plants with BWR reactors are also in operation and the construction of reactors with an output of 1000 MW(e) is envisaged. The waste waters are sorted and collected separately according to their salt content and level of contamination with radionuclides. The characteristic composition of the wastes (5) is as follows: sodium (0.6 - 15) x 10- kg/m3 iron (0-1.5) x kg/m3 corrosion products Cr, Ni, Co hardness (0-0.3) x kg/m3 kg/m3 salt content (3-50) x The total amount of wastes generated per day is 2760m3 of liquid wastes, of which 360 m3 has a salt content higher than 7 kg/m3 (regeneration solutions, desorption solutions, laboratory wastes, etc.), and 2400 m3 has a very low salt content, i.e., up to kg/m3 (washout water, water from the spent fuel elements cooling tank and leaks from the reactor). The processing of these wastes generates an annual 1100m3 of concentrates from the evaporator with a salt content of 500 kg/m3, 180 m3 of sludges with perlite and 150m3 of ion-exchange resins. The 1000 MW(e) reactor discharges a daily effluent containing the following radionuclides: 15-19TBq (400-5OOCi) rare gases (Kr, Xe and Ar isotopes) 1311 190 MBq (5 x Ci) 400 MBq (lo-' Ci) aerosols with a half-life of less than 24 h 1.9 MBq (5 x lo-' Ci) "Sr and 90Sr Table 5 and other data show that the BWR generates a greater amount of waste-saturated sorbents and an increased amount of concentrates from liquid waste processing, whereas PWR reactors generate larger amounts of solid wastes. There are no major differences between the other types of wastes. The third widespread type of nuclear power plant is that with the CANDU heavy-water reactor. The characteristic wastes from the CANDU reactor are given in Table 8. The Rajastan nuclear power plant in India, with an 220 MW(e) reactor, has a n annual discharge of 2 PBq (53 000 Ci) from an 89-m stack. The annual permissible limit is 74 PBq (2 x lo6 Ci) of rare gases, 3.4 TBq (91 Ci) of halogens and aerosols and 900 PBq (24 x lo6 Ci) of tritium in the form of water, The actual amounts of effluent discharged reach 2.5 % of the limit set for rare gases, and 0.01 % of the limit set for the other radionuclides. Heavy water serving as the moderator and coolant is returned to the reactor following treatment. Liquid wastes consist of

43

cooling water from the storage of spent fuel and various evaporation solutions. Concentrates are solidified with cement and solid wastes are compacted. The Atucha nuclear power plant in Argentina has a 320 MW(e) reactor that has been in operation since 1974. Gaseous wastes are discharged from a 40-m stack and the annual effluent limits are 410TBq' (1.1 x 104Ci) of 85Kr, 480TBq (1.3 x lo4 Ci) of 133Xe,56 TBq (1.5 x lo3 Ci) of 41Ar, 36 GBq (0.96 Ci) of I 3 ' I , 4 GBq (0.1 Ci) of aerosols and 240 TBq (6500 Ci) of tritium. The actual amount discharged is 10 % of the permissible limit for rare gases, 1 % for l 3 I I , 3.5 % tritium and 0.06 % for aerosols. Liquid wastes are chscked for radioactivity and are then either discharged or are processed in evaporators and by ion exchangers. The used ion exchangers are stored in tanks, and solid wastes are compacted into drums with a volume of 0.1 m3 (1001). The largest heavy-wa!er reactors are in operation in Canada, where the Pickering nuclear powzr plant ha; an installed capacity of 508 MW(e). The annual discharge from a 45-m stack is 930 TBq (2.5 x lo4 Ci) of tritium, 160 TBq (4.4 x lo3 Ci) of rare gases, 150 MBq (4 mCi) of I 3 l I and 1.3 GBq (34 mCi) of radioactive aerosols. Liquid wastes from this power plant may be classified into three groups: conditionally active wastes with a volume activity of less than 40 kBq ( Ci/m3): the non-active laundry, laboratories, showers, the reactor building; active wastes with a volume activity of up to 400 MBq Ci/m3); the decontamination unit, laboratories, washout waters from areas in which radioactive materials are handled, the active laundry, safety showers; chemical radioactive waters with a volum: activity of up to 4 GBq/m3 (0.1 Ci/m3): from the decontamination unit, laboratories, spent fuel storage area. Waste waters are checked for volume activity and are then either discharged with non-active waste water or are stored for a period of time needed for their volume activity to decay to a safe level, and are then discharged with the cooling water. Wastes generated by heavy-water treatment contain 6 m3 of ion exchangers and 2 m3 of filtration materials per annum. Solid wastes are sorted into combustible (0.54 m3/d) and non-combustible (0.04 m3/d). They are compacted and incinerated in the central incinerating unit. Tables 5 and 8 show that, compared with PWR and BWR reactors, heavy-water reactors generate a considerably larger amount of tritium, which is discharged in gaseous and liquid wastes. Heavy water is used as moderator and coolant and its total amount is returned to the reactor. After treatment, the amount of liquid wastes and hence also the amount of radioactive concentrates are lower. The most recent type of nuclear power plants have gas-cooled reactors and are mainly in operation in Great Britain. The Tokai nuclear power plant in Japan has an installed capacity of 166 MW(e) and waste effluents are discharged from an 80-m stack. The total annual activity of the effluents is 230TBq (6300Ci) at an annual average volume flow-rate of

-

-

-

Table 8: Characteristics of Wastes from Nuclear Power Plant with Heavy-water Reactors

Source of waste

Gaseous emissions

volume m3/d

activity

3.7 m3/s

5.4 x 103

GBq/d

tritium Discharged liquid wastes

ion exchangers

incinerable non-burnable

50-70 m3/s

2 x 102

2.9 x 103

9.1

4.1 x lo2

GBqtd

27 12 x 10-3

volume m5/d

19 16 x 10-3

2.58 x 1 0 - 7 -12.9 x 10-4 (C/kg/h)

activity GBqtd

685 0.10

2.5 x 10-3

0.51

5.5 x 10-3

19 x 10-3

1.3 x lo-'-5.16 x 10-4 (C/kg/h)

.-

Solid wastes

activity

19

21

waste waters sludges concentrates filters

Canada 500 MW(e)

volume m3/d

26

other radionuclides Liquid wastes

Argentina 320 MW(e)

India 220 MW(e)

Type of waste

64 x 10-3

5.5 x 10-3

16 x 10-3

180

0.55 41 x 10-3

0.51

~~

2.6 x lo-'-3.9 x lo-' (C/kg/h)

33 x 10-3 2.7 x 10-3

80 m3/s. Gaseous wastes mainly contain 41Ar and are purified using filtration. Liquid wastes may be classified into two groups: chemical wastes (0.65 m3/d) and waste waters from laundries and showers (15.4 m3/d). The processing of liquid wastes generates an annual amount of 0.2 m3 of concentrates and 1 m3 of ion exchangers. Solid wastes (0.01 m3/d) mainly contain paper, textiles and graphite residues. In Great Britain, modern nuclear power plants have an installed capacity of 820 MW(e), using two 410 MW(e) reactors. Gaseous wastes mainly contain 41Ar generated by the activation of natural argon. Purification of the coolant gas is carried out with glass-fibre filters, ceramic filters and absolute filters. Ion exchangers (0.004 m3/d) are stored in a lagoon and solid wastes (0.27 m3/d) are incinerated. Compared with other types of reactors, gas-cooled reactors have a high level of gaseous wastes, radionuclide 41Ar being the major component. On the other hand, they have the smallest amount of concentrates from liquid waste processing. Solid wastes contain graphite and their content is higher than with a PWR of the same output. In the EEC the production of radioactive wastes from nuclear power plants for 1970 was estimated as follows: a) liquid wastes: lo6 m3 of tritium with an activity of 400 TBq and 70 TBq (2 x lo3 Ci) of activation products, b) gaseous wastes: 70 PBq (2 x lo6 Ci) of activation products with a short half-life, 400 TBq (lo4 Ci) of fission products of rare gases, 4 TBq (100 Ci) of halogens and 40 GBq (1 Ci) of radioactive aerosols. Considerations of the long-term development of nuclear power projxts envisage the use of fast reactors. In view of the non-existence of an industrial nuclear power plant with a fast reactor, considerations on wastes from such power plants will have to rely on experience with experimental fast reactors. The economic operation of a fast reactor nuclear power plant is closely linked with the reprocessing of the spent fuel from such a plant. Considering that this will be high-level waste and with regard to attempts made to process the fuel as soon as it has been removed from the reactor, projects have been designed for onsite reprocessing of spent fuel. Wastes generated by the reprocessing of spent fuel from fast reactors are described in Section 2.3. The operation of fast reactors generates radionuclides by fission or by neutron activation. If there is no fuel element leak, all activity of the primary circuit is determined by radionuclides generated by the neutron activation of sodium, the impurities and argon present therein. In the Russian BOR reactors the steady-state volume activities in the primary circuit and in the argon blanket reach the following values: 24Na (half-life 900 s) 1.70 PBq/m3 (46 Ci/l) "Na (half-life 8 x lo7 s) 150 GBq/m3 (4 mCi/l)

46

23Ne (half-life 38 s) 44 TBq/m3 (1.2 Ci/l) 41Ar (half-life 6.5 x lo3 s) 11 GBq/m3 (0.3 mCi/l) Other radionuclides may be formed from impurities present in sodium, the most frequently occurring being iron, nickel, chromium, and hydrogen. The only radionuclides important from the point of view of radioactive wastes are 54Mn and WO. A leak from the fuel elements must be considered even in the normal operation of fast reactors. This will cause the release of fission products, uranium and transuranium elements into the sodium circuit. It is assumed that in the BOR reactor a leak from one fuel element will cause the sudden release of 93 (GBq ((2.5 Ci) of 85Kr, 330 GBq (9 Ci) of "Kr, 700 GBq (19 Ci) of "K.r, 110 GBq (3 Ci) of 131Xe, 9 TBq (240 Ci) of '33Xe, 3 TBq (90 Ci) of '35Xe, 560 GBq (15 Ci) of I 3 l I and 63 GBq (1.7 Ci) of 137Cs.These fission products mostly pass into the gas blanket. The activity of the released fission products will increase considerably when sodium penetrates into the fuel element and will radically increase when it melts. The other fission products remain in the primary circuit and the cold trap will trap most fission products together with the non-active impurities. The amount and nature of the radioactive substances trapped by the cold trap depend on the system of traps and on the frequency of fuel element leaks. Deactivation waters are another source of waste. Considerable amounts are generated by the decontamination of spent fuel elements using a deactivated solution, hot water or steam. The removal of spent fuel elements will contaminate handling equipment and the respective spaces, whose deactivation will generate intermediate- and low-level wastes. The amount and composition of these wastes cannot yet be estimated.

2.3 Wastes from Spent Fuel Reprocessing Spent fuel from nuclear power plants contains economically important amounts of unused nuclear fuel and newly generated fuel. For this reason, the reprocessing of spent fuel has been included in the fuel cycle, the aim of which is to isolate the components of unused fission products. The reprocessing consits in the mechanical cutting of fuel elements into smaller parts, which are then dissolved in acidic solutions and the individual basic components are separated. The isolation of uranium, plutonium and certain transuranium elements and fission products takes place. This method is known as aqueous or liquid extraction. The most important industrially used method is the Purex system, which is used by most reprocessing plants. Dry methods, such as the distillation of fluorides, are being developed and are often considered for spent fuel from fast reactors.

47

Reprocessing plants are the source of the greatest amount of a wide range of high-level radiotoxic materials. The reprocessing technology will then depend on the composition and form of the wastes. Table 9: Specific Activities of the Most Important Fission Products in High-Level Wastes Radionuclide

90Sr 90Y 93Zr 5Zr

95Nb 99Tc lo3Ru Io6Ru 1 0 6 a

'"Sb lZ5Te 127mTe 1191

134cs '35Cs '37Cs IQ4Ce 144pr 147Pm lslSm ls2Eu Is4Eu lssEu

Half-life (S)

8.8 x 2.3 x 4.7 x 5.6 x 3.0 x 6.6 x 3.5 x 3.2 x 3.0 x 8.5 x 5.0 x 8.6 X 5.4 x 6.3 x 9.5 x 9.5 x 2.4 x 1.0 x 8.2 x 2.8 x 3.9 x 5.0 x 5.7 x

Total specific activity of fission products

Activity in Bq per kg of reprocessed fuel for a period of 10 years

100 years

10l2 10l2 107 10" 10"

2.2 x 10'2 2.2 x 10'2 7.0 x 107

2.4 x 10" 2.4 x 10" 7.0 x 107

01'

5.2 x 10'

109 10'' 10l2

1.6 x loro

1 year 10' 105 1013 lo6 lo6 106 107 loo 107

lo6 lo6 1013 107 1013 10' 107 103 107 109 lo8 10'

107

2.8 x 2.8 x 7.0 x 2.1 x 4.8 x 5.2 x 5.6 x 7.8 x 7.8 x 2.3 x 2.6 x 2.2 x 1.4 x 5.6 x 1.1 x 3.7 x 1.2 x 1.2 x 2.8 x 4.0 x 4.0 x 2.4 x 1.6 x

loL1 1O'O 10'0 lo6 10l2 107 10l2 1013 1013 1OI2 1OO ' lo8 10" 10"

6.3 x 1013

-

-

1.6 x 1O1O 2.3 x 1O1O 9.3 x 109

-

1.4 x 2.7 x 1.1 x 3.1 x 3.7 x 3.7 x 2.6 x 4.0 x 2.4 x 1.6 x 5.2 x

lo6 10" 107 10'" 109 109 10" 1O1O 10' 10" 109

1.1 x 1013

-

5.2 x 10' -

-

-

1.4 x lo6 -

1.1 x 107 3.7 x 10"

-

1.9 x loio 1.3 x lo6

3.3 x 107

-

1.3 x lo'*

The wastes from reprocessing plants contain fission products resulting from the fission of the nuclear fuel, transuranium elements resulting from nuclear reeciions and from the gradual decay of mother isotopes during the operation of the nuclear reactor, and radionuclides generated by nuclear reactions of structural and other materials with neutrons. In the proportion of the total activity, fission products are the most important component. The generated nuclides of fission products differ considerably in half-life, in the character and energy of released radiation and in their chemical nature. A list of the most important fission products

48

contained in high-level wastes from reprocessing 1 kg of spent fuel from fast reactors (from burning up 2850 kJ/kg (33 MWd/t) with a specific output of the reactor 50 kJ/s kg (50 MW/t)) is given in Table 9. Table 10: Specific Activities of Uranium, Important Transuranium Elements and their Products in High-Level Wastes

Radionuclide

234U 235U 236U 238u

237NP

P

2 3 9 ~

236PU Z38PU

Z39PU 24OPu 241Pu '41Pu 241Am 242Am 2 4 2 m ~ ~

243Cm 242Cm 243Cm 244Cm

Total specific activity

Half-life

(4 7.6 x lot2 2.2 x 10'6 7.3 x 1014 1.4 x 1017 6.5 x 10'3 2.0 x 105 9.2 x 107 2.8 x 109 7.6 x 10lL 2.1 x 10" 4.4 x 108 1.2 x 1013 1.4 x loLo 5 8 x 10' 4.7 x 109 2.3 x 1.4 x 107 9.5 x 108 5.7 x 108

Activity in Bq per kg of reprocessed fuel per year

1.5 x 105 3.2 x 5.2 x 5.9 x 1.3 x 6.7 x 4.8 x 3.3 x 5.9 x 9.6 x 1.8 x 2.6 x 5.6 x

103 104 104 107 lo8 104 109 107 107 1O1O

105 109 1.5 x lo8 1.5

x lo8 x lo8 x 10"

6.8 1.5 2.0 8.9

x 108 x 10''

2.7

x 10l1

Concerning long-term disposal, the nuclides of Cs and Sr, nuclides with a high yield fission reaction and a relatively long half-life and relatively high radiotoxicity, are especially important. Nuclides generated by reactions other than fission reactions mainly belong among the transuranium elements. The amount of which is considerably smaller than that of fission products. Owing to their long half-lives and the nature of their radioactivity (alpha emitters) they comprise a special category from the point of view of waste disposal. Here are mainly concerned with the nuclides a neptunium, plutonium, americium and curium, whose activity in the wastes decreases considerably with increasing atomic number of the element. Table 10 shows the characteristics of the most important transuranium elements in high-level wastes on the condition that in the reprocessing the losses of uranium

49

and plutonium will reach 0.5 %. Induced radionuclides include the nuclides of iron, cobalt, chromium, nickel, manganese, etc., i.e., nuclides generated by the activation of elements contained in stainless steels. Depending on the type of fuel element can, the nuclides also include those of zirconium, niobium, titanium aluminium and certain other elements which are part of the fission products (after Nb and Zr this group includes Mo, Tc, Ru, Sr, Y,etc.). During spent fuel reprocessing the groups of radioactive materials are involved in different technological processes and concentrate in different forms. Depending on the nature of the thus generated wastes, these fractions should be concentrated, isolated or mixed. Three types of wastes are always processed separately, viz., solid wastes, liquid wastes, gaseous wastes. Solid wastes from reprocessing plants include: a) insoluble remnants of fuel elements, b) solid materials used in the technology, c) structural parts of the equipment. Depending on the type of nuclear fuel and technology, solid wastes from the first stage of the production process may include parts of fuel element cans manufactured from various metal1 alloys or sintered materials and structural parts of elements that have been mechanically separated. In the process of solution an undissolved residue may remain, which will become part of the high-level solid wastes. In the process of separation, ion exchangers (organic or inorganic) or sorbents are used, which become depleted after a certain period of time and must be removed. This also applies to packings of filtration units (activated charcoal, metal filters silica-gels) which are used for the purification of discharged gases and for arresting aerosols. Parts of the operating equipment become radioactive wastes, i.e., when faulty parts are replaced, whole equipment is dismantled or as a result of an accident. With these parts deactivation is always considered a necessity and their resulting activity depends on the process. Liquid wastes create the greatest problems with regard to both processing and waste disposal. Plants reprocessing spent fuel using the process of liquid extraction generate approximately 0.5 - 1 m3 of high-level liquid wastes per 1000 kg of spent fuel from light-water reactors. This amount includes 35 -40 kg of oxides of fission products and more than double this amount of oxides of non-active materials, mostly products of corrosion. Depending on the reprocessing technology, the composition of these wastes may differ considerably owing to the proportion of dissolved parts of the fuel element can or chemical additives used in the process. The composition of liquid high-level wastes is varied and depends on many factors. The most important include:

50

a) the type of fuel and reactor, fuel element burn-up, the duration of the storage of spent fuel in the interim storage area (cooling of spent fuel elements), b) the type of process including the used chemicals, structural materials, equipment, etc. The reactor type and its operation affect the ratio of fission products in the generated wastes. This ratio is also affected by the time that elapses between the removal of the spent fuel from the reactor and its processing. In general, it may be said that the chemical composition of the fission products is not affected by burn-up and cooling mostly changes only the ratio of radioactive to stable nuclides. The energy spectrum of the neutrons of the fission reaction has only a minor effect on the ratio of certain nuclides. This mainly applies to fast reactors, where fission takes place by neutrons with a higher mean energy and where the mixture of fission products is enriched with nuclides of medium fission yield, i.e., radionuclides of Ru, Rh, Pd, Ag, Cd, In, Sn and Sb. All of these deviations from the rule have only a slight effect on the chemical composition of the fission products. The composition of the waste products is much more affected by the reprocessing technohgy. The most widely used extraction method is the Purex technique with modifications to certain stages. The fuel element is treated with high-molar nitric acid, which converts uranium, transuranium elements, fission products and part of the structural materials into a solution. The substances contained in this solution are then separated in a series of extraction processes, purified and isolated for purposes ensuing from the logic of the fuel cycle. Radioactive substances with different specific activies may then be processed as wastes, either as a mixture or separately. Volatile fission products containing rare gases, iodine, part of the ruthenium and part of the fission products are released in the first stage of the process and are processed as gaseous wastes. Following the first extraction cycle, most of the other fission products and radionuclides generated in the reactor by nuclear reactions involving structural materials become concentrated high-level wastes. These wastes should be processed separately because their composition has a fundamental influence on the technology selected for processing the highleve I wastes. Substances forming this group of high-level wastes may be classified as follows: a) fission products, b) radionuclides from the structural components of the fuel element, c) non-active dissolved parts of the fuel element, d) chemical additives used in the reprocessing: - nitric acid, - salting-out agents (NaNO,, AI(NO,), , etc.), - oxidizing agents, reducing agents and catalysts, - substances with a highly efficient neutron cross-section (gadolinium), e) non-active corrosion products, f) products of radiolysis.

51

The proportions of fission product components are relatively stable whereas the specific activity and concentration of the other components is variable. Following suitable b a t i o n , those waste products are important which form a considerable part of the total amount of generated wastes and thus significantly affect their nature. Significant concentrations are formed of residues of nitric acid, salting-out agents, dissolved non-active corrosion products and remnants of the structural materials of fuel elements. The concentration of nitric acid may be up to 6- 7 M . For the interim storage of such wastes it is necessary to use high-grade chromium-nickel steel. It is also possible to neutralize the nitric acid, which will considerably increase the content of ballast (sodium) in the wastes designed for reprocessing. It will be mentioned in Section 4.1 on high-level waste processing that other methods of interim storage assume the decomposition of nitric acid prior to waste solidification. Salting-out agents, namely sodium nitrate and aluminium nitrate, may reach high concentrations, but on the other hand may be totally absent from some wastes. Gadolinium, which in view of the high concentration of rare earths in fission products does not affect the qualitative composition of the wastes, is used as a neutron scavenger to prevent the critical level of fission materials in radioactive wastes being exceeded. One of the common components of the wastes is dissolved parts of stainless-steel equipment used for the reprocessing of spent fuel elements. This consists mainly of iron, which in most wastes occurs in concentrations an order of magnitude higher than the concentrations of nickel and chromium. Radiolysis wastes consist of various compounds of phosphorus, i.e., acids in higher oxidation states, chlorides, etc. Various dissolved metals, such as stainless steel, zirconium, aluminium, magnesium and molybdenum, also pass into liquid wastes from the remnants of spent fuel elements. Considering other types of extraction methods of reprocessing, the Thorex technique will generate non-active impurities, such as fluorides, sulphates, phosphates, thorium and aluminium, the Redox technique will generate aluminium nitrate and the Darex technique will generate chlorides, etc. Several typical compositions of hig-level waste solutions are given in Table 11. Next to the high-level wastes from the first extraction cycle, other liquid wastes from spent fuel reprocessing also contain a large amount of radioactive wastes. A wide range of components of fission products pass into other technological flows and the wastes therefrom should therefore be considered as high-level wastes, even though the specific activities of the radioactive substances will be several orders of magnitude less than those mentioned abcve. Special attention should be devoted to wastes with a high content of transuranium elements, which should be processed separately because most of the nuclides are long-lived. The action of radiation on the solvent extraction medium used causes its radiolytic decomposition. With the commonly used tributylphosphate, a mixture of carboxylic acids, nitro compounds and organic nitrates is generated, together with

Table 11: Typical Composition of Solutions of High-Level Wastes (6) A - Non- Fission Products

Concentration (moi/m3) Na NH4 Mn Zn Hg Al Fe Cr Ni Mo Th U Pu HNOi NO;

s0:Po:-

FB0:-

Thorex

Purex Handforc

ICPP acidic

0.2

1.o

0.05

0.001 0.3 0.06 0.05

0.1 0.7 0.02 0.01

Purex Tarapur

0.03 0.003 0.003

0.06

0.01

0.1 0.01

0.012 1.6 0.05

ICPP stainles! Darex

-

Marcoule 0.97 0.02

0.01 0.07 0.02 0.01

1.3 0.4 0.02

0.09

0.02 0.03

0.005 i x 10-4 0.003

3.4 x lo-'

0.017 3.4 x 10-

7.0 0.07 0.05 0.03

0.5

2.49 1.1 0.01

H+/1.0 Ht/3.2 2.6 5.8 0.6 0.01

Ht/0.9 6.4

1.96 4.43 0.01

0.012

CI -

0.003

B - Fission Products Se

Rb Sr

Y

Zr

Mo

Tc Ru Rh Pd Tc cs Ba La Ce Pr Nd

0.2 2.0 6.0 1.3 20.0 20.0 4.0 7.6 1.o 1.o 1.7 13.0 5.0

3.5 5.5 5.0 10.0 10.0

10.0 4.8

22.0 45.0 44.0 9.0 16.0

5.9

25.0

4.3 5.0

10.0

-~ Rare earth

34.0

2.3 1.3 4.8 4.7

3.6 1.72 1.6 3.7 1.7 3.8

81.0

53

a wide range of other types of organic compounds in lower concentrations. These compounds Qften have complex-forming properties and cause the retention of a number of metal cations in the organic phase. The thus degraded extraction medium has to be regenerated. This is done by vacuum distillation, which generates intermediate-level liquid wastes, sorption by organic sorbents, which generates solid wastes (AI,O,, SiO,, MnO,, TiO,, etc.), or by chemical decomposition, e.g., by hydrolysis using mineral acids, pyrolysis, dealkylation, etc. All technologies that have so far been developed result in the generation of solid or liquid radioactive wastes, with an activity higher than that of wastes from the first extraction cycle. The situation is completely different with the so-called dry technologies of spent fuel reprocessing, where the fluoride volatility process deserves special attention. Following the fluorination of mechanically treated fuel and the distillation of volatile fluorides, there usually remains a residue containing the less volatile fluorides of fission products (Cs, Sr, rare earths, Y, etc.), in a relatively pure state or mixed with partially fluorinated aluminium oxide, which forms the bed of the most commonly installed fluidized bed reactor. Volatile fluorides of the other fission products are adsorbed on sorption columns with a packing of fluorides of alkali metals or alkaline earths metals, which in the process of chemisorption form complex fluorides or addition compound. The wastes from the process have an extremely high specific activity owing to the small amount of ballast materials and the high burn-up of the fuel of fast reactors. This type of waste is solid, but with Table 12: Gaseous Radioactive Wastes from Spent Fuel Reprocessing Reactor Fuel burn-up (TJ/kg) Capacity of reprocessing plant* (kg/d) Cooling period (d) Nuclide

PWR 2.6 4,800 150

3H

1291

1311

7.5 73

550 5.4 x 103 2,350 8.6 x lo4

130 x 103 1.3 x lo6

270 x 103 2.7 x lo'

Radiation dose** exposure @Gy/day) (mrad/year)

2 71

328

520 1.9 x lo4

Required decontamination factor* * *

20

50

2000

Discharged effluent (GBq/day) (Ci/year)

9

;

2000

* is related to the reprocessing of spent fuel from the fuel cycle of nuclear power plants with an output of 50 OOO MW(e) ** with 85Kr,related to the skin exposure from external irradiation; with 3H, related to wholebody exposure from internal contamination by inhalation and ingestion of tritium; and lz9I, related to the exposure of the thyroid gland owing to iodine ingestion. with

54

respect to its origin and composition (containing most of the fission products) and to the technology employed for its disposal it is very similar to high-level liquid wastes from the first cycle of spent fuel reprocessing. Spent fuel reprQcessing generates not only high-level wastes but also larger amounts of intermediate-level liquid wastes. These are mainly used decontamination solutions, rinsing solutions, residues from the generation of the organic phase, etc. These solutions are collected and processed using various solidification met hods. Reprocessing spent fuel is by far the largest source of gaseous radioactive wastes in the fuel cycle of nuclear power plants, causing the quantitative release of all gaseous radionuclides that accumulate in the fuel element during fission. The gaseous radionuclides are released in the first stage of spent fuel reprocessing, i.e. during the removal of the can from the fuel rods and the destruction of the fuel itself by dissolution in acid or by the dry method, i.e. pyrometallurgical technologies. The principal radionuclides which may be present in the gaseous wastes of the reprocessing plant are the nuclides of iodine, xenon, krypton and tritium. With regard to the short half-life of '''1 (8.05 days) and '33Xe (5.27 days), the amounts of these radionuclides in the gaseous wastes from the reprocessing plant depend on the time which elapses between the removal of the spent fuel from the reactor and its reprocessing. After 150- 180 days from the removal of the fuel from the reactor, the activity

High-temperature

3H

8.6 1000 150 B5Kr

160 x 103 1.6 x 106 1.6 x lo6 1.6 x lo7

Fast

1291

4 40

1311

3H

85Kr

440 15 x lo3 1.6 x lo6 4.3 x lo3 1.5 x lo6 1.6 x lo7

2.5 95

5 193

275 1700 1.0 x lo4 6.1 x lo4

2.5 90

5.5 200

100

100

3000

20

100

for critical group of population for dispersion coefficient = s/m3 *** calculated on assumption that D (30 mrad/year)

=D

(3H)

6.9 1400 90 1291

6 58

1311

133Xe

120 X lo3 900 1.2 x lo6 8.9 x lo4

410 550 x lo3 1.5 x 104 2.0 x 107 5 x 10'

+ D (8sKr) + 1/3 D ("'I + I3'I)

0.1 3.8 -

0.8 pGy/d

55

of the gaseous wastes from the reprocessing plant is determined by the content of long-lived gaseous radionuclides in the wastes, i.e., "Kr, I2'I and tritium. The variation of the activity of the individual gaseous radionuclides which are released from reprocessing spent fuel from the fuel cycle of a nuclear power plant with an output of 50 000 MW(e) using light-water, fast or high-temperature reactors with the degree of the fuel burn-up, the cooling period and the capacity of the reprocessing plant is shown in Table 12.

-

coohny period (days)

Fig. 3. Dependence of lJII decontamination factor on cooling time of spent fuel

The technologies used in reprocessing plants assume a cooling period of 150 to 180 days. Under such circumstances the activity of 13'Xe in the spent fuel and thereby also in the gaseous wastes is negligible and allows their discharge into the atmosphere. The amount of 13'1 which passes into the gaseous wastes during the reprocessing of thus cooled fuel is not so high as to become a limiting factor for the discharge of these gaseous wastes into the atmosphere, provided that a suitable technology is used for removing 1 3 1 1 from the wastes. The situation is completely different when the spent fuel cooling period is shortened to less than 100 days, i.e., as is required by the fuel cycle of fast reactore. Table 12 shows that in such circumstances both 133Xeand especially "'I contribute significantly to the total activity of the gaseous radioactive wastes and thereby also to the potential contamination of the environment of the reprocessing plant. The values of the decontamination factor for gaseous wastes for reducing the dose commitments to 0.82 Gy/d (30 mrad/year) will reach values for '"I with which currently used technologies are unable to cope. This situation is illustrated in

56

Fig. 3, which represents the dependence of the decontamination factor for l3'I on the time needed for cooling spent fuel from a light-water reactor with a specific output of 40 kJ/kg s (40 MW/t) at an average burn-up of 3.5 TJ/kg (40 000 MWd/t). The required effectiveness of the decontamination process is given for reprocessing plants with capacities of 300 and 1,500 t of fuel per annum given the highest permissible discharge of '''1 effluents into the atmosphere at 628 kBq/s (1.67 Ci/s) or 5 GBq/d (50 Ci/year). Of the gasous radionuclides, "Kr, 1291and tritium are very special. The cooling period of the spent fuel is so short compared with their half-life that it does not affect their accumulated activity in the fuel elements, i.e., their activity, which may potentially escape into the atmosphere during reprocessing. Owing to the vertical and horizontal circulation of air currents in the atmosphere, these radionuclides represent a continuous source of atmospheric contamination not only in the area of the reprocessing plant but practically on a global scale. I

I I

167.8 I

I 122.2 I

I I

I 14.81 I

24

953

-

year

-

GWe

1663

4 500

"R

Fig.4a. Growth of total world production of 05Kr

On the basis of the technological and economic interdependence of nuclear power plants and reprocessing plants, and on the basis of demands on power sources and the ensuing indispensable growth of nuclear power, production estimates may be made of the amounts of "Kr, 1291and tritium generated and potentially released from the fuel cycle of installed nuclear power plants. Many analyses have been made in this situation (8). The annual and total amounts of "Kr, 3H, '291

57

generated comply with the envisaged growth of the installed capacity of nuclear power plants (Fig. 4). The bottom curve in Fig. 4 shows the annual prcduction of these radionuclides and the upper curve their total amount discharged into the atmosphere at normal pressure and temperature if all of the 85Kr, 12'1 and tritium generated during spent fuel reprocessing were to be discharged into the atmosphere. All such forecasts are only approximate, yet they do provide basic data which should be used for evaluating the potential hazard caused by these long-lived radionuclides on both a national and an international scale. This is so even though the model used does not adequately consider the fact that approximately 75 % of the discharged effluents will be released in the northern hemisphere and that the circulation of the air currents in the atmosphere does not guarantee the even distribution of these effluents in the atmosphere on a global scale. With regard to the considerably varied physical properties of krypton, iodine and tritiated water vapour the concentration of iodine 12' vapour and tritium vapour in the atmosphere will be considerably less than the values given in Fig. 4. owing to their washout by precipitation or deposition in solid or liquid aerosols. The fact that plants for reprocessing fuel with a short cooling period use equipment for removing 13'1 from gaseous wastes and that 80-90 % of the tritium from the reprocessing of spent fuel passes into liquid wastes makes "Kr the dominant component of gaseous wastes from the reprocessing of spent fuel and long-lived fission products discharged into the atmosphere. The fact that radiation exposure of the body organs of the world's inhabitants to "Kr will not in the near future reach the maximal permissible dose does not justify its discharge into the atmosphere. It also appears that the contamination with 85Kr of krypton and other rare gases separated from the atmosphere may become a serious limiting factor for their practical application. It is therefore desirable that appropriate separation methods be applied in all spent fuel reprocessing plants.

2.4 Wastes from Research Centres and from the Production and Use of Radionuclides Radioactive wastes from research and development work, the production of radioisotopes and their use in the national economy are generated in small amounts and are characteristic of the wide range of materials and radionuclides present in these materials. They mostly originate from contact between various materials and radionuclides and their nature and amount are highly dependent on the operations conducted. Radioactive wastes of this type consist of solid materials, such as filter-paper, textiles, plastics, laboratory glass and injection syringes, carcasses of experimental animals and litter from biological experiments. Liquid wastes mostly

58

-year

Fig. 4b. Growth of total world production of 'H

'0

I

I

1980

7990

24

1

-year

Fig. 4c. Growth of total world production of '''1

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include materials used for washing and cleaning solid materials, remnants of radioactive solutions and animal excreta. Gaseous wastes may contain aerosols and gases generated from the evaporation of radioactive solutions or the incineration of solid materials. According to the methods used for the sorting, processing and disposal of these materials, and their properties, types and radionuclide content, the wastes may be classified into the following groups: 1. the content of radionuclides, into high-level, low-level and with a conditional activity level, 2. the type of radionuclides, into long-lived and short-lived, 3. the type of radionuclides, into radionuclides with a high level of radiotoxicity, a medium level of radiotoxicity and a low level of radiotoxicity, 4.chemical properties, into acidic, neutral and alkaline aqueous solutions organic liquids, aggressive substances and non-aggressive substances, 5. their physical properties, into solid, liquid and gaseous, compactible and non-compactible, combustible and non-combustible, 6. according to safety level, into sharp, brittle, explosive and flammable, 7. according to disposal method, into wastes disposable on site, on heap or dump, by pipeline and by central off-site burial. Regarding the stages of development of the use of radionuclides, most countries undergo two stages, which differ considerably in the amount and character of the radioactive wastes involved. In the first stage only small amounts of relatively short-lived radionuclides are used, in health care, agriculture, industry and applied research. The amount of wastes generated is small and their disposal does not pose any difficulties. In the second stage of development, nuclear research institutes are established with research reactors, the production of radionuclides and other similar activities which require the appropriate processing and disposal of wastes. Such installations may be used to advantage for procsssing wastes from small workplaces. There are two widely differing methods of applying radionuclides, i.e., the open emitter and safely sealed emitters. The wastes from these two types of emitters differ considerably. In industry, various types of instruments are used with built-in emitters for measuring the thickness, density and level, the removal of electrostatic charges, the sterilization of foodstuffs and products, etc. The use of these instruments with built-in emitters usually does not generate any radioactive wastes. The emitters do, however, require special handling. If the service life of the equipment is shorter than that of the emitter, the producer will have to remove the emitter and will re-use it. In other cases, the service life of the equipment is much longer than that of the emitter. If the decay of the radionuclides causes the radiation abundance of the emitter to decrease below the permissible limit, the emitter will have to be replaced. The old emitter will be disposed of. Only a leak in the emitter due to corrosion,

60

incorrect handling or damage may cause contamination of the equipment and of the workplace environment. The contamination is, however, restricted to a small area and decontamination will generate only a small amount of waste. Similar considerations apply to fire detectors containing 226Raor 241Am. In contrast to the above equipment they are used on a much wider scale and may not only be affected by failure, leakage, etc., but may also be damaged or destroyed in a fire. It is usually possible to detect and remove the emitter using simple instruments; when this is not possible, the presence of the emitter must be taken into consideration in the removal of the whole equipment. Another group of radioactive objects are various luminescent signs or adveriisements on buildings. In the past radium activation was used for these purposes, but nowaday activation by tritium and 147Pmis used exclusively. No special precautions need be taken in the demolition of buildings and the remnants of the colours from the signs may be removed and disposed of with the building material. Sealed emitters which are not connected to the equipment are a much greater hazard. These consist of emitters used in hospitals, for gamma radiography in field tests and neutron sources used for geological surveying. Originally radium emitters were used in medicine, but became a hazard owing to leakages and radon escape. This problem was solved with the introduction of artificial radionuclides for these purpose. The use of sealed emitters does not generate any wastes under normal operating conditions. Only if there is a considerable decrease in the activity of the emitter due to decay will it be necessary to remove and dispose of the emitter as radioactive waste material. These solid wastes are characterized by their small volume and considerable activity. Only very rarely do these sealed emitters cause environmental contamination which results from the leakage of the emitter or its surface contamination during manufacture. A relatively large amount of radionuclides in the form of sealed emitters is used in the radionuclide sources of electric power. Terrestrial sources contain 90Sr in ceramic form and plutonium is often used in space research. The technique for the removal of the radioactive fuel from the source following shut-down is usually determined before the assembly of the source. This radioactive material is usually re-used. The application of radionuclides in open emitters usually generates radioactive wastes. The removal and disposal of these wastes is usually concurrent with the application and the amount of waste is therefore minimal and disposal is easy. Radionuclides are sometimes used directly in the environment itself, e.g., in monitoring the movement of ground and surface waters using tritium, monitoring the movement of sand and mud using glass particles labelled with 46Sc or 19'Au and monitoring the distribution of phosphorus in fertilizers. These applications do not generate radioactive wastes, but they do require careful environmental control. The remaining radionuclides should either be allowed to decay in one place, e.g., radionuclides used for monitoring the environmental hazards of fertilizer applica-

61

tion, or should quickly be dispersed to such an extent as to prevent their becoming a hazard to man and the environment. Another frequent application is the monitoring of the movement of materials in various industrial processes. The most frequent applications in industry are monitoring the process and efficiency of mixing materials, the duration of material retention in equipment, the effectiveness of filtration, the distribution of salts in crystallization, etc., i.e. applications which serve to optimize the technological process. Negligible amounts of radionuclides suffice for large amounts of tested materials, which means that the radionuclides will be sufficiently diluted in the equipment and no radioactive wastes will be generated. Radionuclides are also used for materials testing, i.e., to check craks in materials and to study the movement of the medium in pipes. In gas pipelines the most frequently used radionuclides are "Kr and the liquid phase of 24Na. The latter is short-lived so that not even the use of a relatively high activity will generate radioactive wastes; the low radiotoxicity of "Kr allows the discharge of large amounts into the environment. Only very low activities will be used for research on and the development of new industrial applications. In most instances the behaviour of one element or component is monitored and for this purpose a large number of diverse radionuclides are sometimes used. In biological research mainly tritium and substances 1ab:lled with I4C are used; the radionuclides characteristic of other research fields are difficult to determine. Higher activities are used in research involving irradiated samples, such as activation analysis or the investigation of materials fatigue. The irradiated material should be disposed of as solid radioactive waste because it contains induced activity, i.e., 6oCoand "Fe. The use of open emitters is of long standing in medicine, I3'I, I9'Au, 32P and 90Yare used for therapy and a wide range of small amounts of radionuclides are used for diagnostic purposes. The radioactive wastes generated (liquid and solid) ar; mxtly short-lived and decay during short-t-rm storag:. Liquid wastes are then checked for radioactivity and discharged; solid wastes are either incinerated or disposed of using other appropriate technologies. Luminescent colours are a very widespread form of the use of radionuclides, and are used on the dials of various equipment. The initially used 226Rahas been replaced with 90Sr. Equipment and instruments with such dials show a considerable intensity of radiation above the surface and are therefore disposed of as solid radioactive waste. Recently, tritium and promethium have been used to excite luminescence in these colours. These elements are characterized by low-intensity radiation and a much lowzr radiotoxicity. The problems of radioactive wastes from workplaces involved in the development of reactors and of technological processes of the fuel cycle and workplaces producing radionuclides are completely different.

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Radionuclides are produced either at independent workplaces or within larger radiochemical research complexes. Currently more than 100 radionuclides are produced and the amount increases by 20 % annually. The most important radionuclidesproduced are tritium, 32P,6oCo,90Sr, 1311, 13’Cs, 19’Ir, I9*Au and isotopes of transuranium elements. The production of radionuclides generates solid, liquid and gaseous radioactive wastes contaminated with almost all radionuclides produced. The amount and nature of the wastes depend on the amount of the radionuclide produced, its physical and chemical properties and the production technology. It is therefore very difficult, if not impossible, to characterize the amount and nature of the wastes. The separate collection and disposal of concentrated wastes containing the radionuclides will be highly significant for waste processing and disposal. Gaseous wastes from radionuclide production mainly contain tritium, radioactive iodine, activation and fission products and certain alpha emitters. The development of nuclear power production and its fuel cycle, namely the production of fuel and spent fuel reprocessing, is under way in various nuclear research institutions. These institutions tackle such problems as improved design and operation of existing reactors, the development of new and better types of nuclear power plants, the development of nuclear fuels and structural materials, spent fuel reprocessing and radioactive waste processing. In most instances nonradioactive materials are used, but in some instances radionuclides, often highlevel, have to be used to test the methods developed. This will generate a considerable amount of various radioactive wastes containing a wide range of radionuclides in various concentrations. Most of these are low-activity liquid and solid wastes. The volume and activity of wastes from research institutions vary with the scope and objective of the research work; the method of waste collection and sorting has a decisive importance for low-level wastes. In some instances cooling and waste waters from non-active workplaces are not separated at all or are only partly separated, which in turn generates large volumes of low-level wastes. The largest amounts of radioactive wastes are from research institutions with a high proportion of radiochemical laboratories, whereas theoretical research institutes obviously generate smaller amounts of radioactive wastes. The operation of many research institutes makes it possible to specify the amount of radioactive wastes generated. Those workplaces in which cooling water and waters from non-active workplaces are separated generate annually 10- 100 m3 of low-level liquid wastes per worker handling radioactive substances, which corresponds to 2-30 m3 annually per employee of the research institution. Given adequate waste sorting, only 5 - 10 of these wastes will require processing. At workplaces in which cooling waters and water from non-active workplaces are not separated, the volume of low-level wastes will reach 200-300 m3 per worker handling radioactive substances, corresponding to 50-200 m 3 of wastes per employee of research instituticn.

63

As for low-level solid wastes, the annual amount is 0.30-0.8 m3 per worker handling radioactive substances. Of these wastes, 40 - 60 % are various combustible materials, such as paper, wood, textile, rubber, plastics and biological materials, and the remainder are non-combustible materials, of which 30-60 % may be reduced in volume by crushing and compacting. In addition to these wastes from routine operation, larger amounts of solid wastes of a different nature are sometimes also generated by the reconstruction of active areas, replacement of equipment, dismantling of laboratory and pilot plant equipnient, etc. These solid wastes usually contain contaminated building and structural materials and large equipment, such as fume chambers, dust-tight cupboards, containers, and laboratory equipment, whose deactivation and volume reduction are difficult. Gaseous wastes from research centres mostly contain 41Ar, other rare gases, tritium and other activation and fission products. The amount and composition of intermediate level-wastes depend entirely on the nature of the research work conducted at the particular institution. The main sources of such wastes include metallurgical research on the properties of irradiated fuel from which the major part of the radioactive material remains in the solid state, the development of spent fuel reprocessing and the separation of rare products which generates liquid wastes. In the EEC the production of radioactive wastes in for 1970 has been estimated as follows: (9) a) liquid wastes; - production of radionuclides: total volume of liquid wastes 1000 m3, total activity 40 TBq (1000 Ci) of tritium, 4 TBq (100 Ci) of other radionuclides, - use of radionuclides in medicine: total volume of liquid wastes 100 m3 total activity, 40 TBq (1000 Ci) of short-lived beta emitters, - other uses of radionuclides: total activity 4 TBq (100 Ci) of tritium, 3 TBq (100 Ci) of other beta emitters, - nuclear research centres: total volume of liquid wastes lo6 m3, total activity 400TBq (lo4 Ci) of tritium, 400TBq (lo4 Ci) of other beta emitters, 40 GBq (1 Ci) of alpha emitters, b) gaseous wastes; - production of radionuclides 70 TBq (2 x lo3 Ci) of tritium, 190 GBq (5 Ci) of activation and fission products, 10 GBq ( 1 Ci) of iodine isotopes, 10 GBq (1 Ci) of alpha emitters, - research reactors: 4 PBq (lo5 Ci) of 41Ar, 4 PBq (lo5 Ci) of other rare gases, 40 TBq (lo3 Ci) of tritium, 4 TBq (100 Ci) of other activation and fission products, - research nuclear centres: 40 TBq (lo4 Ci) of tritium, 400 GBq (10 Ci) of rare gases, 400 GBq (10 Ci) of other activation and fission products.

64

3 Processing Liquid and Solid Radioactive Wastes

The processing of all types of low- and intermediate-level wastes is based on processes that have long been used in various industrial fields, mainly in waste water treatment, industrial waste water treatment, gas emission control and solid waste incineration or compacting. In contrast to these processes, in nuclear energy higher efficiency is required and new technologies have therefore been investigated and applied. The main difference, however, rests in the fact that radioactive concentrates are formed in radioactive waste processing (e.g., chemical sludges, evaporator concentrates, saturated sorbents, air-technology filters, incinerator plant ash) which must be further processed into a solid homogenous form and safely buried until the radionuclides decay. Although exaggerated care is sometimes devoted to radioactive wastes, other, far more dangerous, industrial wastes are often treated in a completely unsatisfactory manner. The technologies developed for radioactive waste processing could thus also be applied to the processing of different industrial wastes and harmful substances.

3.1 Processing Liquid Wastes from Radioactive Raw Material Mining and Treatment Mine waters tend to be re-used as much as possible, thus reducing the need for fresh water and the amount of liquid wastes. Several techniques have been used for mine water quality improvement, viz.: - the use of electrolytes reduces the amount of rock released in the hydraulic sluicing of tailings into the worked-out space; - oil-containing waste waters are collected separately; - losses of unused explosive fractions containing ammonium salts are reduced;

65

- accumulation in mines of crushed sulphite ores in oxidation conditions is minimized in order that the possibility of acid formation and increased metal dissolution is reduced. If the amount of mine water exceeds the amount requirzd for re-use, the excess water is discharged without further treatment into a closed interception system. This may only be done when a system is in opxation with an evaporation capacity that exceeds the amount of mine water discharged. If the capacity of the interception system is insufficient, the excess water may be discharged into surface waters, i.e., following checks and treatment. The discharged water quality should comply with valid standards. The specific method of treatment and discharge depends on local conditions, such as the type of rocks, climatic conditions, topography and the distance from the dressing plant. Treatment can include the separation of uranium, radium and other metals using different processes, e.g., sedimentation, lim: neutralization, ion exchange, coprecipitation with barium salts and cementation on iron surfaces. In Czechoslovakia, mine waters are treated prior to discharge. An ion-exchange column packed with a cation exchanger is used for treating mine waters with low salt contents that are contaminated with radium. The regeneration column solution is decontaminated with finely ground baryta, a selective sorbent based on activated barium sulphate or by coprecipitation in mixer-settlers. The decontamination factor of this simple process is a b m t 10, while a volum: reduction (the ratio of treated mine water to dry concentrate) of lo6 is obtained (10). Research and development in this fi:ld has mainly been focused on selective sorbents and the use of reversible organic gels (sulphonated styrenes). Activated barium sulphate-based selective sorbents cannot be used in c o m m m sorption columns in view of the fact that their structure is altered in the process and their sorption properties deteriorate. Facilities of the sedimentation tank type are therefore used for these sorbents. Operating trials have been conducted with a two-stage apparatus of this type, having a concentration factor of 2 x lo6 and a maximum decontamination factor of 20; sorbznt regeneration, however, has been significantly difficult. Mine waters sometimes present problems even after the closing of the mine when their composition gradually alters and the amounts of suspznded substances and contaminants decrease. Ion exchangers have recmtly bzen used for treating mine water from closed mines containing mainly sulphuric acid and mstal ions. The same method of water treatment as for mine waters is also used for seepage water from waste rock heaps or ore tailings impoundments. The most efficient method of reducing the environmental hazards by chemical treatment plant operation is minimizing effluents. In some instances, unless the technological process is infavourably affected, the use of water in a closed system is possible. This technique is often employed in alkaline leaching. Tailings are discharged in tailings impoundments and slurry water is drained and recycled. The procedure, howxer,

requires detailed water balancing throughout the process and occasional optimization of water management during operation. Water recycling is mostly affected by process wastes. The technologies and chemicals used therefore be selected so as to permit maximal re-use of water. It is also useful to retain separately water with a higher level of contamination. Provided that the climatic conditions are favourable, it is possible to replace the waste water drainage by the evaporation of excess water in the tailings impoundment. The msin method of treating acidic waste waters is neutralization with lime or a lime-limestone mixture (1 l), in which the following processes take place: - sulphuric acid neutralization, - precipitation of some sulphates, - precipitation of heavy metals, - partial removal of 230Th,232Thand ,"Pb, - removal of most 226Ra;neutralization to p H 8 may reduce the radium content to 3.7 -7.4 kBq/m3 (100-200 pCi) and less, - removal of ammines from the precipitate. Opxating experience in France has shown that the efficiency of radium removal depends greatly on the neutralizing agznt used, i.e., lime or limestone. In radium removal, limestone is less effizient than lime. It was therefore recommended that limestone should only be used for neutralization to pH 2, while further neutralization, up to p H 8, should b: carried out with lime. This procedure will considerably increase the amount of radium removed in neutralization. Barium is commonly used for radium removal in clear solutions. Also, the addition of a BaCI, solution to decanted water from lagoons neutralized to p H 8-9, with subsequent addition of aqueous solutions of the salts of long-chain acids (oleic, stearic or palmitic acid) in a n amount of kg/m3 (10 mg/l) of waste water has led to sulphate precipitation with mixed oleates of calcium, barium and radium. In this situation, relatively rapid sedimentation takes place, and the resulting precipitate is less soluble than the sulphates themselves. The discharge of slurry water sluiced into the impoundment with the tailings will result in a considerable particle suspension and thus also in the reduction in specific actikg/m3 (10 mg/l) of sodium oleate kg/m3 (10 mg/l) of BaCI, and vity. are added to waste water containing 4-40 kBq/m3 (100- 1000 pCi/l) of radium, thus reducing the content of radium to 70- 190 Bq/m3 (2-5 pCi/l). This value is comparable to a concentration of 4-400 Bq/m3 (0.1 - 10 pCi/l), which is the natural concentration in water flows. Should part of the slurry water be drained from the tailings impoundment, it will usually be neutralized prior to sluicing. In spite of this, 4-40 kBq/m3 (100lOOOpCi/l) of 226Ra, sulphates and other contaminants remain in the soluble form. Therefore, the 226Ra content i n water to be discharged is reduced by coprecipitation of BaSO,. To achieve this, in the exscess of sulphates BaCI, is added at p H 8 -9. Water must not contain tailings residues because the presence

67

of these solids would considerably reduce the process efficiency. The reduction is due to the exchange of radium present in the tailings for barium. Precipitation using BaCl, and efficacious clearing reduce the radium content to 100 Bq/m3 (3 pCi/l); for reasons that have not yet been elucidated, such a high efficiency has not been achieved in operating conditions. The precipitate has a very low solubility and does not present any immediate danger. Nevertheless, new problems arise in long-term operation or as a result of operational changes. After long-term operation, the precipitate accumulates and becomes a potential source of contamination. It contains a much larger amount of radium than normal tailings. Its discharge to tailings impoundments should be restricted and, depending on the capacity and the nature of the technological process, the precipitate should be removed from the impoundment. Should the sediments be kept permanently in their original state, precipitations may, during or after operation, reduce the sulphate content and thus cause gradual precipitate dissolution and radium release. The solidification of the precipitate and the methods of disposal are discussed in Section 3.2. In Czechoslovakia, removal of radium from slurry waters drained from tailings impoundments has been tested using ground natural baryta. The baryta, with a grain size of 1-3 mm and a total volume of 18 m3, was packed in four concrete columns connected in series of two. The bottom sand layer was separated from baryta with a metal mesh. Each column had a capacity of 0.01 m3/s (10 I/s) and counterwashing was carried out using half the flow-rate for at least 15 min. The purified water flowed out into a 55-m3 tank and from there into a 19-m3 tank. From this tank it was pumped into the water flow. Wash water was supplied to a 32-m3 tank and was then returned to the impoundment. This simple procedure reduced the radium content 3 - 5-fold but the operation was not reliable enough. Water seepage from tailings impoundments forms another group of liquid wastes. Where no water is discharged from the impoundment, seepage water is usually intercepted during operation and pumped back. Given a suitable slope, seepage water may be intercepted in a tank in the slope below the impoundment. The seepage water will then evaporate in the tank. Seepage water may be treated together with slurry water drained from the impoundment. If, for instance, a large amount of discharged alkaline slurry water is mixed with seepage water, the resulting mixture will either be alkaline or neutral and will therefore only have a low heavy metal content. Only acidic seepage water will remain, which may contain a considerable amount of heavy metals when the impoundment is filled. It would thus be useful to neutralize seepage water even after the impoundment is filled.

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3.2 Processing Solid Wastes from Radioactive Raw Material Mining and Processing Solid wastes from radioactive raw material mining are characterized by a large volume and a small content of natural radionuclides. Waste rock heaps are reclaimed in order to restore the devasted landscape. The most appropriate method of removing the waste rock from the environment is to return it to worked-out mine spaces. This, however, requires suitable hydrogeological, radiological, technical and economic conditions and is not always possible. Heaps therefore sometimes remain on the surface. Heaps should have a minimal adverse environmental impact and their structure should be such as to ensure long-term mechanical stability. Having passed through the heap, the percolated precipitation will have a certain activity due to natural radionuclides. It is therefore useful to divert the percolated precipitation, monitor it and possibly also treat it. After the heap has been put out of use, its height and slope will be reduced, the rock covered with earth and the whole area reclaimed. Contaminated building and structural materials are usually discarded into suitable spaces in mines, waste rock heaps or tailings impoundments. After operation has been discontinued, the used equipment may be decontaminated and re-used elsewhere; unusable parts are disposed of using the same procedure. Access to the material should be bsrred and environmental contamination minimized. There is no generally applicable technology for the disposal of tailing from chemical dressing plants. A thorough geological survey takes place before the siting of a tailings impoundment. Boreholes are drilled to investigate the structure of bedrock while the horizontal and vertical permeability of rocks is tested in the laboratory. The data obtained are used for estimating the annual seepage rate from the tailings impoundment into the environment. The best solution would be to return the tailings to worked-out mine spaces: this would, however, involve considerable technical difficulties and significant costs. While tailings contain sand and fine-grained precipitates, only the sand fraction is suitable for filling deep mine spaces. The remaining fine-grained tailings contain most of the radium, show very poor properties for disposal into lagoons and are difficult to reclaim. Sometimes tailings are disposed of in open-cast mines. Ground and surface waters may, however, be contaminated with radionuclides, heavy metals, and harmful chemical substances. Most frequently, therefore, the usual tailings disposal areas are built. Tailings impoundments are usually dammed; the design and dimensions will differ considerably. The characteristics of a tailings impoundment are given by the capacity of the chemical processing plant, the type of processed ores and processing

69

method, topography of the area, water evaporation rate, bedrock and dam permeability, etc. A new design of tailings impoundment has been developed in Canada by Robinsky (12). This type of impoundment does not have steep dams, does it have the top layer of liquid. In lieu, the tailings area is cone-shaped and has greater erosion resistance. Slurries are discharged from one place in the middle of the tailings area. Every 3 days the slurry pipe with the discharge outlet is raised. The conical shape with a’slope gradient of 5 - 6 % is achieved by discharging concentrated slurry with a solid phase to water ratio of 1 : 2. At first the slurry is discharged so as to cover the whole area evently. The water content is then gradually reduced. The tailings settle homogeneously and do not part into coarse and fine particles. The tailings area is surrounded with an impermeable dam which retains excess liquid and fine precipitates. The dam height is much lower than in conventional tailings impoundments. No decanting towers or decanting pipes are required, a shallow channel being satisfactory. This new type of tailings impoundment has been used by ore dressing plants but not, however, by uranium ore dressing plants. Suitably sited and correctly operated and maintained tailings impoundments do not present a significant environmental hazard. Nevertheless, accidents do sometimes occur, such as dam failure, pipe rupture, dam overflow owing to heavy rain, etc. As a result, a large amount of tailings may escape and contaminate the neighbouring area. The siting of the tailings impoundments should therefore be such as to preclude significant tailings release into the surrounding area. Slurry water remains in the impoundment and excess water is discharged. Slurry water is removed by evaporation or seepage, which should be as low as possible. Accumulated water is pumped back to the tailings impoundment. Where the evaporation capacity of the impoundment is insufficient, excess water should be discharged. Another method of tailings disposal is to dump the tailings into lakes which are not used for drinking water supply and water management. In almost all instances, tailings are discharged into deep water. The density temperature and chemical composition of the water in the disposal area must be known. It is sometimes necessary to drain the area around the lake and to regulate the flow in their tributaries. Tailings impoundments should be monitored and maintained long after they are no longer in use. A number of problems are encountered, viz.: - as a result of rainfall and water seepage, the fine precipitate may remain in liquid form for a considerable time and may act hydrostatically on the tailings impoundment dams. Erosion may damage the dam, releasing a large amount of tailings; between 1942 and 1972, clay dams in twenty tailings impoundments containing phosphate tailings failed in this way in the USA; - the reclamation of tailings areas is difficult, as is the maintenance of a vegetation cover. Long-term water and wind erosion may therefore occur;

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- the tailings disposed of are suitable for building purposes and i t is difficult to prevent their illegal exploitation. People living in houses built using these materials are then exposed to increased radon concentration and to external irradiation; - the natural air flow above the tailings areas prevents the formation of dangerous concentrations of radcn or its decay products; buildings constructed on the area will disturb the natural air flow. Housing developments in the former tailings areas should therefore be permanently prevented; - tailings impoundment dams usually retain rainfall from areas larger than the impoundment proper. This water is then discharged using ditch and dam systems. After the tailings impoundment is filled the systems dilapidate and large amounts of rainfall may then infiltrate the tailings impoundment, dams may be scoured and erosion irmeased; - seepage water may infiltrate through bedrock or dams and contaminate ground or surface waters used for irrigation or drinking water. The decontamination of tailings is not considered to be possible and the respective area is expected to remain affected by wastes for a long time even after the impoundment has been filled. The permanent stabilization of tailings areas is therefore indispensable, improving their appearance and protecting the area from water and wind erosion. Reclamation, which is the best method, is not possible in all climatic conditions. In some instances, tailings are covered with a layer of crushed rock (usually waste rock), o r chemical substances, cement or bitumen are applied. This protection, however, may serve only as a provisional precaution facilitating reclamation. A number of problems should be considered prior to the stabilization of the impoundment, including: preserving the depression in the middle of the impoundment, - in case of depression is preserved, the water lagoon must be maintained and its level controlled, - surface levelling or sloping, - reducing the steepness of the slopes of dams. The solution should always respect local conditions. The fundamental criterions is the maximum stability of the dams under all predictable circumstances. In view of the long half-life of radium and its decay products, the tailings areas should always be isolated and marked so that it is obvious even after a long time that they are not suitable for uncontrolled use. Fences and warning signs have only a short-term character and the areas should therefore be recorded in cadasters, the public should be informed and the area monitored. Waste concentrates from chemical processing plants are mostly disposed of in tailings ponds. The Ba(Ra)SO, precipitate is usually accumulated in large sedimentation tanks where it is not sufficiently isolated from the environment. The precipitate should be disposed of in ponds where i t is dispersed in the tailings. Another

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suitable technique is cementation and burial on-site or in tailings ponds. Filter cakes from monazite processing are encapsulated in metal containers with a concrete jacket and buried in a suitable demarcated area. Contaminated parts of chemical processing plant equipment are usually buried in tailings impoundments. The site selected should allow the material to be covered with slurry very quickly. The slurry layer above the material should be at least 1 m thick when the impoundment is filled. The same technique is also used for the disposal of solid materials from the deactivation of rooms and equipment in the dressing plant after the end of operation. Leaching residues on heaps are solid wastes which to a certain extent resemble chemical dressing plant tailings but may contain the residues of leaching solutions. It is therefore advisable to wash the residues with water, which is then treated or discharged depending on contamination. The remainder of the process is the same as with waste rock heaps. The heaps are given a mechanically stable shape, are covered with a layer of earth and reclaimed. In some instances, on-site leaching is carried out. This requires that the rock is washed with water to remove the leaching solution residues. No other treatment is necessary.

3.3 Processing Liquid Wastes from Nuclear Power Plants, Research Centres and from Application of Radionuclides Liquid wastes from nuclear power plants and research establishments and from radionuclide applications range from small volumes of considerably active solutions to different amounts of low- and intermediate-level waste water and to very low-level wastes whose direct discharge into the environment is associated with only minimum hazards. The same variety may be found in the number and concentration of radionuclides present in the total volume of waste waters and the concentration of non-radioactive fractions. If the increased volume activity of radionuclides in liquid wastes precludes the direct discharge of the wastes, it is first necessary to decrease their radioactivity. The objective of each processing step is to concentrate the radioactive fraction in a minimum volume and to separate the concentrate from the remainder of the waste water volume. The volume activity is thus reduced to such an extent that the waste water may be discharged following one or several stages of treatment. The treatment, however, never eliminates the entire volume of activity and a small part of it escapes into the environment with the discharged waste water. In addition, different concentrates are formed in radioactive waste water treatment (chemical sludges, evaporator concentrate, used ion, exchange materials, mineral sorbents, etc.)

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containing most of the radior,uclides. They should therefore be further processed and safely disposed of. The basic processes in liquid waste treatment include filtration, chemical precipitation, evaporation and ion exchange. In some instances biological treatment, solar evaporation, electrodialysis or direct cementation are also used. The methods of liquid waste treatment are almost as numerous as there are workplaces employing them, which shows that there is no suitable universal process for liquid waste treatment. The optimum technology should always be selected very carefully. Factors affecting the methods of liquid waste treatment include local water quality, waste water composition, radionuclide concentrations, the presence of small amounts of detergents and complex-forming substances and regulations governing the permissible concentrations of radionuclides and harmful chemical substances in waters.

3.3.1

Filtration

Filtration is the simplest method of liquid waste processing. Prior to discharge waste waters must often be neutralized. The precipitate formed is removed, together with part of the activity. The decontamination factor obtained, however, is very low. Filtration is much more frequently employed as part of a complex technological scheme together with ion exchange or chemical treatment. Water filtration preceding ion exchange increases the volume of liquid wastes processed in a single cycle, extends the life of ion-exchange resins, facilitates process control and allows more efficient regeneration and washing. Following chemical treatment, clean water is partly separated from the precipitate by filtration. Gravity filtration through a sand bed is the oldest filtration method. It is little used for radioactive waste treatment because it is suitable only for a few types of precipitates, in the process sand is choked with sludge, the filtration rate is very low and a relatively small sludge concentration is obtained. Two types of filters are used, depending on the filtration rate. Low-performance sand filters give a filtration rate of 2 - 5 m3/mz d. They are relatively efficient and can be easily manufactured from locally available materials. However, they are too big and are therefore not suitable for processing large amounts of waste waters. Rapid sand filters having a filtration rate of 3 - 20 m3/m2 h are much more common. In addition to the chemical composition of the filter material, the filtration rate depends on the size and relative distribution of the particles. Rapid sand filters or anthracite filters are most frequently used after chemical treatment, including precipitation, flocculation and sedimentation. The operation of gravity filters is very simple and almost no maintenance is required. Sludge is removed manually and therefore only very slow specific activity sludges may be treated using this

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technique. At Harwell, the technique is used for processing sludges from low-level waste waters with a specific activity of 4 MBq/m3 Ci/m3); the dose rate at a distance of about 0.05 m above the sludge is around 360 nA/kg (5 mR/h). The filtration rate may be increased by applying pressure. Pressure filters are most frequently used for treating waste water prior to ion exchange or adsorption. Pressure filters have the same performance as rapid sand filters, and usually have the form of a sealed container. The precipitate-clogged filter bed is removed by countercurrent washing. At Harwell, pressure filtration is employed for sludges containing calcium phosphate and iron hydroxide or calcium phosphate, copper hexacyanoferrate (11) and a small amount of iron hydroxide. The sludges mainly contain fission products and small amount of uranium and plutonium. Their total volume activity is around 19 GBq/m3 (0.5 Ci/m3). A candle filter is operated at Karlsruhe with a total filtration area of 4 m2 operating at a pressure of 7 x lo5 Pa (7 atm). Pressure filters are relatively simple and cheap. Sludges treated with pressure filters have a lower water content than those obtained by gravity filtration, but the volume of sludges is increased by auxiliary filtration and therefore a smaller volume reduction is attained. Vacuum filters are the commonest type of filtration equipment. Vacuum filters with a layer of infusorial earth have proved best for most radioactive sludges. At Marcoule, this technique has been used for intermediate-level sludges (40 MBq/m3-40 GBq/m3 - 1 Ci/m3)) with an 8- 10 % solid phase content. Sludge is obtained containing 54 % of water and an average of 5 % of infusorial earth. Prior to disposal it is bituminized. Another vacuum filter has also been used for processing higher level wastes with an activity of 370-740 GBq/m3 (10-20 Ci/m3) containing 35 % of iron hydroxide, 35 % of nickel hexacyanoferrate (11) and 30% of infusorial earth. These colloidal sludges are more difficult to process and after filtration they contain only about 18-22 % of solid phase. At Los Alamos sludges formed by the precipitation of iron hydroxide from low-level waste waters are filtered. A vacuum filter with an infusorial earth layer is used. The sludges contain 1 % of solid phase, which is increased to 5 - 10 % by sedimentation. After filtration, sludges containing 50-60 % of water are obtained. At Trombay, chemical precipitation sludges from low-level waste waters contain calcium phosphate and iron hydroxide in amounts of 1-2 %. After vacuum filtration, a product is obtained consisting of 15 % sludges, 19 % filter material and 66 % water. Vacuum filters are characterized by reliable operation and a low failure rate. They can easily be batch operated and the product formed is of a suitable composition for further solidification. Crystalline sludges are easy to filter whereas the filtration of colloidal materials is far less effective. In the chemical treatment of waste waters, colloidal sludges are often produced which are difficult to filter. Different techniques have therefore been tested, leading to colloidal particle precipitation and water removal. At Harwell, the use of a freez-

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ing process prior to filtration has been devised. Sludges were slowly but competely frozen and large ice crystals were formed. Filtration proceeded within 8 h after defreezing. Defrozen sludges do not have to be rigorously stirred. The cooling system used a CaCI, solution. After vacuum filtration the solid phase concentration increased from 1 - 2 % to 20-25 %. The equipment was in use for 8 years. The only disadvantages were a relatively low concentration factor and considerable welded joint corrosion caused by the CaCI, solution. Similar freezing-out equipment has been used at Mol for processing BaSO,, Cu,Fe(CN), , Fe(OH), and Ca,(PO,), sludges containing of up to 20 % of organic matter. Freezing has been carried out in two I-m3 tanks cooled using a Freon/glycol system. After sedimentation the sludge has contained 4 % of solids. Following freezing and vacuum filtration an increase to 35 -40 % has been recorded.

3.3.2

Chemical Treatment

Large-scale chemical treatment of liquid wastes is being applied in many nuclear facilities (13). The decontamination factors achieved are not significantly high, but the process is economical and is suitable for large volumes of low-level waste waters. Where a higher decontamination factor is desired, chemical treatment is often used as the first stage of processing prior to evaporation and ion exchange. The advantages of chemical treatment mainly include low costs, a sufficient decontamination factor for most low-level liquid wastes, the relative insignificance of the presence of non-active salts, the fact that no special equipment is required, easy and simple operation and the simultaneous removal of ions and colloidal radionuclides. The disadvantages of chemical treatment include a relatively low volume reduction, a relatively low decontamination factor for intermediate-level wastes, difficult processing of formed sludges and a considerable amount of laboratory work necessary for determining the optimum proportioning of chemicals. Chemical treatment is the commonest technique of liquid radioactive waste processing. It uses the insolubility of hydroxides, carbonates and phosphates which precipitate with cations of many fission products, thus removing most of them. Efficiency is increased by the addition of a carrier. The decontamination factor obtained in chemical treatment varies widely depending on the initial concentration of radionuclides, the type and concentration of non-active salts and the method of waste processing. For instance, the decontamination factor for Ru and Cs is usually less than 2 while for Sr it is 10-100 and for some alpha emitters it is higher than 10 000. The choice of the precipitation technique depends on many factors, of which the most important are the type and activity of radionuclides present, the required decontamination factor and the type of non-radioactive fractions present. The

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radionuclide activity in waste waters is normally very low and is thus insufficient both for the formation of insoluble precipitates of these radionuclides and for the coagulation of finely dispersed particles without which the precipitate cannot be separated. Various coprecipitation agents are therefore added, allowing coprecipitation of some radionuclides together with the non-active ions present or the adsorption of part of the radionuclides on the formed precipitate. Most waste waters also contain dispersed solid particles on which radionuclides are adsorbed. Chemicals causing heavy flocculation are necessary for the removal of the dispersed particles from the solution, One of the main reasons why chemical treatment of low-level waste waters is so widespread is the possibility of using the process with a wide range of non-radioactive fractions. There are, however, certain substances which disturb the process and, although their presence in waste waters cannot be entirely precluded, measures should be taken so as to minimize them, viz.: - liquid wastes containing interfering substances are collected separately i n special containers; - the use of these substances is minimized and the substances are replaced with other materials that do not affect waste processing. The substances that have an unfavourable effect on liquid radioactive waste collection and processing include: - solid wastes (paper, glass, textiles) which may cause clogging of pipelines, pumps and filters, or which may pollute mixers and disturb the technological process; - oils, fats and organic solvents responsible for foaming which adversely affects the process. Solvents may accumulate in pipelines, pumps and containers and lead to the danger of explosion. Heavy solvents such as CCI, may infiltrate through the walls of plastic pipes and containers; - strongly acidic solutions which cause corrosion of pipelines, pumps and containers and require a considerable amount of chemicals for neutralization. In contrast, alkaline solutions are much more convenient because processing mostly takes place in an alkaline environment; - detergents, mainly those containing alkylbenzene, sulphonates, which unfavourably affect flocculation at a concxtration as low as 5 ppm. At a concentration above 15 -20 ppm. intensive foaming occurs, preventing flocculation entirely. The effect of detergents may be partly alleviated by the use of larger amounts of precipitating and flocculating agents. This, on the other hand, increase the amount of sludge and process costs. Reducing the detergent concentration in wastes therefore appears to be the most suitable solution. It was found that more detergents are often used than is necessary for efficient treatment. The optimum detergent concentration is very low and it therefore suffices to observe the correct solution composition, thus assuring that waste water does not contain excessive amounts of detergents;

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- complex-forming substances, e.g., sodium salts of ethylenediaminetetraacetic acid and polyphosphates whose effict on flocculation is similar to that of detergents; the required concentrations are higher; complex-forming substances are often used in large amounts for deactivation of equipment. It is sometimes useful to encapsulate small amounts of used detergent solutions with these substances in separate containers. The reasonable use of complex-forming substances does not usually cause problems in waste water treatment. Chemical waste water treatment takes place in three stages. First, precipitating chemicals are added to the liquid. The chemicals are added while the whole liquid volume is rigorously stirred to attain rapid and homogeneous chemical distribution. The rigorous stirring of precipitating chemicals is very important: slow stirring would cause the precipitant to disperse very slowly and the reaction would only be limited in area. Adverse reactions might also occur. The second stage is coagulation, i.e., a series of chemical reactions and physicochemical processes and changes in which a finely dispersed precipitate is formed. The last stage consists of stirring the liquid in which flocculation takes place, i.e., the finely dispersed particles agglomerate. This insoluble precipitate settles, entraining colloidal materials present in the solution. The floccules may entrap colloidal materials in three ways, viz.: - by simple mechanical wrapping, - by adsorption of colloidal materials on floccules, and - by neutralization of positively charged colloidal particles with floccules which have negative charge at the time of formation. The region of optimum precipitation is mostly very narrow and is therefore determined by laboratory tests. The most important precipitation factors include the pH ionic strength, stirring and flocculation, temperature and the presence of suspended substances. The pH range in which the precipitant is most effective and the concentration of ions present are to a certain extent interconnected. Strongly coloured water with a low concentration of solids and low turbidity is most difficult to treat. The pH range suitable for precipitation is very narrow. The optimum pH for the precipitation of soft, coloured water by means of alum is about 5 and it is even lower when iron salts are used. Depending on the composition of waste waters, alum may be used up to pH 7.5 and iron salts will show the precipitating effect at pH as high as 9 and above. Stirring and precipitation should be sufficiently long for the reaction to be complete. Generally, the higher the salt content or the content of the added precipitant, the shorter is the time required for stirring and flocculation. Lime-soda flocculation is commonly used for reducing process water hardness. The addition of lime and soda ash will cause the precipitation of all cations except alkali metals. If hydrogen carbonates are present in waste water, the addition of lime will precipitate carbonates, including calcium carbonate. If sulphates are present, lime and excess of soda ash should be added to remove magnesium and

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calcium in the form of carbonates. In hard water treatment an equivalent amount of soda ash is first added and lime is then proportioned to reach the optimum pH value. Strontium is removed at a higher rate than would correspond to the solubility of its salts. This is probably due to the formation of strontium and calcium carbonate mixed crystals. The decontamination factor is improved by the use of excess of soda ash. For a mixture of beta emitters the decontamination factor usually ranges between 5 and 10. The precipitation of aluminium hydroxide or iron hydroxide is another common process. Aluminium or iron salts are added to waste water and the pH is increased by the addition of lime, soda ash or sodium hydroxide, resulting in precipitation of metal hydroxides. Alum, sodium aluminate and iron salts are used for treating low-level waste waters. The basic reaction taking place when any of the precipitants is used requires an alkaline environment. The resulting precipitate is either aluminium or iron hydroxide. Hydroxides and alkaline carbonates of many multivalent cations are precipitated; alkali metals and alkali earths remain in the solution. The efficiency of the process is improved by the presence of sulphate ions. Soda ash yields better results than sodium hydroxide because strontium and to a considerable extent also other elements are removed in the form of alkaline carbonates. The efficiency is further improved by the addition of clay minerals, such as kaolinite and bentonite. The decontamination factor is in the range 5 - 10. The method of precipitation of phosphates has been developed for low-level waste water processing. Sodium phosphate is used as the precipitating agent, and 99 % of the total alpha activity and about 90 % of the total beta activity are removed from waste waters containing a mixture of fission products. The removal of ruthenium depends on its chemical form. Cesium is very difficult to remove. The methods of chemical waste water treatment described so far do not always yield a sufficiently high decontamination factor and do not permit discharge of waste water without previous treatment. Ruthenium is one of radionuclides which are most difficult to remove. The best results have been obtained using coprecipitation with copper sulphide in an alkaline medium. Another procedure is based on the formation of iron sulphide precipitate after the addition of iron sulphate and sodium sulphide at pH higher than 8. The degree of ruthenium removal may differ depending on the form of the ruthenium in the waste water. The process, however, is very efficient for removing colloids containing radioactive substances. Cesium is partly removed by ferrocyanide coprecipitation. Routine processes are satisfactory for the removal of strontium.

3.3.3

Ion Exchange

Ion exchange is a relatively simple means of deactivating liquid wastes containing only small amount of dissolved salts (14). The salt content should be lower than 2500 ppm.; liquid wastes having a salt content below 1000 ppm. are most

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suitable for this method of processing. A wide choice of materials can be used for ion-exchange waste waters treatment, organic, inorganic, natural and synthetic, which differ in their sorption properties and cost. The main advantage of ion exchange is the simplicity of the equipment and the high volume reduction achieved. The radioactive fractions may also be concentrated in a form suitable for simple packing and storage. Regeneration solutions which should be processed further are formed in the regeneration of ion-exchange resins. The decontamination factors obtained in waste water treatment using ion exchangers depend significantly on the type and composition of the waste, the radionuclides present, the type of ion exchanger, the regeneration technique, and the character of operation. Decontamination factors from 2 to lo5 have been cited in the literature. Higher values are obtained when multi-stage columns with mixed beds are used. Two fundamental processes are employed, viz., intermittent anti column operations, both with a number of variants. Mixed-bed columns contain ion-exchange resins for both anions and cations in a single vessel. Strongly acidic and strongly alkaline resins in a ratio of I : 1 are most frequently used. Other combinations have been described for a number of applications. Mixed-bed columns are used for treating reactor coolants and waste waters from various nuclear facilities. For fission product mixtures, mixed-bed columns are far more efficient than cation exchangers alone. Two series-connected columns are also often used, one being packed with a cation exchanger and the other with an anion exchanger. Compared with mixed-bed columns, their regeneration is simpler; acid-resistant structural materials should be used. Efficiency is improved by the use of several pairs of columns. Many natural materials show ion-exchange properties, e.g., polysaccharides, proteins, lignites, coal, clays and zeolites, which are often used in developing countries. Natural organic materials (brown coal, anthracite, wood, cotton, tar, nut shells, olive stones, etc.) are not stable enough and are therefore seldom used in their original form. Their disadvantages include a low ion-exchange capacity, excessive volume increase with a peptizing tendency, limited radiation stability, low mechanical strength and decomposition by alkalis. Natural inorganic sorbents (clay materials, zeolites, etc.) also show relatively low sorption capacities and zeolites also show low abrasion resistance. These materials tend to increase the volume and clay materials also show susceptibility to peptization. Zeolites are difficult to classify mechanically by grain size. Other characteristics include partial decomposition by acids and alkalis and limited stability in solutions with low salt or silicon concentrations. Recently, synthetic ion-exchange materials with an improved capacity or selectivity for certain radionuclides have increasingly been used. Even these materials, however, have disadvantages. Synthetic ion exchangers have limited radia-

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tion stability and high radiation doses result in capacity reduction, colour changes, grain size changes, shape changes, etc. Their cost is high; nevertheless, only small losses are recorded in service. On the other hand, the cost of chemicals needed for regeneration is high and processing the regeneration solution is also expensive.

3.3.4 Evaporation Evaporation is one of the most frequently used processes in the food industry and is very efficient for the removal of radioactive materials from radioactive wastes (15). Generally, evaporation is suitable for intermediate- and high-level wastes where a low-cost heat source is available, where a high decontamination factor is necessary, or where no other method of waste water treatment is feasible. Compared with chemical treatment and ion exchange, evaporation requires a high capital expenditure. When the equipment is used to capacity, the specific cost of waste water treatment is usually not very high. The average decontamination factor between the concentrate and the condensate is 104and in many instances lo6, lo’, or even more. These values are reduced by the presence of volatile radionuclides, entrainment and foaming. The maximum obtainable volume reduction depends mainly on the amount and properties of salts dissolved in wastes. Crystallization of the dissolved salts in cooler places of the evaporator, foaming, deposit formation and corrosion are the limiting factors of the technique. A small amount of dispersed solids does not usually limit the volume reduction obtainable. At high volume reduction, a high activity may accumulate in the condensate. The whole operation or part of it may have to be shielded. Increased concentrate activity is also associated with increased activity in the distillation product. The use of simple boiler evaporators or submerged-coil evaporators has been discontinued on an industrial scale. However, they are still suitable equipment for processing wastes from smaller facilities. The most frequently used types include jacked evaporators with a heating jacket or stem-heated evaporators in which steam passes through pipes of various shapes. This type of evaporators is suitable for processing waste waters which easily form sediments. The sediments are removed from the pipe surfaces by rapid cooling or mechanically after withdrawing the heating coil from the evaporator. At Chalk River a boiler-type evaporator with a submerged coil has been in use since 1958. The capacity for low-level waste water processing is 2.5 x m3/s (9001/h). Some US nuclear power plants use boiler-type evaporators in intermitted operation for treating waste m3/s (160 l/h) to 7.5 x water. The evaporators have a capacity from 4.4 x x m3/s (2700 l/h). A simple evaporator with a heating coil 2.5 x lo-’ m in diameter and 113 m in length has been in use at Harwell. The healing coil is mounted on a lever and is easily removable.

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Natural-circulation evaporators are manufactured in three variants. Horizontaltube evaporators are suitable for low-ceiling plants and may be used for processing smaller amounts of waste waters which do not tend to crystallize or settle. Verticaltube evaporators show relatively high heat transfer coefficients at high temperature gradients. Heat transfer is poor at low temperature gradients. Extended-heater evaporators ensure the maximum circulation of the evaporated liquid and in comparison with other types they may handle lower volumes of liquid, which in turn reduces the dimensions of the equipment. In forced-circulation evaporators, the liquid is circulated in heating tubes by a pump. The liquid moves with sufficient speed, which makes this type of evaporator suitable for water which forms sediments. The main advantages include high heat transfer coefficients, sufficient circulation of the liquid and resistance to sedimentation, crystallization and clogging. The disadvantages include higher costs and high energy requirements for the operation of the pumps. Problems encountered include the clogging of the tube inlets with salt deposits detached from the evaporator walls, insufficient circulation due to high pressure losses, increased corrosion and erosion and the release of salts from the tubes during boiling. Most economical is the repeated use of evaporation energy as the heat source for further evaporation. In waste vapour compression evaporators the energy of discharged low-pressure steam is increased by compression. This allows the use of a higher heat of condensation and increases the overall heat efficiency. Two types of equipment are produced, viz., evaporators with mechanical compression using a compressor and evaporators using a steam injector for compression. Electric motor- or steam turbine-driven centrifugal compressors are mostly used for mechanical compression. The liquid supplied to the evaporator is preheated to a temperature approaching the boiling point. The evaporator operates at a very small temperature gradient of 5 - 10 "C, requiring an increased heating surface. The energy requirements depend on the compression ratio and optimum conditions must therefore be determined. In France, equipment is used in which the output steam temperature is 100 "C and the temperature of steam behind the compressor is 145 "C. At Risa forced-circulation evaporation system is used with mechanical compression for concentrating waste waters to a salt content of 10 %; concentration proceeds in an electrically heated evaporator. Steam compression by a steam injector is possible only where high-pressure steam is available. Steam injectors have a relatively low efficiency (25 - 35 %), which is further reduced owing to the large pressure difference between low-pressure and high-pressure steam. Steam injectors have a lower efficiency than mechanical compressors. In spite of this, they are widely applied because the capital cost is low, they are capable of processing large amounts of steam and they have no moving parts. Multi-stage evaporators use steam from one stage for heating the liquid in the

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following stage, where boiling takes place at a decreased temperature and pressure. This technique is mainly suitable for processing large volumes of steam. A detailed feasibility study is always required for determining the number of stages and the operating conditions to be used. Film evaporators of the vertical or horizontal type are also widely used. The liquid flows down the inner wall of the evaporator and is spread by mixer blades or by several scrapers into a very thin layer. The rotating blades also break the foam and reduce the escape of particles with the vapour. Film evaporators will produce a concentrate with a higher salt content than other evaporators. The main advantages of film evaporators are the possibility of continuous output adjustment from zero to maximum and the high heat transfer coefficients. The disadvantages are higher capital costs and, in the horizontal type, the restriction of the maximum heating surface area to 20 m2. Efficient operation of the evaporators is conditional on different auxiliary equipment, such as drop separators, condensers and pre-heaters. Various separators are used for drop removal using the principle of the sudden change in the gas phase movement, cyclone separators, packed columns, wire-mesh separators, plate columns, etc. Their efficiency depends on the type of equipment and drop size. Foaming is the most serious problem encountered in evaporator operation. Foam is formed as a result of the presence of colloidal particles, particles decreasing the surface tension, or finely dispersed particles, dissolved gases, etc. Foaming significantly reduces the decontamination factor. Diverse means are therefore used to reduce foaming; the evaporator is operated at a lower output to achieve the same end. Foam breakers in the shape of a spiral or a ring are placed in the evaporator immediately above the liquid level. Foam is broken but not completely removed by a sudden decrease or increase in temperature. The same effect is achieved by the operation of an evaporator with a minimum level of liquid. In some types of evaporators, foam is removed using water or steam jets, sometimes with the addition of defoaming agents. Defoamers are also added to some evaporating solutions. Defoamers include alcohols, fatty acids and esters, amides or silicones. Foaming may also be reduced by pH control. Scale formation on the evaporator heating surfaces considerably reduces heat transfer and special attention is therefore paid to methods for its prevention. Special crystallizing agents may be added on which the scale will form to prevent it from forming on the walls. The addition of magnesium sulphate to the evaporated solution results in the formation of a calcium sulphate precipitate, which can easily be removed from the surfaces. Ethylenediaminetetraacetic acid forms soluble complexes with calcium and magnesium, thus preventing scale formation. Scale formation can also be reduced by intensive circulation of the liquid.

3.3.5 Comparison of Liquid Waste Processing Methods Liquid waste processing is a significant part of the whole system of radioactive waste processing and may considerably affect the cost of processing. Special attention should therefore be devoted to the choice of the appropriate method. Waste water treatment itself and all liquid waste processing produce concentrates (chemical sludges, evaporator concentrates, used ion-exchange materials, regeneration solutions, etc.) which should be further processed and suitably disposed of. Chemical treatment is suitable for waste waters with a relatively high salt content where a high decontamination factor is not required. The decontamination factor is normally around 10; if specific techniques are used, a value of 200 and more may be obtained for certain radionuclides. The cost of chemical treatment itself is relatively low. It does, however, achieve only a small volume reduction and further processing and disposal may be considerably expensive. A considerable amount of laboratory work is required for determining the optimum proportions of chemicals. Ion exchange is the most suitable method of processing waste waters with very low salt contents (preferably under 1000 ppm.). The simplicity of the equipment, the significant volume reduction and activity concentration is an easily disposable form are the main advantages of the method. The main disadvantage of synthetic ion-exchange materials is their expense. These materials are therefore repeatedly regenerated. The regeneration solutions must be processed and in some instances it is cheaper not to regenerate the materials at all. Ion exchange is often a suitable method for removing activity which has remained in waste waters after chemical treatment. Filtration should always be applied prior to ion exchange. Evaporation is used when a high decontamination factor is required or with waste waters that cannot be treated chemically or by ion exchange. Compared with other methods, ion exchange requires relatively high capital costs. The three processes are sometimes used separately; in most instances they are used in combination.

3.3.6 Some Other Methods of Liquid Waste Processing Other methods used for processing liquid wastes include biological treatment, electrodialysis and solar evaporation. Biological treatment reduces the contents of organic substances in liquid wastes before their discharge into the environment. It also partly eliminates inorganic ions, by filtration and adsorption or by their introduction into the metabolic cycle. Algae and bacteria may be used for the decomposition of complex organic molecules. This may be done by biological filters consisting of a thin layer of sand or finely ground rock deposited on a layer of pebbles and provided with a n outlet through

83

which the waste waters slowly seep. The layer is either aerated from below, or oxygen is supplied to the bottom layers by microorganisms in surface layers. Organic substances are removed from waste waters partly by filtration and partly by biological metabolism. The second method of biological treatment uses activated sludge. Waste waters are mixed with 15 % (w/w) of sewage sludges or with a corresponding amount of synthetic sludges and are allowed to aerate for several hours. After settling, the supernatant is discharged and sludges are disposed of. The standard activated sludge treatment will remove 96 % of activity from waste waters containing plutonium. Of other radionuclides, uranium and rubidium accumulate in biological sludges. In a normal situation, the volume activity level in waste waters is so low that biological processes are not affected by radiation. For inorganic waste treatment synthetic sludges and nutrition for bacteria should be added. Inorganic sludges may also be mixed with a corresponding amount of sewage, which serves both as sludge and nutrition. Biological treatment can be used successfully for washery water pre-treatment. This will avoid the excessive consumption of chemicals and the production of a large amount of sludge in the subsequent chemical treatment. The removal of active substances by biological processes also proceeds in biological ponds, i.e., large open-air tanks for waste water containing microorganism colonies. Electrodialysis is a relatively new method. Significant attention has recently been focused on it in connection with desalination. Units are available that reduce the salt content to several hundred ppm.; the application of electrodialysis to liquid radioactive wastes processing is still very limited. Electrodialysis is electrolysis and dialytic diffusion in combination. Ion-exchange membranes are used which form barriers between cells with different electrolytic solutions. Recently, selective membranes have been developed which theoretically only permit the passage of either cations or anions. As a result of electric potential in the cell, ions migrate to the cathode or to the anode and hit the membrane. If the membrane charge is opposite to that of the ion, the ion passes through the membrane; if it is identical, the ion is repelled. Simple cation or anion membrane cells have attained a high degree of separation of some ions (95 -99 %). In desalination and in the development of electrodialysis application for radioactive waste water processing the cells containing a minimum of one cation and one anion membrane are either used in series, in parallel or with different techniques of recirculation. The basic designs applied in practice consist of a series configuration of the individual cells, where each element contains several cells and a pair of electrodes, a parallel configuration with continuous feed recirculation and product dircharge, and a parallel configuration with recirculation, intermittent feed, recirculation and product discharge, and a parallel configuration with recirculation intermittent feed and intermittent product discharge. In electrodialysis the processed solution is passed through the cells, which are

84

separated from the electrode space on one side by a cation-exchange membrane and on the other by an anion-exchange membrane. On the application of potential, ions pass through the respective membranes to the electrode space. The membranes do not allow movement in the opposite direction to ions with the opposite charge. The salt content in the feed solution is thus reduced. However, with a reduced salt content the cell resistance often increases until further desalination becomes uneconomical. If, on the other hand, the middle cell is packed with a mixed cationand anion-exchange resin bed, the cell resistance remains almost constant and deionization may proceed down to much lower concentrations. Under suitable conditions demineralized water may thus be obtained. Electrodialysis makes it possible to obtain pure water and concentrated wastes. The degree of disposal and the concentration factor depend on cell design, current density and flow-rates. The maximum obtainable concentration factor is limited by the value at which the membranes cease to be selective and where the efficiency of the movement of ions from the electrodes towards the desalination cell starts to decrease. In countries with suitable climatic conditions, concentration by solar evaporation may be used for processing low-level liquid wastes or sludge. The process was first used at Lucas Heights (Australia) and later by the Lawrence Radiation Laboratory (USA) and at Trombay (India). It is considered suitable for a number of developing countries. At Lucas Heights sludge with aluminium hydroxides are thus processed containing 5 - 10 ppm. of beryllium, 0.37 - 37 MBq/m3 Ci/m3) of total alpha activity and 0.037-37 GBq 1 Ci/m3) of total beta activity. A concrete evaporation tank is used that has a usable capacity of 23 m3 and a n evaporation surface area of 86 m2, If required, it can be covered with a light metal roof. Sludge is supplied on a layer 0.10 m thick. When a paste-like consistency has been obtained, the sludge is cut into squares with sides 0.075 m in length. The evaporation surface area is thus increased and the irregular cracking of the sludge which results from drying is avoided. The dust which is usually generated by the packing of sludge into drums is reduced. Solar evaporation is very simple, sufficiently safe and economical. However, it requires suitable climatic conditions and a suitable sludge composition. It will probably not be sufficiently effective in areas with an average annual evaporation of less than 0.75 m. The maximum activity which may be obtained using this technique depends on the properties of the dry sludge. Practical experience has shown that solar evaporation may be used only for sludge of a crystalline nature with a very low specific activity.

3.3.7

Processing Liquid Wastes from Nuclear Power Plants

The method of waste processing depends on the type of reactor used and on the local conditions of radioactive materials disposal into the environment. The fundamental aspects include the amount of liquid wastes, possible water recycling, the presence of undissolved solids, the chemical composition and p H of water, the total activity level and the activity levels of the idividual radionuclides, permissible activities for discharge, the presence of non-radioactive harmful substances and water temperature.

m rnonitorinj

chemical

precqitafion

tank

I

_ _ _ _ _ _ _J sohdification

Fig. 5. Flow chart of radioactive liquid waste processing in pressurized water reactor power plants

Liquid wastes are collected via collecting pipelines into tanks. The basic prerequisite for the pipeline is its absolute tightness. Jacketed pipelines are often used when the collecting pipeline is placed under the surface. Leak monitoring and the collection of any escaped wastes from the pipes are provided. Waste waters which may contain oils, surfactants or other materials that unfavourably affect treatment are collected and processed separately. The liquid waste tanks are provided with water-level gauges and spillways connected to stand-by tanks. The collecting tanks are placed in a water-tight tub capable of retaining all escaped liquid in case of an accident. Figs. 5 - 7 show the basic flow charts for the processing of liquid wastes from pressurized, graphite gas-cooled, boiling-water and heavy-water reaclor nuclear power plants (16). Although liquid radioactive wastes are relatively easy to process, wastes from many nuclear power plants, especially those situated near the sea or large water flows, are discharged directly into the hydrosphere. To reduce the specific activity of radionuclides in the discharged liquid wastes, non-active cooling water from the secondary circuit is often used for dilution. This technique, however, cannot

low-salt

---______ evaporatop 1

-

2-

I

Iiyui d wades

--%-

1 -

moni forrnj

col/ectiny tanks L- - -

tank

conditwnally a c h e l i p i d wasf a

Fig. 6. Flow chart of radioactive liquid waste processing in boiling-water reactor power plants

Heavy-

water

flon i tori ny

lipid wasfes

tank Reacfor

-

Disti lla t'e

-

Discharge kav -and

lyhbl- water

lipid wastes

_ j

Collecting fanks

-

Enrichment column

Reac for

Evaporator

U

Collecting fanks

' -SolrdiTicat'ion

Fig. 7. Flow chart of radioactive liquid waste processing in heavy-water reactor power plants

be perpetuated, as the growing number of nuclear power plants increases the environmental burden caused by radioactive substances which have been discharged in this manner. In addition to increased total radionuclide activity in water flows, radioactivity increases in mineral sediments and in plant and animal organisms. Concentration factors of 1000- 10 000 are not exceptional in selected types of

87

water flora and fauna. Radionuclides cumulate in fish organs, plankton, algae, etc. Liquid radioactive waste discharged without prior treatment can therefore be only a provisional measure. In pressurized water reactor power plants (Fig. 5 ) , waste waters with high and low salt contents are separated. Wastes with a low salt content are sometimes only mechanically filtered and after checking they are returned to the primary coolant circuit or are discharged. Sometimes they are treated using an evaporator. High-salt liquid wastes are treated using evaporators. The condensate is sometimes further treated by mixed-bed ion-exchange columns and after checking it is mostly returned to the primary circuit. In some instances, liquid wastes with a high salt content are only treated by chemical precipitation. Evaporator concentrates, regeneration solutions and chemical sludge are solidified by cementation or bituminization prior to burial. In the Voronezh-type Czechoslovak nuclear power plant (a 440 MW(e) pressurized water reactor) liquid wastes are treated separately using several treatment units, viz.: - the primary circuit water treatment unit maintains the desired water quality in the primary coolant circuit at nominal ratings, and should operate reliably even if there is an increase in primary circuit activity due to a fuel element failure. The primary circuit is filled with demineralized water containing boric acid at a concentration of 0 - 6 kg/m3. Treatment proceeds in cation and anion exchanger-packed columns. The plant capacity is 20 m3/h and after saturation the ion exchangers are disposed of in tanks without regeneration; - the treatment unit for deactivation solutions regenerates solutions for the deactivation of the reactor primary circuit. The unit’s capacity is 400 m3/h and it is estimated that it will be used once in 4 years. After deactivation the solutions are discharged from the primary circuit into drainage tanks from where they are transported to the treatment plant. Soluble and insoluble corrosion products are separated using cation- and anion-exchange columns. Used filtration materials are hydraulically transported to waste storage tanks; - the primary circuit drainage water treatment plant is used for the treatment of primary coolant circuit water from operations associated with boric acid concentration changes, refuelling, draining of loops and circuits for repairs, etc. All controlled leakages from the primary circuit are collected and are returned to the primary circuit after degassing. If required, ion exchangers are used for treatment. The water deficit is replaced with pure concentrate. The concentrate containing boric acid is re-used after treatment for feed water treatment. The plant capacity is 45 m3/h and is equipped with mechanical filter and cation- and anion-exchange columns; - treatment plant for uncontrolled leakages from the primary circuit and for deactivation waters is designed for processing such leakages from the primary circuit, water from the deactivation of rooms and equipment, ion exchanger

88

regeneration solution, water from filter bed lightening, water from the transport OJ used-up sorbents to storage tanks, water from radiochemical laboratories and water from active laundries and active closures. Water is pumped fromstorage tanks via mechanical filters to an evaporation plant. The evaporation plant houses two systems, each consisting of an evaporator, a final evaporator and a condenser with a degasifier. One system is used for waste water processing and the other is designed for boric acid regeneration. If required, the second system may also be used for waste water treatment. Heating steam from the secondary circuit is used as the energy source for evaporation. After treatment in the evaporator separation equipment, waste steam is condensed and degasified. The condensate is then transported via a mechanical filter, a heat sink and ion-exchange columns. The treated condensate is used for replacing losses in primary circuit water. The salt content of the concentrate is increased in the final evaporator to 200 kg/m3 and the concentrate is transported to storage tanks; - the spent fuel storage tank water treatment plant is designed for processing water from the storage tank for spent fuel elements and water from emergency tanks. The water is treated using a mechanical filter and columns containing cation and anion exchangers. Treated water is returned to the storage tanks. In normal operation, the plant operates intermittently for 800- 1000 h a year. The maximum capacity is 40 m3/h. The used filter material and ion exchangers are stored in waste tanks. In case of an accident in the primary coolant circuit, the plant is also used for treating water from emergency tanks; - the treatment unit for steam generator surface blowdown maintains water purity in the secondary circuit within the range specified by the boiler water standard. Water is treated using a mechanical filter and ion exchangers with a capacity of 12 m3/h. Periodical blow-off and washing will increase the capacity of steam generators to 37 m3/h. Most of the liquid waste is re-used after treatment and only small amounts are discharged. The discharged water, i.e., evaporator condensate, contains only negligible radionuclide specific activity and, compared with natural radionuclides, it does not present an environmental hazard. Most radionuclides are retained in concentrates which are solidified and buried following temporary storage in waste storage tanks. Fig. 6 shows a schematic diagram of liquid waste processing in nuclear power plants with boiling-water reactors. Low-salt waste waters are mostly treated by filtration and are treated using ion exchangers. After treatment and monitoring the waste waters are discharged or recycled. In some instances, evaporation is used as a treatment method. High-salt waste waters are treated using several methods. Solids with part of the radionuclides are mostly removed by filtration. The water is discharged after checking. In some power plants the high-salt waters are treated using evaporators or chemical precipitation; evaporator concentrate and chemical sludge are solidified prior to disposal.

89

In the Leningrad nuclear power plant all liquid radioactive wastes with a high salt content are collected in 200-m3 tanks and are treated in several stages, including iron hydroxide chemical precipitation, distillation in a three-stage evaporator, rare gas removal and tertiary treatment of the filtration condensate using activated charcoal and cation- and anion-exchange columns. The deactivation plant capacity is about 8.3 kg/s with a minimum decontamination factor for volatile radionuclides of lo5.After treatment, liquid wastes are discharged or recycled. The waste volume is reduced in the evaporator by 5-6-fold and the salt content is increased to 50 kg/m3. Further concentration to 500 - 800 kg/m3 proceeds in a concentration evaporator with a capacity of 1.4 kg/s. Waste waters with a low salt content are filtered and treated by cation- and anion-exchange columns. This is an intermittent operation, with a maximum attainable capacity of 28 kg/s. Treated water is mostly recycled. All concentrates, auxiliary materials and used ion exchangers are stored temporarily in waste storage tanks for future bituminization. For this complex processing the costs of solidification and removal of liquid waste processing concentrate are relatively high. The treatment of radioactive waste from nuclear power plants with graphitemoderated gas-cooled reactors is relatively simple. Almost all waste water is collected in collecting tanks and, after filtration and checking, it is either discharged or recycled, Only rarely are 1iq;id wastes treated by evaporators or chemical precipitation. On the other hand, heavy-water reactor power plants are provided with a relatively complex system of liquid radioactive waste treatment (Fig. 7). In view of the high price of heavy water, the whole system of liquid waste retention and processing is designed so as to minimize heavy water losses and dilution. Heavy water discharged from the reactor is collected in tanks and is treated by filtration and ion exchange. After treatment, heavy water is returned to the reactor. Liquid wastes containing a mixture of light and heavy water are filtered and treated using a mixed-bed ion-exchange column. After treatment, they are further processed using an enrichment distillation column. The enriched heavy water is transported to storage tanks and from there back to the reactor. The distillate consists mostly of light water and is discharged after cooling. Waste water not containing deuterium is treated by evaporation and the distillate is checked and discharged. Washery and spray water is either discharged without treatment or is chemically treated. Evaporator concentrates and chemical treatment sludge are solidified and disposed of.

90

3.4 Processing Solid Radioactive Wastes from Nuclear Power Plants and Research Facilities and Application of Radionuclides Solid radioactive wastes are mostly treated using methods similar to those employed in processing corresponding non-active wastes. The most frequently used method include fragmentation, compacting, and incineration (17). The main objective of solid radioactive waste treatment is to reduce the waste volume, which considerably reduces the costs of waste disposal and facilitates handling and transport. The basic prerequisite for the safe treatment of solid wastes is their correct sorting. The diversity of materials and the difficulties of measuring their radioactivity require that the wastes be sorted in-situ. The sorting technique used depends on the processing techniques chosen and is often carried out before processing . Depending on the nature of the operations, the methods of solid waste processing are classified into mechanical methods (fragmentation and compacting), incineration and various special methods. Fragmentation only indirectly reduces the volume of wastes by changing the shape of the waste and thus facilitating the storage of wastes in containers or compacting or incinerating. Compacting results directly in a reduction in volume.

3.4.1 Fragmentation Fragmentation is used for reducing the volume of sizable objects or for separating equipment parts with different degrees of contamination. Operations proceed in special rooms depending on the degree of waste contamination. Personnel are clad in pressurized clothing. Small articles may be crushed in dust-proof cupboards. The following tools are mostly used for fragmentation: electric or oxygenacetylene torches, circular, reverse and chain saws, grinding wheels, shears and crushing and tearing machines. Fragmentation is carried out by many facilities. Tools are selected depending on the nature of the wastes and the method of further processing. For instance, at Saclay electric torches and shears are used, a t Marcoule underwater cutting with a plasma torch is employed, a t Harwell man-made materials are treated by tearing machines while glass is crushed and at Karlsruhe filters are cut with chain or circular saws.

91

3.4.2 Compacting Low-level wastes contain mostly materials, such as paper, plastics, textiles, rubber and small glass and metal articles, which can be compacted. Compacting is not suitable for thick and stiff materials, materials containing flammable or explosive components and for wastes with a liquid content. The efficiency and safety of compacting therefore depends considerably on the quality of sorting. At many facilities sorting carried out directly at the source of wastes is unsatisfactory and wastes have to be sorted again prior to compacting. Double sorting is disadvantageous for safety reasons and is uneconomical. Sorting tables and presses are placed in rooms provided with suitable ventilation system and operators wear protective clothing. Efficient ventilation is especially important in compacting wastes containing alpha emitters or in work with highly toxic materials (beryllium). Wastes may directly be compacted into the drums in which they are stored. A pressure of 3 -4.5 MPa (30-45 kg/cm2) is usually chosen, at which almost maximum volume reduction is achieved and ruptures of drum bottoms or walls are avoided. The maximum volume reduction of the individual types of material may be obtained using the following pressures: polyethylene 7.8 MPa (80 kg/cm2) filters 6.9 MPa (70 kg/cm2) 4.9 MPa (50 kg/cm2) textiles and rubber metal materials 108.0 MPa (1 10 kg/cm2) normal wastes 6.4 MPa (65 kg/cm2) According to the pressure used, the following volume reduction may be expected for different materials: material

origin

pressure MPa

easy to compact

laboratory wastes

3

difficult to compact

equipment dismantling

4.9

volume reduction 2-5

1.5-3

In some facilities, wastes are compacted together with containers. This technique requires costly presses and is therefore suitable only for processing large volumes of waste. Nuclear centres operate various pressing machines, ranging from the simplest manual presses to sophisticated horizontal presses with several pistons. Their operation is adapted to the type of wastes being processed and to the methods of transport and disposal.

92

Presses with an operating pressure of up to 10 t are used by some facilities for compacting small volumes of wastes, i.e.,0.02 -0.05 m3. Table 13 lists the characteristics of such machines. Table 13: Small Equipment for Solid Waste Compacting

1

Holland S. Africa Japan Norway UK

I

1

Facility

Force

Lawrence Radiation Laboratory Petten NNRC Tokai-Mura Kjeller Aldermaston

4.16 x lo4 8.80 x 1 0 5 5.90 x 104

I

I 1

Volume reduction

0.56

4 10 5 5 5

1.02

3.40 x 104 9.80 x 104

I

Piston diameter (m)

2.5 - 5

I

Larger machines are employed, e.g., in France for processing large volumes of solid wastes. At Marcoule, a 60 t vertical hydraulic press is used for compacting wastes directly into 0.2 m3 drums. For compacting, the drums are placed in a solid metal case preventing their deformation. A 250 t vertical hydraulic press is used at Cadarache for compacting wastes into a reinforced concrete container. At Saclay, a special 400 t hydraulic press was built for compacting 0.25 m3 metal drums and objects of up to a volume of 0.5 m3 resulting from the dismantling of a plutonium unit. Six 0.25 m3 drums containing waste are gradually compacted into an approximately 1 m3 reinforced concrete container. Similar equipment is also used in a number of other countries.

3.4.3

Incineration

Much low-level solid waste consists of combustible materials. Incineration reduces the volume 100-fold and the weight 13 -25-fold. Solid materials incinerated include: - cellulose materials: paper, cardboard, wood, wood-wool, cotton-wool; - plastics and rubber, and - animal corpses. The basic materials properties affecting the incineration process include the physical and chemical state of the materials and their calorific value. The physical state determines the incineration rate. Compact materials burn slowly and imperfectly while powder materials may burn too fast. The waste chemical state affects the degree of corrosion of the equipment and the calorific value determines the

93

incineration temperature in the incinerator. The minimum calorific values of the commonest waste materials are as follows: rubber polyethylene pol yvinylchloride paper, cardboard wood

41 -42 MJ/kg (9800- 10 000 kcal/kg) 36-38 MJ/kg (8500-9000 kcal/kg) 25 - 27 MJ/kg (6000 - 6500 kcal/kg) 17 - 25 MJ/kg (4000 - 6000 kcal/kg) 14.6 MJ/kg (3500 kcal/kg)

Incineration of cellulose materials is the simplest. A simple incinerator furnace equipped with a suitable smoke purification and ash removal device is satisfactory for this purpose. The incineration of plastics generates strongly corrosive gases, which very quickly damage the incinerator furnace and the smoke purification and ash removal devices. In many plants, the incineration of PVC has therefore been discontinued. Most complicated is the incineration of the corpses of animals. These biological materials have a low calorific value but a high water content and in the process odorous gases are released. In some instances, liquid organic wastes such as mineral oils and solvents are also incinerated. The liquid should be suitably conditioned prior to its injection into the furnace. The adjustment of viscosity by heating or dissolution and the removal of solid particles and possibly also water are most important. Incinerator furnaces are either single-purpose or multi-purpose. In the latter all types of materials may be incinerated; single-purpose devices are much less costly. The choice depends primarily on the overall amount of wastes and the relative proportions of the individual materials. Evenness of combustion is an important requirement in incineration, and varies with the type of wastes and their calorific value. Explosive components or materials whose combustion gives rise to toxic gases should be removed. Correct sorting prior to processing is therefore even more important than for compacting. Wastes are usually sorted manually in dust-tight cupboards or enclosed areas. The staff wear protective clothing. The incineration process is followed by the collection and removal of ash, and the cooling and cleaning of fumes. Wastes are introduced into the furnace through a hermetically sealed closure whose inner door is made of a refractory material or is provided with water cooling to prevent the overheating of materials. While a batch system is much simpler, a continuous screw feeder ensures even operation, especially at low capacity. The technique of incineration depends on the thickness of the waste layers and on the system of air supply. Air may be supplied from below through the grate; pre-heated air is introduced into the top of the furnace. The air passes through the layer of waste and is exhausted together with flue gases from the space below the grate. The supply of air from above and a thicker layer of waste facilitate the

94

incineration of volatile substances while the non-incinerated fraction increases. The supply of air from below reduces the non-incinerated fraction but it causes entrainment of ash by flue gases. Incinerator furnaces are mostly built from refractory bricks lined from outside with a steel jacket and their design is thus identical with that of incinerator furnaces for industrial wastes. The heat required for incineration is supplied by gas or oil burners. Low-capacity furnaces with high incineration temperatures are electrically heated. Furnaces are heated to the operating temperature at the start of opxation. During incineration most energy is obtained by waste incineration and the burners are thus only auxiliary. The optimum operating temperature is around 900 "C,the actual furnace temperature ranging between 700 "C and 1100 "C. Somztimes, additional incineration at 1100 "C in separate furnaces is used to reduce the amount of non-incinerated particles, soot and fume. Ash is arrested in a refractory material chamber and after cooling it is removed to drums using manual or mechanical raking devices. In some facilities, ash is washed out with water. Dust is eliminated and the resulting liquid wastes should further be processed. The outlet temperature of flue gases is 1100 "C. The gases still contain a certain amount of radioactive fractions which are mostly bound to solid particles. Prior to discharge, they should therefore be cooled and treated. Three methods of cooling are employed, viz.: - adiabatic cooling by spraying with a countercurrent of a cold water, - gas dilution with the ambient air, and - gas passage through a water-cooled tube heat exchanger. Water cooling is most efficient, removing free acidity but generating liquid wastes. Dry cooling methods are less efficient. The capacity of filter equipment should be increased for dilution with ambient air; tube heat sinks are exposed to condensation and increased corrosion. Cooled gases are either dry or wet cleaned. For wtt cleaning, diffusion scrubbers, cyclone scrubbers and spray columns filled with ceramic spheres of Rasching rings are used. Gas scrubbing always requires auxiliary equipment for water neutralization and treatment prior to recycling. The scrubbed gas:s may have to undergo tertiary treatment. Water is first removed and the gas is heated to the dew point. Electrostatic filters or absolute filters are then used. In dry cleaning, gases are collected in bag filters from heat- and corrosionresistant materials. Filters are cleaned pneumatically or using mechanical shakers. Electrostatic filters with continuous particle removal may also be used. Tertiary treatment is identical with that used in wet methods. Dry gas cleaning is more advantageous. At higher temperatures, corrosion is lower and no liquid wastes result. The disadvantages include the poor quality of currently manufactured high-temperature filters and the fire risk resulting from soot accumulation. The wide range of incinerators currently in operation include small standard incinerator furnaces without gas cleaning often used for incinerating small amounts

95

of low-contamination wastes, mainly for biologieal wastes. The complexity of equipment increases with the amount and activity of wastes. Various types of incineration equipment have been described in detail in the literature (17). Table 14 gives a survey of well known incineration plants.

r

Table 14: Radioactive Waste Incineration Plants

I Country

India USA GFR France Great Britain

USSR

Facility

Trombay Rocky Flats Marcoule Cadarache Fonteneay-aux-Roses Harwell

Capacity (kg/h) 25 25-35 10 80 30 50 27 5 25

Notes

I

Wood Biological wastes Solid and liquid wastes Solid and liquid wastes

3.4.4 Other Methods of solid Wastes Treatment Special techniques are applied for small-scale treatment of some solid wastes. The most frequent technique is the melting of lead with suitable admixtures selected according to the nuclides present (sodium carbonate, borates or silicates). Lead melting proceeds at 450 "C. Radionuclides are bound to the slag while lead losses are around 1 %. Very simple equipment suffices for the process, but the operators should be protected from toxic lead fumes. The same technique could be also used for treating iron and its alloys. Temperatures of 1500- 1600 "C are required for this application and the melting equipment is more complex. Wet incineration in sulphuric acid in the presence of a catalyst has been developed for biological waste disposal. The incineration proceeds in a quartz vessel at 250 "C. A 60-fold volume reduction has been achieved. The simplest way of treating small amounts of biological radioactive wastes is preservation with formaldehyde. Pre-treatment of biological wastes involves drying in a stream of hot air (mummification), which results in a weight reduction of 75 %, and a volume reduction of 25 % has been used for increasing the efficiency of preservation.

96

3.4.5 Processing Solid Wastes from Nuclear Power Plants Fig. 8 shows a flow chart of the processing of solid radioactive wastes from nuclear power plants. Following temporary storage in decay tanks, the used ion exchangers are removed into suitable containers (drums) and are disposed of. The introduction of a suitable solidification process is envisaged; however, cement is not suitable for this purpose and the application of bitumen has not yet been sufficiently tested.

7-

saturated sorbents

Decay fanks

Non - compactable

I I Solrdificafron adrnufures

wasfes

B

Cornpacfing

I I I

I-.

Sludge, concentmks charcoal filters

fWiPhou!

Pvcp

.

, *

Disposal Fixation

I

Incineration o m ~ 1 6 fble w sfes

.

Flue gases

-----

Fig. 8. Flow chart of processing radioactive solid wastes from nuclear power plants

Sludge and evaporator concentrates are solidified with cement or bitumen before disposal. Most solid compactable wastes are compacted. In large power plants compacting is replaced with incineration. Non-combustible wastes are crushed or shred and put into drums. Free space is sometimes filled with cement or bitumen.

3.4.6

Final Waste Treatment

The shape of the container varies with the activity of the wastes and the conditions of storage or disposal. For higher activities the container walls also serve as a shield. Reinforced concrete is therefore the most frequently used material. Usually, metal or plastic drums 0.05 -0.25 m3 in capacity are used. Metal drums are lined with asphalt and provided with an anti-corrosion coating.

97

4 Processing Wastes from Spent Fuel Reprocessing The rapid development of nuclear power in recent years and the prospects of the further construction of nuclear power plants has led to the re-assessment in most advanced countries of their initial approach to radioactive waste disposal. A 1000 MW(e) nuclear power plant requires approximately 30 t of nuclear fuel for 1 year's operation. Fuel with a burn-up of 2.6 TJ/kg (30 000 MWd/t) results in the production of about 1 t of fission products by a light-water reactor. The fission products contain a number of high-activity nuclides, some of which are relatively long-lived (see Section 2.3). The estimated amount of fission products accumulated presents a grave social hazard for the current decade, one which is qualitatively different from the problems posed by low- and intermediate-level wastes. The amount and biotoxicity of some fission product isotopes (137Cs, "Sr, 147Pm, lS1Sm) and the long half-lives of some transuranium emitters make specialists search for unusual and radical solutions in order to avoid passing on to future generations the disastrous heritage of a disturbed biological equilibrium. Of possible solutions, the following have recently been seriously studied: a) the separation and application of radioactive and stable fission product isotopes in industry and science; b) the conversion of long-lived radionuclides into non-active nuclides or nuclides with shorter half-lives; c) transport and disposal of waste in extra-terrestrial space; d) burial of more or less treated high-level wastes. From the point of view of economy, the first of the named solutions is most attractive because it changes unpleasant waste into a commercial raw material. Basically, high-level wastes may be utilized in their initial form or in a slightly modified form as intensive beta and gamma emitters; after partial or complete separation of the individual elements the most important elements may be used for specific purposes. The methods of fission product separation and their possible applications are discussed in detail in Chapter 9. The classification of radioactive waste fractions according to the type of emitters and half-life is interesting with regard to burial. The method allows

98

a differential approach to be taken to the problem of wastes and makes it possible to process the individual fractions as economically as possible, mainly because the overall amount of wastes for long-term controlled burial (transuranium elements) and for relatively shorter periods of time, as is the case with wastes containing longer lived fission products (I3’Cs, 90Sr, etc.), is significantly reduced. Classifying radionuclides by the type of radiation released is advantageous for the choice of a suitable waste treatment technology, bearing in mind the different effzcts of radiation damage by the final products. The chemical classification of radionuclides may also be suitably used for the most radical radioactive waste disposal, i.e., for the transmutation of radionuclides to stable or short-lived nuclides by irradiation with neutrons or other particles (e.g., transuranium element recycling in the fuel cycle). A complex method of the disposal of radioactive substances by nuclear reactions or in thermonuclear reastion facilities is being considered. This also applies to another radical solution of the problem of radioactive waste disposal, viz., the transport of wastes outside the Earth’s gravity field. The main constraint to the solution, be in a missile-load of wastes launched towards the Sun or to its own orbit nat intersecting the Earth’s orbit, is the high risk of a possible accident during the launch or travel of the missile within the Earth’s field of gravity. In terms of cost, this solution seems to be feasible in comparison with the cost of the fuel cycle. The most closely studied and so far the only practically implemented variant is the burial of non-processed or processed wastes in a selected disposal site. Fig. 9 shows the flow chart of the possible treatment of radioactive wastes from a reprocessing plant. Branch A represents the possibility of separating some waste fractions (transuranium elements and some fission products) with a view to their practical application or separate processing and storage. Branch B shows the methods of converting the wastes to a solid product suitable for long-term burial, and branch C shows the variant that has SO far been most frequently used in practice, viz., conditional burial of liquid wastes, either non-treated or pre-treated in tanks. Interim or permanent burial of the various forms of the product is conducted according to principles listed in Chapter 10. Interim burial is used in two cases: 1. The plant producing high-level wastes has not found more suitable methods of waste treatment and envisages future final treatment, 2. Interim burial is considered as being an intermediate stage in the technology of solidifying wastes or valuable nuclides. An alternative waste burial form is being considered which would guarantee sufficient safety even in long-term controlled storage and would allow processing and the application of some fractions in the future. Depending on the degree of processing, the following principal solutions are possible:

a) burial in the liquid form, i.e., without or with conditioning (adjustment of pH, salinity, etc.), b) evaporation of part or the entire water content, c) calcination, and d) sintering associated with the transformation of wastes to a crystalline or glassy material. Seemingly the simplest way of storing liquid- high-level wastes in controlled tanks turns out in the next stage of nuclear power development to be too costly, impractical and hazardous. In view of the growing number of nuclear power facilities, the capacity of tanks would soon increase to an unacceptable level. Fig. 9. Flow chart of processing radioactive wastes from spent fuel reprocessing

L,[=7

L

Calcination

L Y

I

I

I

I

1

burial

Possible accidents associated with the release of radioactive solutions into the biosphere preclude the use of this alternative, especially in densely populated areas. In spite of this, this simplest method of high-level waste disposal is most widespread and much experience has been gained with tank design and the method of radiation safety control. Wastes thus stored are possible intermediate products for further processing. We shall therefore discuss some aspects of tank storage at this point. Chemical pre-treatment and corrosion products from tank structural materials, fuel and

100

technological equipment may significantly affect the overall waste composition and thus also the choice of a suitable treatment technology. The tanks for high-level waste storage (self-boiling solutions) are typically of a reinforced concrete design, with a cylindrical shape (18). The total capacity of the tanks is several thousand cubic meters with a cylinder diameter of several tens of meters and a depth of around 10 m. They are plated with steel or stainless steel to prevent leakage and are provided with equipment for filling and discharging, mixing, cooling, ventilation, etc. Sometimes, the tanks are designed as free- standing steel tanks in a concrete protective barrier. The most important aspects affecting the fail-safe operation of the tanks include: a) long-term absolute tightness of the tank, b) outlet of volatile substances, aerosols and radiolytic hydrogen and their disposal, c) perfect cooling. Tank tightness is conditional on the choice of a metal material suitable for the given solution composition and on the quality of welded joints. Additional protection is safeguarded by the surrounding concrete barrier and a suitable storage site selection, considering the permeability and sorption properties of soils in the area (the factor of radionuclide migration in the soil). The tightness of the system is monitored by precision measurement of the solution level in the tanks. L eve1 yauje

Concrete Steel

Fig. 10. Design of tank for self-boiling waste storage

For reasons of economy, strongly acid solutions are sometimes neutralized. This treatment then facilitates the use of cheaper grades of steel as tank structural materials. For perfect cooling, the waste solution is agitated, which also ensures the mixing of liquid wastes containing insoluble substances in the form of suspensions or sediments, and is cooled using a system of tubes with circulating cooling liquid. Before the waste tanks are filled with self-boiling wastes, they should be warmed,

101

thus preventing stress which may occur in the reinforced concrete walls. Vapours and aerosols entrained by them are discharged from the space above the surface via condensers and glass-wool filters into the atmosphere. Fig. 10 shows an example of the design of a self-boiling waste tank. A method of solidification in the tank has been developed that is aimed at reducing the original volume of radioactive wastes (19). At Hanford (USA), where long-term experience has been gained with the storage of large amounts of high-level wastes, the process was carried out by evaporating most water from liquid wastes in a tank. The solution was heated to a suitable temperature with air at about 800 "C being forced into the bottom part of the cylindrical tank. The air was pre-heated in an electrically heated heat exchanger and its temperature at the tank inlet was about 400-500 "C.After passing through the liquid the air was transported via a system of condenser and filters to the atmosphere. In the process the solution was heated to around 1000 "C, and water was evaporated until a salt cake of crystalline waste substances formed. After the evaporation of the main proportion of water, the cake was allowed to cool. This process was slowed down by decay heat. After cooling to the steady-state temperature a solid cake and a small amount of moisture remained in the middle of the tank. Compared with liquid wastes, waste in this form offers an advantageous volume reduction and moreover requires minimum monitoring. Operating costs are lower and the storage method is used to advantage for the disposal of wastes in which significant heating due to radioactive decay does not take place. Such wastes were typical of former reprocessing technologies used mainly in the USA in the early period of nuclear power development. Modern extraction methods used for spent fuel reprocessing generate wastes of high specific activity, causing the self-heating of stored liquid waste, in which sludge sediments on the bottom cause uncontrolled steam release in the form of small explosions. The wastes should be stored in liquid form in forced ventilation tanks.

4.1 High-Level Waste Disposal in Deep Geological Formations One of the possible methods of high-level waste disposal is burial in relatively deep layers of selected geological formations. The problems of siting and the methods of burial will be discussed in Chapter 9 in more detail. In this section we shall discuss the process which precedes the injection of the waste into the soil, or which is part of the waste conversion to a form suitable for permanent burial. The injection process is curently being studied very intensively in the USSR. In addition to site selection conditions, other conditions should be considered associated with liquid waste processing. The waste solution should meet certain

102

qualitative criteria to preclude pore clogging during the filling of underground spaces, and to avoid the formation of channels which could route the leakage of radioactive substances in an undesirable direction. All suspended substances should therefore be filtered from the solution. pH adjustment to a neutral or alkaline state is recommended. This will reduce the possibility of the solution entering into chemical reactions with the soil. Solutions with lower salinity and specific activity are more suitable for injection. Such solutions may be obtained by deliberately mixing high-level wastes from the first cycle with solutions from other technological streams of spent fuel reprocessing. More complex chemical processing is sometimes recommended, such as coagulation with the compounds of iron, chromium, calcium, etc., with subsequent clarification in sedimentation tanks and filtration. Secondary wastes, however, are formed, such as settled precipitates and precipitates from filters. The fundamental shortcoming of this solution is the risk of uncontrolled leakage of radioactive substances during permanent burial. An interesting variant of high-level waste burial is so-called on-site incorporation (or on-site melting). It consists in injecting a certain amount of liquid wastes into a cavity in a silicate massif where water evaporates as a result of heat liberated by radioactive substances. The process may be controlled by the supplied rate of high-, intermediate- or low-level wastes or of cooling water. The recycling of condensed water from the first filling stage may be used for this purpose. As soon as the water or solution supply is stopped, the temperature increases, which melts the waste matter proper and later the surrounding silicate rock layer. Heat dissipation by convection and a gradual activity drop result in cooling and in the solidification of all matter. During melting, the molten radioactive matter cannot leak outside the area of the process because any smaller amount of melt will immediately solidify in a cooler medium. On the other hand the extraction of the melt in the liquid state is precluded by the high temperature of the medium which immediately causes water evaporation. In time, the process of burial may be classified into three stages. The first stage is that of filling the cavity. Using the above-mentioned control and cooling, this stage may be extended to equal the estimated time of operation of the reprocessing plant, i.e., to a period of 25 years. The melting of the waste proper and of surrounding rock is the second stage. According to preliminary estimates, this stage may last several decades (up to 100 years). The third stage is the cooling and solidification of matter, which again may last for decades. With respect to economy and safety, and with respect to environmental control, this method is more suitable than liquid waste injection or tank storage. As long as a suitable geological formation can be found in the vicinity of a reprocessing plant, the method will probably be preferred to the methods discussed above. The disadvantage of the process is a certain loss of control over the buried radioactive .wastes and the impossibility of reprocessing and re-using the buried material.

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4.2 Processing Intermediate-Level Wastes from Spent Fuel Reprocessing High-level wastes are generated by the first extraction cycle as well as by reprocessing plants. The latter have relatively lower activities and contain different ballast materials. They include aqueous solutions from further extraction cycles, used and degraded organic solvents, deactivation solutions, etc. Aqueous solutions are mostly processed using routine methods developed for processing intermediate-level wastes (see Section 3.3), or are mixed with high-level wastes and are processed together. This latter technique, however, complicates high-level waste disposal and increases its cost. It is therefore more profitable to apply one of established methods of liquid waste processing and concentrate fixation. In reprocessing, radiolytic decomposition of the used extraction reagents and solvents takes place owing to the high radiation of the fission products. In the commonest process of the Purex type, it involves tributylphosphate (TBP) dissolved in aliphatic hydrocarbons, which is regenerated after several extraction runs by the removal of degradation products. When products reach a certain given concentration, regeneration becomes too costly and organic substances with a high content of radioactive admixtures should be disposed of. Such wastes amount of dozens of cubic metres per 1000 t of reprocessed spent fuel. The wastes were either incinerated or incorporated in solid organic materials, such as bitumen or polyethylene. Incineration generates a number of other unpleasant wastes. The release of radioactive aerosols is a particularly difficult problem. Solidification in organic matter increases the volume of wastes which are to be buried and which moreover, lack the fundamental properties suitable for safe burial. Several procedures have therefore been proposed that try to separate the organic matter from most radioactive substances. The Eurowat process is one of the solutions. It envisages the separation of the solvent from the complexing phase in which most fission products concentrate, and the subsequent decomposition of tributylphosphate (dealkylation) to volatile hydrocarbons (20). On the completion of the process, phosphoric acid and most radioactive substances remain. This inorganic remainder is a high-level waste, which is then processed together with the waste from the first extraction cycle. The phosphoric acid present may be used to advantage as a glass-forming additive. Tributylphosphate may also be decomposed by the action of lye (saponification). This reaction results in the formation of butanol and the respective phosphate. Radioactive substances are again concentrated in the aqueous phosphate phase. Other methods have been proposed, such as the reaction of TBP with phosphoric acid, all with the objective of separating most fission products in the form of an

104

aqueous solution which may be further processed together with rather high-level wastes. The tritium problem, which has so far not been satisfactorily resolved, deserves particular attention. Fractional distillation, which has so far been the single method of separating tritium from the other hydrogen isotopes, is a very costly process. The isolated and safe burial of tritium also presents a complex problem.

4.3 Processing Wastes Containing Alpha Emitters Some technological flows in the reprocessing chart contain a number of transuranium element radionuclides, of which many are long-lived alpha emitters (see Section 2.3). After several hundred years, the radiotoxicity of these nuclides represents the major proportion of their environmental impact. This hazard is minimized after roughly 1000 years, when solidified wastes containing transuranium elements reach a specific activity to that of uranium ores in their natural deposits. The best method for the disposal of these wastes is their separation from other high-level wastes (mainly fission products) with the subsequent chemical separation of transuranium elements into the individual elements or groups of elements. These may then be applied in practice, disposed of by reactions in a reactor, or they may be solidified and buried separately from the remainder of the radioactive wastes. Techniques are used which have been developed for the separation of transuranium elements and fission products, based on extraction or ion-exchange methods (see Chapter 8). If transuranium elements thus separated are disposed of using reactor reactions, the period during which the wastes present a risk may be reduced by up to three orders of magnitude. For solidification technology, wastes containing transuranium elements do not pose a special problem; any of the methods discussed in Chapter 8 can be used for their disposal. Particular attention should only be devoted to the different effects of alpha radiation on the fixation product material and the high demands on the resistance of the fixation medium to the leaching of incorporated transuranium elements in an aqueous medium. With regard to the estimated long period for controlled burial, a special regime of deep disposal is considered. After around 1000 years the wastes thus buried present a lower risk than uranium deposits near the Earth’s surface.

105

4.4 Radioactive Wastes from Fluoride Reprocessing The so-called dry methods of reprocessing were developed primarily for fast reactor fuel. The most promissing method is based on the fluorination of chemically treated fuel and on distillation, utilizing the different degree of volatility of uranium, plutonium and fission-product fluorides. Radioactive wastes containing most nonvolatile fission product fluorides and also possibly the fluidized-bed packing, and volatile fission products arrested using sorption columns, are treated separately. In addition to simple storage of loose wastes in metal containers which are intensively cooled, methods similar to those developed for liquid waste fixation may be used for the disposal of these wastes. It is again favourable to use silicate or phosphate systems forming fluorosilicates and fluorophosphates with waste compounds at high temperatures. The technological advantage of processing the said wastes is that the water need not be removed and the calcination stage is thus eliminated. The disadvantage of the method is the relatively high specific activity, which may reach the order of thousands of PBq/m3 (cca lo4 Ci/l). Technologists should therefore apply processes in which significant dilution of the original activity is achieved with inert matter of the fixation medium. For these reasons, a method of incorporating powder wastes from the fluorinator into prepared sintered glass having certain optimum properties has been suggested. The molten glass matter surrounds the waste particles, in this particular instance the partly fluorinated alumina with fission products, thus forming a compact monolith which does not differ in its properties from the original glass. Another method, developed in the USSR, utilizes the original composition of wastes and, using suitable additives, converts them into a material whose composition is similar to that of cryolite (sodium fluoroaluminate). The disadvantage of this products is the relatively high leaching rate of cesium ions. Sorption column packings (pellets of alkali metal or rare earth fluorides) form complexes with fission-product volatile fluorides. The complexes are stable only up to about 550 "C. As fission products with a high specific activity are involved, with a high specific thermal output, only columns of small size should be used. They may be placed directly into special containers sufficiently cooled with an air current in storage areas. To improve long-term safety, it is advisable even in this instance to convert, the sorption column matter into a fixation product that satisfies the criteria discussed in Chapter 8. The appropriate method of processing these wastes is to incorporate them into a suitable fixation medium, similar to fluorinator wastes. In view of the relatively long-term prospects for the industrial use of fast reactors and the fluoride processing of spent fuel elements, no large pilot plant for processing these wastes exists.

106

5 Processing Gaseous Radioactive Wastes Almost all facilities handling radioactive substances are equipped with ventilation and air purification systems whose fundamental objective is to prevent workplace and environmental air contamination and to protect the personnel and population from internal contamination and external irradiation. The technical aspect of air purification have currently been considered in great detail, mainly with respect to aerosols. It is considerably more difficult to remove gaseous radionuclides efficiently from the contaminated atmosphere owing to the very low volume activity of the radionuclides and, with radioiodine, to the existence of some of its compounds whose presence in the contaminated atmosphere under certain circumstances reduces the overall efficiency of its removal. In spite of this, the current technological standard of air purification methods in nuclear facilities is generally able to safeguard the maintenance of the required level of radiation protection of the staff and population.

5.1 Fundametals of Gaseous Radioactive Waste Processing In the processing of gaseous radioactive wastes, a number of parameters are taken into consideration, such as half-life, isotope ratio and activity, the total volume and the presence of other non-active fractions, and the criteria and standards of radiation protection of the population. As shown in Chapter 2, fuel reprocessing and nuclear power plants are the main sources of radioactive wastes in the fuel cycle. Gaseous radioactive wastes from nuclear power plant operation are mainly characterized by the presence of long-lived radionuclides, such as the radioactive isotopes of rare gases, 83mKr, 85Kr, "Kr, l3lrnXe,'33Xe, 135mXeand I3'Xe, in light-water reactors and 4 ' Ar in gas-cooled reactors. The principle of delay and decay is applied in their processing. A suitable design of part of the nuclear power plant ventilation system, i.e., the delay line and the

107

choice of suitable operating conditions of the line, may result in such a time delay in the discharge of the short-lived radionuclides Kr and Xe that in the continuous passage of gaseous wastes through the line the activity is practically eliminated from the emissions. Figures 11 and 12 are block diagrams showing alternative delay line designs for boiling-water and pressurized-water nuclear power plants. Turbine

a ) condenser

atmosphere

elector

Turbine

c) cqndemer vector

Atmosphere

L-

-

Rare pses storage

Fig. 11. Flow chart of technology for purification of boiling water reactor waste gases

Uncondensable gases with short-lived gaseous radionuclides exhausted from the steam turbine condenser in boiling-water reactors and from the degasifier of the primary circuit and steam generators in light-water reactors are transported through a delay time consisting, in older power units, of a long pipeline or a decay gas tank (variant (a)), and in newer units of an activated charcoal adsorption bed or of a low-temperature distillation or absorption unit for the separation of gaseous radionuclides from the wastes. In view of the presence of aqueous radiolytic products in waste gases, a catalytic recombiner of the oxygen-hydrogen explosive mixture and a drying line for the removal of water vapours are inserted upstream of the delay line. Prior to stack discharge from the nuclear power plant, gaseous wastes undergo tertiary treatment using a high-efficiency aerosol filter (the high-efficiency particulate and aerosol HEPA filter). The principle and operation of delay line adsorption beds are discussed in detail in Chapter 5. Compared with primary circuit gaseous wastes, gaseous radioactive wastes from a ventilated reactor hall or from a reactor containment and auxiliary facilities in normal power plant operation typically have large volumes and gaseous radionuclides of low volume activity are present. Following passage through an iodine and an aerosol filter they may be discharged into the atmosphere without further processing.

108

High specific activities and long half-lives typical of radionuclides from waste reprocessing render unacceptable the concept of radioactive waste processing based on the delay and decay principle. The current techniques for processing gaseous wastes from this stage of the fuel cycle of nuclear power plants are therefore based on the concentrate and contain principle using any of the separation and concentration processes discussed in Section 5.2. The choice of a suitable process Atmosphere

Stack

HEPA aerosol

filter

LRare pe.s

- slamye

Fig. 12. Flow chart of technology for purification of pressurized water reactor waste gases

for the separation of gaseous radionuclides from waste gases generated by fuel reprocessing and their concentration depends on the type of fuel element, fuel composition, the procedure used and the conditions of reprocessing, the presence of other non-active substances in waste gases, on their dilution, etc. Section 2.3 indicated that gaseous radionuclides with the exception of tritium are released practically quantitatively into the gas phase in the initial stages of reprocessing, i.e., during fuel rod decladding and fuel dissolution. Table 15 shows the relationship between the amount of gaseous radionuclides released during these operations and the amount of radionuclides transformed into the liquid phase. In view of the presence in gaseous wastes of a number of admixtures formed during the initial stages of reprocessing (mainly nitrogen oxides), the separation proper of gaseous radionuclides is proceded by the removal of all fractions which could affect it unfavourably. In continuous operation of the delay line, gaseous radionuclides are considerably concentrated in waste gases. In order to prevent undesirable dilution with air, gaseous radionuclides are released from the solution of dissolved fuel and acid by blowing with helium which, after the removal of nitrogen oxides by reduction and

109

Table 15: Relative Amounts of Gaseous Radionuclides Released in the Individual Stage of Reprocessing -~

Amount of gaseous radionuclides (%) released into gaseous phase in

Type of spent fuel

Light-water reactor

Carbide-hightemperature reactor

Method of reproces- Radionuclide fuel element sing cutting Purex

Thorex

iodine rare gases tritium iodine rare gases tritium -~

Oxide-hightemperature reactor

Thorex

Fast

Purex

iodine rare gases tritium iodine rare gases tritium*

fuel crushing

matrix incineration

fuel element dissolution

1-5

95 95-99

1-10

1-2

1

-

10 50-90

5

1 1 5

30-50 90-95 97

40-60

Amount in aqueous phase (%)

5

0 88-98

10

10 0

1

2

3-5 90

75-85 90

10 0

2

2 85 10-50 1

70-85 5 0 5-1 0

* Approx. 90 % of tritium is released in the reactor through the fuel element can. drying, is transported through the gaseous radionuclide separation unit and returned to fuel dissolution. Waste gases from intermittent fuel dissolution diluted with air require a different technique of processing. The fundamental stages of gaseous waste processing include: - the removal of nitrogen and oxygen oxides from air by reaction with hydrogen resulting in the formation of elemental nitrogen and water; - waste gas drying on molecular sieves; - gaseous radionuclide separation. Table 15 shows that in some instances considerable amounts of gaseous radionuclides are released from the decladding of spent fuel rods. This applies to the release of gaseous radionuclides from the incineration of the fuel graphite matrix in the reprocessing of carbide fuel from high-temperature reactors or from the decladding and cutting of fast reactor spent fuel rods. These operations are a source of heavy contamination of the waste gases with solid aerosols. Aerosol separators and filters are therefore used in the waste gas processing line. Waste gases from this process have a high C 0 2 content and their low-temperature absorption in

110

liquid carbon dioxide is used in the separation and concentration of gaseous fission products from gaseous wastes. These techniques of processing gaseous wastes from spent fuel reprocessing use a cooling time which prevents any significant contamination of gaseous wastes with radioiodine. If contamination cannot be prevented, a unit for the removal of radioiodine from gaseous wastes is included in the processing line forward to gaseous fission product separation. The unit mostly consists of a scrubber and an iodine filter in combination. It should be added that the recuperation of nitrous gases alone, which results from the continuous dissolution of spent fuel in a scrubber sprayed with nitric acid, represents the first stage of gaseous waste decontamination from iodine vapours.

5.2 Separation Methods for Processing Radioactive Gaseous Wastes from Nuclear Facilities 5.2.1 Gaseous fission products In view of the inert nature of rare gases, the methods for separating krypton and xenon radionuclides from nuclear facility gaseous wastes are based on physical processes. They include adsorption on solids at normal or low temperatures, absorption in liquids, and low-temperature distillation and diffusion in permselective membranes. The capability of rare gases to form clathrates and, in certain conditions, also chemical compounds does not have significant importance for their separation. The half-life of the given radionuclide and its fission yield should be taken into consideration when methods are selected for the practical removal of gaseous fission products from nuclear facility wastes. It has already been shown in Chapter 4 that the retention of gaseous fission products with a relatively short half-life, such as krypton isotopes with mass numbers 83m, 85m, 87 and 88 and xenon isotopes 133, 135m and 135 on solid absorbents in the so-called continuous delay lines may be used for their removal. I n the passage of a gas mixture containing, in the case of contaminated air, mainly nitrogen and oxygen and rare gas radionuclides through the delay line adsorption bed, the chromatographic separation of the individual fractions of the mixture takes place. As a result of the stronger bond of rare gases during their physical adsorption on the surface of adsorbents in comparison with the other fractions of the gaseous mixture, the transport of krypton and xenon isotopes through theline is

111

delayed relative to the other fractions of contaminated air. The time delay, i.e. the retention time of krypton or xenon in the delay line, may be expressed for very low volume activities of the sorbed gaseous fission products as

m f=KF where m is the amount of the adsorbent in the delay line bed, F is the volume $ow-rate of the contaminated gas through the line and K is the so-called dynamic adsorption coefJicient.

Kr 0Sufdiffe+eakinen

203C(Great, &&in, vpe not stated (GFR) e Desorex J8-2bchdovakia) e&perswbon lxr-f(crrchan/oha, OSKT-28 USSR) OJKT- 6A fUSJR) A TSURUUI GLS (Japan)

R4 (GJk') WTK 14 (SOR)

I

-uu

0

I

80

1

760

I 2*0

Fig. 13. Dynamic absorption coefficients of Kr for different activated charcoals

As shown by Equation (5.1), at the given temperature the dynamic adsorption coefficient represents an unambiguous param:ter of considerable practical importance which may be used for evaluating the retention and thus also decontamination properties of different delay adsorption packings. The relationship between the decontamination factor for gaseous radioactive wastes travelling through the adsorption bed of the delay line and the dynamic adsorption coefficient is evident from the relationship for the decontamination factor when the delay and decay principle is used in radioactive waste processing:

112

2 t e" = 0.693 - = 2 T T1/2 which when combined with Equation (5. I), gives

D

=

where A and are the decay constant and the halJllife, respectively, of the given isotope of gaseous fission products.

Xe

o SuLcliffe Speakmen 203CfGreai Britain a hpe not sfafed (GFR) 8 Dworex

08-2 (Czechoslovakia)

0 Supersorbon HS-I (Czechosiovakiu, 0 SKT -ZB (USSR)

QSKT- 6A (USSR)

Fig. 14. Dynamic absorption coefficients of Xe for different activated charcoals

It follows from Equation (5.3) that the degree of contamination of gaseous wastes of given volume and isotope ratio in the delay line at the given temperature depends primarily on the dynamic adsorption coefficient of the gaseous fission product for a certain type of adsorbent, and on the amount the adsorbent in the delay line. Figs. 13 and 14 show the temperature dependence of the dynamic adsorption coefficients of krypton and xenon for different kinds of activated charcoal of

113

Czechoslovak and other makes such as are used for the adsorption packing of delay lines. The figures show the significant effect of the temperature at which the delay line operates, not only on the delay time of gaseous fission products but also on the separation of krypton from xenon. In addition to temperature, the composition of the processed gaseous wastes also affects the operation of the delay line with the adsorption bed. While the presence of other inert fractions in waste gases, such as H,, N,, 0, and Ar, does not significantly affect the retention of Ar and Xe in the delay line, water vapour in the gas to be purified or on the surface of activated charcoal considerably reduces the degree of decontamination of gaseous wastes during their passage through the delay line. For example, the time delay in the passage of gaseous fission products through a n activated charcoal bed containing about 5 - 10 % of water decreases by about 25 - 50 %. The presence of C 0 2 in gaseous wastes has a similar effect on the function of the delay line, although to a lesser extent. The optimum design of a delay line always represents a compromise between the amount of the adsorption packing (line size) and operating temperature relative to the desired degree of decontamination of gaseous wastes, i.e., a compromise between capital and running costs. The Grundremmingen nuclear power plant (GFR) may serve as an example. In the plant, cca. 80 m3/h of gaseous wastes are processed by a delay line packed with several tons of activated charcoal and operating at a temperature of - 10 "C.The time delay of gaseous fission products in the equipment reaches 17 -24 h for krypton and 9 - 15 days for xenon (21). Activated charcoal-packed delay lines are simple facilities widely used for processing radioactive gaseous wastes from nuclear power plants. Their main disadvantage is that their operation involves the risk of fire as a result of selfignition of activated charcoal, especially in the presence of oxidizing agents in the waste gas. Charcoal ignition may be induced by an increased ambient temperature and by the energy accumulated in the adsorption bed during the decay of adsorbed gaseous fission products. Delay lines operate on the basis of the physical adsorption of gases on activated charcoal, and physical adsorption is also the basis of one of the methods for separating long-lived "Kr from gaseous wastes, i. e., low-temperature adsorption on solid adsorbents. Low temperatures approaching that of liquid nitrogen, at which the adsorption beds for "Kr removal operate, are the prerequisite for a high separation efficiency and for the attainment of a high decontamination factor, which is one of the constraints of the safe and economical disposal of the long-lived radionuclide. The principle, adsorbent and function of low-temperature adsorption beds are similar to those for processing gaseous fission products at normal temperature. The only difference is that the low-temperature beds operate intermittently in parallel pairs. The reliable functioning of the equipment at low temperatures requires much more thorough pre-treatment of waste gases and the removal of

114

$4-

I

Krypfon

frap

J r p j bed

Condenser

1

‘I.

I/

high-boiling admixtures which may condense in the bed, thus reducing or entirely preventing passage. In addition, their accumulation in the bed leads to the hazard of ignition of activated charcoal. The bed design should be such that it will withstand a sudden pressure rise should the supply of the cooling medium be discontinued. Fig. 15 shows a block diagram of a low-temperature adsorption line for the removal of 85Kr from gaseous wastes from reprocessing. The waste gas containing N,, NO,, N,O, H,O, H,, Kr and possibly Xe in short-term cooled fuel is transported to an alkaline scrubber sprayed with KOH in which NO, is removed. In a converter packed with a rhodium catalyst N 2 0 and 0, are removed by reaction with hydrogen at 550-650 "C in the presence of N,, NH, and H,O. Water vapour and NH, are removed from the condenser waste gases by a parallel pair of beds packed with silica gel, molecular mesh or activated alumina. The lowtemperature separation proper of 85Kr proceeds in a pair of parallel adsorption beds packed with activated charcoal and cooled with liquid nitrogenat cca. - 180 "C. When the capacity of one bed is exhausted the gas is passed to the second column. In the meantime, the first column is regenerated by heating and the released krypton (or xenon) is arrested by a trap at the end of the line at -210 "C. After a few sorption-desorption runs, krypton in a highly concentrated form, having a specific activity higher than 740 TBq/kg (20 Ci/g) is pumped into pressure cylinders for final burial. Decontaminated

L tauid

N2

Fig. 16. Schematic diagram of waste gas purification by low-temperature distillation

Although the separation of 85Kror '33Xe from gaseous wastes by low-temperature adsorption on activated charcoal shows a high concentration factor and separation efficiency, the wide practical application of the method is constrained by high running costs due to the excessively high requirements for liquid nitrogen for cooling the massive adsorption beds. Fail-safe line operation is also significantly dependent on the efficient removal of all admixtures from waste gases.

116

Low-temperature distillation and rectification is another process for separating 85Kr from radioactive gaseous wastes and its concentration. The principle of this separation method is identical with the now classical fractional distillation of liquid air according to Linde. Waste gases from which nitrogen oxides, CO, and H,O have been removed using the same techniques as in the low-temperature adsorption are pre-cooled to temperatures between - 160 and - 180 "C and proceed to a rectifying column sprayed with liquid nitrogen (see Fig. 16). Krypton or xenon is separated from waste gases on rectifying column plates operating at - 187 "C and 4 - 6 x lo5 Pa (4-6 atm) and accumulated as a high-boiling fraction on the bottom of the column, while non-liquified gas containing mainly nitrogen is discharged as waste through the top of the column. If the waste gases contain xenon in addition to krypton, the concentrate on the bottom of the column is further rectified with the objective of obtaining separate krypton and xenon fractions. Table 16 shows the composition of the final products in processing gaseous wastes by low-temperature rectification. Table 16: Waste Gas Cornposition in Gaseous Fission Product Separation by Low-temperature Distillation

Fraction

Waste gas wlw)

a*

Krypton fraction

(%, wlw)

Xenon fraction PA, wlw) ~~~

0.004 0.03 1.oo 1.oo 1.o 1.o 1.o 17.5 74.8 4.6 2.1

78.8-85.7 0.14-0.4

-

0.2 0.2

-

1.4-6.9 0.05-10.6 1.o

~

0.14-0.4 90.7-93.6

-

0.2 0.2

-

0.3-0.6 3.7-5.2 0.05

The major benefit of this technique for separating gaseous fission products from waste gases is its high technological level. Facilities for air liquefaction and fractional distillation which have now been in operation for several decades have yielded much experience o n design, construction and operation, allowing the practical application of the process to the processing of radioactive gaseous wastes. These factors are probably the reasons why low-temperature rectification has so far been the only one of the described separation processes to be used on a wider scale for processing waste gases from spent fuel reprocessing plants. Certain disadvantages of the process include the radiolysis and concentration of ozone

117

and acetylene in the rectifying column heater and the associated explosion hazard. The separation of rare gases from radioactive gaseous wastes may also be achieved using the different solubilities of the individual gaseous fractions of the wastes in organic solvents. The first studies on this subject, in 1958, showed that some organic solvents, such as tetrachloromethane, dichlorodifluoromethane (Freon-12), trichlorofluoromethane (Freon-1 I), the kerosene fraction from oil distillation and some other substances, such as liquid CO, and N,O, showed remarkable selectivity in dissolving krypton and xenon over nitrogen, oxygen and argon and other inert gases. At lower temperatures, the steady-state amount of rare gases in the solvent is about 10-100 times higher than the steady-state amount of nitrogen, oxygen or argon. The high selectivity and solubility of gaseous fission products are not the only criteria for the choice of a solvent for practical application. The other criteria include the high boiling point and associated low vapour pressure of the solvent in delay line operation, non-combustibility or at least a high flash point, sufficient thermal and radiation stability, low viscosity and low cost. Allnosphere

Decontaminated Concenfrafedgaseous ~frssrbnproducts

'r

r"

I

I I

I I I

1

I

1 I Fig. 17. Schematic diagram of a line for the separation of krypton and xenon from gaseous wastes by absorption in organic solvents

As shown in Fig. 17, the separation of gaseous fission products from waste gases proceeds in three stages. In the absorption packing column sprayed with a solvent rare gases are separated from the gaseous waste, which is then discharged into the atmosphere. In addition to gaseous fission products, small amounts of other gaseous waste fractions, mainly nitrogen and oxygen, also pass into the solvent in the absorption stage and the solvent is thus transported from the absorption column to a partition column where co-absorbed gases are removed. The gaseous phase from the partition column still contains a certain amount of gaseous fission products and

118

is therefore returned to the absorption column while the solvent is passed to a displacement column where concentrated gaseous fission products are released and the solvent regenerated. With regard to the above-mentioned criteria for the choice of solvents, the fluorinated solvents Freon-I 1 and Freon-I 2 have found the widest application in absorption lines, on the pilot-plant scale. The absorption line operated at O R N L using these solvents has the following specifications: Amount of processed waste gas: 16.1 -37.8 m3 (STP)/h Amount of solvent: 0. I7 -0.284 m3/h Absorption column: temperature -60.6 to -29.5 "C pressure 1.1 -3 MPa ( I 1-30 atm) Partition column: temperature 0 to - 2 "C pressure 0.3 MPa (3 atm) Displaced column: temperature -7 to - 5 "C pressure 163 kPa ( I .63 atm) Inlet krypton concentration: 42 - 8800 ppm. Under these conditions the decontamination factor is in the range 400- 1000, depending on the solvent used (Freon-I2 allows a higher efficiency of waste decontamination) and the temperature and pressure conditions of the facility's operation. The high CO, content in the waste gas may be used to advantage for removing rare gases from the reprocessing of high-temperature reactor fuel (cca. 100 "/, when the graphite matrix is incinerated in pure oxygen cca. 20 % v/v when air is used). At normal temperature and a pressure of about 7 MPa (70 atm), the liquid carbon dioxide takes over the function of the solvent in the absorption column and is capable of removing rare gases from this type of gaseous waste, achieving a decontamination factor of about 500. The design of the absorption line using liquid CO, as a solvent and the operation of the individual stages is the same as in lines using organic solvents. The running cost of the technique based on the absorption of gases in solvents is lower than that of low-temperature processes of separating gaseous fission products from wastes. The requirements for pre-purification of waste gases are not as high as in the former methods. On the other hand, higher operating pressures pose the hazard of uncontrolled leakage of the processed gases from the facility. The removal of krypton and xenon from gaseous radioactive wastes and their separation from each other by means of permselective membranes is based on the different velocities of the passage of the gaseous mixture fractions through a permselective membrane. If the pressure (concentration) gradient of the i-th fraction of the gaseous mixture forms on the membrane, the velocity of the passage of the i-th fraction through the membrane is characterized by the relationship (5.4)

119

where A is the membrane area, d thz membrane thickness, Api the difference in the partial pressures of the i-th fraction on both sides of the membrane, and P i the coefficient of permeability of the i-th fraction through the membrane, defined as Pi

=

KiDi

(5.5)

where Ki is Henry’s constant and Di the diyusion coefficient. It is evident from equations (5.4) and (5.5) that gas separation using permselective membranes depends primarily on the solubility of gases in the membrane, determined by the value of Henry’s constant and the diffusion rate of the gas fraction in the non-porous permselective membrane material, i.e., not in the pores, as is the case with conventional porous membranes. Synthetic membranes from dimethylsilicone rubber, whose coefficient of permeability is 30-fold higher than that of other membrane materials, because of their considerable strength, small thickness and sufficiently high selectivity, best meet the requirements for the practical use of permselective membranes for removing gaseous fission products from nuclear facilities wastes. The development of lines using permselective membranes for the removal of rare gases from radioactive gaseous wastes has been taking place in the USA for a long time, on both laboratory and pilot-plant scales. Such lines consist of several tens of cascade-connected membrane units which achieve a sufficient capacity and a n adequate decontamination factor. Development results have shown that a cascade arrangement of an overall membrane area of between 4000 and 5000 m2 operating at a pressure gradient of 1 MPa (1 atm) on the membrane can process cca. 1 m3 of gaseous wastes per minute with a decontamination factor of up to lo4 at a concentration factor ranging between 10 and 500. The limitations of the method for practical application are the high capital costs (high membrane cost), considerable power consumption in operation and the long-term effects of radiation, temperature and pressure shocks on the behaviour and function of permselective membranes.

5.2.2 Iodine Radionuclides Iodine may occur in gaseous radioactive wastes in different physic0 chemical forms, viz., as elemental iodine vapour, as vapour or organic and inorganic iodine compounds (methyl iodine and higher alkyl iodides, hydrogen iodide and hypoiodous acid) and as a dispersed condensed phase consisting of solid or liquid particulates with physically or chemically bound iodine. These different forms also require different approaches for the removal of iodine from gaseous wastes. In light-water reactors, the amount of methyl iodide in gaseous wastes does not exceed 5 % of the overall elemental iodine content in the wastes, in both normal operation and in case of an accident.

120

In gas-cooled reactors, the methyl iodide content may be slightly higher (10- 15 %). The factors increasing the yield of methyl iodide formation from elemental iodine include a high concentration of organic compounds, reduction atmosphere, large metal surface areas, elevated temperature and- ionizing radiation. Gaseous radioactive wastes from spent fuel reprocessing may also contain a certain amount of methyl iodide and even higher alkyl iodides. The proportion of this particular form of iodine in the total iodine content in gaseous wastes from reprocessing remains a problem. The properties of chemical iodine and its high affinity for sodium suggest that iodine radionuclides leaking from the fuel elements of sodium-cooled fast reactors will quantitatively be chemically bound to the cooling medium in the primary coolant circuit. Fast reactor ventilation systems are therefore not provided with apparatus for removal of radioiodine from gaseous wastes. The aerosol separators and filters installed in fast reactors for the removal of sodium aerosols guarantee that no leakage of iodine radionuclides will occur into the environment, even in a major accident. In view of the chemical properties of iodine, a number of natural mechanisms contribute to its removal from the gaseous phase, such as adsorption and deposition on the surfaces of the reactor primary circuit and the containment, the pipes of the ventilation systems of nuclear facilities, capture in condensing water steam, and adsorption and absorption on solid and liquid aerosols. The action of these natural mechanisms on iodine removal from gaseous wastes has a specific character depending on the conditions in which the individual mechanisms take place. It cannot be assumed that this action will be such as to allow the abandonment of the treatment of the contaminated atmosphere before its discharge into the environment. The removal of iodine radionuclides has attracted much attention since the 1950s, as evidenced by many studies published in the literature and in the proceedings of many conferences and symposia. The current technology for the removal of iodine and its compounds. mainly of methyl iodide from gaseous radioactive wastes, consists of two basic processes, viz.: a) adsorption of iodine vapour on solid adsorbents; b) absorption of iodine vapour in aqueous solutions. Activated charcoal is currently the most widely employed adsorption material for filter beds used for removal of radioiodine from gaseous wastes. The capture efficiency for elemental iodine is remarkable (higher than 99.9 %), even at a high relative humidity of the gas which is to be decontaminated and at high temperatures. The removal of organic compounds of iodine radionuclides, such as methyl iodide, is much less efficient, especially at a higher relative humidity of the gas (see Table 18) The trend to decrease the depth of the activated charcoal adsorption bed and thus the time of contact of contaminated gas with the adsorbent in large-capacity adsorption lines processing large volumes of gaseous radioactive wastes may also lead to a reduced efficiency of removal of elemental iodine.

121

A solution to these two problems is offered by impregnating the activated charcoal surface with a substance that binds iodine radionuclides and methyl iodide by a chemical reaction or by isotope exchange. Triethylenediamine has been shown to be most advantageous for use in the former case. Its action can be characterized by the reaction: + N /'

CH,

I

CHZ

'\I/

I '\

CH,

CHZ

I

I

CH,

CH,

+ 2 CH,I(I~)

I; CH,

CH,

(5.6)

CH,

/

N

"

+

In the latter, potassium iodide or elemental non-active iodine are mostly used for activated charcoal impregnation: '"I

(charcoal)

+ CH,13'I

-,

(gas)

(charcoal)

+ CH,

'"I

(gas) (5.7)

Both types of impregnated activated charcoal, with a degree of impregnation between 0.5 % and 5 %, are commercially available. In view of the ease of ignition of activated charcoal and the relatively low temperature of captured iodine, the applicability of activated charcoal is limited by an upper temperature of 150 "C. The presence of impurities in waste gases, such as organic vapours, nitrogen oxides and carbon dioxide, leads to the poisoning of activated charcoal and significantly reduced the life of the adsorption bed. Impregnated inorganic adsorbents are recommended for the removal of iodine from gaseous wastes containing strong oxidants or for gas decontamination at high temperatures (up to 500 "C). Molecular mesh 13X in the Ag form and catalytic carriers ( S O z , Al,03) impregnated with AgNO, show high efficiency for the removal of both elemental iodine and methyl iodine over a wide range of gas temperatures and relative humidities (22). Molecular meshes remove iodine and methyl iodide from the gaseous phase in the presence of exceptionally stable Ag I incorporated in the mesh structure. The same product is formed in the capture of iodine and methyl iodide vapours on adsorbents impregnated with AgNO, , viz.: AgNO, 2 INO,

I22

+ I,

+ AgNO,

-+

-+

AgI

AgIO,

+ INO, + 3 NOz + 1/2 1,

(5.8) (5.9)

INO,

+

NO,

+ 1/2 0, + 1/2 I,

(5.10)

The so-called silver reactors containing porous ceramics impregnated with AgNO, are widely applied for the removal of radioiodine from gaseous wastes from reprocessing at Hanford and at Savannah River (USA). The equipment runs at temperatures of 110-200 "C. The lower limit is given by the temperature at which HNO, vapour in the reactor condenses, and the upper limit is given by the melting point of the impregnant. The efficiency of iodine removal ranges between 99.5 and 99.9 %. The silver reactors may be regenerated by spraying with AgNO, solution at 67 "C with subsequent drying at 110 "C. The gas to be decontaminated must not contain NH,, which reacts with AgNO, to form explosive products.

o

elemenlal

CHJI carrier ym

r2

0

air temperature 43' "C relative humidit 71+1% test durafion d'h -

\

I

15

0.05

I

I

1

$5 60 de th(rnm) 0.1 A5 0.2 -contact &me /s)

30 --bed

Fig. 18. Dependence of capture efficiency of elemental iodine and methyl iodide vapours on contact time of contaminated air with nonimpregnated activated charcoal bed

The practical application of silver-impregnated inorganic absorbents is considerably limited by their high cost with respect to the type of impregnant and its w/w), although their properties as adsorbents for removal of amount (30 -40 radioiodine from gaseous wastes are comparable to those of impregnated activated charcoal. The design of an adsorption bed for iodine removal is mostly derived from common types of filters. Most frequently, a filter with sorbent in cartridges is used. Boiler and plate systems are widespread in the USSR and in the G F R . If an aerosol

x,

filter is used, it is always located before the desorption bed in view of the possible desorption of iodine vapour from aerosols entrapped by the aerosol filter. The second technique for removal of iodine from the gaseous phase is based on the absorption of iodine vapour in aqueous solutions. The solubility of iodine kg) of I, per 0.1 kg water at in water is not very high; it amounts (29 x 25 "C. On dissolution, most of the iodine reacts with water to give hypoiodous acid and hydrogen iodide:

I,

+ H,O z

HI0

+ I- + H +

(5.1 1)

This is comparatively high degree of hydrolysis (at the initial p H of 7 water contains less than 10 % of elemental iodine while the overall iodine concentration is 2 x lo-' M ) , which equilibrates almost instantaneously at low concentrations, partly explains the high partition coefficients in the system water-iodine vapour-air. The high partition coefficients, ranging between lo3 and lo4 a t iodine concentrations in water of lo-' to lo-* M , which roughly characterize the decontamination effects of water cannot be used in practice for reasons of kinetics and capacities. The kinetics of absorption of iodine vapour in water are very slow owing to the resistance of water to mass transfer in a liquid and the high values of the partition coefficients are conditional on a very low iodine concentration in water. This makes the running and capital costs unacceptable. The technique is also unfeasible for the processing of liquid wastes. A practical solution to this problem consists in the utilization of aqueous solutions containing NaOH, LiOH, sodium thiosulphate and other compounds whose presence in water shifts the equilibrium of equation (5.1 1) to the right and significantly accelerates the kinetics of absorption of iodine in the solution. Absorption of iodine in aqueous solutions (limited to elemental iodine; the very low solubility of methyl iodide and its relative inertness to all common absorption solutions precludes the use of this technique for the removal of methyl iodide from the gaseous phase) is being increasingly used for the removal of iodine from gaseous radioactive wastes generated by reprocessing and for iodine capture in spray systems of reactor containments. The containment spray system is an important component of the systems of reactor engineering safety. Its main objective is to maintain pressure and temperature conditions during a reactor accident within the containment and to prevent the disturbance of containment integrity. Two fundamental types of absorption solutions are used in spray systems, viz., a boric acid solution (0.3 M ) adjusted by the addition of NaOH (0.15 M ) and a boric acid solution (0.3 M ) adjusted by the addition of NaOH (0.15 M ) containing sodium thiosulphate (0.06 M ) . The efficiency of removal of elemental iodine from the containment atmosphere may be characterized by the iodine capture half-life, ranging between 50 and 400 s depending on the containment properties. The washing of iodine from reprocessing wastes by scrubbers sprayed with

124

alkaline aqueous solutions usually represents the first stage of the processing line of gaseous wastes from a reprocessing plant. In addition to iodine radionuclide removal, this stage removes considerable amounts of nitrogen oxides released into the waste during fuel dissolution. NaOH solutions at a concentration of cca. 1 M containing 0.1 M of sodium thiosulphate are used for spraying scrubber beds. At high nitrogen oxide concentrations in the waste gas the absorption capacity of the alkaline washing solution is quickly exhausted owing to nitrate salt formation. The gas is therefore pre-washed in a scrubber sprayed with dilute nitric acid. The decontamination factor of alkaline scrubbers for elemental iodine and hydrogen iodide is not very high, ranging between 50 and 100. Adsorption filters are therefore placed downstream of alkaline scrubbers to meet the high demands on iodine removal. An example of such a layout is the line for iodine removal from reprocessing using the RALA technique, installed in Idaho (USA). Waste gases are passed through a scrubber where, in a countercurrent process, they are washed with a 5 % nitric acid solution containing 0.001 M of HgNO, and 0.001 M of Hg(NO,),. This particular technique was chosen for the abovementioned reasons. The presence of mercury at low concentrations and its complexing effect favourably affect the efficiency of iodine removal. After passage through the scrubber the gas is further decontaminated by two parallel beds containing about 0.2 m3 of activated charcoal with a capacity of cca. 20 TBq (560 Ci) of iodine. Installed at the end of the line are aerosol filters from sintered stainless steel (pore size 20 pm) for the removal of activated charcoal fly ash from the adsorption bed. The decontamination factor of this iodine removal line is lo4 to 105. Spent fuel reprocessing after a short cooling period gives rise to the specific problem of iodine radionuclide removal from gaseous wastes. As shown in Section, 2.3, current criteria for the radiation protection of the area surrounding the reprocessing plant require for iodine radionuclides the decontamination factor of these gaseous wastes should reach lo6 - 10'. It is obvious that a new technology for gaseous waste processing will have to be developed to meet this requirement, because in this respect the current methods of iodine removal from gaseous wastes are unsatisfactory.

5.2.3

Radioactive Aerosols

Radioactive aerosols may be defined as the dispersion system of a solid or liquid radioactive substance suspended in air or another gas. This concept comprises both radioactive substance particulars and particulates consisting mostly of a non-radioactive substance which contains a certain amount of a radioactive substance. Apart from the specific activity of the radionuclide, the aerosol size and its volume concentration are the fundamental qualities that characterize the

125

radioactive aerosol and pre-determine the method for its separation from the gaseous phase. Radioactive aerosols may be formed as a result of a number of operations and processes, including mainly: - processing and destruction of solids; - chemical processes in the liquid and gaseous phase; - adsorption of radioactive gases and vapours on non-radioactive aerosols; - neutron activation of non-radioactive aerosols. It is obvious from the above that all stages of the fuel cycle of nuclear power plants are potential sources of radioactive aerosols. The characteristic amounts of radioactive aerosols present in gaseous wastes from the fuel cycle facilities of nuclear power plants are highly dependent on the type of operation taking place in the facility, The size of radioactive aerosols in gaseous wastes from nuclear facilities may vary from 0.01 pm to several tens of pm and their concentration may range from lo-’’ to lo-’’ kg/m3. Compared with the other industrial sources of aerosols, nuclear facilities produce radioactive aerosols at concentrations lower by several orders of magnitude, which also explains the difference in the technique of arresting aerosols generated by nuclear facilities and by other industrial sources. Radioactive aerosols are mostly separated from gaseous wastes of nuclear facilities by filtration using fibrous filters, which is the only technique capable of trapping small particulates with sufficient efficiency. If the radioactive aerosol concentration in the waste is too high, any of the aerosol arrestors listed in Table 17 should be inserted in front of the filter used for gas pre-treatment. The arrestor will prevent premature exhaustion of the filter capacity or filter clogging. The effects of interception, inertness, diffusion and electrostatic effects contribute to the removal of radioactive aerosols during their passage through a fibrous filter. Their impact on the separation of aerosols from the gaseous phase depends primarily on the aerosol size, the filter fibre size and the operating conditions of the filter. In view of its considerable practical applications, aerosol filtration is theoretically well-founded. For details the reader is referred to specialist monographs (23). Commercially available aerosol filters may be classified into three groups. The first group includes aerosol filters packed with glass, metal or plastic (polystyrene, polyethylene, polyamide) fibres 1-2 pm in diameter, sometimes sprayed with oils with a low congealing point. The second and the third groups include the so-called cloth filters with glass or man-made fibres of different thicknesses. These filter media have a higher pressure drop than the filters in the first group and the filter cloth is therefore folded into a labyrinth. This arrangement increases the overall filter area, thus reducing the linear rate passage of gas through the filter and its pressure drop. The so-called high-efficiency filters (HEPA filters) form a separate category. These aerosol filters are provided with a special filter-paper made from ultra-thin glass fibres (diameter 1 pm) in

126

Table 17: Methods of Radioactive Aerosol Separation in Different Stages of Waste Processing Type of equipment

Aero: siz

(w

Cyclones, large diameter

5

Separation efficiency

Flow rate (m/s)

Pressure dro (Nm-*)

40-85

10-18 (at inlet)

loodO0

40-95

10-18 (at inlet)

90-95

0.5

250-1 300

Fume chambers, chemical separation processes

1-3

250-2500

Gas absorption and pre-treatment of waste gases with dispersed acids

1.500-8000

Removal of pyrophoric materials in ore processing: pre-treatment of waste gases from radioactive waste incineration

(%)

5

500-1000 ...

0.1-

Sprayed filters Packed columns (gases and decay particles)

5

90

Cyclone scrubbers

5

40-85

Venturi scrubbers

1

_____

Airconditioning filters

lo-

60-70 (99 for H2S04 mist)

75-85

10-18 (at inlet)

70-140

2

(at inlet)

2-3

2-5

1

Single-stage electrostatic separators

Two-stage electrostatic separators

1-

Waste gas pre-treatmen in ore mining and processing

-

-

Cyclones, small diameter

Application

Waste gas pre-treatment in ore mining and processing

Part of lines for processing gaseous wastes from radioactive waste incineration Air conditioning and ventillation of working areas

9699

60-200

Uranium metal processing, part of final stage of gas purification in chemical separation processes

85-99

60-100

Not widely used for decontamination of gaseous radioactive wastes

127

a labyrinth configuration offering high efficiency of removal of even very small aerosol particulates at a relatively low pressure drop (cca. 250 N/m2). Their superiority among aerosol filters is shown in Table 18, which compares the efficiencies of different filters in removing aerosols of various sizes.

i I I1 111

HEPA

low medium high very high

min.

I

I 0-2 10--40 45-85 99.97

10-30 40-70 75-99 99.99

40-70 85-95 99-99.9 100

90-98 98-99 99.9 100

Not many companies manufacture filter-paper for HEPA filters. The problem of availability of 0.5 - 1 pm diameter ultra-thin fibres of satisfactory quality, can, however, be resolved by replacing the paper fibres with other materials, e.g., a combination of asbestos (70 %), cellulose (30 %) and glass fibres (20 %) gives a filter medium with a sufficient efficiency, filter capacity and strength. Table 19: Fibrous Filter Size Filtered air flow-rate

45 85 215 850 1700

Dimensions (mm)

203 203 305 610 610

203 203 305 610 610

78 150 150 150 292

The design of HEPA filter sizes is determined primarily by the volume flow-rate of the gas to be processed with respect to the pressure drop in the filter. Table 19 shows the basic dimensions of the standard HEPA filter series for different flow-rates of gases which are to be decontaminated. The tabulated values relate to a filter pressure drop of 250 N/m*. Pressure drops for flow-rates not listed in Table 19, may be estimated from the linear dependence of the gas flow-rate and the pressure drop.

128

The requirements that an aerosol filter should meet depend on its operating conditions, i.e., mainly sufficient resistance to heat, moisture, chemicals and mechanical shocks. Almost all discussed filters satisfy these requirements with regard to radioactive gaseous waste processing. The design of filter units for radioactive aerosol separation and the incorporation of the filters in ventilation pipelines do not differ in any way from those for normal filters used for non-radioactive aerosols.

129

6 Fixation of Radioactive Concentrates

Many methods of low- and intermediate-level radioactive waste treatment produce concentrates that are not suitable for permanent safe burial without further processing. They include mainly evaporator concentrates, chemical processing sludges, regeneration solutions and ion-exchange saturated adsorbents, ash from solid-waste incineration plants, etc. The concentrates may be stored in tanks but all structural materials have only a limited service life and do not provide permanent environmental protection. The conversion of radioactive materials into the homogeneous solid phase improves storage safety because the choice of a suitable solidification process significantly restricts the possible release of radionuclides into the environment. In many instances further handling and transportation are also facilitated. Although safety is the primary requirement in solidification, economic aspects must be considered. In selecting a suitable solidification process, a number of aspects should be considered besides the amount, composition and activity of concentrates, such as the technological level of the solidification process, the degree of solidification of radionuclides, the production and nature of gaseous and liquid wastes and the method of final disposal. The most common solidification methods include cementation and bituminization. Other techniques, such as the incorporation of wastes into plastics, sulphur, etc., have very limited significance.

6.1 Cementation of Radioactive Wastes Of all solidification methods the incorporation of radioactive wastes into cement blocks is the simplest and has been applied for many years in various forms (24). The process is based on mixing liquid radioactive concentrates with cement to produce a solid material whose basic structure is formed by crystalline compounds of calcium hydrosilicates and hydroaluminates. The salts present in the radio-

130

active wastes are adsorbed on the surface of the cement particles and are retained in the block by a solid crystal lattice. Portland cement is the most commonly used material i n cementation, and its chemical composition is shown in Table 20. Other characteristic properties include: Table 20: Chemical Composition of Portland Cement Component

Czechoslovakia

I

22.99-25.60 6.16-9.96 2.64-2.86 54.22-59.77 2.84-5.40 1.24-1.63

USSR 17-25 3-8 0.3-6 6-67 0.1-4.5 0.3-1 0.5-1.3 0.2-0.5 0.14.3

Table 21: Dependence of Cement Consumption and Cement Waste Volume on Salt Concentration in Cemented Solution

1

Salt concentration

50 100 150 200 250 300 500

1

Ratio of solution volume to cement weight 0.80 0.75 0.65 0.50 0.30 0.17 0.15

I

Increase in block volume relative to solution volume

I

1.25 1.35 1.so 2.00 3.30 5.80 6.80

annealing losses 0.90 - 1.37 % insoluble residue 0.76 - 1.47 % start of hardening 105-295 min hardening time 295 -390 min amount of water added 25.3 -27 % compression strength after 28 days 386 -454 kg/cm2 tensile strength after 28 days 33.9 -43.6 kg/cm2 The optimal cementation process technology should ensure a sufficiently high strength (safe transport and handling of materials) and the best possible fixation

131

of radionuclides (low leaching rate). The mechanical strength is conditional o n the attainment of a certain optimal ratio of alkaline to acidic oxides. Thus, one of the fundamental requirements is to limit as much as possible the amount of salts added. The acceptable minimum block strength (50 kg/cm2) is relative to the salt concentration of 0.13 kg per kilogram of cement. On these assumptions, the dependence between salt concentration in radioactive wastes, the consumption of cement and the volume of the final material can be determined (Table 21). As shown in Table 21, it is not desirable to increase the salt concentration in wastes above 150 kg/cm2 in the cementation of liquid radioactive wastes. In addition to the overall salt content, the limits of the maximum concentrations of some compounds in cementation should be taken into consideration. The product’s mechanical properties deteriorate significantly when the following concentrations are exceeded: Ca(NO,), 10 kg/m3 NaNO, 150 kg/m3 NaCl 30 kg/m3 Na,SO, 25 kg/m3 Na3P0, 15 kg/m3 Na,CrO, 5 kg/m3 soap 1 kg/m3 0.5 kg/m3 et hylenediaminetetraacetii acid detergents 1 kg/m3 surfactants 0.1 kg/m3 The presence of radiation-unstable substances, such as water, nitrates and hydrates, results in the formation of gaseous radiolytic products. In view of the large surface area of the cement, however, the gaseous products remain absorbed in the cement and only when the absorbed dose is increased above loJ Gy (10’ rad) does the pressure start to increase in the tank containing cement wastes. Virtually no release of gaseous radiolytic products takes place in blocks with a volume activity lower than 40 TBq (lo3 Ci/m3). The relatively high leachability of cement blocks is the main constraint on the use of cementation. Leachability depends on the type of radionuclide, the solubility of salts present and the age of cement blocks. New cement blocks show the highest leachability, and block ageing reduces leachability. The reduction in leachability is most marked with cesium; with strontium, which has a relatively low leachability, the process occurs to a much smaller extent. These changes may be attributed to the fact that in cement hydration colloidal particles swell and crystalline products grow between them. Pores in the cement becomes progressively narrower, which leads to a reduced leachability. It may also be assumed that in the recrystallization of some components of the cement crystal lattice certain radionuclides penetrate into the crystal lattice while others may diffuse in gel particles whose surface acts as a semipermeable membrane.

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The relatively high initial leaching of cement blocks results from the high dissolution of soluble salts on the surface of the block. After the initial stage the leaching rate significantly decreases because further leaching is only possible following radionuclide diffusion from the inside of the block to its surface. Experimental studies of overall radionuclide leaching showed that ageing for 9 months results in the leaching of I % of strontium and 3 % of cesium previously present in the cement block. Cement blocks may therefore be stored after ageing for 6 months without any surface treatment. Cementation is a suitable process for solidifying low volumes of liquid wastes and radioactive concentrates whose volume activity does not exceed 46 Bq/m3 (0.1 Ci/m3) and which contain the maximum of 150 to 200 kg/m3 of salts. The maximum salt content in a cement block is 130 g/kg and the solidified liquid wastes must not contain free acids. In contrast to bituminization, a significantly lower volume reduction is achieved in cementation and radionuclides can be leached more easily from cement. Cementation is most frequently used for processing low-level wastes from chemical waste-water Itreatment. The cement mix is usually stored in metal or concrete containers, preventing the radioactive cement from coming into direct contact with ground or surface waters. The method is also suitable for sludges containing 89Sr, 90Sr, 239Puor 242Ambecause these radionuclides are firmly bound by cement. On the other hand, cesium and ruthenium can easily be leached and certain wastes should therefore be suitably treated prior to cementation. The addition of clay to the cement mix gives the optimal improvement in the retention of strontium and cesium. Apart from chemical sludges, cement fixation has been employed for evaporator concentrates, saturated sorption materials and small volumes of high-level wastes. The types of equipment used for radioactive waste cementation differ significantly. Different types of cement or mortar mixers commonly used in the building industry are mostly employed. The equipment is operated continuously or intermittently and the c?m:nt mix is dischargzd either into transport containers or large storage bunkers. One of the simplest cemzntation techniques was originally used by the Nuclear Research Institute at Re?, near Prague. Prior to cEmentation, small volumes of sludges with a volume activity of 0.4-4 GBq Ci/l) contained 20-25% of the solid phase (kaoline, Kieselguhr, iron (11) hydroxide, barium sulphate, hexacyanoferrate (11) compounds, phosphates, etc.). The sludges wzre discharged into drums of capacity 0.1 m3 (100 I), sealed with a slide-cover to which a mixer, a sludge inlet hose and a cement inlet hose were attached. Following the addition of the appropriate amount of cement the mix in the drum was agitated and the mixer was removed. After the cement had hardened the drum was covered and painted with a protective coating. The cementation of 1.5 m3 of sludges produced 4.5 m3 of cement matter annually and this final volume corresponded to 900 m3 of annually processed liquid wastes. This simple equipment operated with relative

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reliability; the only problems were the sealing of drums and cleaning of the mixer. Large-scale cementation has only been used in France. The cementation equipment used in Grenoble is similar to that at NRI Re?, Czechoslovakia. Cement and vermiculite are mixed in 0.4 m3 (4001) concrete containers or 0.22 m3 (2201) metal drums. Both cement and vermiculite are stored in containers provided with equipment for pneumatic transport and batching. The equipment processes evaporator concentrate containing 400 kg/m3 of salts, of which 98 % is in soluble form. The activity of wastes ranges between 4 and 190 GBq/m3 (0.1 and 5 Ci/m3). The mix consists of 0.25 m3 of sludge, 300 kg of Portland cement and 40 kg of vermiculite. The solidification of 0.25 m3 of sludges produces an overall volume of 0.4 m3 of cement. The cement and vermiculite are weighed, mixed by a screwmixer and proportioned together with sludges from the mixer to the container. The mix is agitated with a hypocycloid mixer. For proportioning and mixing the concrete or metal container is sealed with a slide-cover provided with a mixer, a sludge inlet, an inlet for the supply of the cement-vermiculite mix and a dust exhauster. The cement mix is gradually discharged into 4 containers placed on a turnable and each container is gradually filled in 4 operations. In view of the fact that in solidification the volume is reduced by I5 %, the technique allows the better use of the whole container capacity. The main advantages of the equipment include simplicity and low capital cost. The cement also serves as a shield and therefore, up to a volume activity of 40 GBq/m3 (1 Ci/m3), no other shielding is necessary for container transportation and storage. Disadvantages include the increased volume of stored material and, in some instances, relatively high leaching, requiring the use of an extra container. At Fontenay-aux-Roses, France, cementation involves mixing in drums placed, in a concrete cell. The process is fully automated and includes the following operat ions : - a 0.2 m3 drum containing a small amount of pebbles is introduced into the concrete cell fitted with an air seal; - the drum is filled with the appropriate amount of cement and concentrate from the evaporator; - the drum is covered and the lid is flanged to it; - the drum tilting equipment is started; - the drum is removed from the mixing equipment and placed in a concrete container with walls 0.1 m thick; - cement is injected into the space between the inner wall of the concrete container and the drum. This equipment also operates reliably. Its disadvantage is an insufficient use of the drum capacity because the space formed as a result of the hardening mix remains unused. At Saclay, France, sludges containing calcium carbonate and nickel hexacyano-

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ferrate (IT) are cemented. Their volume activity is as high as 4 TBq/m3 (100 Ci/m3). The mixture consists of 83 kg of sludges, 55 kg of Portland cement and 0.01 m3 (101) of water. The mixture is mixed by a rotary drum concrete mixer made of stainless steel. Cement is stored in a container and is proportioned into the mixer together with sludges. From the mixer the mix is discharged into concrete barrels, which are filled gradually in several separate operations. After the mix has hardened, the concrete barrel is sealed with a layer of cement containing no radioactive constituents. As a result of solidification the volume of the mix increases from 0.5 m3 of sludges to 0.8 m3 of cement product. At Los Alamos, Calif., USA, a cement-vermiculite mix is used for solidifying wastes containing up to kg/m3 of plutonium and 10-4-10-3 kg/m3 of americium in 6-7 M nitric acid. Of other ions, especially A13+, Fe3+, Mg2+, CaZ+and F - are present. The high content of nitric acid prevents direct cementation and waste waters are therefore first neutralized with a 50 % lye solution. Neutralized waters are mixed with cement to which 3 % (w/w) of vermiculite is added. The cementation process consists of the following operations: - acidic wastes are transported in 1 m3 stainless-steel containers; - 0.378 m3 of acidic waste solution is proportioned into a 0.85 m3 stainless-steel neutralization tank; - acidic wastes are neutralized in the tank with a 50 % lye solution; - during agitation the neutralized liquid is cooled with water passed through a cooling coil; - 126 kg of Portland cement, 4 kg of vermiculite and two halves of a normal brick to facilitate agitation are proportioned into a 0.2 m3 steel drum. The drum is covered with a lid provided with a rubber seal; - 0.078 m3 of neutralized wastes is air-pumped into the drum; - the drum is tilted for 15 min using equipment for mixing in two drums a t a time; - the mix in the drums is allowed to solidify for 24 h and is then transported to a storage site. The cement volume activity ranged between 3 and 40PBq/m3 (8 x lo4 and 1 x lo6 Ci/m3) and the leached amount annually was 7 x %. At Los Alamos, cementation is also used for the solidification of wastes from spent plutonium processing. The wastes, with activity levels around 40 TBq/m3 (lo3 Ci/m3), contain different fission products, mainly 13'Cs and "Sr. In order to reduce radiation hazards to operators, a shielded container is prepared in advance. A 0.06 m3 metal can is placed inside a 0.2 m3 metal drum and the space in between is filled with concrete. The inner drum is filled with a maximum of 120 kg of cement and two pieces of brick are then added to facilitate mixing. The inner can is covered with a metal lid provided with two outlets, for vacuum and liquid wastes, and the lid is sealed with concrete. The appropriate amount of

liquid wastes is then air-pumped into the prepared container and the mix is agitated %). For improved for 15 min. Leaching of 90Sr from cement is relatively low I3'Cs capture, 20 % (w/w) of clay is added. This method of solidification using cement is relatively costly. It would be very difficult, however, to solidify liquid wastes by any other method. At Brookhaven, USA, a mix of cement and vermiculite is used for solidifying evaporator concentrate of a volume activity of 1.9 TBq/m3 (50 Ci/m3) containing about 20 % of dissolved salts. The radionuclides present include mainly 137Cs, 90Sr and 6oCo. The radioactive concentrate, vermiculite and cement were mixed in a mixing vessel and discharged into 0.2 m3 drums. Later, reinforced concrete blocks with dimensions of 1.5 x 1.5 x 1.9 m and a wall thickness of 0.15 m were introduced for this purpose. Vermiculite (2.7 m3) and Portland cement (0.68 m3) are proportioned into a block. Sludges are discharged through a perforated pipe. This technique ensures sufficient mixing and the overall cost of waste solidification is reduced. Medium-level wastes from a fuel reprocessing pilot plant were cemented in Norway. The concentrate from the evaporator, with a mean volume activity of around 1.3 TBq/m3 (35 Ci/m3), contains a mixture of fission products (mainly "Sr, 95Zr, 95Nb, lo3Ru, 137Csand '44Ce), trace amounts of uranium and plutonium and larger amounts of non-active salts. A 0.125 m3 (125 1) volume of the concentrate is proportioned into a 0.21 m3 steel drum lined with polyethylene and the drum is transported to a cementation bunker. In the bunker, the necessary amount of cement is added from a container, the drum is sealed and the contents are mixed using a vibrator. Approximately 5 m3 of the concentrate and 'of the organic solvents, whose proportion in the concentrates may reach 15 %, are cemented. At Karlsruhe, concentrates were cemented from an evaporator for processing medium level liquid wastes. A 0.1 -0.1 1 m3 volume of the concentrate, containing 30-40 % (w/w) of salts, was proportioned into a 0.2 m3 steel drum filled with 200 kg of cement. Mixing by tilting the drum was not satisfactory; the mix was not homogeneous and contained dry cement and liquid concentrates. A new cementation facility has therefore been built. Cementation proceeds in a conical vessel equipped with a mixer, a concentrate inlet and a cement screw feeder. The cement mix is discharged into drums. Thus, the homogeneity of the product was increased while the volume of cement necessary for the disposal of 0.1 m3 of the concentrate was reduced from 200 to 150 kg. The strenuous cleaning of the equipment after the work shift is the only constraint. Various inorganic sorbents are added in the cementation of concentrates containing radionuclides in a soluble form. Operating experience in Czechoslovakia showed that similar results could be achieved when chemical sludges and concentrates are cemented together. The best results were achieved using a mix of the following composition:

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22 kg 10 kg 5 kg Unlike most other countries, in the USSR the cmentation of liquid radioactive concentrates takes place on the storage site and the radioactive cement is stored together with compacted solid wastes in reinforced concrete tanks of capacity 200-600 m3. The tank is partitioned into several compartments, which are filled gradually and are watertight. The disposal of solid and solidified liquid wastes on one site allows the better use of the storage site and prevents precipitation from penetrating the liquid wastes. Fig. 19 shows a schematic diagram of the cementation equipment, consisting of: Portland cement 350 chemical sludges (70 of water) evaporator concentrate (200 kg/m3 of salts)

Fig. 19. Schematic diagram of equipment for radioactive waste cementation

- a 13 m3 cement container with a prism-like bottom part. Attached to the oblique container walls are vibrators facilitating a continuous flow of cement. The container is filled through a pipe 0.08 m in diameter and is discharged via a control valve; - a 4 m long screw feeder for transporting cement from the container to a mixer; - a vertical cement mixer of a cylindrical shape, 1.3 m in diameter and 1 m in height, with a conical bottom 0.3 m in height provided with a 0.15 m diameter outlet. The mix is agitated with a screw-shaped mixer 0.2 m in height at a speed of 5.8 rps. The mixer is powered with a 4.5 kW electric motor via a V-belt in a ratio of 1 : 4. The mixer is sealed with a cover provided with inlets for cement and liquid waste supply. The bottom discharge opening is controlled by a n electric motor; - a centrifugal pump and a piping system for liquid waste supply; automated process control b y a time relay. The automatic operation can be adjusted to one cycle or to continuous repetition of the respective cycles,

-

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The cementation equipment has a capacity of 15 m3/h; intensive agitation for 2 min is sufficient to obtain satisfactory homogeneity of the mix. When the concrete tanks are filled, the respective compartment which :is being filled is covered with concrete slabs, leaving only a small filling hole. After the compartment has been filled with solid wastes, the cement mixer is placed above the filling inlet so that the cement mix is discharged straight into the tank. When the tank is filled with cement, the cementation equipment is removed, the filling hole is covered with a concrete slab and the joints between the slabs are sealed with cement that does not contain radioactive constituents. The entire surface of the tank is then covered with a layer of hot asphalt and a 1 - 1.2 m thick layer of earth. The cost of the solidification and disposal of 1 m3 of liquid concentrates in 1977 was about 315 US $6, of which the amortization for the capital cost of reinforced concrete tank construction amounted to 25 % and the cost of materials needed for cementation amounted to 27 %. The final product is a monolith of volume 200 - 600 m3, the surface of which is insulated against precipitation and infiltration of surface water. The technique has several advantages: - the storage area can be used more efficiently than in the disposal of a number of smaller containers with solidified liquid wastes or in the separate disposal of solid and solidified liquid wastes, - high radiation safety is ensured as the radioactive material has the shape of a large concrete monolith placed in a reinforced concrete waterproofed container and radionuclide migration into the environment is made practically impossible. The disadvantage of the technique is that radioactive concentrates should be transported from the source to the storage area in liquid form. Recently, several new variants of the cementation process have appeared whose objective is to reduce leachability and to increase the mechanical strength of the cement product. In Italy, the possibility of impregnating cement with an organic polymer has been tested (32). The process consists of the following stages: - incorporation of radioactive products in cement, - cement dehydration in a vacuum at 165 "C. Up to 20-30 % (w/w) of water is thus removed from the cement block and large number of free pores are obtained, - impregnation of the dehydrated cement with an organic polymer using a catalyst. Styrene and methyl methacrylate have been tested, - heating and polymerization. Styrene requires a temperature of 85 "C for 40 h and methyl methacrylate requires a temperature of 75 "C for 19 h. This technique more than doubles the mechanical strength and reduces the leaching of radionuclides more than 10-fold. Compared with common cementation, the process is relatively complex and its applicability will depend primarily on economic considerations.

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The Nuclear Research Institute in Rei has improved the process of cement fixation of radioactive wastes by introducing the bottom evacuation of water. The advantages of the improved process are a reduced block volume, increased mechanical strength and reduced radionuclide leachability. The fundamental principle has further been improved by the V$zkumn$ ustav mechanizace, automatizace a technologie v$roby stavebnich dilcci (Research Institute of Mechanization, Automation and Technology of Building Element Production) in Prague, which replaced evacuation by the more efficient use of pressure. This made it possible to increase the dimensions of the cement block and to shorten the time required for the completion of cementation. The Institute has also developed and non-actively tested a pilot line for the fixation of radioactive wastes using cement with a capacity of processing 2.8 x m3/s of chemical sludges or 8.6 x l o q 5m3/s of concentrates from the evaporator. The salt content of sludges and concentrates is 400 kg/m3. In contrast to the commonly used cementation of radioactive wastes, the new technique is characterized by a high water coefficient of the mix. In the compacting process excess water is extruded by positive pressure via a filtration system and is transported to a collecting tank by negative pressure. The necessary amount of liquid wastes is proportioned from radioactive concentrate and cement containers into a rotary activator and cement is then added. Homogenization proceeds for 3 min and mix is then discharged into a mould provided with press cloths on the bottom and on the upper press face. Negative pressure in the chamber and the press face is secured by a water pump. The cement mix is compacted at a pressure of 2.32 MPa; the compacting time varies depending on the type of waste between 8 and 18 minutes. After compacting the mould is immediately removed and the block is transferred to an air-conditioning conveyor belt. Depending on the composition of radioactive concentrates, the cement blocks showed compression strength between 49 and 58.8 MPa and leachability of the order of m-' d - ' g cm-' d-'). The leachability could further be reduced by the addition of a bitumen emulsion to liquid wastes, because a cement block prepared in this manner is impregnated with bitumen. I n vacuum cementation the volume of materials which should be stored is reduced while the leachability attained is the same as in bitumen. The whole process takes place at ambient temperature. Its disadvantage is the formation of liquid wastes in water evacuation from the block which, especially in the process of the solidification of concentrates with a high content of soluble salts, may require further processing.

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6.2 Incorporation into Bitumen The bituminization of radioactive wastes is the technological process of converting liquid or solid radioactive wastes into solid blocks suitable for safe long-term burial. The process is applied to a number of medium-level wastes, mainly chemical sludges, evaporator concentrates, ion-exchange regenerant solutions, organic solvents, used natural and synthetic sorbents, ash and plastics. The mechanism of incorporation of various materials into bitumen involves several processes, such as chemical reactions between bitumen and substances present in wastes, chemisorption of polar groups, physical sorption and the coating of solid particles or objects with a layer of bitumen. The incorporation of radioactive wastes into bitumen has been the subject of reasearch since the early 1960s. The first industrially important installation was started up in 1965 in Belgium. Bituminization has since come to be used in many countries, mainly in the United Kingdom, Czechoslovakia, France, the GFR, Poland, the USSR and the USA.

6.2.1

Categories and Properties of Bitumen

The term bitumen covers a mixture of high-molecular-wzight aliphatic and aromatic hydrocarbons obtained during petroleum processing. Its composition and origin give bitumen a wide range of forms, from an elastic solid to a viscous liquid. It can be separated by solvent extraction into solid asphaltenes and viscous oil malthenes. Bitumen has main properties that are advantageous for solidifying radioactive wastes: it is highly adhesive, waterproof, resistant to acids, alkalis and salt solutions, has good miscibility with a considerable volume of solid particles and can be converted into the liquid phase at a slightly elevated temperature. Its drawbacks are limited radiation stability, flammability, poor thermal conductivity and relatively low tensile strength. Bitumens can be divided into several categories on the basis of the methods of production : - bitumens obtained by petroleum distillation (softening point 34 - 65 "Ci penetration at 25 "C, 2 x 10-3-22 x m); - pitches - bitumens liquefied by addition of a solvent; - oxidized bitumen formed by blowing air through petroleum (softening point 70-140 "C; penetration at 25 "C, 0.7 x IOd3-4.5 x m); - cracked bitumen formed by pyrogenic breakdown of heavy molecules (softening point 77-85 "C; penetration at 25 "C, less than 0.5 x m); and - bitumen emulsion formed by emulsification of bitumen in soapy water. Anionic emulsions contain alkaline soap as emulsifier whereas cationic emulsions contain amine salts.

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A type of bitumen should be selected whose properties, best suit the chosen technology. In selecting bitumens, the following factors are mainly considered: - penetration - determines the relative hardness under known conditions (temperature, loading and time); - viscosity - determines the flow characteristics of molten bitumen and is especially tested in the range of the envisaged operating temperatures; - flash point - indicates the temperature to which the material may be safely heated on an open flame. In applying bituminization, it must be borne in mind that individual batches of bitumen are likely to vary considerably in both composition and physical properties. Various countries use different types of bitumen for radioactive waste processing. In Western Europe, the following types are preferred: Mexphal l0/20 (Belgium), Mexhalt 15 (GFR), Mexhalt 40/50and Mexhalt R 90/40(France). Table 22 shows Table 22: Types of Bitumen Used in Radioactive Concentrate Solidification in Eastern Europe

Country

Bitumen

Softening point ("C)

Penetration a t 25 "C m) 1.3 2.6 1.6

Bulgaria

"Rubrax" insulation bitumen battery bitumen

130 78 113

Czechoslovakia

Romashin4aratov bitumen

43-45

B-45

54-60

GDR

NB-30

UB-45

Romashin bitumen Poland

P-60 D-35

I II

14.3 3.5-5.0

70 40 58

60-69 @ 56-

1.5-4.0 3.M.O

~~

USSR

BNK-2 BN-2 BN-3

BN-4

40 40 4s 70

14.0 8.1-12.0 4.1-8.0 2.14.0

bitumens used in Eastern Europe (33). The A-I00 asphalt used in Czechoslovakia is the residue from the vacuum distillation of Romashin - Saratov - Mukhanov crude oil. Table 23 lists the basic properties of various types of bitumen.

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Table 23: Basic Properties of Some Types of Bitumen Bitumen

Mexphalt 10/20 Mexphalt 20/30 Mexphalt 40/50 Mexphalt SO/lOO A-100 R-60

Softening Point ("C)

60-75 59-69 47-60 41-51 47 60-69

Specific gravity (ks/m3)

1020-1 070 1020-1070 1010-1060 1010-1 050 1030 1010

Penetration (lo-")

10-25 20-30 40-50 80-100

95 1540

Solubility (%)

99 99 99 99.5 99 99

6.2.2 Composition of Bitumen Block Suitable properties of the final product of radioactive waste bituminization must be maintained in the bituminization of radioactive wastes, which also limits the amount of wastes permissible in a unit volume. This constraint mainly concerns the total volume of solid matter, the content of nitrates, specific activity, liberation of radiation heat and flushing of the buried block. The major constraint results from radiation effects of the radionuclides present, as bitumen is a high-molecular-weight mixture of aromatic and aliphatic hydrocarbons and is easily subject to radiolysis. The irradiation of bitumen can result in a series of radiochemical processes: - radiolysis of hydrocarbons with escape of gaseous products; - chemical reaction of bitumen and the.solids present with radiation-produced radicals; and - oxidation of the bitumen. Each of these processes proceeds at a certain rate and the relationship between these rates defines the overall radiation effect. The radiation stability of bitumen was studied by the irradiation of bitumen using a 6oCosource. Experiments have shown that bitumen can safely be used when the radiation dose .accumulated during the storage time does not exceed lo7 Gy (lo9 rad). Increasing the dose above lo8 Gy (10'' rad) considerably changes the chemical structure of the material, changes the C : H ratio and results in the formation of gas bubbles. Radiolysis leads to a volume increase of between 14 and 30 %, the softening point increases and penetration decreases. Different types of 'bitumen show different radiation stabilities. Generally, the higher the level of oxidation of the bitumen used, the lower is the rate of formation of radiolytic products. Together with the formation of gaseous radiolytic products, oxidation and some polymerization reactions occur. The intensity of these processes

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increases with increase in the fraction of unsaturated hydrocarbons in the bitumen. When oxidized bitumen is irradiated, the radiolytic products cause tension, resulting in the formation of pores and thus in the release of g a s s In elastic typis of bitumen, the formation of pores does not occur and the radiolytic products remain in the bitumen, causing an increase in volume. Irradiation experiments with bitumen yield only approximate data on the behaviour of bitumen blocks that contain radioactive wastes. The effect of relatively short-term intensive gamma radiation may differ considerably from that of less intensive long-term action of all types of radiation. It should be added for comparison that irradiation with doses of up to lo6 Gy (lo8 rad) from a cobalt source is equivalent to the storage of wastes containing a mixture of fission products with an initial volum: activity of 41 TBq/m3 (1.1 x lo3 Ci/m3) for 100 years. The effects of internal contamination were studied in samples of bitumen containing 60 % (w/w) of waste solids, which consisted of a mixture of fission products with a volume activity of 4-260 TBq/m3 (102-7 x lo3 Ci/m3). A comparison of the changes induced by irradiation with a dose of 4.1 x lo5 Gy (4.1 x lo7 rad) from a cobalt source or with a similar internal irradiation dose showed much lower effects of internal irradiation. The dose did not result in theformation of gaseous radiolytic products, nor did it cause an increase in volume. The rate of leaching of insoluble fission products reached kg/m2 d (1 - 5 - lo-' g/cm2 d), while the leaching rate for soluble Na and Cs was about kg/m2 d (1 -4g/cm2 d).

+02Q .

9

_----_---

-_----57

G59/kg

_/--

........ ............ .570GEf /kg

Fig. 20. Gaseous product release from bitumen materials with different specific activity for "Sr

Long-term studies have been conducted of the release of gaseous radiolytic products from samples consisting of 60 %, of BN-3 bitumen and 40 of sodium nitrate with a different specific activity of 90Sr. Fig. 20 shows pressure changes in a sealed vessel containing bitumen during storage for more than 4 years. Material with a specific activity of 5.6 GBq/kg (0.15 Ci/kg) b x a m e radiation stable after storage for 2 years; the pressure increase due to gaseous radiolytic products, however, was lower than the pressure reduction due to the adsorption of atmospheric oxygen. In samples with specific activities of 57.6 GBq/kg (1.54 Ci/kg) and 570 GBq/kg

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(1 5.4 Ci/kg), the pressure increase continued but its rate gradually decreased. This also is proof of increasing radiation stability. The composition of radiolytic gases has been studied in experiments with irradiating 60% BN-3 bitumen and 40% sodium nitrate. The materials were irradiated using a cobalt source with a dose rate of 2.1 x lo3 Gy/h (2.1 x x lo5 rad/h), which is equivalent to 2.7 TBq of 90Sr in 1 kg of material. The gaseous phase contained 15.2 % of H,, 2.5 % of CH, and 0.14 % of CO,, while in irradiating the BN-3 bitumen alone the composition of gases was 10.6 % of H,, 2.5 % of CH,, 0.9 % of C,H, + C,H, and 0,13 % of CO,. When the dose rate was reduced to 24 x 10' Gy/h (2.4 x lo4 rad/h), i.e., 370 GBq (10 Ci) of 90Sr in 1 kg of material, the hydrogen concentration decreased to 3.5 % (w/w) and the content of other radiolytic gases was negligible. The radiation stability of bitumen blocks containing wastes is also affected by the content of solids in the bitumen. Sodium nitrate, which is a common component of liquid wastes, may serve as an example. Sodium nitrate has a higher radiation stability than bitumen and its presence reduces radiation effects in the bitumen. For instance, irradiation of a 6 : 4 mixture of bitumen and sodium nitrate results in a decrease of almost 50 % in the amount of radiolytic gas released. The effect of irradiation on the leaching rate of bitumen blocks. has been thoroughly studied. It was found that the leaching rates of irradiated and nonirradiated samples differ only slightly, ranging between values of the order of 10-3-10-4 kg/d (lo-, and lo-' g/cm2 d) for blocks containing up to 60 % (w/w) solids. The experiments have shown that the maximum permissible activity of bitumen blocks is limited by radiolytic gas formation while the leaching rate effect is much lower. From the radiochemical point of view, wastes with a volume activity of 370 GBq/m3 (10 Ci/m3) can be incorporated into bitumen. Radiolysis also results in the liberation of heat, which causes a temperature increase in bitumen blocks and, under unfavourable conditions, may result in the melting of the blocks. The internal temperature increase is affected by the thermal conductivity of bitumen, temperature, radiation and the dimensions of the bitumen block. The thermal conductivity of bitumen is relatively low (not exceeding 0.25 W/m "C); it changes slightly with temperature and also increases with increase in the solids in bitumen. Calculations show, however, that for bitumen blocks with a volume activity of 37 TBq (lo3 Ci/m3), the specific heat release is only 3 W/m3 and self-heating is not a limiting factor in the storage of bitumen blocks. A serious problem encountered in the bituminization of radioactive wastes is the presence of nitrates and other oxidizing agents, which might, under certain circumstances, lead to ignition or explosion. Nitrates lower the flash point of bitumen. The content of 21 % (w/w) of sodium nitrate in Mexhalt 15, for example, lowers its flash point from 420-440 "C to 320 "C and 26 % (w/w) of calcium nitrate lowers it even more, to 270-280 "C.

144

Although nitrates facilitate and accelerate combustion, they are not able to increase the combustion rate significantly. A series of experiments also verified that the presence of heavy metals and nitrates does not enhance the sensitivity of bituminous mixtures to shock in the range of operating conditions envisaged.

6.2.3 Process of Bituminization Several technological problems have to be resolved, such as the properties of bitumen-radioactive waste mixtures, heat transfer in bituminization facilities and leaks of gaseous and volatile substances, before the incorporation of wastes into bitumen is effected. Salts may show various behaviours when incorporated into bitumen; according to their properties of mixing with bitumen they can be classified into the following groups: - salts that form a mixture with bitumen easily; - salts that are readily miscible with bitumen but which tend to agglomerate and concentrate in the lower part of the mixture on account of their higher density; - salts that affect the structure of bitumen unfavourably or increase the leaching rate significantly on account of their reaction with the bitumen. Table 24: Properties of Bituminous Materials Containing 80 % of BNK-2 Bitumen and 20 % of Anhydrous Salts Process temperature

NaNO, &(N0,)2 . 4 H20 Fe(N03), . 9 H20 AI(N03)3. 9 H20 FeS04. 7 H20 Fe2(S04), . 9 H20 Al~(S04)3 18 H20 FeCl, . 6 HzO AIC13 . 6 HZ0

.

I

150 "C

Salt

softening point ("C)

penetration (lo-, m)

45 48 126 114 47 66 55 58 84

13.6 14.7 0.8 1.1 14.2 8.0 3.2 11.8 4.6

230 "C point

penetration

(lo-, m) 57 128 203 200 57

138 98 141 93

9.7 2.0 clotting clotting 7.1 1.2 1.1 1.1 2.0

Salts incorporated into bitumen may considerably affect the physico-chemical properties of the resulting product. Table 24 shows changes in the softening point and penetration a t 150 "C and 230 "C for BNK-2 bitumen containing 20 % of different anhydrous salts.

145

Heat transfer in bitumen-radioactive waste mixtures is very poor and causes a number of operating problems. The heat transfer must usually be determined experimentally as the basic properties of thermal conductivity, specific heat, viscosity and density vary significantly with the type of bitumen used and with the composition of wastes. The incorporation into bitumen proceeds at elevated temperature and thus some gaseous and volatile substances are released into the environment. The rate of release depends on the operating temperature and on the composition of the waste. Scrubbing of the gaseous phase is effected with condensers, electrostatic filters and activated carbon filters. The results showed that in the bituminization of sludge with a volume activity of 3 GBq/m3 (8 x Ci/m3) at an operating temperature of 240 "C, condensed water retained 0.02 % total activity while uncondensed gases retained 0.01 %; after passing through electrostatic filters, the activity was reduced to 0.015 %. Decreasing the operating temperature also decreases the activity present in waste gases. At 150 "C, for instance, condensed water contains 6 x % activity, uncondensed gases 5 x % activity and there is % activity behind the electrostatic filter, all at the same activity of the processed waste. Soluble salts can be washed out of bitumen more easily and therefore the bituminization of evaporator concentrates containing additives of suitable sorption materials, such as various natural adsorbents and sulphur, was tested. For example, the addition of 2 % of clay halved the leaching rate of soluble 13'Cs. It is also necessary to use thickeners, such as clays and cement, in order to obtain acceptable products in the bituminization of organic solvents.

6.2.4 Technological Installations Although there are many bituminization installations they can be classified into three groups according to the principle employed (25, 27, 28): - the first group includes installations where waste sludges or solutions are fed into stirred bitumen that has been previously melted. The installations are either heated externally by an electrical heater or by another heating medium (steam oil, etc.), or electrical heating elements may be located inside the installation. Water is evaporated and solids contained in the wastes are mixed with the bitumen. When the required contents of solids in the bitumen is reached, the mixture is discharged into vessels and allowed to cool before burial. The gaseous phase is passed through a cleaning system which must entrap oil aerosols and gaseous and volatile radionuclides. Such installations are operated in Belgium, Great Britain, GFR, Poland and the USSR; - the second group includes installations that use methods based on the prior removal of moisure. The dehydrated material is either mixed with bitumen in the

146

same installation (screw extruder), or the process takes place in two separate installations (drum drier - screw extruder). This process is used especially in France, GFR, the USSR and at EUROCHEMIC; - the third group uses a suitable bitumen emulsion rather than bitumen. Waste water is eliminated mostly with film evaporators, but other types of evaporators or a screw extruder may also be used. Bitumen emulsions are used mainly in France, Czechoslovakia, the USSR and the USA. Concentrates

n

i-'

S(ud es

&

r-----T---Stack

L

Bitumen heater

1

Condensate from l i p i d waste processinj

Fig. 21. Flow chart of hot butirninization

The incorporation of chemical sludges into bitumen at 200-230 "C was developed in Belgium. A pilot plant of capacity 0.025 m3/h was put into operation in 1962. Since 1964, a bituminization plant has been operating with an output of 0.1 m3/h (Fig. 21). The basic appslratus is a mixer-evaporator divided into two zones for mixing bitumen and for mechanical foam quenching. The apparatus is equipped with automatically controlled electric heating (with power inputs of 25 kW for the mixing zone and 10 kW for the foam quenching z'one). At the bottom of the mixer-evaporator is an electrically heated outlet valve which can be cleaned with compressed air. The vessel is fitted with a portable turbine-type mixer with blades adjustable from 0 to 80". The intensity of mixing is increased by increasing the angle of the blades. The speed of the mixer motor is 25 to 50 revis, varying with the viscosity of the mix. The mixture of vapour and air passes through a water-sprayer scrubber and an electrostatic filter prior to its discharge into the atmosphere. The mix from the bituminization vessel is discharged into drums under slightly reduced pressure, which facilitates degassing of the mix.

147

For safety reasons, the bitumen is pre-heated in a separate vessel under a minimum negative pressure of 1000 N/mZ. A similar installation with an operating temperature of 220 "C was commissioned in Great Britain in 1968. The installation consists of a bitumen pre-heater, a bitumen and sludge receiver, a mixing oven and an off-gas clean-up system. In view of the difficulties posed by the transport and dosing of some sludges, an oil-heated drum drier is used for evaporating the moisture from the sludge. The sludges are dewatered to 50 - 60 % and broken down to pellets of diameter between 0.005 and 0.01 mm. The pellets are easily transported and batched. A pilot plant with a 0.1 m3 bitumen oven with a stirrer was commissioned in 1964 in the GFR. The oven was loaded with 80 kg of bitumen and heated to 220 "C. The concentrate was added to the bituminous mass at the rate of 8 x m3/h until a level of salts of 40 % (w/w) was reached. The melt was then discharged into 0.2 m3 drums. The vapour from the mixing oven was passed to a column with a wire mesh on which condensate was sprayed from the top and the uncondensed gases were removed by a vacuum pump through a filter. The method was used successfully for sludges but it was found to be less suitable for the processing of evaporator concentrates. Based on Belgian experience, a pilot plant for the bituminization of radioactive waste was constructed in Poland between 1969 and 1971. Its capacity is 0.025 m3/h and it is used in sludge bituminization. In the USSR bituminization at elevated temperature was first tested at a Moscow radiation protection station involved in radioactive waste solidification and burial in a pilot plant with a capacity of 0.025 -0.040 m3/h. The results have made it possible to build an industrial plant with a maximum output of 0.07 m3/h. Bituminization proceeds in a reactor-mixer, into which the molten bitumen is fed from the melter together with evaporator concentrate. The evaporated water is condensed in a condenser and is further scrubbed together with the uncondensed gases. As soon as the required salt content has been attained, the bitumen mix is discharged into a storage container. Bituminization at elevated temperature is relatively simple and the installation is easy to operate. Its disadvantages are the need for large heat transfer surfaces as a result of the poor thermal conductivity of bitumen and the requirement of maintaining the temperature in a narrow range in order to prevent overheating of the melt and formation of incrustation. The bitumen mix may contain up to 4 0 % (w/w) of solids; if more solids are present discharge of the melt from the bituminization vessel is difficult. In view of its discontinuous nature, the process has only a limited capacity and is suitable mainly for the bituminization of small amounts of sludges using installations with a capacity of up to 0.04 m3/h (40 l/h). Most continuous bituminization facilities in the USSR use film evaporators with molten bitumen. In a number of locations, different installations are operated

148

with capacities of 0.05, 0.15, 0.23 and 0.5 m3/h. A standard bituminization line consists of storage containers with molten bitumen, tanks with radioactive wastes, steam-heated pipes and fittings for bitumen, a film evaporator, condensers, separators, packing columns and absolute filters, dosing pumps and equipment for drum filling and handling. Concentrates and the molten bitumen are separately proportioned by dosing pumps into the top part of the vertical film evaporator and evenly spread by rotor blades on the heated evaporator wall. Water is removed in the bottom section of the evaporator at 160 "C while water vapours are passed via the separator to the condenser. The resulting product contains 50 - 60 % of salts. The bituminization of radioactive wastes including the mechanical removal of water at temperatures up to 100 "C has been developed in France. The process consists of three stages: - mixing of the sludges, the reagents and the bitumen; - separation of the water released from the coated material; - complete dewatering of the coated material. In the first stage, surface-active agents are added to the mixture of bitumen and sludges. The surfactants lower the surface tension of both the aqueous phase and the boundary between the water and the bitumen, thereby emulsifying the bitumen in the aqueous sludge suspension. The emulsion formed is unstable and breaks down almost instantaneously, while most of the water (75 - 83 %) is removed from the solids-bitumen mixture. When low-level sludges are processed, water is separated and discharged without further treatment. When wastes of higher activity are processed, however, the water must bz treated prior to discharge. The material still contains 7 -20 % of water in the second stage. The temperature of the mixture is therefore increased to 130 "C in order to evaporate the residual water and to attain better homogeneity of the mixture. At temperatures above 130 "C volatilization of the isotopes Io6Ru and 13'Cs would result. The bituminization process is carried out in equipment which was developed from a conventional plastic extruder with two screws rotating in the same direction. The extruder consists of four sections (Fig. 22). The cold sludge, the surfactant and the hot bitumen are mixed in the first section, and the coating of the material and the separation of the water are carried out in the second section, which consists of kneading and transport elements. In the third section the water is released. A mixture of sludges with bitumen and water is passed successively through lowand high-pressure zones. In the low-pressure zone, the material is moved with a large-pitch screw at a faster rate and the thread space is therefore never completely filled. The high-pressure zone consists of a shorter screw with a reverse pitch i n addition to a normal transport screw. The accumulated material acts as a seal. The water is forced back from the high-pressure zone to the low-pressure zone, from which it flows to a collecting reservoir. In the fourth section the material is heated to 105 - 110 "C and then fed to the dehydration machine, which is based on the same principle and consists of two cylinders rotating in the same direction inside a metal

149

housing. The heat produced by friction of the mixture against the screw blades and the walls of the chamber, together with the steam heating of the housing and screws ensures a continuous flow of the material at a temperature of 130- 140 "C and, almost complete dehydration of the material (less than 0.5 % of water). The material is discharged from the dehydration machine into 0.25 m3 drums. In France, several such units are being used for treating wastes of various activity. The capacity of the low-level bituminization unit was 600 kg/h. Sludges

ag ents Bitumen

Electric motor

Fig. 22. Flow chart of cold biturninization

As good experience was gained in operating the extruders, an installation of the same type was also used in the bituminization of intermediate-level wastes with a volume activity up to 3.7 PBq/m3 (lo5Ci/m3). The installation consists of a hot chamber and non-active zone and uses Mexphalt R 90/40. The hot chamber is divided into two cells: evaporator concentrates are treated in the first, and the second is used for the bituminization of sludges. The non-active part consists of a bitumen section and a section for vapour and water supplies, containing also the control panel. The installation operates at 130 "C and processes 70 kg/h of concentrates. The hot part of the bituminization machine consists of nine steel sections provided with two screws rotating in the same direction. The rotation speed is continuously variable between 0.5 and 5 revjs. The upper part of the casing has four openings for the extraction of vapour in condensers. At the end of the machine is a steam-heated device for discharging the product. An industrial extruder-type bituminization plant is also in use in the GFR, consisting of ten successive housing, each of which is separately heated by steam. Supported in the housings are two helices of length 5 m and diameter 0.12 m. Concentrate and bitumen are dosed into the first housing simultaneously, and the mixture is disturbed as a thin film on the threads and conveyed into the equipment. The vapours are removed into a condenser and the mixture of dehydrated wastes

150

and bitumen, containing up to 60 (w/w) of salts, is extruded into drums by the end part of the machine. At the EUROCHEMlC plant at Mol, extruder-type bituminization equipment is in operation for the solidification of intermediate-level concentrates from the chemical decanting of different types of spent fuel elements. The concentrates have a volume activity of 15-37 TBq/m3 (400- 1000 Ci/m3) and are chemically pre-treated prior to bituminization. A mixture of sludges and water is passed together with bitumen into an extruder with a capacity of 0.14 m3/h of evaporated water. The extruder is provided with a steam-heated case and is longitudinally divided into three separately heated zones. Above each zone is a dome for accumulation of steam. Steam is bled and condensed in each zone separately. The condensate contains 0.05 % of bitumen oil, which is removed by filtration. After dosimetric measurement, the condensate is discharged or further processed. The extruder is also provided with a supply pipe for decontamination with tetrachloroethylene and water. The bitumen product is discharged into 0.22 m3 metal drums, specially modified for a minimum storage period of 50 years. A turntable under the extruder houses six drums. In the USSR the concentrate is dried separately before i t is bituminized using an extruder. Chemical sludges are first dried in a rotary drum drier at 130 "C and the dried product is introduced into the extruder together with bitumen heated to 160 "C. The final product contains 70 % of salts and is discharged into storage containers at 130 "C. Another two-stage bituminization unit uses a rotary drum drier at 650 " C . The gaseous phase passes successively through a cyclone separator, condenser and absolute filters. The dried product is introduced into an extruder together with bitumen heated to 160 "C. The bitumen product contains about 80 % of salts and is discharged into transport containers with a capacity of 0.5 m3. The initial application of extruders in cold bituminization had several advantages. The use of various amount of different types of surface-active agent makes possible the processing of various crystalline and amorphous sludges with a higher water content. This results in major savings in running costs compared with bituminization at elevated temperature. The major disadvantages are the high cost of the equipment and the presence of surfactants in the product. The surfactants increase the water content and accelerate leaching from bitumen blocks. These types of bituminization plants are mainly suitable for processing large volumes of sludges whose compositions vary slightly. The advantage of the extruder method of bituminization is the completion of the whole process in one operation, small dimensions of the machine, a very short bituminization time (1.7 min), thus avoiding destruction of the bitumen, and a high salt content in the product (60%, w/w). Evaporation proceeds on a thin film, which precludes violent reactions and formation of foam. Unlike the situation in cold bituminization, concentrates can also be processed. The main disadvantage is the high cost of the equipment.

151

The third category of bituminization plants is based on the. use of emulsified bitumens. For small waste volumes, a batch process using an evaporator and a mechanical mixer is employed and large volumes of waste are processed continuously using a film evaporator. Both of these processes were developed in the USA. They have, however, never been employed there in view of the very high activity of wastes. The batch process includes the following operations: - mixing wastes with emulsified bitumen, - water evaporation at 160 "C, and - draining the bitumen product into suitable containers. Equipment was used in the USA that consisted of a stirrer (1.7 revls), a system for heating to 160 "C, bitumen and waste inlets and a discharge outlet. In continuous operation, liquid wastes and bitumen are introduced into the upper part of film evaporator where the mixing and evaporation of the material take place. The mixture flows down the walls of the evaporator, which are heated to 160 "C, and is spread by mixer blades rotating at 300 rev/min. Hence the heat transfer from the evaporator walls to the bitumen mass is improved. In the USA, experiments were first carried out using a wiped-film evaporator with 0.37 m2 of heat transfer surface and a capacity of 0.015 m3/h. Mixtures were produced containing approximately 60 % (w/w) of incorporated materials. A pilot plant was later constructed for testing the bituminization of non-active concentrates and later still also of actual wastes with a volume activity of 48 TBq/m3 (1.3 x l o 3 Ci/m3). The basic apparatus of the pilot plant consists of a wipe-film evaporator with 4.65 m2 of heat transfer surface, a feed and condensate surge tank, a bitumen storage tank and pumps. Emulsified bitumen and concentrates are brought separately to the top part of the evaporator, where they are mixed by rotation of the distributor plate, which is part of the evaporator rotor. Centrifugal forces move the mixture to the perimeter of the distributor plate, from where it is sprayed over the walls of the evaporator. The wiper blades are made of plastic and are provided with downwards-directed slots so as to facilitate the movement of the material in the evaporator. Water vapour passes through the centre space of the evaporator, where entrained drops and particles separate, and then proceeds to the condenser. Film evaporator bituminization was originally developed for high-level waste processing but could not be used owing to the insufficient radiation stability of bitumen. It has not so far been used in the USA for the solidification of intermediate-level wastes. In Czechoslovakia, a pilot plant for the bituminization of sludges and concentrates is in operation. The basic apparatus is a LUWA P 210 film evaporator with 0.8 m2 of heat transfer surface. The evaporator is steam-heated to 145-155 "C at a pressure of 0.5 -0.6 MPa (5 - 6 kp/cm2), The evaporator is fitted with a fourblade rotor powered by a motor with a speed of 24 of 16 rev/s. A special aerosol separator is part of the upper section of the rotor. Emulsified bitumen is pre-heated

152

to 60-85 "C and is proportioned into the evaporator separately from concentrates. Water vapour passes to an 8 m3 condenser while the uncondensed gases are discharged into the atmosphere from a stack via a filter. The bitumen product is discharged from the evaporator into 0.1 m3 drums. The application of emulsified bitumen instead of bitumen makes it possible to reduce the temperature from 230 "C to 150 "C,thereby eliminating some disadvantages of bituminization at elevated temperature, including the risks of incrustation and fire. In addition, foam formation and release of radioactivity with the gaseous phase encountered during evaporation can be considerably limited. Installations with a higher output can therefore be built. The disadvantage is that the emulsified bitumen contains 60 % (w/w) of water, which must be evaporated together with the water present in the wastes, thereby reducing the useful capacity of the installation and increasing the operating costs.

6.2.5

Evaluation of Bituminization

The bituminization of intermediate-level wastes meets the requirements of radiation protection and its costs are acceptable. The main benefits are: - the starting material (bitumen) is reasonably cheap, - the temperature applied in the various method of bituminization is in the range 130-230 "C, so that the process is relatively simple in comparison with vitrification, - bituminization can be used for wastes with a wide range of activity, i.e. from wastes with a threshold activity that cannot be safely discharged to wastes with an activity of up to 370 TBq/m3 (10' Ci/m3), - the resistance of bitumen to leaching of the activity of incorporated radionuclides is relatively high and costly plants therefore need not be built for storing the bitumen blocks. On the other hand, bituminization has certain constraints: the radiation effects decrease the maximum activity that can be incorporated into bitumen, the required properties of the final product limit the total amount of solids in bitumen. The selection of the equipment and of the bitumens always depends on the type and the amount of wastes involved, on local conditions and on economic considerations. The choice of a suitable bituminization process can be based on operating experience. The bitumen process at elevated temperatures is suitable for processing small amounts of chemical sludges solidified to a salt content of 20-40 % (w/w).

-

-

153

Emulsified bitumens are satisfactory for large amounts of evaporator concentrates or chemical sludges. Cold bituminization allows the processing of large amounts of sludges of similar composition. Extruder-evaporators or extruders combined with suitable types of driers for sludge dewatering have proved to be the most suitable apparatus for the bituminization of large amounts of sludges and of concentrates with a variable composition.

154

7 Solidification of High-Level Radioactive Wastes

Large volumes of solutions from fuel reprocessing plants containing radioactive wastes are the major hindrance to their economic and safe disposal. A 1000MW(e) power plant produces annually approximately 30 t of irradiated uranium, containing about 1 t of fission products and 250 kg of plutonium. In this case, wastes from a fuel reprocessing plant amount to approximately 30 - 60 m3 of a high-molecular nitric acid solution containing about 1000 kg of fission products and 1 kg of plutonium and other active and non-active additions. Thus, after treatment, the waste solution contains concentrations of roughly 100 - 500 kg/m3. Solidification in the broad sense of the word is the conversion of liquid effluents into solid form. In most instances this means in practice the removal of water present in the effluents. Processes using mere evaporation of water were discussed in Chapter 4. The resulting matter, i.e., a mixture of inorganic compound salts, is readily soluble in water and is not, according to current views, suitable for longterm storage. All currently studied fixation technologies for radioactive wastes use the conversion of materials contained in the initial solution into a relatively insoluble compact form that meets a number of requirements based on safety and economic criteria. In these waste treatment methods chemical conversion occurs mostly of the initial compounds associated with the formation of crystalline melts or vitrious materials, or of combinations thereof.

7.1 Properties of Fixation Products The choice of criteria applied to the properties of the fixation products is conditional on a number of aspects. In view of their different significances it is very difficult to find a basis for their comparison. Two major groups of criteria should be considered, viz., safety and economics. The economic criteria include such aspects as the production cost of the equipment (the cost of construction materials and the cost of production itself) and running costs (including the cost of transport

155

and the cost of long-term burial). The safety criteria include aspects related to the safety of transport and the risk of long-term controlled or uncontrolled burial. In connection with environmental control, the safety aspect has of late become a widely discussed item, affecting the basic criteria applied to the choice of the fixation product and the technology itself. A time span of about 700 years, considered as a reasonable time limit for high-level radioactive waste burial during which no major accident should occur, does not permit the use of materials or storage of wastes which do not provide sufficient safeguards against the release of large amounts of activity or toxic substances into the environment. Similar considerations, also based on current technological possibilities and on the knowledge of the materials concerned, have finally led to the following requirements: a) Volume reduction. This mainly economic aspect requires that the volume and wzight of the product to be buried should be as low as possiblz.The most significant volume reduction is achieved by dewatering; further processing however, leads to volume and weight increases owing to admixtures or structural materials generally (container, product matrix, filler materials, etc.). The product geometry, which is affected by the amount of thermal energy released as a result of radiation absorption, allows only a limited reduction in overall weight. The maximum permissible product dimensions affect the geometrical configuration and thereby the use of the burial area. b) Hydrolytic resistance. The possibility of accidents requires that the product should have a low overall solubility and a low leachability of critical radiotoxic nuclides. c) Thermal conductioity. A high thermal conductivity of the product makes it possible to increase the specific activity of radioactive materials and overall product weight. d) Compactness and tensile strength. Impact resistance and high mechanical coherence ensure handling and transport safety and reduce the risks of accidents in the storage area. The properties of stored products are also related to demands on the properties of packaging materials. In some instances the poorer quality of the product can be offset by the quality of the container so that the packaged product meets the requirements on the product. The specific properties of a good container are as follows: - high corrosion resistance, - design ensuring leak resistance of the system, - good adhesion of the product to the container, ensuring high thermal conductivity, - low weight and low cost of the container, - sufficient mechanical strength.

186

7.2 Solidification Process Requirements Of the general criteria, safety and economics should again be emphasized. The effort applied to reducing the cost of high-level waste disposal is often excessive. This is evident from the trend to restrict the cost of wasteLdisposa1 to such an extent that it should not exceed cca. 1 % of the overall energy production cost, or 10 % of the fuel cycle cost. In view of the requirement for minimizing environmental contamination these small sums appear to be insufficient. The handling of large amounts of radioactive materials whose activity in one block of the final product reaches several EBq requires simple remote-control equipment. The design of this equipment should observe the principles that apply in other workplaces where high activities are handled. The product resulting from this process should have certain optimal properties and it should not become the source of excessive amounts of other radioactive substances which would have to be re-treated; this mainly applies to the escape of gases, vapours and aerosols, which may be reduced by modifying the temperature regime or by changing the process.

7.3 Final Form of Fixation Products The chemical composition of radioactive wastes and the criteria determining the optimal properties of products for burial are the basic factors affecting the selection of a suitable system given by the chemical composition and structure and thus the choice of a certain technology. The chemical composition of waste solutions is given by the ratio of the following basic constituents: a) fission products with a relatively constant proportion of radionuclide activities, b) corrosion products; their qualitative and quantitative compositions differ depending on the materials used for the manufacture of the apparatus and on the aggressivity of the environment, c) compounds added in the course of fuel reprocessing operations, d) in some instances, dissolved structural parts of fuel elements, e) additives converting the waste mixture into a material (product) of a vitreous or crystalline nature. Chemically, all of the above mentioned groups primarily contain metal cations of elements that form soluble nitrates. Additives are used in the fixation process, because in almost all studied technologies nitrates or their oxides formed by calcination are converted into compounds with more appropriate properties.

157

The resulting material can be classified into three groups according to its macrostructural parameters: a) homogeneous (or, more accurately, macrohomogeneous), b ) heterogeneous (dispersion), c) compound systems. The flow chart below shows a more detailed review of the most intensively studied products: -synthetic -silicate

systems

rock-based glass

~

-glass -glass

A macrohomogeneous systems

--I

--

-phosphate

glass

1-combined

glass (BSi, PSI, PbSi, etc.)

,I -cermets

silicate or phosphate melts or sinters (ceramics)

-glass-ceramics

waste oxides-

B heterogeneous systems __

~

-glass

ceramics

~

Emetal matrix in: ~-----ceramics -graphite glass

C compound systems using simple or compound container systems

The most frequently studied systems have for many years now included amorphous (vitrious) or crystalline (ceramic) silicate- or phosphate-based materials. Owing to the varied and complex initial composition of wastes, numerous solutions exist, each having a wide range of benefits and constraints. Gsnerally, silicates and phosphates have been selected for the relatively low solubility and stability of the compounds which they form with most cations present in wastes. For the conversion of wastes into silicates, experience has mostly been gained in the glass-making industry with silicate-based glass formation; in some instances the process is based on analogy with rock or mineral cornpasition. Silicate glass has a crystal lattice formed by a continuous lattice of Si04 tetrahedrons. Other elements present in wastes or added during the process serve as so-called modifiers and interm:diary elemmts. Modifying elemznts (mostly alkali metals) do not take part in lattice formation but only fill vacancies in the skeleton; they partly break bonds (by saturation of parts of oxygen bridges) and thus affect the physical characteristics of the glass. An increase in the concentration of most modifiers results in a decrease in the glass forming temperature and in the viscosity of glass in the operating temperature range, but also in a reduction of mechanical

158

and chemical resistance. Intermediary elements are not glass-forming but take part in the formation of the basic polymer skeleton in association with glassforming anions. Boron oxide plays an important role in the technology of silicate glass formation. Its presence reduces the glass melting temperature and crystallization capacity, which is a significant factor for the reduction of devitrification capacity. This is affected by the relatively high product temperature and by radiation, B 2 0 3 has some effect on increasing chemical resistance and reducing the thermal dilatation coefficient. Some studies have cited the unfavourable effect of B203 on the escape of ruthenium (29). Boron glass is very sensitive to the concentration of alkali modifiers present. Boron oxide is formed by BO, triangles. On increasing the alkali concentration, boron is converted into more stable BO, tetrahedrons, which are bound via B 0 3 groups rather than directly to one another (as in SiO,). The skeleton starts to be broken only when the Me,O concentration (where M,e is a metal cation) exceeds 16.7 mol-x, when non-bridge oxygen begins to be formed and when the properties of boron glass change rapidly. The presence of SiO, partly alters this situation, as both tetrahedrons link. However, the so-called B,O, anomaly is also manifest in this glass. This is used such that suitable B 2 0 3 concentration serves as a melt in glass making and, after cooling by transformation to BO, reinforces the basic borosilicate skeleton. Interesting and varied systems are formed in the presencc of aluminium oxide. Depending on the Na,O : Al,03 ratio, aluminium can take part in the formation of the glass lattice in the shape of AIO, or it can serve as a modifier in a hexagonal coordination (AIO,). When present in wastes, such as those from dry fuel reprocessing using fluoride technology, aluminium can be used in the formation of relatively resistant silicate or borosilicate glass. Otherwise, aluminium is mostly added together with silicon in order to form glass or a synthetic rock with a vitrious or crystalline character. Pure chemicals or a mixture of chemicals with natural raw materials are used as starting materials. The aluminosilicate glass systems are very complex, owing to the presence of alkali metal and alkaline earth metal modifiers. They have been thoroughly studied in the glass-making industry. The ratios (Me,O - Al,O,)/B,O,, (Me,O + BaO - Al,O3)/B,O3, etc., determine the presence of AlO,, AlO,, B03 and BO, groups, of which groups with a tetrahedral configuration enhance chemical stability. Good properties, suitable for the fixation product, have been observed in a number of aluminoborosilicate systems. An example is a glass with following molar ratio of components: 4 SiO, : 2 B,03 : A1203 : 3 Na,O : 2 CaO. This system has been chosen as the basis for glass preparation using the natural raw material Filtrolite (polagonite tuff from the Eifel Mountains in the GFR), which in addition to B 2 0 3 contains the above components in a similar ratio. The presence of lead oxide in silicate glass reduces the melting temperature and

159

crystallization capacity (devitrification) and, to a certain extent, also gives increased chemical resistance. Reduced solubility of lead glass has been observed up to concentration of PbO of 30 %, when it reached the maximum. The major constraint with lead glass is its high specific gravity and volatilization of lead oxide in the manufacture of glass. This type of glass may find application in waste processing technology as a matrix for the vitrification of wastes in the form of another glass, ceramics or calcines. Of the most commonly occuring anions, phosphates have a similar effect to the silicate anion in the preparation of glass or crystalline products, i.e., the capacity to cause polymerization and low solubility of most compounds. The chemistry of phosphates and polyphosphates is very complex and less well known than that of silicates, mainly from the point of view of glass making. Some phosphate properties have, however, attracted attention for use in obtaining fixation products with suitable properties. The advantage of phosphate systems, especially phosphate glass, include a low processing temperature (often less than 1000 "C), low viscosity in the operating temperature range and the capacity of phosphate systems to bind different anions without demixing the glass, as occurs with silicate glass. This is due to the structure of the PO4 tetrahedron, due to which phosphate glass can include intermediary elements in its structure. In contrast to the S O 4 group, relative to the phosphorus + 5 charge, one of the oxygen atoms is a non-bridge oxygen because it is bound to phosphorus with a double bond. There is therefore one free valence with a positive charge on phosphorus, which allows linking to a number of other elements or ions (fluorides, sulphur, etc.). On the other hand oxygen which is bound by a double bond to phosphorus is not capable of forming a bridge bond, which partly reduces the mechanical strength of phosphate glass. Phosphate glass is used to advantage for the removal of inert A1,0, in fluoride fuel reprocessing when the PO4 and AlO, charge defects are compensated; the fluorides present are readily bound to the positively charged phosphates and fluorophosphate glass is formed. Information on the formation of phosphate glass is still scarce. However, the metaphosphates of cations formed from the original orthophosphates via the diphosphate stage by heating at 700 - 800 "C seem to play a major role in the production of the glass. In addition to the above advantages, phosphate glass produced from wastes has one inconvenient property, viz., it is subject to devitrification in a temperature range as low as 550-700 "C, which is the temperature of the central parts of stored product blocks. Devitrified products show low chemical resistance; in some instances their hydrolytic resistance is .reduced by as much as three orders of magnitude. Technologically, phosphate glass has another inconvenient property, viz., its significant corrosive effect. This effect is so great that only a few commonly used building materials are resistant to it. Phosphate glass could more be generally used in systems reducing the above

properties. The introduction of SiO, groups into the skeleton of phosphate glass to reduce corrosion and increase chemical resistance have been partly successful. The presence of B,03 in phosphate glass further reduces the melting temperatures and the widening of regions where glass is formed. By the application of this technique phosphate glass would, however, lose the technological advantage consisting in the use of a single additive in the form of a solution. The main disadvantage of all of the above glasses is their capacity to change from the unstable amorphous state to the stable crystalline state, the so-called devitrification. In most glasses the devitrification rate is negligible under normal conditions. The conditions in fixed wastes (heat and radiation effects), however, significantly accelerate this process. Crystalline materials that do not basically change their structure even after long-term exposure to elevated temperature and radiation have therefore been studied and tested. These crystalline materials include a number of silicate-, aluminosilicate- and phosphate-based melts, both synthetic and natural. Fine-crystal silicates containing S O 2 ,A1203,B 2 0 3 , Me,O, MeO, and eventually TiO, , ZnO, etc., obtained by controlled crystallization show outstanding qualities, mainly mechanical strength and hydrolytic resistance. Similarly, phosphate ceramic materials are also used as the final product of fixation. These systems have the relatively low melting point of approximately 900 "C and are formed by a mixture of orthophosphate crystals. The measurement of their properties has shown that these ceramic materials consist of a finely crystalline mass similar to glass of a similar composition. A particular type of phosphate ceramics is the product of the LOTES process, consisting of a mixture of wastes incorporated at the relatively low temperature of 350 "C in AlPO, granules. The crystalline (ceramic) materials also include natural rocks, of which the greatest attention has been paid to basalt. Extensive experience has been gained with basalt in industrial use. Basalt, in both the crystalline and glassy forms, has a high chemical and mechanical resistance. It is an extremely cheap raw material. In order that the proportion of wastes incorporated in basalt-like structures should increase, synthetic basalt and other rocks of similar composition are being developed. Research is following the same lines as that into aluminosilicate ceramic materials. As yet, little research has been carried out on cermets, i.e., a combination of metals with ceramics. The advantage of such systems would be increased thermal conductivity and hydrolytic resistance; the disadvantage would be a higher product price. Otherwise, this type of product resembles dispersion systems, which will be discussed below and from which it differs only in the size of the domains. The group of macrohomogeneous materials also includes glass-ceramics, i.e., materials with both phases present. In practice, this mass is often produced anyway during the preparation of the above-mentioned types of glass or ceramics which almost always contain a proportion of the latter component. It may be said that

161

the amorphous glassy proportion of such a mass deliberately prepared usually has a lower chemical resistance than the crystalline component. Technologists have tried to eliminate the disadvantages of all of the above systems by combining different masses. Such heterogeneous systems combine the optimal properties of solidified wastes and the ideal properties of the “extrinsic” masses. The fundamental bmefit of dispersion system is that the matrix material in which waste is dispersed may be selected so as to meet the most significant criteria of properties that are not significantly affected by further treatment. Masses are therefore chosen showing high thermal conductivity, chemical resistance and mechanical strength. High resistance to elevated temperatures and radiation is also desirable. Scheme on p. 158 shows basically all possible combinations of dispersion systems. However, only a few of them are being studied. A typical example is the use of a metal matrix from fusible alloys in which the granules of the waste fixation product (glass or the intermediate product, i.e., calcine) are dispersed. Their unfavourable properties, such as low mechanical and chemical resistance, are “shielded” by the action of the metal. In addition, the thermal conductivity of the metal matrix is higher by approximately one order of magnitude, which makes it possible to increase the overall dimensions of the product block or the concentration of radioactive substances. Another advantage is the lower operating temperatures needed, while disadvantages include the complexity of remote-controlled equipment and the cost and the weight of the matrix material. This method seems to be a logical solution to the solidification of wastes from fluoride fuel reprocessing in the form of a mixture of fission product fluorides and inert materials used in the equipment for technological purposes (e.g., A1,0,). The use of a suitable container represents a technologically more feasible technique, leading to the shielding and sealing of solidified wastes with high-quality material. For this reason, it is more appropriate to consider the product-container system as a whole, as is the case in dispersion systems. Principally, product containers can be classified into two types, viz. one that is part of the technological equipment and in which the whole solidification process or at least its final stage., i.e., melting, proceeds, and one that is a container for the molten product mass. Contact with the glass or any other melt (and eventually with all intermediate products in the course of solidification) places high demands on the corrosion resistance of the material used. When the container should compensate the poorer qualities of the product, it should withstand the possible corrosion resulting from the action of mineral waters on the storage site and radiolytic products affecting its surface. Adverse to these demands is the economic factor, which induces most technologists to use inexpensive metal materials. In pilot plants and during trial operation common steels with increased chromium and nickel contents have so far been used. Apart from their relatively low cost, metal materials can easily be machined and hermetically sealed using the welding technology developed for work in hot

162

laboratories. With the exception of steels with a higher content of carbon or silicon, these materials mostly have very high mechanical strengths and are highly resistant to impacts, e.g., in accidents during transport. Metal materials with ceramic cladding, mostly coatings of different refractory oxides or chemically stable ceramic compounds, are interesting from the point of view of corrosion protection during processing and storage. Suitable technological methods have been developed for these materials. The application of pure ceramic materials is not suitable owing to their low impact resistance.

7.4 Fixation Technologies There is a wide choice of products with properties suitable for long-term burial. It is difficult to compare the advantages and disadvantages of the described fixation systems and specialists have so far not agreed on the optimum fixation product. The asessment is influenced by the approach of society (e.g., consideration of the hydrological conditions of particular country) and by the possibilities of implementing a certain process. A number of technologies are therefore being developed in many countries (see Table 25). The fixation technology itself may be simplified depending on the main chemical and physico-chemical processes taking place simultaneously with solidification. Solutions of high-level wastes from reprocessing and mixtures of several types of wastes are mostly strongly acidic, with high nitric acid concentration (see Section 2.3). In view of the high corrosion effcct of nitrates and for reasons of economy, the denitration process is the first stage in almost all technologies. After nitric acid removal, heat treatment of wastes takes place in which water is first removed by dehydration, the nitrates of all waste fractions except for some thermally stable anions are converted into oxides by calcination, with subsequent melting in which a reaction is induced by high ttmperature between the waste fractions and additions used, resulting in glass or ceramics formation. Some processes include in the final stage dispersion in a suitable matrix. The principle of the process with possible logic pathways is shown schematically in Fig. 23. The stages shown in che boxes represent processes and not specific technological equipment. Some of the technologies and equipment associate some of the above processes while others conduct them separately. A comparison of two completely different solidification technologies will szrve as an example: first a very simple process of mzlting glass or ceramics in a crucible or in a cylindrical metal container heated to a temperature corresponding to the transition from calcination to fixation. A waste solution with suitable additions is

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Table 25: Survey of Solidification Processes for High-Level Wastes Name of process 1

Calcination in crucible Calcination in fluidized bed Spray calcination Film evaporator calcination Spray solidification Phosphate glass PHOTO Continuous single-phase process ESTER Two-phase process Rising level method Spray solidification Two-phase process ESTER PIVER AVM continuous process FINGAL HARVEST VERA Rising level method Single-phase continuous process LOTES PAMELA FIPS THERMITE THERMALT STOPPER

Country 2

USA USA USA USA USA USA GFR USSR Italy USSR USA USA USSR Italy France France UK UK GFR India USA BELGIUMEurochemic BELGIUMEurochemic GFR GFR USA USA

Product 3

calcine granulated calcine calcine calcine phosphate glass phosphate glass phosphate glass phosphate glass phosphate glass phosphate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass borosilicate glass phosphate ceramics phosphate glass in metal matrix silicate ceramics ceramics aluminosilicate silicate or aluminosilicate

Operating % temperature (w/w) of oxides (K) 4

5

specific density

Thermal conductivity

Leachability

(g/cm3)

(J/SmK)

(P/cm' dm)

6

7

a

90 1.123-1.173 673-773 100 8CL100 1.073 80-100 1.073 21.473 30 1.473 1.273 25-35

1.2-1.4 1.O-1.7

1.4-2.4 0.9-2.4

3.0 2.7-2.9 2.6-2.9

4.8-8.4 = 10-4 5.7 x 10-4

20-25

1.173

2.3-3.5

30-50

1.173

2.9-3.1

2&25 20-30 20-30 2 5 4 25 20-30 22-28

30 25-35 25

up to 500

1.273 1.423 1.373 1.323 1.173-1.323 1.373-1.473 1.173-1.323

723 1.273, 673 2.273 2.273

2.7-3.0 2.5--.2.9 2.5-2.9 2.8 2.6 2.5-2.7 2.5-3.0

2.1 5

2.9

x 10-4

x

5 x 10-1 5 x 10-1

10-2-10-4 10-4-10 - 6 10- 5-10- 7 10- 5 10-5-10-6

10-6-1 0- 7 10-5-10-7 10-5-10-7 10-5-10-7 10-5-10-7 10-5-10- 7

6.7-8.6 6.8-8.6 6.0-9.6 6.6-9.6

x

6.0-8.4

x

7.2-9.6

x

10-6-10-7

x lo-+

10-6-10-7

5.7

x x 10-4 x 10-4

10-6

10-7-10-8 10-7-10-8

put in the crucible. Denitration, water evaporation and calcination with subsequent melting of glass or ceramics thus take place simultaneously and in a single vessel, which may then be used for storage purposes. The combined action of temperature and highly corrosive chemicals places high demands on the material used for the crucible. The implementation of the process in four separate steps simplifies the problem of materials; however, in such a case the whole equipment is more complex and the probability of failure is higher.

Hi h

level

wipsfar

gENITRA TION

ICONCENTRA nod

++CALCINATION

' i ME1 TING

Fig. 23. Flow chart of solidification technology for high-level wastes

A compromise solution is mostly sought whereby denitration, dehydration and calcination take place in one equipment while melting takes place in another. All of these processes can be intermittent or continuous; naturally, designers strive to utilize the more economical continuous process. Table 25 shows a surveys solidification processes including some variants avoiding the high-temperature stages.

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7.4.1 Denitration The high concentration of nitric acid in waste solutions presents a great problem, especially as concerns the highly corrosive effect of the acid. In addition, the return of nitric acid to the technological process is advantageous for economy reasons. The effort to achieve controlled denitration at relatively low temperatures has led to studies suitable denitration agents. Recently, attention has for many reasons focused on formic acid, formaldehyde and some other denitration agents specific for certain technological processes, e.g. red phosphorus, which is also an additive for phosphate glass formation. The starting concentration of nitric acid in high-level wastes varies and depends on the type of reprocessing. It can, however, reach 8 - 1 I M. Nitric acid reduction with formaldehyde in high-level wastes proceeds with the release of the same reaction products as in the use of other reduction agents. Depending on the starting concentration of HNO, , the following reactions prevail: for more than 8 M HNO,:

+ HCHO

=

4 NO2

+ CO, + 3 H 2 0

(7.1)

4HN03+3HCH0

=

4NO+3C02+5H20

(7.2)

=

HCOOH

4 HNO, 1-lOM:

for less than 1 M : 2HN0,

+ HCHO

2 HNO,

+ 3 HCOOH

=

+ 2N02 + H20 2 NO + 3 C O , + 4 H,O

(7.3) (7.4)

The reaction rate depends on both HNO, concentration and temperature. The presence of a small amount of NO2 is sometimes advantageous for the commencement of the reaction. Denitration with formic acid also depends on the reaction conditions; it proceeds with the formation of CO,, N20, NO, N, or NO2. According to some authors, the following reaction prevails: 2 HNO,

+ 4 HCOOH

=

N20

+ 4 CO, + 5 H 2 0

The two commonest variants have been investigated in detail and can be applied in both the continuous and intermittent modes. The reaction can be implemented in a simple heated vessel or column with mixing. A design has been published of denitration taking place inside a spray calciner. Denitration with red phosphorus, which is oxidized to a phosphate during the process, is associated with the subsequent formation of phosphate glass. The advantages of the process also include the high denitration reaction yield and the reduction effect on ruthenium compounds, which significantly reduces ruthenium escape during fixation. The denitration reaction is considerably exothermal; its heat of reaction amounts to 690 MJ/kg atom (approx. 165 kcal/g . atom), HNO,

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decay products include 39 % N,, 29 % NO2 and 32 % NO (with traces of N,O, and the course of the reaction can be controlled by maintaining the system at a suitable temperature. In some instances, lead oxide or molasses has been used for denitration and nitric acid has also been removed by extraction with polar solvents or by electrolysis (anion membrane).

7.4.2 Concentration and Calcination Even after the denitration process, most cations of waste fractions present in the solutions are electrostatically balanced with nitrate anions present. Such a solution will concentrate into a powder of high-solubility nitrates. Although a significant reduction in volume will be obtained, the resulting product will be highly soluble in water. The conversion of this intermediate product to the final form requires the dehydration and decomposition of nitrates into the corresponding oxides. The decomposition is achieved by heating the intermediate product in calcination facilities of different types to a temperature of several hundred degrees Celsius. Released nitrogen oxides may be trapped using wet filters, and, like nitrous gases from denitration, they may be returned to fuel reprocessing. The final product of calcination is an oxide powder of a polydispersive nature with a different grain size. The design of calciners may vary but is based essentially on drying and calcinating equipment used in the chemical industry. In spite of the obvious benefits and constraints of current calciners, the application of each specific type depends on the properties of the solution to be processed and on the availability of equipment or operating experience. Recently, several types of equipment have been used for high-level waste calcination. The pot calciner is the simplest type of the equipment. The solution is fed to the top part of a vertical cylinder heated to the calcination temperature. Escaping gases, fumes and aerosols are conducted to the gas management facility. The crucible may also be used as a container for the resulting calcine. If this is the case, the procedure for sealing and testing the vessel is similar to that used in can mefting process, which will be described in the section 7.6. The versatility of the equipment and process which may be employed with solutions of different chemical compositions is a great advantage. Spray calciner. Spray drying and calcination is based on atomizing the supplied high-level waste solution by air, steam or other medium. During its drop into the heated calciner space, the fine spray is dried and the fine salt powder settles at the bottom. These calciners may have different shapes, mostly conical at the stack end. Calciners are heated with electric resistance heaters, steam or hot air. The advantages of a spray calciner are its high performace relative to its size, rapid starting and possible shutdown and a low dependence on the composition of the solution.

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The basic disadvantages include significant dust emission, requiring the use of cyclones or filter systems. Fluidized-bed calciner. In this calciner the high-level waste solution which is to be dried and calcined is passed into the so-called fluidized bed, i.e., a layer of powder material driven by a flowing gas medium. Aluminium oxide or silicon oxide of a certain grain size maintained in a pseudo-fluid state with a stream of air is most frequently the fluidized-bed inert material. The inert bed particles may be used to advantage as an addition to the final fixation product. Again, the disadvantages include the formation of fine particulates and aerosols. The fluidized-bed space may be heated either indirectly using electrical resistance or any other heating medium or directly by combustion of the liquid or gaseous fuel in the bed. The dried calcine accumulates at the bottom and is transported to the next operation. Finer particles are separated from gases and aerosols by cyclones. Similar to the spray-type calciners, the fluidized-bed calciners are advantageous for their continuous operation, the possibility of automated control and their high capacity. Drum drier. The principle of drying by means of a film (drum) evaporator is based on collecting moist matter or the solution on a slowly rotating heated cylinder, on whose surface water is evaporated. In less than one ratation of the cylinder around its longitudinal axis the matter is dried and falls off spontaneously or is mechanically wiped off the drum. There are several operational variants, classified according to the cylinder design, its location and the type of heating. A vertically mounted cylinder where the solution to be dried flows in the direction of its longitudinal axis has a slightly different function. The fundamental advantage of this type of equipment is a relatively small level of aerosol emission. The survey of solidification processes (Table 25) includes the process and the equipment in the FIPS process developed in the GFR.

7.5 High-Temperature Process The proposed technological processes mostly require reactions that proceeding at temperatures around or exceeding 1000 "C for the formation of a product with suitable final properties. The initial nitrates or oxides from calcination are converted into complex polymer systems which, after cooling, form glass or crystalline materials. Waste materials with additions in the solid or in liquid state are carried to furnaces operating either continuously or intermittently. The furnaces are mostly of the crucible type. The fixation process may also be carried out in disposable vessels which are later used as the final product container.

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The mix can be heated by: a) electric resistance, b) electric induction, c) dielectric means, d) chemical reaction. Electric resistance furnaces are heated indirectly or directly. Indirect heating is achieved using resistance wires of platinum or kanthal, crusilite rods, superkanthal elements, etc. Direct heating of the melt in a continuous-type furnace has so far been utilized only in pilot-plant operation. In this type the melt is heated by the resistance of a passing current (of the order of lo3 A) which is supplied using molybdenum electrodes. The disadvantage of electric resistance furnaces are the difficulties associated with their operation in hot cells. This mainly applies to the replacement of defective parts, such as heating elements and to the disposal of the whole equipment, e.g., owing to wear or accident. This disadvantage mostly does not apply to the induction heating technique. Most of the equipment (frequency changer, transformers, etc.) is placed outside the hot cell while inside it are placed a relatively simple inductor (cooled copper tube coil) and the heated vessel. The inductor failure rate is very low, A current frequency of up to tens of hertz is used for the common types of the equipment. The constraint of induction heating is the problem of suitable metal materials for the manufacture of a vessel, which should resist the concurrent action of high temperature and of a highly corrosive medium (nitrogen oxides, phosphorus pentoxide, chlorides, fluorides, etc.). This type of heating can be used to advantage for a zonal furnace configuration and for heating disposable vessels. Another type of heating showing similar advantages is high-frequency current (dielectric) heating. In this instance, current generators of a frequency of the megahertz order are used which allow the heating of non-conductive ceramic or glass materials. It is thus direct heating of the solidified material itself, permitting the use of corrosion-resistant non-metallic structural materials for the manufacture of the vessel. This type of heating equipment using a high-frequency current, however, may cause electric shock hazards. Heating might also be effected by certain chemical reactions, e.g., aluminothermic reaction, which is being seriously considered for some solidification processes. This technique will be discussed in more detail in the next section. The possibilities have also been considered of utilizing heat released as a result of fission product radioactivity for melting ceramic or glass materials. The high specific activity of the final product would probably make heat removal from t h e stored high-level waste blocks very difficult.

169

7.6 Main Solidification Processes for High-Level Wastes 7.6.1

HARVEST Process

The rising level technique, envisaging gradual water evaporation, nitrate decomposition, silicate or phosphate formation and their transformation to a glass melt in a single device, represents a very simple solidification process. In Great Britain the process was developed under the name Fingal (30). The commercial variant of the process which evolved from it is called Harvest (31). In practice, a solution of high-level wastes together with additives (glass-forming materials, such as silicon oxide or sodium tetraborate) are passed into a heated cylindrical vessel where layers are gradually formed in which the process of solidification takes place. While the vessel is being replenished the reaction layers rise and the vessel itself is gradually filled from the bottom with a glassy mass. Depending on the type of high-level wastes involved a solidified glassy product is formed in the vessel with a maximum radioactive waste content of 25 % (w/w); the remainder is additives, sodium, lithium, boron and silicon oxides. The greatest problem encoutered in this type of process is the material for the reaction vessel, which is eventually used as a container for the fixation product, Alloys with high chromium and nickel contents, such as materials of the HK-40 and Incaloy 800 L types, were found to be most suitable. A toroidal shape of the vessel has been designed, allowing the cylinder diameter to be increased to 1.2 m. The objective is to make possible the storage of a larger amount of wastes in a single block while maintaining the optimum storage temperature. A container of the new design has a capacity of up to more than 2 tons of final product.

7.6.2

Continuous French Solidification Process

A continuous vitrification process which in France is in the stage of industrial plant operation (32) has been relatively well investigated. It essentially consists of a two-phase process with continuous calcination and intermittent melting. Calcination proceeds in a rotary inclined cylinder, electric-resistance heated to a temperature of about 600 "C.Material disintegration during calcination, which accelerates the calcination process and helps form homogeneous material for melting proper, is enhanced by the addition of gas-generating substances, such as azodicarboxamide (NH2 . CO . N = N . CO . NHJ, which also prevents the formation of sticky layers on the calciner walls. Material is also loosened from the calciner walls mechanically. The finely dispersed calcine (particle size smaller

170

than 0.002 m) is then mixed with fritted glass and drops to the melting equipment. Melting may take place either in an induction-heated metal crucible or in a ceramic container where the glass melt is directly heated by a high-frequency current. The molten solidification product is discharged from the furnace via a temperaturecontrolled valve located in the bottom of the melting container. The containers which are also used for storing the final product are made of stainless steel. Currently equipment, is being built a t Marcoule having an annual capacity of 150 m3 of wastes, which is to be used for the disposal of all high-level wastes accumulated so far. Further equipment of this type is being designed for a reprocessing plant in The Hague with a projected capacity of 500 - 600 m3 of liquid wastes per year.

7.6.3 LOTES Low-temperature Process Disadvantages associated with high-temperature waste processing (volatility of some radioactive and non-active substances during the process) are eliminated by a low-temperature process called LOTES(LOW TEmperature Solidification) (33). It consists essentially in the conversion of nitrates present in high-level wastes to the appropriate phosphates by the addition of a stoichiometric amount of orthophosphoric acid, with subsequent distillation of nitric acid. The reaction mixture is heated gradually to 150, 250 and 300 "C, during which process the mixture is transformed into a syrupy mass and finally to a solid cake. Most elements are present in the final product in the form of orthophosphates which are of lower solubility than the initial nitrates. The product is thus suitable for temporary storage or it may be used as an intermediate product for other procedures. It is especially suited for incorporation in metal matrices in the PAMELA process (34). For this purpose, the granulated form is most suitably formed as a result of the interaction of phosphates present in the waste and aluminium phosphate at about 350 "C. The resulting granules are 0.005-0.015 m in diameter, 1600 kg/m3 in specific density, 150 W . m-' "C in thermal conductivity at a content of up to 30 % (w/w) of the initial active wastes in the final product. The hydrolytic resistance of the granules is four orders of magnitude higher than would correspond to the appropriate calcine. Either crucible, fluidized-bed and spray calcination techniques can be used for the LOTESprocess.

7.6.4 PAMELA Process The advantages of dispersion systems led the EUROCHEMIC company to design the incorporation of the granules of a calcine, glass or a ceramic material in a metal matrix. The process, called PAMELA, is still at the stage of the development of the

171

fundamental technological procedure; nevertheless, its prospects, some advantages and disadvantages may already be assessed. A simplified flow chart of the process envisages the production of aluminophosphate granules (see the preceding process), calcine granules or phosphate or borosilicate glass granules. The granules are introduced in a molten metal or alloy by negative pressure in special equipment. With a suitable configuration a product can be obtained in which the metal fills only the gaps between the spherical granules of the intermediate product, thus protecting them from environmental effects, removing heat and shielding fission product radiation. The final product is also mechanically more resistant to impacts. The main benefit, however, is the exceptional thermal conductivity of the metal, whose value has so far been best tested for a lead alloy (84 % (w/w) Pb, 12 % Sb, 4 % Sn) and amounted to 30 W/m "C.Together with wastes, a dispersion product is thus obtained with a thermal conductivity coefficient of 10 W/m "C. Should the removal of liberated heat be increased, other metal materials, such as aluminium, may be used. With aluminium, the thermal conductivity of the final product would be around 30 W/m "C and the incorporation would be possible of wastes having a thermal output of lo6 W/m3 in a block 0.3 m in diameter. The temperature difference between the walls and the centre of the cylindrical block would, in this instance, be approximately 200 "C. Subsequent melting of the metallic material would result in wastes in their initial form, which would allow their reprocessing.

7.6.5

Aluminothermal Process

Another variant of high-level waste fixation is a process utilizing the strongly exothermic reaction of the combustion of powdered aluminium in. g mixture with a strong oxidizing agent (35): Similar reactions have been thoroughly investigated and found to be of practical use, e.g., in welding technology. When the Fe,O, or MnO, is used, more than 840 kJ (200 kcal) of reaction heat is obtained per 2 gram-atoms of aluminium. In the case of active wastes from spent fuel reprocessing the oxidizing effects of the nitrates present can also be considered. The starting material for the aluminothermal process may be a waste calcine obtained using any of the possible methods. Good results have been obtained in experiments with manganese dioxide and sodium nitrate as oxidizing agents in a mixture with a calcine, aluminium and silicon dioxide. With MnO, a product has been obtained showing properties similar to those of borosilicate glass-based products. The main disadvantage of the process is a significant effluent of dust and aerosols, which required the use of arrestors and filter systems and the disposal of other

172

high-level wastes. At the Nuclear Research Institute at Karlsruhe the problem has been resolved by using special coagulators, cyclones and filters. The equipment where the reaction takes place is manufactured of refractory ceramic materials. Some of the disadvantages of the process could be eliminated by carrying out the aluminothermal reaction in a bore-hole drilled in a suitable geological formation.

7.6.6 SANDIA Process It appears that for technological and economic reasons the most advantageous solution would be to separate the high-level wastes radioactive materials which are designed for burial from ballast admixtures, thus obtaining concentrated radioactive waste. Fission products could best be separated using the methods which have recently been developed for separating radionuclides, be it extraction, ion-exchange or co-precipitation methods. A disposal technique for high-level waste has been designed in the USA based on the use of inorganic synthetic ion exchangers with self-fixing properties (36). They include several hydrated oxide-based sorbents of the general formula of M(M:O,H,),, where M is the exchangeable cation and M' is titanium, niobium or zirconium. The NaTi,O,H titanate sorbent has proved most suitable. Its exchange capacity approaches the theoretical value of 5 mequiv/g. This cation exchanger is a good sorbent for almost all fission product cations with increased sorption capacity for ions of higher valency. The process of separating radioactive substances from liquid wastes consists in the initial adjustment of the solution acidity to cca pH 1 with subsequent sorption in a column packed with the sorbent. The escaped anions of radionuclides (e.g., TcO;) are removed using standard Dowex 1 type anion exchangers, and part of the cesium ions by synthetic zeolites (Zeolon 900 Na). The process sludges and used sorbents are mixed and further processed into solid waste. Heating to above 600 "C results in the decomposition of the sorbent to titanium dioxide and the respective titanates of fission product elements which form a polycrystalline ceramic mass. Powder sintering is facilitated by the use of pressure (pellet compacting) or by the addition of glass powders, which also improve the hydrolytic resistance of the final product. The compacted and sintered pellets may be used for waste disposal, for incorporation in a suitable protective material (metal matrix) or for further processing. The described process is especially advantageous for waste containing ballast with a high salt content (old neutralized waste in tanks) or for large amounts of less active wastes from fuel reprocessing plants.

173

7.7 Methods of Monitoring Product Properties and Testing Fixation Process The relatively complex system of radioactive wastes (a wide range of substances, changes in their composition) and high demand on the quality of the fixation process require that different analytical and measurement methods be applied both in developing a technology and in operation. If specialists are to give at least minimum guarantees of product quality for several hundred years burial, they should have the greatest possible knowledge of the microstructure and macrostructure of products, to test for several years their properties and extrapolate and estimate the product’s behaviour in long-term burial from the detailed results of testing. The methods employed may be classified according to whether they are used for the development of materials and technology of production (laboratory and pilot-plant research) or for analytical testing on an industrial scale and for quality control of the final product. The choice of methods is also influenced by their availability and by local experience. Chemically or physico-chemically, the fixation process consists of a number of reactions between many components of initial radioactive wastes and additives. Several elements in the studied systems and relatively little known solid-state reactions make the fixation process a complex analytical problem. The properties of the final product studied may be classified into the following groups: a) microstructure b) macrostructure c) chemical composition d) chemical resistance e) physical properties

7.7.1

Microstructural Analysis

The final outcome of the fixation process is a more or less homogeneous mass of an amorphous or crystalline nature. Absorbed thermal energy and radiation energy induce alternations in the structure of the mass; the best known example is devitrification. Information on the state of the individual waste components can be obtained using the following analyses: X-ray structural analysis mostly involves the study of a powder sample (DEBYE - SCHERER) and permits the ascertainment of the extent of amorphousness or crystallinity, estimating the size of crystals and in a number of instances making a qualitative and quantitative determination of the composition of the crystal phase. It is thus possible to determine the presence of unreacted

174

initial substances (e.g., Fe,O,), the presence of mineral-like crystal phases of the aluminosilicate types, crystal formation as a result of devitrification, etc. A more detailed view of the composition and the distribution of phases of the product is provided by the study of microsections using conventional microscopic analysis, electron microscopy (slides) and electron microanalysis. Microscopy in normal or polarized light allows the determination of the presence of typical compounds or crystals by a comparison with known specimens, measuring the refractive index and the hardness of separated areas. In addition, it yields information on the size of crystal phases and the presence of defects (cracks, pores, bubbles, etc.). Electron microscopy allows the determination of the shape and size of crystals using sections or replicas. Magnification is one order of magnitude higher than in optical microscopy. The use of electron microanalysis, whether with a microprobe or a scanning electron microscope, offers a wide choice of possibilities of analysing the final product or the various stages of treatment. Thanks to the interaction of a narrow electron beam and the studied specimen (metallographic section, fracture surface, etc.), microanalytical instruments make it possible to record recoil, primary or secondary electrons, characteristic X-radiation, sample luminescence, the emf produced and AUGERelectrons. Using this detection technique two-dimensional sample images are obtained of samples with distinctly different surfaces, depending on the contents of elements differing in atomic weight or in the wavelength of the characteristic X-radiation. An analysis along the chosen straight line of the sample (linear analysis) makes possible the semiquantitative analysis of selected element contents in the regions studied. A microprobe allowing the determination of elements with a higher atomic number (greater than 11) with an accuracy of up to 0.01 % of the occurrence of a certain element in a mixture is used for a more accurate quantitative microanalyser can detect defects (bubbles, cracks, etc.) and gives information on the condition of the contact surface between the container and the product. The technique is also suitable for studying the mechanisms of leaching and devitrification of glass. For more detailed studies of the product mass state other analytical methods can be used that are commonly employed for the determination of the structure of materials, such as infrarzd and visible light spectroscopy, emission spectroscopy and measurement of magnetic properties.

7.7.2 Macrostructural Properties The macrostructure can be tested using both des ruc ive and non-destructive methods. Information on the presence of defects in the product, on the compatibility of the product with the container, and on the condition of the container (corrosion, etc.) can be obtained from total sections through the final fixation

175

product (microscopy and visual observation). This technique, especially when used for high-level wastes, is very demanding on the equipment and procedures used. For this reason, ultrasonic testing or neutron radiography should preferably be employed. Ultrasound mainly permits testing of the condition of the container, finding corroded spots and partly also product macrodefects. Neutron radiography offers the possibility of handling highly radioactive material and gives information on the non-homogeneous distribution of substances in the product, on defects in the product, and on container conditions. The method uses the differences in neutron interaction with atom of a different effective cross-section. It is especially suitable for determining the distribution of some typical fission product nuclides in the studied mass. Neutrons passed through the studied specimen are recorded using suitable converters (e.g., gadolinium) from which the induced beta or X-radiation is transmitted to a sensitive film. X- and gamma radiography are suitable only for testing non-active specimens. They are also used for the quality inspection of used metal containers. Some information on the conditions of the fixation product may be obtained by the accurate monitoring of its dimensions (distance between selected points o n the product) using periscopes, prior to and after heat tratment. It is thus possible to determine significant container deformations due to corrosion, thermal effects o r product material expansion.

7.7.3 Chemical Composition The composition of the product or its components can be found by routine methods of qualitative and quantitative analysis, which mostly include destructive testing methods in which the solid mass of the product is converted into a solution. Local analysis of the composition is possible using electron microanalysis or a laser probe (spectroscopy) or typical (mostly colour) reactions on the metallographic specimen surface. It is advisable to combine the methods of chemical composition analysis with structural analysis.

7.7.4 Chemical Resistance Chemical resistance to an aggressive medium is one of the most closely followed and most important product characteristics. Although in the broad sense of the word we are concerned with the environmental impact (moisture content, radiolytic products, etc.), the effect of the aqueous medium is considered to be most significant. If there is an accident during transport or destruction of the disposal area which, considering the duration of disposal, cannot be excluded anywhere

176

on the earth, the aggressivity of the aqueous medium, be it surface, mineral or sea water, will always be involved. A number of methods have been developed by various laboratories and several standard methods have been taken over from industries (e.g. standards for glass testing) or specially approved for the fixation products of high-level wastes which are used for the determination of hydrolytic resistance. These are methods of determining the overall solubility of the mass under study or examining the leachability of certain specific radionuclides. They mostly include long-lived highly radiotoxic radionuclides which are selectively retained by human body organs or tissues (e.g. 13'Cs, "Sr, rare earths and mainly long-lived transuranium elements). For overall solubility determination, a method is used based on leaching a ground sample with water at room or elevated temperature. Following a certain exposure, leached substances are determined by titration (in the case of alkali glass leaching) or by weighing the evaporation product. Depending on the results of these analyses materials are clasified according to their hydrolytic resistance. In addition, the kinetic dependence of the dissolution of the product sample (with a precisely defined area) is sometimes determined; the dissolution rate expressed in kg/m2. day is measured at given time intervals. The leaching of the individual elements is studied either by tracer techniques or by atomic-absorption spectroscopy (this method is suitable for non-active samples when a greater number of leached elements is to be monitored). For this purpose, the International Atomic Energy Agency has recommended a standard static method; a Soxhlet-type extraction apparatus is used, which simulates the dynamic leaching conditions. The leaching process is sometimes very complicated and is the sum of several mechanisms. The leaching rate is affected by the character of the surface, the diffusion on the sample-liquid interface, the diffusion of solid-state ions, the solubility of the individual sample parts, temperature, the experimental set-up, etc. Although many attempts have been made to provide an accurate explanation of the leaching of selected elements, it is convenient in practice to express the kinetic dependence by empirical relationships. According to the type of accident assumed, various aggressive media may be used for model tests which can be conducted at higher temperature corresponding to the actual conditions under which the high-level waste products are stored. Because of the combined effect of many factors, no model can substitute for the actual situation with the actual fission product concentration.

7.7.5

Physical Properties

Certain characteristic properties are determined for the final solidified product which are important from the point of view of safety and economy, or simply serve as information for the control of the technical process.

177

Common properties for which sophisticated methods are not required include the measurement of sample density, porosity, and thermal expansivity. Thermal conductivity is one of the limiting characteristics of the product and is therefore very important. It affects the overall dimensions of the solidified block, the conditions of storage (cooling), state of the product at different points of the unit, etc. It is usually measured as a function of temperature using techniques which are routine in glass-making and ceramics production. Other methods, such as those for determining the softening point or the melting point of the product, the transformation point, and the temperature dependence of dynamic or kinematic viscosity, are techniques originally used in the glass and ceramic industries. The data not only characterize the resulting products but are also important for technology. Masses with a lower viscosity flow more easily from operating vessels and thus fill the containers better. The product impact strength test is important from the point of view of hypothetical accidents (especially during transport), Although the activity of incorporated radionuclides is known in advance and although methods exist for calculating absorbed energy and heat transfer, the calculation for the actual configuration is comparatively complicated. Therefore, calorimetric Aeasurements to determine the amount of heat liberated are conducted in products containing actual high-level wastes. A partial solution may be obtained by placing thermocouples on the block surface. Methods for measuring the properties of the container or the product as a whole mainly include methods for determining escaping gases by measuring the pressure inside the container, leak testing methods, etc. For the operating conditions, the following data will obviously be required: - contamination of container surface, - system tightness after the lid has been welded, - container wall thickness, - outer dimensions, - thermal output, - homogeneity inside the product (distribution of active mass). In the course of solidification (during the high-temperature stage) several chemical reactions take place simultaneously, and transformations, sintering and melting take place together with other transformation processes during mass cooling or during controlled crystallization. Under the experimental conditions, the weight reduction of the product (thermogravimetry) and changes associated with exothermic or endothermic effects (differential thermal analysis, and in some instances also ‘dilatometry) during temperature changes are studied. An’ increasing contribution is made by measurements using emanation thermal analysis based on measuring the temperature dependence of the escape of an inert gas (radon) from the substance under study. In addition to the effects of

178

chemical reactions, it is possible to observe heat liberation or consumption, the sintering process, transformation processes with minimum heat effect, etc. The method is a suitable model for simulating all processes during which volatile substance are released (radon is usually used for this method). The release of volatile fractions, i.e., solidification reaction products, is one of the most significant process characteristics. It consists essentially in the liberation of water vapour, nitrogen oxides (or P,O, in processes producing phosphate glass or ceramics) and fission product volatile compounds, mainly ruthenium (RuO,) and cesium. Volatile fluorides (SiF,, fission product fluorides, etc.) can also escape in fluoride waste reprocessing. In addition to the vapours of volatile substances, a greater or smaller amount of aerosols will escape into the gaseous phase during the process. The measurement of the escape of these substances is one of the basic measurements, both in the development of the process and in its industrial application. The methods for the arrest of these substances are discussed in Chapter 5.

179

8 Use of Radioactive Wastes as Raw Material

In nuclear power production, as in the other industrial fields, by-products must not present an environmental danger but should be reprocessed to the maxima1 extent (37, 38). Considering their weight and the level of activity, the wastes from spent fuel reprocessing are the most substantial source of radioactive wastes. One of the reasons why reprocessing is desirable is the accumulation of fission products in fuel. Some of the fission products have a large absorption cross-section for neutrons and gradually affect the reactor neutron balance. Their removal after some time of operation therefore becomes a necessity. The currently prevailing technology for spent fuel reprocessing is based on liquid extraction. This technique has been developed for thermal reactor fuel, but it appears that when modified it will also be applicable for fast reactor fuel reprocessing. Methods for this purpose have been developed and tried in pilot-plant operation based on the controlled distillation of the volatile fractions of nuclear fuel. The methods consist in dividing spent fuel into fractions containing fissionable nuclides re-usable in the reactor and nuclear reaction ballast products. The composition of the latter fraction depends on the type of fuel and on the reprocessing technology. The differences in composition, especially in so far as macrofractions of mixtures are concerned, are significant. They always represent a varied mixture of non-active and active nuclides of elements originating in the fuel itself and its cladding, from the equipment corrosion products, chemicals and their radiolytic products, and finally, from the products of fission and other nuclear reactions. For more detailed information on this subject, see Section 2.3. The amount of fission products and their proportion in spent fuel depend on several factors, including the degree of burn-up and the time in fuel cooling. Table 26 (39,40) gives some data applying to the British power programme in the year 2000 for an installed capacity of 100 GW(e). Other nuclear reaction products are transuranium element radionuclides formed by reactions other than fission. For practical purposes plutonium, although formed in this manner, is classified among usable fissionable materials. In view of this fact,

180

which is also taken into consideration in reprocessing technologies, plutonium will not be discussed here. Table 27 gives information on the amount of transplutonium isotopes in spent fuel. It applies to a light-water power reactor with a burn-up of 2.16 TJ/kg (25 000 M Wd/t). Table 26: Annual Production of Some Fission Products from an Installed Capacity of 100 GW (e) Activity (Bq x lo1*) Radioactive 85Kr 90Sr(+ 13'Cs (+IJ7'"Ba) 14'Pm 144Ce Stable or very long-live Tc Rh Pd Xe

Weight

0)

0.56 4.4 ($4.4 7 (+7) 28 > 37

0.1 1.2 2.7 0.4 0.8

-

0.9 1.3 3.5 7.9

-

Table 27: Contents of Transuranium radionuclides in 1 t of Light-water Reactor Spent Fuel; burn-up 2.16 TJ/kg (25 OOO MWd/t) Isotope 237NP 24LAm 243Am 242Cm 244Crn

Weight

kg)

384 56 57 13 11.3

For nuclear power production the fraction of ballast products of nuclear reactions containing the group of fission products and transuranium radionuclides is an inconvenient and dangerous waste. Concentrated in the waste is almost all of the activity from the nuclear reactor fuel cycle, whose escape into the biosphere is extremely undesirable from the point of view of protection control. The long-term controlled disposal of the waste in either its original or modified form is the simplest solution and is technically feasible. Obvious shortcomings encounter-

ed in liquid waste disposal (considering the waste from the so far most frequently employed extraction technologies) lead to the development of solidification technologies which concentrate waste in a stable solid state (see Chapters 4 and 7). These technologies undoubtedly meet the demands for the protection of the biosphere. However, they share two fundamental shortcomings: 1. The number of radioactive decays in a unit volume of waste from fuel reprocessing and their average energy represent a time-dependent variable. Shortlived radionuclides with high decay energy, whose proportion makes up the major part of the mixture consisting of up to several hundreds of radionuclides with half-lives ranging from tens of days to tens of years, gradually disappear from fresh concentrated wastes. After the relatively short period of several years, only a fraction of the original total activity remains, corresponding to a few long-lived radionuclides. These radionuclides then are the factor determining the time over which the buried wastes should be controlled, i.e., a period of the order of hundreds of years. The waste processing technology should take into consideration the high initial total activity and its tolerances will therefore be many times greater than those which are actually needed. 2. Efficient waste disposal results in the irreversible conversion of the wastes to a form that is physically and chemically as stable as possible and which will preclude their spontaneous proliferation. Such waste disposal, however, will hinder any later application of the removed nuclides which are of technical value, be it for their properties as emitters or as rare elements. An analysis has shown that, e.g., the amount of rhodium and palladium in wastes today is comparable to the amount extracted from ores. A paradoxical situation may thus occur in the near future when the amount of rare elements returned to underground disposal areas in form of wastes will be greater than the amount obtained from the earth. Such a situation may be precluded by the pre-treatment of high-level wastes before disposal. The pre-treatment consists in the separation of long-lived and radiotoxic fractions from the initial mixture and separate processing of the proportions thus formed. The significantly greater proportion containing short-lived radionuclides and ballast substances is then processed using a simpler and less demanding variant of the solidification technology while the smaller proportion, both in volume and in weight, of pure long-lived radionuclides is either used as a raw material for the separation of valuable components or is solidified using separate low-capacity equipment. The principle of this solution has been known for a number of years (41). A more detailed analysis, however, has revealed some weak points. One is the virtually impossible quantitative separation of long-lived and radiotoxic nuclides from the entire initial volume of the mixture. The decrease in the total activity will thus differ from an ideal situation represented by curve 4 in Fig. 24 (37). Previous considerations were focused mostly on radioactive fission products,

182

and attention has been devoted to alpha emitters only in recent years (42). The amount of alpha emitters in spent fuel wastes from thermal reactors is small compared with fission products, yet alpha emitters present a serious problem owing to their radiotoxicity and very long half-lives. Wastes with a higher alpha emitter content produced by high neutron-flux reactors require many thousands of years for spontaneous decay. This also applies to wastes from which plutonium has not been satisfactorily separated.

?

Fig. 24. Waste toxicity; 99.5 % of uranium and plutonium removed 1 - Total, 2 - Actinides, 3 - Sr Cs, 4 - Fission P,s Peak at lo6 years caused by daughter products not present originally

+

The controlled storage of the main part of waste may be shortened depending on the completeness of separation of dangerous fractions; the problems of safe storage will always be decreased by separation. The difficulty exists of establishing stable relationships between supply and demand for large amounts of radionuclides (43). Unless deliveries of pure radionuclides are guaranteed, there will not be any interest in application; unless demand is guaranteed, the producer has no interest in extending production capacity. The crux of the matter is primarily in the! low level of knowledge on the part of prospective customers and sometimes in their reluctance to introduce new technology. Nevertheless, some progress has been made, especially in cases when the use of radionuclides is economically beneficial.

183

8.1 Possible Application of Fission Products and Transuranium Elements Radioactive and non-active fission products and transuranium elements may be used as sources of electrical, mechanical, thermal and light energy, as radiation source and as technological materials. For instance, "Sr and 144Ce, and to some extent also lo6Ru and 137Cs, are suitable for the manufacture of larger power installations designed for unattended operation in remote areas; portable power units employed in space technology and in medicine utilize 14'Pm, 238Pu,242Cmand 244Cm.The last two radionuclides are especially attractive for applications where high performance and low weight are required: 242Cm,has a specific output of 122 W(t)/g and 244Cm2.83 W(t)/g. asKr may bc used in luminescencs sDurc2 utilizing the direct conversion of ionizing radiation to light energy for filling lamps with a very long service life which do not require an external power supply. 147Pm is used as the activator of luminescent colours. The affects of ionizing radiation on various materials may lead to other fission product applications. With fission products, some technological processes may be simplified or made less costly, e.g., in the controlled polymerization of macromolecular substances; materials may also be manufactured that are unobtainable by other means. In the food and pharmaceutical industries and in medicine ionizing radiation is used for sterilization and disinfection. Increased attention has recently been devoted to food sterilization. Experiences in many countries have shown the efficacy and economic feasibility of the method. Of the available fission products, mainly 137Csis used. Other possible applications are based on the energy of alpha particles and on the spontaneous fission of most transuranium elements. 241Am is a convenient part of the Am-Be neutron source and is also a low-energy gamma source applied for measuring material thickness or atmospheric density. 252Cf is a neutron source many times more efficient than the more commonly used Pu-Be source. It is envisaged that it will be applied in neutron radiography, activation analysis and in point irradiation in medicine. Spent fuel wastes also contain a group of rare elements which can be considered as non-active. They include rhodium, palladium, technetium and xenon. Rhodium and palladium are valuable industrial materials. Rhodium is used as a catalyst in the oxidation of ammonia to nitric acid, and palladium may replace platinum in many instances. Both metals, formed as fission products, contain small amount of the radimuclides lo2Rh and Io7Pdand it is therefore necessary to find by an analysis of the situation whether or not in a particular case their application is permissible within valid safety standards. Technetium, which ba3ically occurs only as the "Tc nuclide with a half-life

184

of 2 x lo5 years, emits soft beta radiation and for its chemical similarity to rhenium may be used as an addition to special alloys. Xenon released during fuel reprocessing in the stage of fuel element dissolution represents a mixture of stable and radioactive isotopes. The radionuclides, however, decay very quickly and after cooling for 1 year xenon is non-active material with versatile applications in electrical engineering and medicine. The high cost of this rare gas so far has been an obstacle to its wider use. Table 28: Properties of Long-lived Radionuclides and Stable Fission Product Nuclides in Wastes

Element or nuclide

Type of radiation

Half-life (days)

in spent fuel at 216 GJ/kg

Applications

Wt)* 85Kr

B

3,800

16

beta emitter, light source

(275 g for all Kr)

90Sr 99Tc

Ru

B B B. Y

104

77 x 106

stable

+ lo6Ru

400

beta emitter, heat source

628

catalysts, alloys

1707

heat source

370

(56 g for Io6Ru)

Rh

**practically stable

337

chemical industry, electrical engineering

Pd

**practically stable

876

chemical industry electrical engineering

Xe

stable

3987

I37CS

104

B1

Y

B

285 950

medicine, electrical gamma emitter (sterilization), heat source

***gg ****67

gamma emitter, heat source beta and X-ray emitter, heat source, luminescence

* Data are approximate and depend on the reactor energy spectrum. ** Rh contains a small amount of active 'O*Rh; Pd contains lo7Pd soft beta emitter. *** After 1 year's cooling. **** After 2.6 year's cooling. The application of fission xenon may dramatically change the situation; its amount obtained from a reprocessing plant with an optimum capacity of 5 t of spent fuel per day is one order of magnitude higher than its amount produced by a large oxygen plant with a capacity of lo3 t of oxygen over the same period of time. The applications of the individual radioactive and stable waste fractions listed above are not exhaustive and serve only as examples. As far as further development

185

is concerned, it is estimated that a decisive role will be played by economic, social and safety aspects. Tables 28 and 29 give some characteristic data on the individual fractions (38). Table 29: Properties of Transuranium Elements in Wastes ~

Nuclide

Types of decay

Half-life of main type of decay (days)

x lo8

237Np

a

8

238pU

u , SF

3 x 104

239pu

a, SF

9

2 4 1 h

a, SF

l.'

x

x

1013

heat source ~

lo5

3 x 106

a, SF

242Cm

a, SF

163

244Cm

a, SF

6.6 x lo3

252cf

a, SF

SF

1.8

Application

target material for 238Pu production

-

2 x 10'8

106

2 4 3 h

-4Cf

Half-life of - spontaneous fission (SF) (days)

I 1

'z I

nuclear fuel, target material for transplutonium elements production

7.3 x 106

neutron source,alpha and gamma source

1.2 x 106

target material

2.6 x 109

heat source, 238Pumother element

5 x lo9

heat source

3 x 104

neutron source

61

neutron source

8.2 Separation of Long-Lived Radionuclides and Separation of Valuable Radionuclides Although a large amount of nuclides is present in the initial mixture of nuclear reaction products shortly after the end of the fission reaction in the fuel, for practical purposes attention is focused on only some of them. Excluding the isotopes of elements formed with only a small fission yield, short-lived radionuclides whose decay has advanced so far that they have nearly disappeared from the mixture, and finally the stable isotopes of elements which are also abundant in other sources, three smaller groups remain. The first includes transuranium elements from neptunium to californium, the second contains the radionuclides of cesium, strontium, promethium, cerium, zirconium, niobium, ruthenium and krypton, and the third consists of stable or slightly active isotopes of xenon, rhodium,

186

palladium and technetium. As regards radiotoxicity, the elements of the first group and cesium and strontium from the second group are the most significant. Methods for the separation the individual fractions from the mixture and their isolation in the pure state have mostly been derived from previously known separation procedures, often from procedures used in analytical and preparative chemistry. The choice of method will depend on whether we wish to remove a difficult or valuable fraction or to obtain a pure product. Both criteria may be applied to the same fraction. In the former instance, the fundamental requirement is complete separation while chemical and radiochemical purity are of secondary importance. The separation of the valuable 239Puwhich is carried out during reprocessing will serve as a good example. If the objective is to obtain a pure product it is essential that the resulting fraction be as pure as possible; the requirement for purity is related directly to the standard quality of the product designed for a specific application. Good examples of this are pure beta emitters, which must not be contaminated with a gamma source, and stable xenon, which does not contain radioactive isotopes of krypton. The choice of the method should be based on its efficiency, economy and feasibility. The radiation and chemical stability of reagents and their corrosion effects on equipment, their explosiveness and flammability are significant constraints. Neglect of these criteria might at best result in protracted operating difficulties, at worst in serious accidents caused by leakage of radioactivity. Compared with fission products, the handling of transuranium elements has certain specific features. It is necessary to prevent the accumulation of the critical amount of these elements, which for some of them in an aqueous medium is only a few hundredths of a kilogram. The increased radiolysis of aqueous and non-aqueous system fractions in a strong alpha radiation field should be taken into consideration and, compared with the handling of fission products which require shielding from gamma radiation, protection is required against spontaneous fission neutrons. The handling of radionuclides has one common feature, viz., their variability with time, unusual for common chemical operations with stable elements. This may on the one hand cause complications, e.g., where for time reasons it is difficult to prevent the formation of undesirable products, and on the other hand it may be used to advantage for spontaneous intermediate or final product purification by the decay of the undesirable admixture. Despite these limitations, a number of methods now exist which have been proven on an industrial scale. They mainly include methods for the separation and isolation of radioactive and stable fission products which have in the past three decades been well documented. Information is scarce on higher transuranium elements whose isolation from spent fuel has been studied for a much shorter time. Considerations in this respect were stimulated by technical progress in the development of light-water power

187

reactors whose operating regime is a prerequisite for the formation of larger amounts of transplutonium elements. As stated above, this does not apply to plutonium, whose proportion of spent fuel greatly exceeds that of all other transuranium elements and whose value as fissionable material for nuclear reactors has been obvious from the start.

8.2.1

Separation and Isolation Methods

In the past great attention has been devoted to the principles and applications of separation methods in radiochemistry, and hundreds of studies have been summarized and reviewed in a recent monograph (44). Of major interest are solvent extraction and chromatographic methods, which are rapid and efficient for the isolation of minute amounts of elements from complex mixtures. Precipitation and coprecipitation methods, which previously were very common, have in view of their lower efficiency and sometimes slow rate given way to the former two methods. Other techniques which should be taken into consideration include sorption, electrochemical distillation and foam separation methods. Extraction methods are often selective and allow the separation of a radionuclide in a single operation. Their advantages include a high reaction rate, a small amount of waste, continuous operation and high productivity; there is a wide choice of automatic and remote-controlled industrial extraction apparatus. A significant constraint on high-capacity equipment is the flammability of extraction liquids and their toxicity. In addition, radiolysis, the product of which reduce the efficiency of extraction, may cause other unexpected complications. The choice and design of an extraction process should consider the interactions between the extraction reagent, solvent, salting-out agent and complex-forming agents, which together form a wphisticated and sznsitive system. Extractants include organic phosphorus compounds, aminzs, substituted phenols, and ketones. Ion-exchange chromatography, although not as rapid as liquid extraction, has been wzll proven for separating elemmts with similar chemical properties, such as rare earths and transuranium elements. The advantage of the method, which also was the immediate reason for its introduction in nuclear chemistry, are its easy operation, high reliability and easy adaptability to various operating modes, all of which facilitate remote control. The initial drawback presented by the strong radiolytic effects on organic ion exchangers has been overcome by the replacement of organic materials with inorganic ion exchangers, which from this point of view are far more stable. Ion-exchange chromatography also has constraints; a lowzr reaction rate and the necessity to use in certain instances large amounts of elution solutions, which presents problems, especially in the technological implementation of the method. One constraint of ion-exchange chromatography stems from the principle of

188

ion exchange itself. This is unsuitability of applying the method to substances without charge and for colloidal particles whose presence in the solution may cause clogging of the pores of ion exchanger grains. Gas bubbles generated by the radiolysis of water may have the same effect. Inorganic ion exchangers, both synthetic and natural, often operate only within a limited pH range whose exeedance results in the destruction of the ion-exchange bed; reversibility of the ion-exchange reaction is not always guaranteed. Synthetic, commercially produced organic resins and inorganic materials consisting of hydrated oxides, aluminosilicates, salts of acids with multivalent metals, complex cyanides and heteropolyacids are used in practice. Precipitation, coprecipitation and crystallization based principally on separating the low-solubility component from the solution are methods which have been employed since the very beginnings of nuclear chemistry. The methods may thus be implemented in laboratory and operating conditions. Also, the large-scale use of this method causes fewer difficulties than the use of other methods and radiation effects interfere very little. Their disadvantage is that impurities are entrained into forming or already formed precipitates or crystals; also, the solubility of these substances is not always sufficiently low to avoid losses. Despite this, precipitation is still routine for separating groups of related fractions from the initial complex mixture and crystallization is widely used for the purification of certain products. Precipitation mostly involves sulphates, carbonates, oxalates, heteropolyacid salts and complex cyanides, while crystallization is used for alums. A separate group of methods are those used for the uptake and separation of rare gases. These elements do not form chemical compounds except with fluorine and the selection is therefore restricted to physical methods, such as adsorption and absorption, distillation and diffusion (see Chapter 5).

8.2.2 Separation of Elements 8.2.2.1 Strontium From the point of view of radiotoxicity and stability, the most significant strontium isotope in spent fuel is 90Sr which, together with 137Cs,represent longterm fission product fractions. A number of methods, including precipitation and coprecipitation, ion exchange and liquid extraction, have been used for obtaining strontium from the wastes of the Purex reprocessing method. Coprecipitation with lead sulphate used as the carrier in the simultaneous extraction of strontium sulphate and rare earth sulphates has been applied in technology developed at Hanford (USA). The precipitate is then converted into a mixture of carbonates. Lead is removed and strontium is separated from rare

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earths by precipitation using oxalic acid. Although precipitation methods have a number of disadvantages, as shown above, their usability as a simple method for obtaining the crude fraction of strontium has been well proven in practice. Ion exchangers, especially synthetic Dowex 50W,have been used for refining the crude fraction. The refinement has proceeded in several columns connected in series where strontium has gradually been purified of undesirable admixtures. Strontium thus prepared had a chemical purity of greater than 98 %, whilc radioactive admixtures, mainly 95Zr-95Nb and 144Ce-144Pr, made up of the order of lo-' of the Sr activity. Inorganic ion exchangers developed for strontium isolation include polyantimonic acid, barium sulphate with artificially introduced lattice defects, complex cyanides and combined ion exchangers. Of the reagents proposed for the liquid extraction of strontium, the most frequently used is di-2-ethylhexyphosphoric acid (HDEHP or D2EHP) with tributyl phosphate (TBP) in kerosene as the organic solvent. Remarkably, this combination was not the outcome of efforts to achieve a synergic effect using HDEHP and TBP, but' of the necessity to increase the HDEHP concentration in the solvent with the aim of avoiding the formation of the third phase, which is undesirable and difficult to eliminate in extraction technology. This extraction technique, alternatively with ion exchange, has in practice been used for refining the strontium crude fraction and for direct extraction of strontium from waste. The resulting product obtained using the m?thod is strontium nitrate, Sr(NO,),.

I - 4 M HNO3

Dissolution

concentrate

-4

Oxalic acid

Fig. 25. Separation of strontium, lathanides and actinides

8.2.2.2 Promethium and Cerium The most significant isotope of promethium occuring in considerable amounts only in spent nuclear fuel is 14'Pm. Cerium, represented by 144Ceand the 144Pr daughter, is responsible for the major part of the total activity of fresh and mediumage wastes.

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Both elements are rare earths (lanthanides) with characteristic chemical similarity, owing to their trivalency. Cerium differs from most of the other elements in its capability to occur in the tetravalent state, which is a property used in many separation methods. The actinides are chemically similar to lanthanides. They consist of transuranium elements and thus, in group separation of the fission product mixture, the fraction including both lanthanides and actinides is obtained at the same time. This i s also the case with strontium separation, shown in Fig. 25. After oxidation to Ce (IV), cerium is obtained by extraction with HDEHP; promethium is obtained and refined by ion exchange using diethylenetriaminepentaacetic acid (DTPA) or nitrilotriacetic acid (NTA) as eluents. Promethium is thus also separated from the transuranium elements americium and curium, which are processed in the following stage. The resulting products are the Ce,O, and Pm,OJ.

8.2.2.3 Cesium Of the cesium isotopes generated during fission processes, particular attention has been devoted to I3'Cs because of its high occurance and its properties. It is a nuclide whose long life (together with "Sr) is a limiting factor for the duration of the safe waste storage, i.e., for the storage of wastes not containing substantial amounts of transuranium alpha emitters. Alkaline wastes

pH I0 - 12

Fig. 26. Separation of cesium from alkaline and acidic wastes

The radionuclide can be removed or obtained from fission products using specific methods based on precipitation, ion exchange or extraction. Cesium has been isolated directly from alkaline wastes by ion exchange using synthetic zeolites and from acidic wastes by precipitation with heteropolyacids, as shown in Fig. 26. Cesium has also been separated using multibasic acid salts,

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mainly zirconium phosphate and hexacyanoferrates. Inorganic materials have been proven as ion exchangers, both for their radiation stability and their selectivity for cesium. The application of extraction processes meets with a number of difficulties ensuing from the chemical nature of cesium, owing to the low charge of the cesium ion. From the wide range of compounds studied for this purpose a complicated phenol derivative, BAMBP (4-sec-butyl-2(-methylbenzyle)phenol) has been used in combination with HDEHP for the group separation of strontium, rare earths and cesium. CsCl is most frequently the final form of the cesium emitter.

8.2.2.4 Zirconium and Niobium 95Zr and its daughter "Nb are medium-lived radionuclides. Chemically they are very similar in that they have a strong tendency to hydrolyse, polymerise and form colloidal particles which are often uncontrollably sorbed on various precipitates or on the surfaces of equipment. Obtaining these elements from radioactive wastes is therefore difficult. The oldest and best known method of their isolation is based on the sorption of the elements on silica gel with subsequent elution using oxalic acid.

8.2.2.5 Ruthenium Like xenon, ruthenium is a fission product whose proportion in spent fuel is very high. Io6Ru, accompanied by its daughter Io6Rh,is the main representative of ruthenium radionuclides. It is one of the few elements in general which may exist with all theoretically possible valencies. In addition, it easily forms complexes with waste fractions and tends to polymerize. The specific chemical and physical states of its compounds in a similar medium can therefore be characterized only by a complex system of equilibrium states. Oxidation to volatile RuO,, which can be distilled, is a universal and rapid method for obtaining ruthenium. Technetium is distilled togethx with the ruthenium.

8.2.2.6 Technetium, Rhodium and Palladium The common feature of these elements is that their fission products contain either stable nuclides or their long-lived radionuclides. The former is present as a periechnate, the latter two as anion complexes.

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The starting material for obtaining the elements can be, e.g., alkaline waste from the Purex reprocessing from which Cs has previously been removed. All three elements are sorbed on an organic anion exchanger from which they are gradually eluted and refined. The procedure for technetium isolation has been documented and has been applied in technological practice.

8.2.2.7 Krypton and Xenon These two elements are liberated from the process of cutting and dissolving spent fuel elements. In contrast to the above elements, krypton and xenon are not present in radioactive waste containers following reprocessing. It is desirable that the gases obtained are diluted with air as little-as possible and this requirement should be taken into consideration in designing the cutting and dissolution equipment. While almost all of the xenon is stable after the I-year cooling period for fuel, krypton contains a certain proportion of radioactive 85Kr. Further applications of xenon require that the presence of other radioactive gases in mixtures with xenon should be as low as possible. The purity of isolated xenon thus depends on how well krypton has been separated. One of the practical possibilities is adsorption on activated charcoal at normal or low temperatures, desorption being carried out with a stream of helium. More advantageous is deep-freezing of the two gases and their distillation a t liquid nitrogen temperature. Other methods are based on the absorption of the gases in tetrachloromethane or on membrane diffusion. Of chemical methods, the ability of fluorine to form compounds with rare gases can be utilized. Xenon enters into compounds with fluorine more readily than krypton, which can be used in separating the two gases.

8.2.2.8 Americium and Curium The isolation of transplutonium elements from spent fuel was considered much later than the separation of fission products. Information is thus significantly less abundant and is mainly focused on the first two members of the transplutonium series, whose contents in power reactor spent fuel are not negligible. The americium radionuclides involved are 241Am and 243Am. The half-lives of these two alpha-emitting radionuclides by far exceed those of "Sr and I3'Cs, which until recently were considered to be decisive for the determination of waste decay time. Curium is represented by 242Cmand 244Cm. Chemically, these elements resemble rare earths, which is the reason for their occurrence during the separation of wastes into fractions in the lanthanide portion.

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One of the methods, mentioned in the Section 8.2.2.2 is based on ion-exchange elution with a DTPA or rather NTA solution. In the latter instance, the ions are eluted in the order Cm3+ = Y3.+ > Gd3+ > Eu3+ = Am3+ > Sm3+ > Pm3+ > > Nd3+, which is advantageous for obtaining pure preparations. Another method, listed here only as an example of the application of extraction techniques, uses the principle of gradually separating fission and corrosion products from lanthanides and actinides by means of HDEHP; the separation of the last two groups is effected by re-extraction of Am and Cm using DTPA solution. The separation of Am frpm Cm can be achieved by precipitating the double carbonate R5Am,0,(C03), or by elution chromatography. Fig. 27 shows a simplified flow chart of the process. Acidic

wastes

D TPA

Lacfic acid

HDEHP TBP

1anthani de fraction =x trac -

tion

el 1

Waste

crn fraction

separa f ion Purification

Fig. 27. Extraction separation of actinides and lanthanides

The main difficulty in obtaining higher transplutonium elements from spent fuel from current reactor types is the small amount of these elements. Methods for isolating Bk, Cf, Es and Fm from irradiated targets have been developed and tested on a laboratory scale.

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9 Radioactive Waste Disposal

There are two fundamental approaches to radioactive waste disposal, viz., dilution and dispersion, and containment of the radioactive material and its disposal in the grounds, preventing contact with the environment. The latter approach is mainly used in inland and densely populated countries and requires that an adequate number of storage containers be available for liquid wastes or that treatment plants be installed in nuclear power plants with complete technological equipment capable of maximally concentrating radioactive wastes, converting them into a solid, unleachable form and disposing of them so as to prevent health and environmental hazards. These wastes should not be stored at source in liquid form for reasons of economy and hygiene. The possibility should be considered of a failure of the reservoir or pipeline, which might result in the release of radioactive substances into the ground, contaminating the water table. This in turn usually becomes a hazard to the sources of potable and utility water in the environs of the storage site. In densely populated areas and countries, this aspect of the problem is extremely important because the hydrogeological and hydrological conditions of localities selected for nuclear power plant sites are far from suitable, mostly being in the vicinity of large water sources or water reservoirs where the water table is usually relatively near to the surface and comes into contact with surface water sources and ground waters in the local area. On the other hand, such water tables do not have a high permeability so that the radioactive contaminant moves through them relatively slowly. If there is a release from the reservoir the extent of the contamination should thus be detected quickly and adequate measures taken. In order to prevent the occurrence of such releases (reservoir leaks cannot be excluded), storage containers are provided with double walls, are placed in concrete cells with efficient waterproofing and are continuously monitored. This procedure considerably increases the costs of radioactive waste disposal. Current technical and economic analyses have shown that the cost of the fixation of liquid wastes, including the storage of the end product, is 2-3-fold less than

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storage in containers. An intensive search is therefore being made for new and suitable fixation methods which would make radioactive waste disposal more economic and safer for man and his environment. M i g r a t i o n of R a d i o a c t i v e Substances i n t h e Soil The most probable path of radioactive substances from the closed-circuit fuel cycle into the biosphere is by way of gaseous and liquid radioactive wastes. Disregarding the rather improbable case of an accident at the nuclear reactor of reprocessing plant, when vast amounts of fission products would be released into the atmosphere in form of emissions and the controlled discharge of gaseous wastes into the atmosphere, we are left with the typical cases of the local contamination of the soil and ground waters by minor accidents. During the operation of the nuclear reactor or reprocessing plant the containment of the stored radioactive wastes may be damaged, the pipeline through which radioactive wastes are transported may leak, as may a corroded reservoir, or solidified radioactive wastes may be leached by precipitation or ground water, which will result first in the contamination of the soil and then in the contamination of the ground water, allowing the radioactive substances to travel to places accessible to man. The controlled introduction of liquid radioactive wastes into the ground represents a communication between the fuel cycle and the environment. Damage to a container containing high-level liquid wastes and the release of several cubic meters of liquid to a depth of several tens of metres into the ground could, under unfavourable geological and hydrogeological conditions, become a serious danger to the resources of potable and utility water in the local area. There is also the danger of contamination of water courses flowing through the threatened area. Many minerals and rocks are capable of efficiently containing and thus slowing down the travel of the contaminants through the aquifers. When the radioactive substances come into contact with the soil they will form a reservoir of radionuclides, from which they are absorbed for a long period of time by the root systems of plants or will infiltrate into surface or ground waters. The soil significantly influences this movement of radioactive substances through the biosphere and their entry into the human organism. Radioactive substances are sorbed by the soil. The mechanism of the reaction between the radionuclide and the soil may vary, and is affected by a number of factors which depends on the soil properties, radionuclides and environment in which the sorption processes take place. The migration of radionuclides in the ground may be negligible or very significant. The higher the rate of travel the more the radionuclides in the given environment disperse and the more their concentration decreases. The travel of radionuclides in the ground will, on the other hand, be slower if radioactive substances are contained by earths with high sorption properties. A knowledge of the sorption and desorption properties of various soil types is indispensable for making

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estimates of the danger of s i l contamination in each particular instance, for ascertaining the extent of the contamination and for taking the necessary measures to eliminate it. The speed of the movement of radioactive substances in the soil profile depends not only on their properties but also on the properties of the soil, i.e., mainly on the character of the sorption complex and on the physico-mechanical properties of the soil, such as porosity, permeability and the mineralogical and petrographic composition of the rocks of the contaminated formation. The migration of radionuclides is greatly influenced by the soil water regime, because the movement of radioactive substances in the soil basically follows the direction of the soil water flow. The quantitative observation of the movement of radioactive substances (fission products) is extremely difficult, and is made even more complicated by the fact that the most dangerous fission products occur in the form of ions with an ion-exchang: capacity for ions of the soil exchange complex. In ground water the radionuclide travels at rated velocity vR given by

where

v , is theflow-rate of the ground water, K,(R) is the distribution coefJicient of the relation between radionuclides in the solid and liquid stages, e is bulk density, f is the porosity of the medium.

The distribution of the radionuclide in the porous medium depends on the diffusion or dispersion coefficient, D . Dispersion prevails over diffusion in a porous medium and at high flow velocities, whereas at a low flow velocity of the liquid the distribution of radionuclides in the liquid phase is determined by molecular diffusion. The following relationship will apply for D and the standard probability deviation c2 = 2Dt (9.2) where t is time. A modification and solution of the general differential equation will yield the probability function M CX,Y,Z s(t)3/2 (D,D,D,)~ l 2 where M = coVo and expresses the amount of radionuclides at time, t = 0, and X, Y, Z are the floating boundary coordinates. The dispersed radionuclide forms an elliptical or spherical shape in the water table. The size of the formation is given by the boundary concentrations. For evaluating the consequences of the release of radioactive materials into the ground

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as boundary concentrations, the highest permissible concentrations of radioactive substances in potable water should be applied. The area affected by contamination is best determined experimentally directly under the field conditions. For the purpose of making preliminary estimates or in areas where geological conditions do not allow such experimental field observations, the area of contamination may be calculated. In such cases, according to the Gaussian distribution, the maximum concentration will correspond to the multiple of the standard probability deviation which may be determined under laboratory conditions on undisturbed soil samples. For determining the continuous release of liquid into the ground, can be applied at time t' when the face of the dispersed radionuclide reaches (x'y) in a twodimensional configuration written as = cos y

+

X'

sin y'

Y

(9.4)

where

y' =

2g

y,

(9.6)

qo is the specific filtration JEow-rate in m3 . day, Q is the radial speed of the dispersed liquid in the aquifer in m3 . day, x, y are distances from the discharge point and t is time.

The advancing face is usually oriented in the direction of flow of the ground water symmetrically and along axis x at y = 0. The equation can be written as t' = x' - In(1

+ x)

(9.7)

The introduction of the relative rate of the migration of radionuclides, u,/u,, will also depict the retention of radionuclides resulting from the sorption capacity of earths in the given environment. The above relationships may only be applied to homogeneous media, i.e., to conditions with a very simple geological and hydrogeological situation. In heterogeneous media, semiquantitative relationships only should be applied, which may, however, lead to considerable deviations in forecasting radionuclide behaviour. In order to preclude any underestimation of the possible consequences of accidents, the most important parameters should be tested by field tests; in the evaluation, the most unfavourable possible conditions for each individual case of soil contamination should be reckoned with. If even the least favourable conditions so determined do not pose any danger to the human environment, the given area may be considered as being safe and suitable for radioactive waste disposal.

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9.1 Selection of Site for Radioactive Waste Disposal 9.1.1

Criteria for Determining the Suitability of Radioactive Waste Disposal Sites

The siting of nuclear installations is the first of a range of measures for securing their nuclear and radiation safety. This range includes project design, production, construction, installation, commissioning operation, close-down and the safeguarding of quality at all stages. The problem of siting consists mainly in solving the extremely complex systems of inter-relations between the nuclear installation and the surrounding area, guided by the efforts to find an optimal solution. Initially, the approach to the siting of nuclear installations and radioactive waste disposal areas consisted in safeguarding that the population and the human environment in the area were at a safe distance from the installation and the disposal site. This approach does not, however, solve the fundamental issues of the problem and is, moreover, applicable only in certain countries. It therefore became necessary to seek new approaches that could be made in countries with a small area, dense population and highly advanced industry and agriculture. With the development of nuclear power, production conditions were gradually created for securing the safety of nuclear installations by technical measures. This approach has ;educed the importance of the factor of population density, while the importance of certain technical and area factors has increased, such as the safety and reliability of the nuclear installation itself, which is important for environmental control in the whole area, The aspects of reliable operation are inter-related with the general issues of environmental control ensuing from the construction and operation of nuclear installations. For these reasons and with regard to the regulatory function of State bodies, it would be useful if the system of criteria applied to the selection of radioactive waste disposal sites were extended to cover not only the narrow aspects of safety but also in the broader context, regional planning criteria which form an integral component of environmental control. The site selection process therefore also involves economic criteria which are not directly related to the safety of construction and operation of nuclear installations, but which are important for selecting the economically optimal locality. Considerable differences in capital costs and running costs related to the selection of the locality make it necessary to devote appropriate attention to this aspect. Criteria for determining the suitability of radioactive waste disposal sites are based on the following factors: - technical feasibility of the construction and operation of the installation, - environmental control,

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- the social importance of other interests in the exploitation and utilization of the area, - economics of construction and operation of the installation. Criteria applied to determine the suitability of a radioactive waste disposal site may be classified into two groups: exclusionary and comparative (evaluation). There are two types of exclusionary criteria. Phase I exclusionary criteria eliminate areas that are affected by landslides, areas with mining activities, karstic areas, areas with a low load-bearing capability (less than 30 kPa (0.03 N/mmz) in a depth of 7 m) and areas with a seismic activity at grade VIII or higher on the MERCALLI scale. With regard to environmental control, nuclear installations should not be built and operated in the proximity of natural curative springs, drinking water resources or areas in which, under normal operating conditions, it is not possible to ensure that the threshold dose for individuals and the value of the population dose per unit energy output will not be exceeded. As concerns other preferred uses of the area, radioactive waste should not be disposed of in areas with deposits of important raw materials or important communications. Phase I1 exclusionary criteria will allow the selection of sites that are suitable for building nuclear installations provided that they have costly technical equipment, on condition that special legal permission is granted for project implementation or on the basis of a re-evaluation of the priorities for the explotation of the area. This includes areas with unfavourable geological and engineering properties., i.e., affected by landslides, areas with a seismic activity at grades VI and VI! on MERCALLI scale, areas with the water table less than 2 m below the surface, inundation areas which are likely to be flooded in case of an accident involving a hydrotechnical installation, areas with a considerable energy of relief and areas that do not allow transport facilities to be built on the construction site. With regard to environmental control, this includes areas with highly permeable rocks, nature reserves, national parks and conservation areas, landscape conservation areas and their control zones, protected deposits, areas with important cultural monuments and areas that are being surveyed for minerals. As concerns the social importance of other interests, in the use of the area, priority must be given to agricultural use (such areas must be strictly protected from radioactive waste disposal), forest areas, areas that are currently being or will in the future be used for recreation, tourism and wildlife, the planned routes of important roads, oil pipelines and gas pipelines, areas of calm, areas in the proximity of projects that could be the cause of undesirable collisions and areas unsuitable from the point of view of special State interests. Evaluation of economic criteria helps to quantify the costs, i.e. , the suitability of area conditions of the construction and operation of the installation. These siterelated costs include the cost of structures and land, unused raw material deposits, surface pre-treatment, costs and savings for foundations, waterproofing, the water supply, the waste water discharge system, and the construction of roads. Additional criteria include issues that cannot be quantified economically and complement the

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overall evaluation. These additional factors include population radiation doses, threshold doses, the flow-rate and direction of the flow of ground water, the geomorphological and mesoclimatic conditions with regard to the possibility of the occurrence of unfavourable meteorological situations, conditions for the demarcation of control zones and their specification, the current state of the radiation burden of the population in the surrounding area, the general characteristics of the agricultural land, the ways in which food and drinking water are supplied to the population, linkage of the area to amenities in the nearest towns, and distance from the State border. These criteria are applied when selecting the most-suitable site for radioactive waste disposal. Larger and smaller potential sites are identified, which are further classified and evaluated by applying the following auxiliary criteria: geographical, meteorological, hydrological and hydrogeological and geological (pedology and lithology).

Geography The evaluation of possible sites from the area point of view requires the plotting of isohyets, thereby giving the distribution of terrains with different altitudes above sea level. We shall exclude terrains with altitudes of more than 800 m above sea level, and the remainder will be classified into four categories. Also excluded will be areas with a gradient of more than 5 %. The remaining area may be evaluated roughly as follows: terrains at altitudes of up to 200 m will be generally suitable in so far as the locality is not in the proximity of medium-sized or powerful water courses or water reservoir, which usually means that there is a lowland alluvium with a high water table. Terrains located at altitudes between 200 and 600 m will be very suitable, while terrains located at altitudes between 600 and 800 m will be less suitable as they usually abound in steep slopes along rivulets and streams with extensive ground water communication and circulation.

Meteorology Meteorological factors are extremely important for waste disposal, because climatic conditions may affect the properties of the disposed wastes and the amount of precipitation falling on the disposal site. Climatic parameters include the average annual temperature, especially its maximal and minimal deviations, the number of days per annum with snow cover and the number of days per annum when the temperature drops below 0 "C.Less suitable areas will be these with large temperature fluctuations and a large number of days per annum with snow cover. A large number of days per annum with temperatures below 0 "C is also disadvantageous,

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It is recommended that the investigated area be divided into 4-5 categories covering all usual conditions and that a detailed comparison be made of the areas. As concerns precipitation, in general the lower the level of precipitation in any given area the more suitable it is for radioactive waste disposal. For the evaluation of an area on the basis of this parameter the area should be divided according to the level of precipitation, e.g., areas with precipitation of less than 500 mmlyear, 500 - 700 mm/year, 750 - 1000 mm/year and 1000- 1500 mm/year. Other forms of classification may also be used if required by the geographical situation of the area. The intensity and direction of prevailing winds is a parameter that is considered to be complementary and is observed with the aim of investigating the effects of the wind erosion on the given locality. An important fact which should not be disregarded is the possible handling of powder substances on the disposal site. The blowing of such substances and contaminated soil particles by the wind may constitute a health hazard to personnel in the disposal area.

Hydrology and Hydrogeology Hydrological data (classical hydrography) are more or less related to the geographical data discussed above. From the hydrological point of view, sites that serve as reservoirs of drinking water should be excluded. Aquifers are rarely used as sources of drinking water, river water being increasingly used. Therefore, the siting of the radioactive waste disposal area in the proximity of such resources appears to be unsuitable. As concerns ground water, the absence of surface water-bearing horizons, which are usually accompanied by the occurrence of clays, is a positive factor. Often these horizons are of a local character and the ground water in them flows very slowly. The site selection process should therefore proceed as follows: areas with an extensive circulation of ground water and a high permeability of rocks should be excluded. The other areas should be classified into: - areas with a continuous water table which communicates with water courses (alluvia, sands and gravels, conglomerates), - areas with a discontinuous water table in heterogeneous formations with medium permeability, - areas poor in surface water-bearing horizons, with an impermeable bedrock where ground water flows very slowly (cm/d). The last situation will have priority in our method of evaluation. The direction of flow of the ground water does not play an important role in this rough evaluation but it will become an important factor in a more detailed survey of the locality. The velocity of flow of the ground water is closely linked with the permeability

of rocks. Areas with substantial layers of loess, dust, clay and conglomerates are very suitable, whereas gravels and coarse-grained sands are not. The continuous water table should be at a minimal depth of 5 m below the surface of the terrain and the permeability should be within the range lo-* - lo-’ m/sec. Ideal conditions in this respect are represented by a sufficiently powerful unsaturated layer where precipitation and ground water do not communicate, e.g., in localities in which a thick loess layer prevents precipitation from reaching the water-bearing horizon which is fed from other sources.

Geology Soils and quarternary formations may contribute significantly to preventing the migration of radioactive substances into the local area of the disposal area. Suitable materials in this respect are clays for their sorption properties, chernozemic soils, some acid brown soils and weathered rocks. Less suitable are pseudogleyic soils, brown soils and podzolic and illimerized soils. Sands, slates, gravel, calcareous soils and almost all ingenous rocks are unsuitable. Very important from the economic point of view is the cost of the soil, which in high-quality agricultural land may be very high. The division of the area into pedological zones will be of great help in assessing the area not only from the point of view of safety but also from the point of view of economy.

9.1.2

Data Indispensable for Site Evaluation

A more detailed study of suitable sites selected on the bzsis of criteria given in tne previous chapter will allow the selection of sites suitable for ground disposal of radioactive wastes. The subsequent stage should therefore consist in collection of all available archival data and data published in the literature. The collected data should be evaluated and a decision taken as to whether their quantity and quality are adequate to allow a detailed survey and a technical and economic comparison to be made of all considered sites. The site selection process will involve the following procedures: a) Collection of data on the general characteristics of the area from the point of orohydrography, geomorphology, meteorology and climatology, demography, bioecology, data on transport communications, important branches of the economy, agricultural production, etc. b) Collection of data on the geological configuration of the area - strat igraphic lithology, tectonics, evaluation classification, important rocks and raw material reserves. c) Collection of data on hydrogeological conditions, including a description of water-bearing formations and their distribution, the stacking of layers, filtration

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parameters, the chemical composition of the ground water, available water reserves, data on the abundance of ground water resources, hydrogeological boring and description of obtained results, and possibly results obtained by pumping. The choice of data should proceed from the technical requirements placed on the disposal area with regard to the amount and type of wastes designed for storage, the treatment of the wastes and other factors. The following documents are required for the above purposes: - diagrammatic representation of geologically surveyed areas, - mapping; topographic maps to a scale of 1 : 500 000, - 1 : 50 000; geological maps to a scale of 1 : 200 000; hydrogeological maps, special-purpose maps, etc., - tectonic diagram of the area, - geological and hydrogeological profiles of the area, - results of analyses of surface and ground waters, - a list of geological borings carried out in the area with an evaluation of the geological profiles, - other data in response to specific requests. Only in exceptional cases is the survey of the locality carried out so meticulously that it provides an adequate basis for the final selection of the radioactive waste disposal site. The first survey should therefore be followed by a detailed survey, the aim of which should be suplement those data on the area which could not have been obtained from archival sources. Finally, in the third phase the evaluation should be completed with a technical and economic comparison of those sites which maximally meet the demand of radiation safety and environmental control.

9.1.3 Site Evaluation Process In normal operation radioactive substances should not be released from the disposal site to such an extent as to become a hazard to the population, i.e., the population exposure should not exceed to permissible limit. A more serious hazard to the population could be caused by leakages resulting from accidents causing soil contamination and to the contamination of surface and ground waters in the surrounding area. Thorough analyses of these hazards should therefore be based on a detailed knowledge of the geological, hydrogeological and hydrological situation on the given site which, completed with the results of the study of the migration of radioactive substances in the ground, could serve as an adequate basis for a realistic estimate. As concerns the geological situation of the given site, the survey is usually conducted by specialist teams whose main objective is to evaluate the site for construction purposes, the available water supply for the population, etc. Many of the data obtained by these specialist teams may be used for the purpose of site

204

selection for radioactive waste disposal. When completed with a comprehensive geological and topographic description of the area in a broader context, the survey becomes the necessary prerequisite for integrating the disposal site in the surrounding area. The study of the migration of radioactive substances in aquifers and in soils of surface geological formations must include two main aspects: - the study of the mechanical properties of earths, - the study of their physico-chemical properties. The survey proceeds as follows: Earth samples are investigated for their mechanical properties and mineralogical and petrographic composition. In order the determine the mechanical properties of earths a n analysis must be carried out of grain size, on the basis of which the earth is classified into a particular group; it is then necessary to determine the moisture and porosity of the earth and the permeability of the undisturbed samples. The study of the sorption properties of earths includes the determination of the distribution coefficients for selected radionuclides, the determination of the relative rate of migration of radionuclides in earths and the study of the effects of the composition of the wastes on sorption processes in the soil. Basically only a few radionuclides are selected for these determinations, mostly those which have longterm and radiotoxic effects. Field verification of results obtained in the laboratory requires that a certain number of boreholes be drilled, from which samples are taken for laboratory testing. The number of boreholes depends on the geological situation of the area and is usually between 10 and 30. The boreholes serve for the determination of the ground water level on the site and the ascertainment of the direction of flow of the ground water. Boreholes should be sunk and probes inserted into the soil to obtain undisturbed samples for determining permeability. The accurate determination of the direction and flow-rate of the ground water is made in selected boreholes using active tracers or activation methods. On the basis of the above data an evaluation is made of the site, which serves both the operators of the disposal site and the hygiene inspection authority. The evaluation should contain an analysis of the consequences of at least two types of accidents which may cause the contamination of the ground water in the area. It should also contain draft proposals for removing the consequences of such accidents in so far as the evaluation has shown that such measures are absolutely indispensable. In areas with a complex geological situation each case should be treated individually. In practice, this means that data given in the evaluation should be completed with the results of other tests, namely field experiments, and that the following principles should be observed in the evaluation: a) calculations should be performed for only one radionuclide, i.e., that which appears to be the gravest environmental hazard,

205

b) a conservative approach should be taken in the calculation of model accidents, i.e., substitute the most unfavourable basic conditions which may exist in the natural environment of the given site, c) estimates of the effects of the contaminants on the environment must take into consideration that only part of the contaminant will sorb on minerals and rocks forming the unsaturated or saturated zone; d) the rate of migration of the radionuclide into the local area should be considered as approximating the flow-rate of the ground water in the given locality. In a simple geological situation on the site the consequences of the release of radioactive substances into the surrounding area may easily be analysed. In practice, such conditions most frequently occur in the alluvial silts of large water courses and in sandy sediments where simple conditions allow forecasting of the behaviour of all of the most important radionuclides in the ground in arbitrary situations. It is very important that the project design of the waste disposal area and its auxiliary facilities should include considerations of inspection and control measures based on the results of previous surveys. Such measures should include observation boreholes, wells and various types of barriers to prevent the transmission of the contaminant, thereby minimizing the environmental hazard.

9.1.4 Use of Radioactive Tracers for Site Evaluation Considerable interest in tracer methods has been aroused by the need to forecast the direction and rate of the ground water flow and thereby also of the rate of travel of the contaminant through soil medium. The radioactive indicator therefore must: - be qualitatively detectable even in very low concentrations, - be present in the observed medium in negligible amounts, - not react with the ground water in the presence of precipitates, - not be absorbed by the given medium, - not undergo chemical or physical changes or reduce the accuracy of measurement during its passage through the medium, - be low-cost and readily available. The use of radioactive tracers is a very sensitive method of measurement and allows detection even at extremely low concentrations. These substances may be incorporated in a wide range of inorganic and organic compounds. The most frequently used radioactive tracers for studying the current problems of soil hydrology and classical hydrogeology are 3H, I4C1 32P,35S,"Br, 86Rb, I3'I and certain complex compounds containing a radioactive cation, e.g., 51Cr, 6oCo, 24Sb. The main advantage of this method is the low concentration of the tracer,

206

which does not affect the observed medium. The disadvantages of the method are radiation danger, the high cost of the measurement apparatus and the possible interaction of the tracer with the medium (mainly owing to isotope exchange), which cannot, however, be excluded even in non-sorbing tracers. The systematic selection of radioactive tracers for hydrological and hydrogeological uses led to the conclusion that 13'Cs, 21Pand 45Ca are unsuitable because of their low yields. The best results in all types of waters and soils were obtained with 35S,3H, 82Br and 13'1 for clay soils, 51Crfor clay rocks, clays and siliceous sands and 24Na for all types of sands. Views on the suitability of a certain isotope as a tracer may differ, owing to the different approaches to the evaluation. The most widely used tracers in field conditions are 13'1 and "Cr. This is due to their easy applicability, low cost, availability and the relatively low health hazard of their application. In hydrology and hydrogeology radioactive tracers may be used for the study of a wide range of problems, such as the flow-velocity,the discharge, dispersion and the communication between different water bodies. Tracer studies show that water moves in surface soil horizons at a surprisingly slow speed. Tracers may be used to obtain results which in conventional hydrology cannot be obtained at all or which are extremely difficult and laborious to obtain using conventional methods. The use of radioactive tracers still has its difficulties and increased attention will have to be devoted to both methods and equipment. For field application better portable measuring apparatus will have to be developed and routine detection methods will have to be simplified.

Tracer Activation Method In the use of radioactive tracers, generally only a minute fraction of the radioactive substance is used in the experiment. In field experiments the surplus of radioactive substance cannot be removed under controlled conditions, which causes an undesirable contamination of the biosphere. On the other hand, the use of radionuclides with the shortest half-life considerably shortens the time needed for the experiment. It therefore appears to be advantageous to reverse the usual tracer procedure. Instead of the radioactive nuclide a suitable stable element is used, which is identified when the experiment is completed with high sensitivity in the sample after its activation. This will eliminate two important problems, that of radioactive contamination of the site and possibly also the uncontrolled release of the radioactive tracer into the environment. This method, moreover, restricts the time interval between sampling and measurement. Suitable stable nuclides for these uses are mainly nuclides with high activation

207

cross-sections, namely 164Dy, "Mn, "51n, 1931r and '*'Re, certain halogenides and metal complexes with EDTA. The stable isotopes Br, I, Dy and In are the most suitable substances for detecting the ground water flow and the migration of radionuclides in aquifers and unsaturated horizons.

9.2 Land Burial In a number of industrial and densely populated countries, surface disposal of radioactive wastes is the only possible alternative. As numerous drinking water and utility water resources and intensive agricultural production do not allow the discharge of liquid wastes into the ground and into surface flows, it becomes essential to store wastes on surface disposal sites or in landfills. It should be that salt deposits need not be the proper solution (see below). The same is true of dumping wastes into the deep ocean layers, which need not necessarily be economical for a particular country. As concerns wastes from nuclear power plants (not wastes from reprocessed burnt-up fuel), these low- and medium-level wastes should be stored on surface disposal sites or in disused geotechnical installations. Waste disposal on surface sites is relatively economical, especially in the storage of solid or solidified wastes. These wastes can be managed safely and, in case of emergency, they can be transferred to a safer place. The disadvantage of the method is the high cost of transport of any type of radioactive substance. It is therefore advisable to build several regional disposal sites so as to shorten the distance between the sites. This is especially important with disposal sites located near nuclear power plant that produce considerable amounts of radioactive wastes. There are various methods of storing solid and solidified nuclear power plant wastes, depending on the treatment and properties of these wastes. A survey of storage techniques and a description of suitable sites is given below.

9.2.1 Basic Units of Radioactive Waste Disposal Site The main operating unit of the disposal site is the storage facility. The following components are used for the storage of radioactive wastes on surface deposits or in surface formations: - a simple surface disposal site with or without earth cover, - a simple trench with earthfill, - a concrete trench, open or covered and filled, - a roofed building (hall) on the surface or sunk, - the storage trench, - the simple pit or the concrete pitch with filling.

208

Simple Surface Deposit The simple piling of wastes on the ground surface is extremely economical because it eliminates the costs of earth moving and excavation. Such deposits are suitable in mildly sloping terrain and may be used for low-level wastes deposited in blocks with protective packaging or for the storage of drums which need not be shielded (Fig. 28). The deposited wastes are insulated against precipitation by a

SURFACE DEPOSITION OF BLOCKS WITH INSULATION COATING

LOCATING BLOCKS ON SLOPE

turf

cover and rnsulafion

depmifed

ma teraal

base -rock Fig. 28. Surface storage

cover of insulation material and by earth backfilling. Rainfall will run off the slope and should therefore not affect the safety of the site. Another advantage is the easy handling of the deposited wastes.

Simple Trench This is usually considered as one of the basic solutions (Fig. 29). Its dimensions depend on the form and volume of the wastes and on their disposal. The transport and handling equipment is usually installed on the periphery of the trench or travels across the area. In the former case the width of the trench should be such as to allow the arm of the filling crane to reach easily beyond the longitudinal axis of the trench. The transverse parameters of the trench depend on the soil loadbearing capability and on the stability of the banks of the trench to secure safe transport, handling and oprration.

209

The deposited material is covered with earth and the earthfill is compacted and graded. The earthfill may, if necessary, also be covered with a layer of insulation material, capped with a thin layer of earth and planted with turf. An underdrain or a peripheral drainage system with a sump may be installed. The trench should be filled gradually with each layer of deposited waste so as to fill the whole of the storage space.

Longitudinal section arums

+

Blocks

Jrums

I LOO

Fig. 29. Simple trench (dimensions in cm)

The project design of the trench should be proceded by a preliminary technical and economic evaluation, the trench being a structure demanding space. It is first necessary to assess the depth of the trench, which is given by: - the necessary height of the trench bed above the level of the highest point of the water table, - the possibilities of depositing materials in layers, blocks, drums, etc., - the economic utilization of filling machinery and handling efficiency. The optimal width of the trench is given by: - the reach of the filling equipment, - the efficiency of the trench profile, i.e., the width to depth ratio. In relatively narrow trenches there remains a high proportion of unused space, while broad trenches do not allow adequate control of deposited materials in the middle part of the trench. The dimensions of the trench must therefore be optimized.

210

Concrete Trench The variability of wastes and the diverse methods of waste processing and pretreatment may require safer storage than is provided by simple trenches. Much safer and obviously much more costly is the disposal of radioactive wastes into Covered with slabs

concrete trench

Fig. 30. Concrete trench (dimensions in cm) semi-sunken sforage hall with 72-m span. pan file cover

Asbesfos cement

L

1

1200

L

1

Fig. 31. Semi-sunken storage hall with a span of 12 m (dimensions in cm)

concrete containers built in earth ridges (Fig. 30). The concrete container, having suitable dimensions and parameters based on optimization, allows the deposition of wastes with a higher specific activity, allowing the early detection of released

radioactive materials from the storage area, offering easier handling and providing better protection against precipitation by ceiling panels when the nature of the stored material requires such safety precautions to be taken.

Roofed Building A semi-sunken hall with light roofing steel binders (Fig. 31) and asbestos cement pantiles is another possibility for storing radioactive wastes and is especially suitable for certain types of solid radioactive materials, for safer storage of radioactive wastes in less favourable geological conditions or in more complex hydrogeological conditions. The concrete hall may also be provided with an insulating layer or stainless-steel cladding which is not economically beneficial.

Simple plt

with prefubricateed bed

Reinforced lining and reinfirced concreie bed

Fig. 32. Types of storage pits (dimensions in cm)

On the other hand, the spatial arrangement of the hall is acceptable. As for handling, the hall is a top-loader, i.e., wastes can only be deposited from the top using stationary handling equipment in the hall.

212

Storage Pits Storage pits are mostly used for the disposal of wastes with a higher surface radiation that must be shielded for a certain period before they can be handled as low-level wastes. The pits have various linings; the basic lining is shown in Fig. 32. The design and construction of these pits are the most costly of all the described facilities and waste storage is extremely demanding.

Disposal Trenches Disposal trenches (Fig. 33) are basically modified trenches, either simple or concrete, and are suitable for the storage of all wastes of non-standard dimensions. The lining consists of concrete blocks which make up the walls and the bed of the trench. Waterproofing is usually required - this depends on the character of the terrain, and is done by lining the walls with layer of asphalt. The structure and waste handling are very simple. When the pit has been filled it may be covered with earth, concrete slabs, boards, etc. A comparison of the three selected ground disposal and storage techniques is shown in Table 30.

Fig. 33. Storage bunker (dimensions in cm)

9.2.2 Operation of Storage Site The operation of a storage site is illustrated in Fig. 34, which serves as the basic outline for the evaluation of a technologically and technically undemanding operation of such a site. Processing line supply blocks and drums containing solidified wastes which are transported into the handling area, from where they are transported into the sto-

213

Table 30: Comparison of Three Selected Techniques for the Disposal and Storage of Solidified Wastes Ratio of storage Disposal technique

Estimated capital costs in U S $

Storage area (m2)

Specific costs U S $/m3

area and of annual production from 2 x 440 MW(e)

14 OOO 30 OOO 68 OOO

1100 590 2 160

12.8 50.7 31.3

2.00 1.07 3.93

I ~

~~

Simple trench Concrete trench Roofed building

rage area. The handling area may also be used for the disposal of treated wastes from sources which have processing equipment but do not operate disposal sites. Fig, 34 shows the gradual build-up of the disposal area which complies with waste production. In storage area 1 treated wastes are stored in one unit, storage Office building

1

Surface stomye

I

re -treatment

Temporary storaye L

I

Storage shafts

I I

1

Fig. 34. Diagram of disposal area

area 2 serves as a standby space for the storage of fixed wastes, storage area 3 is excavated after the previous storage areas have been filled, and the excavation of storage area 4 is prepared. A vacant site is available for building further storage areas. Their design is based on experience gained with radioactive waste disposal in the filled storage areas.

214

Storage areas of regional character (Fig. 35) include the following units: storage areas, communications, auxiliary non-active facilities (concrete plant, workshops, garages, office building) and auxiliary active facilities (monitoring laboratory,

decontamination hall, laboratory of the dosimetric service). The operation of all these facilities depends on the amount and quality of the wastes which are to be deposited on the regional disposal site. The following conditions should be met by such a regional storage site: - the area should be large enough to store the estimated amounts of wastes, - the storage area should be designed as a complex, - the disposal techniques should be flexible, complying with the different types of wastes to be stored, - the storage area should meet all demands on safety and economy of operation.

9.2.3 Properties of Stored Wastes From surface storage, wastes will be transported to other areas for final disposal. The following properties of the wastes are important with regard to transportation and final disposal: - mechanical strength - radiation stability - leachability.

215

Mechanical Strength The mechanical strength of fixed wastes is affected by the following factors: composition of the wastes, exposure of the material to radiation and atmospheric conditions on the storage site. Regarding the composition of wastes, the effects of various salts, especially nitrates, are the most important. In cement blocks the nitrate content should not exceed 250 kg/m3 and phosphates should not exceed 150 kg/m3. The presence of incorporated ion exchangers may cause the disintegration of the concrete block if it is stored in a moist place and is not provided with a watertight coating. In bitumens additions of this kind are unimportant if the total salt content does not exceed 400 kg/m3. On the other hand, larger amounts of manganese have unfavourable effects on the composition of the bitumen mass because they may reduce the flash point of bitumen and nitrates and because they muse a high leachability of the resulting mass. Atmospheric conditions have a negligible effect on the mechanical strength of wastes. The frequent effect of precipitation on stored concrete blocks may, however, cause cracks, especially in those blocks which have not been provided with protective asphalt coatings or which have been stored in simple trenches or in unlined pits.

Radiation Stability Part of the medium-level wastes from nuclear powzr plants may contain a relatively large amount of radioactive substances, up to a level of 10" GBq. The effect of internal radiation from incorporated sources of ionizing radiation may best be investigated by long-term external irradiation, e.g., with a Co source. The whole process is usually modelled in such a manner as to make the experimental conditions approach real conditions as closely as possible. The irradiation time is therefore usually longer than 365 days and the escape of gases resulting from radiolysis is continuously monitored. Typical results of such experiments are shown in Table 3 1.

Leachability The contact of the product with water, be it precipitation or ground water, usually results in the leaching of radioactive substances and their release into the ground. In order to be able to evaluate the effect of water on the final product designed for disposal, and to be able to make correct decisions on the manner of its final disposal, it is necessary to know the leachability of the product under different conditions and all the consequences thereof.

216

Table 31: Radiation Stability of Stabilized Wastes

I

Total dose (GY)

Type of waste

concrete

10'

vacuum-treated concrete

lo8

vacuum-treated concrete with bitumen

1O8

bitumen

108

bitumen with sorbents

108

Amount of escaped gas (H2) m3/kg x

70

~

no change

I s

I

no change

7

i

Notes

disintegration into dust particles

1

~

Description of product

8

change of rheological properties

trace of CH4

change of rheological properties

trace of CHI

Table 32: Leachability of Solidified Wastes Proportion of leached radioactive substances in

% after contact with water for different lengths of time

t=lOd

concrete, common vacuum-treated concrete vacuum treated concrete-bitumen bitumen RS bitumen RS sorbenti vacuum-treated bitumen sorbents

+

+

1 I

Type of waste

0.1 0.001 o.Oo0 2 o.oO0 1 o.Oo0 01

0.001

I

I t=60d

I t=600d

0.02

0.02

O.OO0 6 o.Oo0 2 O.OO0 08 O.Oo0 004 O.OO0 8

O.Oo0 08 o.oO0 1 o.Oo0 02 o.oO0 002 O.oO0 04

1

Material irradiated with10"Gy t=lOd

1

1 0.001 o.Oo0 2 o.Oo0 1 o.Oooo1 -

Leachability is determined at various workplaces which develop fixation methods r which produce fixed products. Typical values of leachability are given in Table 32, which shows the combinations of fixed products with substances that may considerably reduce leachability. It also shows that with the exception of initial releases, i.e., the wash of the surface

of the product and the dissolution or mechanical removal of substances occurring on the surface or near the surface, the results of the experiments are very promising and show that safe storage of these products can be achieved without great difficulty.

9.2.4 Waterproofing of Storage Areas The necessity to waterproof storage areas is dependent on the leachability of stored wastes. Leachability will be enhanced by atmospheric precipitation in surface sites or trenches which are not waterproofed. The periods of the so-called episodic contact of the waste with water may take several minutes (rain shower) or several days (spring snow thaw). The main principle to be observed in waterproofing disposal sites is that the content of radioactive substances in the ground water should in no instance be higher than the highest permissible concentrations of those radionuclides which are stored in the site. The specific activity of the leachate, i.e., water from precipitation which has come into contact with the waste and which seeps through the surface cover layers in the disposal area, must not be too high. The following equation can be written for the calculation of the specific activity of the leachate:

where s, is the specific activity of the leachate, A , is the specific activity of the fixed waste, R is leachability, t is the time of contact between water and the deposited waste, s is the surface area of the block and V is the volume of water. For typical calculations t = 1 h, s = 1 cmz and V = 10 cm’. In a concrete block with a fixed radioactive substance of total specific activity 37 MBq/kg (lo-’ Ci/kg), and if the leachability of the material is the specific activity of the leachate will be 370 kBq/m3 (lo-* Ci/l). The annual threshold intake is given at 13’Cs = 444 kBq (1.2 x Ci) and if cesium is the sole radionuclide present in the fixed waste then it is highly improbable that the amount leached could cause internal contamination of persons ingesting water from a source in the proximity of the disposal area. When radioactive wastes are stored in water (without containers), the calculation of the specific activity of the contaminant will be more complicated. This is so improbable that it need not be dealt with in detail. On the other hand, brief episodic contacts between radioactive wastes and water are more frequent, and in such an event it is appropriate to carry out a series of in situ model experiments which will yield the most important data on the amount and character of radioactive substances that have escaped from the stored wastes into the surrounding area.

218

9.2.5

Storage of Waste with High-Level Radioactivity in Containers

High-level radioactive liquid wastes are the product of the first extraction cycle and contain more than 99.9 % of all non-volatile fission products. So far preference has been given to their thickening and storage at source in containers made of carbon or stainless steel. These containers are usually provided with cooling equipment for the removal of the heat of decomposition, and possibly equipment for mixing the settled particles. Mixing is carried out either with air or mechanically in order to remove the sources of heat accumulation and thereby the possibility of an accident. In the USA 350000m3 of high-level radioactive wastes were disposed of in this manner in 1970. Although corrosion tests guaranteed a servicelife of more than 36 500 days, in 15 cases failures of the walls of containers made of carbon steel occurred. At Hanford in eleven cases 5.65 PBq (150 000 Ci) of I3’Cs escaped into the ground. Future projects therefore envisage the separation of 90 % of all 90Sr and 13’Cs. These nuclides will be solidified and stored on temporary sites until suitable final storage areas are found. The remaining liquid wastes will evaporate in the containers, forming a massive solid mass with the residual salts. The storage of this mass will be much less hazardous than storage in liquid form. The volume of containers in the USA ranges between 1000 and 5000m3, in Western Europe they are smaller. At Marcoul end in the Hague containers with a volume of 60m 3 are provided with cooling equipment consisting of cooling coils and a condensing unit. Mixing is carried out by air bubbling. At Windscale high-level radioactive wastes are dumped in containers with a volume of 150 m3; at the Eurochemic plant at Mol the containers have volumes of 40 and 210 m3 and are equipped with cooling coils. Containers with high-level radioactive wastes are continuously monitored. The level of the waste liquid is measured by remote control, as is its temperature, the technical state of the container and a number of other parameters. At Hanford vertical and horizontal monitoring probes have been installed around the disposal area with detectors signalling any leakage of radioactive materials from the container. In Idaho the containers are placed in a bunker with a double containment and at Savannah River the container is placed in a steel-clad concrete capsule. At Windscale the walls of the concrete bunker have a thickness of 1.5 m and are lined with stainless steel. In the Hague the containers are buried in the ground in the impermeable clay horizon, which has a thickness of several tens of metres. In almost all instances there are stand-by containers into which the wastes are immediately transferred in case of a failure.

219

9.3 Depositing Radioactive Wastes into Geological Formations 9.3.1 Storing Wastes in Salt Deposits Salt deposits are suitable for the storage of all types of liquid, solidified and solid radioactive wastes. The method is highly promising and meets all requirements for safe and long-term waste storage for the following reasons: a) salt deposits occur in many regions of the world b) salt has very favourable mechanical properties c) salt has very good shielding properties, similar to those of heavy concrete d) the costs of salt mining are low e) the deposits of rock salt with a laminated structure do not contain any circulating water and are isolated from the aquifers by layers of impermeable slates. Table 33: Accumulated Amounts of Certain Significant Long-Lived Nuclides in the Year 2000 Accumulated Radioactivity Nuclide GCi 90Sr 95Zr

lo6Ru "7CS

'44Ce 14'Pm

80 120 100 80 110 50

i

TBq 3.0 x 109 4.4 x 109 3.7 x 109 3.0 x 109 4.1 x 109 1.8 x 109

The question arises, however of whether salt deposits should be used for the storage of low-level radioactive wastes with short-lived radionuclides to the detriment of high-level wastes, which in the future will have to be stored for a practically unlimited period of time. By the year 2000 the estimated installed annual output of nuclear power plants will be minimally 2000 GW(e) which means that 4.107 kg of spent nuclear fuel will have to be processed annually. The annual world production will be around 3000 m3 of solidified waste and by 2000 the total accumulated volume of radioactive wastes will amount to roughly 40 000 m3 with a content of 2 x 10'O TBq (540 GCi) of radioactive substances, of which 5.5 x lo7 TBq (1.5 GCi) will be alpha-active nuclides. The accummulated amounts of certain important nuclides are shown in Table 33.

The prospects for the development of world nuclear power production allow us to make an estimate of the trends in the use of salt deposits for radioactive waste disposal. Exhausted salt deposits with less favourable hydrogeological conditions will be used for the storage of low-level radioactive wastes. The packaging of these materials will be simple and will depend on transport and radiation hygiene conditions rather than on the layout and location of the disposal site. Demands on storage will obviously be more strict for wastes with medium-level radioactivity and even more strict for the disposal of high-level radioactive wastes. In the latter instance a wide range of requirements will have to be met before the waste disposal area is declared completely safe. The following criteria should be applied to the selection of a suitable ground formation for radioactive waste disposal: - the salt deposits should have a laminated structure and an appropriate area to secure the insulation of the deposed wastes, - the depth of the deposit should be at least 150 m and the thickness of the surrounding layers at least 50 m to ensure that the heat is sufficiently dissipated and radioactivity contained, - for economic and operating reasons the deposit should lie at a depth of more than 500 - 800 m, - the deposit should not be located in the immediate proximity of potential deposits of valuable exploitable raw materials, - the deposit should be in a tectonically stable area. Rock salt is a typical monomineral rock based on halite (NaCl) which may have a small content of additions, such as carnallite, sylvite, anhydrite and glauberite, and also trace amounts of free bromine and iodine. The important properties of halite include high thermal conductivity (6.1 W m-' deg-'; 0.0146 Cal/cm.s grad), high solubility in water (37 % at 0 "C) and low electric conductivity (dielectric constant 6.2 at 20 "C). The melting point of halite is 801 "C and the boiling point 1413 "C. The geological survey of localities considered for the burial of high-level radioactive wastes must include an accurate description of the stratigraphic distribution of the area, the petrography of the rocks and the tectonic structure of the area, all in much more detail than is needed for mining areas surveyed for mining purposes. The most important prerequisite is a detailed knowledge of the hydrogeological conditions of the locality, including the number of aquifers, their area and thickness, the source of their saturation and the level of the water table, the chemical properties of the ground water, the leachability of the rocks and the occurrence of the brines. The geophysical survey should include the solution of geothermal and radiological problems, such as the distribution of the thermal field in time and space depending on the type and amount of stored wastes and on the disposal technology, the evaluation of possible secondary factors which could affect on the technology and safety of disposal.

221

The storage of high-level radioactive wastes in salt deposits was demonstrated within the US Salt Vault Project in a disused salt mine at Lyons, Kansas, where 14 spent fuel elements with a radioactivity of 1.5 x lo5 TBq (4 x lo6 Ci) were stored and withdrawn after the experiment. The salt block was electrically heated. The irradiation of the ambient salt with a dose of up to lo9 rad did not have any measurable radiolytic effects, nor did it have any affect on the mechanical properties of the salt. Much useful information was gained on thermal tensions, the migration of cavities filled with brine and on salt flow as a function of temperture. On the basis of the results of the survey a disused salt mine was selected for the storage of high-level radioactive wastes containing alpha-active nuslides. The salt mine had an area of 7.3 x lo5 m2 with an adjoining area of 3.65 x lo6 m2 of available vacant land for the future storage of high-level radioactive wastes which will be produced in the USA before the end of this century. A special building will serve for the delivery of wastes with alpha-active emitters. The wastes will be deposited in the building from containers, measured, packaged in special containers and dumped in the mine. They will be stored in the disused mine, which has a total capacity of 560 000 m3 of wastes, at a rate of approximately 20 - 25 000 m3 annually. The used area will be backfilled with salt and sealed off. The building for the delivery of high-level wastes is to be located approximately 260 m from the delivery building for alphaactive wastes. The containers holding the wastes will be removed from casks weighing up to 100 000 kg and lowzred through a shaft. The maximum capacity of the storage area should be 300 containers per week. The spaces for the storage of high-level radioactive wastes will be mined using routine technology and approximately 50% of the salt will be used for backfilling the mined area after the burial of the wastes. The containers will be transferred to the mined area and will be stored in cavities in the salt in such a manner as to allow the heat released from radioactive decay to dissipate without causing an excessive increase in the temperature of the salt or other ambient rocks. This salt deposit is located in a Permian formztion at a depth of 270 - 370 m and consists of several horizontal strata of salt with intermittent layers of slate. It is covered with a powerful layer of slate while the subgrade is formed by a sufficiently powerful layer of anhydrite and alternating layers of limestone and slate. The deepest fresh water aquifer is at a depth of 90-95 m. Between 100 and 134m there are not water resources. Several salt water aquifers are located at greater depths. The salt deposit suddenly terminates about 40 km east of Lyons, and the nature of this boundary indicates that the salt was dissolved by ground water. As the face of the salt layer has reduced by less than 8 km over the past 10 years, it can be assumed that the storage area will be protected for at least another lo6 years. Protective measures have also been taken to prevent the deposit from being infiltrated with water from nearby drillings. Calculations and measurements which were made to determine the pmnissible

222

values of heat generation showed that the most important factor in this respect is the decrepitation of the salt-filled cavities and the effects of heat on the stability of the mine while it was still in operation. The following criteria were derived therefrom: a) maximum temperature of salt between two containers must not exceed 185 "C, b) maximum temperature at a distance of 0.2 m from the wall of the hottest container must not exceed 250 "C. The heat gradient will cause the migration of cavities filled with brine. Owing to the ageing of wastes and the decrease in their radioactivity as a result of natural decay, this migration will cease within 20-30 years. In this period 2 - 10 dm3 of brine may move up to the container. The settlement of the mine and the thermal dilatation of the ambient rocks will be balanced in the initial period of storage; after 200 years a contraction will occur which will result in the formation of a depression approximately 1.2 m deep. This process will probably continue for several thousand years and the overburden should adapt to this process and should not be disturbed. The most important products of radiolysis, viz., H, and 0,, oxygen compounds of chloride and bromine and HCl, will accelerate the corrosion of the containers. The tightness of containers made of stainless steel will be disturbed after several months in the mine, whereas containers made of carbon steel will probably last several years. It is assumed that operating safety will be maintained because the storage spaces will be sealed before the tightness of the containers is disturbed. The corrosion of the cmtainers would complicate the situation only in case of transfer of the stored wastes. M:thods for extending thz life-span of the containers and the development of technologies for withdrawing wastes in case of necessity are being intensively studied. Hydrogen gas and HCl which escape from the burial area into the transport corridors will be diluted with ventilation air to a safe concentration. The sole possible source of the release of radioactive substances into the environment is air from the ventilation, which will, however, be filtered using highly efficient filters. It is assumed that the amount of radioactive substances released will be less than 0.1 % of the maximum permissible values. Solid wastes which will generate in the storage area will also be converted into a suitable form and stored o n selected sites. Liquid wastes will be recycled and surplus water will evaporate into the atmosphere. Low-level radioactive wastes have been stored in salt deposits in the Federal Republic of Germany since 1965, in the Asse I1 project near Wolfenbuttel. The area was prepared for the storage of all low- and intermediate-level wastes until the year 2000. Radioactive materials are stored at 14 levels in more than 100 rooms, the highest level being 490 m and the lowest 850 m below the surface. The wastes are stored in drums with a volume of 0.2 m3 and the rooms are gradually filled. Since 1971 drums have also been used for the storage of intermediate-level liquid wastes. The drums are placed in containers with lead or steel shielding casks.

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9.3.2 Storage of Gaseous Radioactive Wastes from Reprocessed Spent Fuel The safe storage of the gaseous fission product "Kr after its separation from gaseous wastes from reprocessed spent fuel and possibly from other short-lived rare gases (133Xe) is, with regard to its long half-life and high specific activity, an integral component of the processing of gaseous radioactive wastes from reprocessed spent fuel. In selecting disposal sites for 8sKr the following factors should be considered: the capacity of the area, the thermal and radiation conditions, the capital and running costs and safety of operation. As there is no available technology for separating "Kr from non-radioactive krypton, present as a component of the air in gaseous wastes and thus also in separated 85Kr,the capacity of the storage area is one of the limiting criteria for selecting the area and type of storage of this long-lived fission product. A reprocessing plant for spent fuel from light-water reactors with a burn-up of 2856 J/kg (33 000 MWd/t) with an annual reprocessing capacity of 1500 tons of spent fuel will produce approximately 721.5 PBq (16.8 MCi) of waste "Kr per year, which amounts to 189 m3 (STP) of krypton with a proportion of "Kr of approximately 6 - 7 %. With regard to the world growth of nuclear power production (see Section 2 . 9 , it can be assumed that by around the year 2000 the annual amount of waste krypton from reprocessed spent fuel will reach lo4 m3 (STP). The storage of separated "Kr in steel cylinders at normal or high pressures (10.3-15.2 MPa; (100-150atm)) currently appears to be the best means of burial. The storage of "Kr in pressure cylinders at normal pressure guarantees maximum safety and the equipment required in storage for such disposal is not demanding. The storage of 85Kr in cylinders at high pressure reduces the storage volume approximately 100-fold in comparison with storage at low pressure, but the capital and running costs are much higher. Pressure cylinders have double walls and require continuous cooling owing to the high accumulated radioactivity (approximately 18.5 PBq (0.5 MCi) of "Kr for one cylinder of stored krypton with a volume of 60 m3 (STP) at a pressure of 10.13 MPa (100 atm)). The demands on control systems are considerably higher than in the storage of 85Kr at normal pressure. The storage of 85Krin steel cylinders at low and high pressure inevitably requires the removal of H,O,0 2 ,O3 and N2from the stored gas in order to prevent the occurrence of corrosion products from the radiolysis of the stored gas. An unsolved problem in this respect is the long-term effect of decay products of 85Kr (rubidium) in kilogram amounts on the mechanical stability of the pressure cy1inders. The incorporation of "Kr into the solid phase (clathrates, molecular sieves, kryptonates, compounds of rare gases) is one means of reducing the pressure of the rare gas in the cylinder and moreover forms a barrier limiting the escape of "Kr

224

from the cylinder in case of the damage thereto. The practical application of this method of 85Kr disposal is highly dependent on the capacity of the material and on the thermal and radiation stability of the newly formed “compound”. Experience has hitherto been gained only on a laboratory scale and the results are not very encouraging. The suggested technologies for the incorporation of ”Kr into solid materials are highly demanding from the engineering point of view, requiring high pressures and complicated and costly equipment. The storage of 85Kr in geological formations is based on the same principles as the storage of liquid wastes. The advantage of this potential technology is its considerable capacity to “process” gaseous radioactive wastes without the previous separation of ”Kr. On the other hand, the inert nature of 85Kr and its gaseous state considerably increase the probability of its escape from ground storage areas. The storage of ”Kr in geological formations is technically feasible; practical implementation would have to overcome certain unsolved problems, such as the migration and dispersion of ”Kr in soil layers and in geological formations and the related problems of the contamination of the human environment.

9.4 Storage of Radioactive Wastes in Seas and Oceans The method of dumping radioactive wastes in solid or solidified form into deep depressions and faults on the ocean beds meets most demands of the requirements of safe storage. This method has been used since the 1940s and is especially advantageous for countries with unfavourable geological and geographical conditions. Such operations have to be carried out under strict international supervision and have to be approved. Such approval is based on a very careful estimate of the possible hazards of such storage. The results of several storage operations carried out in recent years have shown that even the most unfavourable events, such as damaged packaging, failure of containers and the release of radioactive substances into water are not hazardous for marine life or human beings. The flow in these deep ocean layers and the exchange of water between the individual layers are negligible. On the other hand, it should be noted that the strict technical conditions set for the dumping of containers into deep sea waters and the technology of the preparation of wastes designed for such storage almost completely exclude any failure of the containers and any major release of radioactive substances. All such operations are strictly supervised by qualified specialists of the International Atomic Energy Commission and the selection of sites sufficiently deep and distant from fishing areas ensure the maximum safety of this type of storage. No packaging can, however, be failsafe. There is no technology of waste solidification which totally prevents the leaching of radioactive substances when such

225

substances are in contact with sea water for a long period of time. For these reasons only radioactive wastes cotaining short-lived radionuclides, i.e., with half-lives of the order of years may be stored in seas and oceans. After several decades their radioactivity will have decreased to such a level that even their release into the ambient medium in large amounts will not constitute any hazard to the population following ingestion. Specialists attending international conferences on marine ecology have often stated that the natural radioactivity of the oceans is itself several orders of magnitude higher (lo9- 10" TBq) than that of radioactive wastes stored in this manner. Dose limits and dose committments recommended by the International Commission for Radiological Protection as being safe for the population would only be attained if the radioactivity of the wastes were in the region of the following values: wastes with alpha-active nuclides 3.7 x 10' TBq (10" Ci) annually wastes with beta-active nuclides 3.7 x lo9 TBq (10" Ci) annually wastes with tritium 3.7 x lot3TBq (10'' Ci) annually In fact, less than IOOOTBq (2.7 x 104Ci) of different radioactive substances are dumped into the world seas and oceans annually. The wastes are solidified by bituminization or cementation and are placed in metal drums which protect the contents during dumping. In order that the container should withstand the high external pressure of sea water it has apertures which help to balance the pressure differencz. In the course of four operations carried out in 1967-1972, a total of 50 000 containers with a total weight of 2.76 x lo7 kg of wastes and a radioactivity of 2.4 x lo3 TBq (64 000 Ci) were dumped in the Atlantic Osean. The area in which these containers were dumped is 450 km distant from a continental shelf, 900 km southwest of Cape Finisterre, where the ocean reaches a depth of 4.5 - 5 km. The countries taking part in these operations were Belgium, the Netherlands, the Federal Republic of Gxmnny, Switzerland, Sweden, France, Italy and Great Britain. Previously this method has only been used on a broader scale by the USA and Great Britain; in the USA six areas were selected for radioactive waste dumping, three being in the Atlantic Ocean and three in the Pacific Ozean. The sites selected were distant from regular maritime routes, from fishing areas and fisheries and from communication cables on the sea bed, at a minimal depth of 2000 m. The wastes were dumped in drums, reinforced with concrete, with a volume of 0.25 m3, or were fixed by a cement mix in the drums and allowed to harden. The highest permissible radiation level on the surface of the drums was 2.8-3.5 nA . kg-' (30-50 mR/h). In 1946- 1956 about 20 000 containers containing solid and solidified wastes including contaminated parts of reactors and hot chamber equipment with a total radioactivity of 600 TBq (16 000 Ci) were dumped in the area. Under British regulations for dumping wastes in the Irish Sea the maximum dose rate on the drum surface is set at 35 nA. kg-' (0.5 R/h), and the maximum annual permitted amount of packaged wastes dump2d in the Irish Sea was set

226

at 5 x lo6 kg. The maximum pcrmitted radioactivity of alpha-emitters was set at 7.5 TBq (200 Ci) and of beta-emitters at 150 TBq (4000 Ci) per year. The great disadvantag: of this storage technique is its high operating cost. Balast, i.e., the steel container and the concrete shielding, usually amount t o 20-70 % of the total wzight of the dumped material and, with high-level radioactive wastes, shielding may make up to 95 % of the total weight. The dumped wastes are transported by truck to the port, then shipped to the area of burial. The costs of the radiological protection of personnel and the eventual decontamination of the ship are not negligible and the specific unit costs are in the range US $ 0.50-0.90 kg. This was also one of the reasons why the USA abandoned this storage technique and why Great Britain is introducing the less costly storage of radioactive wastes in surface geological formstions.

9.5 Waste Effluents Released into the Environment 9.5.1

Radioactive Wastes Released into the Atmosphere

The discharge of gaseous and volatile radioactive substances and radioactive aerosols from nuclear installations into the atmosphere presents a serious danger to the population in the local area, which is exposed to the potential danger of irradiation resulting from the inhalation of contamined air, the ingestion of agricultural products contaminated by discharged radioisotopes, the deposition of radioisotopes in aerosol form on the body surface, clothes and vegetation and in extreme conditions the direct effects of radiation emitted from a cloud of discharged radioactive substances. The evaluation of the environmental impact of radioactive effluents is based on analysis of the travel of radioactiie substances in the atmosphere in timeand sp~c:, which is conditional on the quantitative description of the process:s of the travel of radioactive effluents in the atmcsphere and its deposition on different surfaces in the surrounding area of the source. The radioactive plume has a characteristic shape and its dimensions and travel in the atmosphere depend on turbulent diffusion. Turbulence in the lower layers of the atmosphere results mainly from changes in the temperature of the atmosphere with height above the earth's surface and by unevenness of the terrain, be it natural or man-made. Under normal conditions the temperature of the air will decrease with h i g h t at the rate of 0.6 "C per 100 m on average. As a result, warm air will move upwards. The atmosphere is unstable and enhances the mixing of atmospheric layers, and thereby also the diffusion and dilution of the radioactive cloud. In an adverse case, when the temperature of the air incrzases with height (thermal

227

inversion), the cold air and thus also the radioactive effluent remain at ground level and its diffusion and dispersion is considerably limited. It is thus evident that the temperature profile of the atmosphere in the vertical direction as an indicator of the stability of the atmosphere will significantly effect the travel of emission

-

Temperature STRONG LAPSE CONDITION (LOOPING)

Temperahe

--L

WEAK LAPSE CONDITION (CONING)

Temperature

4

/NVERSION CONDITION (FANNING)

I h

\

E m p afure

-

Bmperafure

INVERSION BELOlt: LAPS€ ALOFT (LOFTING)

-

,

.

I

-\

,..

LAPSE BELO1U: INVERSION ALOFT (FUMIGATION)

Fig. 36. Effect of the temperature profile of the atmosphere in the vertical direction on the plume

Passage

clouds from the source to the ambient area (Fig. 36) as well as the concentration of radioactive substances in the ground layers of the atmosphere. Models for the quantitative description of the travel and dispersion of radioactive emissions from a point source are based on the K theory and on the theories of mathematical statistics.

228

The K theory is based on the partial and diffxential equation for diffusion in space, which can be written as (9.9) where C is the concentration x, y, z are the distances in the three-coordinate system and K, , K,, K, are diflision coefjcients. The values of the diffusion coefficients change considerably with the characteristics of atmospheric flow, from 2 x m2/s for molecular diffusion to lo5 m2/s for powerful cyclone storms. This means that the rate of diffusion is considerably dependent on the movement of the atmosphere at any given moment of the passage of the cloud. As it is virtually impossible to estimate the time dependence of the magnitude of coefficient K , the practical solution of the diffusion equation of the Ktheory has not been formulated (with the exception of several simple, specific cases). The second model for the calculation of the travel of effluent clouds is based on the statistical theory of turbulent diffusion, which, because of its relative mathematical simplicity and practicil applicability, is used in most industrial countries. The basic relationship for diffusion in a homogeneous environment for an effluent from a point source was derived from this model. The equation can be written as

where C is concentration (volume activity) of the radioactive substance at time t and at a location characterized by coordinates x, y , z in Bq/m3 (Ci/m3), Q is the intensity of the source, axuyand o, are standard deviations from the statistical distribution (diffusion parameters) and u is the wind velocity. Equation (9.10) may be converted into Fick’s diffusion equation (9.9) by relating the standard deviations to the diffusion coefficients K . The equation is then written as oi = 2 K i . t .

(9.11)

For steady-state continuous emissions of radioactive substances from a source at height H above the earth’s surface, on condition of complete reflection of the cloud from the earth’s surface (this assumption is fully justified for radioactive rare gases; for other forms of emitted radioactivity it leads to a certain overestimation of concentrations), the concentration C at a point with coordinates x, y , z can be expressed by the equation

The numerical values of the diffusion parameters cry and u, depend on the atmospheric stability of the wind, the wind velocity and the characteristics of the

229

terrain. If a highly accurate analysis of the travel of the radioactive cloud is required it will be necessary to determine the values of the diffusion parameters for the given location experimentally.

Fig. 37. Variation of diffusion parameter 0; with distance from source of effluent

-

Distance horn source (m)

Table 34: Pasquill Stability Classes Surface wind speed mlsec (at 10m)

Day-time insolation high moderate slight

Night-time conditions cloud cover 318

118

A A-B B C C

A-B B B-C C-D D

B C C D D

-

-

E D D D

F E D D

For most practical applications fully adequate tabulated or graphically represented values of uy and a, were developed using Euler’s correlation function for the fluctuation of wind direction and velocity by MEADand PASQUILL (45, 46) for six classes of atmospheric stability. The stability classes are given in Table 34. Figs. 37 and 38 show the dependence of diffusion parameters ay and a, on the

distance from the source of the effluent for each class of atmospheric stability. The suitability of the given model and from it the ensuing variation of atmospheric stability classes with diffusion parameters have been tested in a number of field measurements with regard to its wide practical applicability. The tests showed that this model, in combination with PASQUILL’S classification of atmospheric stability, represents the conditions of relatively low diffusion parameters. Hence its application to continuous emission sources in varied terrain may lead to an underestimation of the diffusion of radioactivity at greater distances from the source.

Fig. 38. Variation of diffusion parameter u, with distance from source of effluent

By relating concentrations of emitted radioactive substances to wind velocity and to the intensity of the source, HILSMEIER and GIFFORD (47) obtained a characteristic set of curves (Figs. 39,40 and 41) which show that the point of maximum concentration travels from the source in the direction of the wind with increasing height, H , and increasing atmospheric stability, while its height decreases. It follows from the solution of equation (9.12) that maximum ground concentration is given by the equation (9.13)

where e is the natural base of the logarithm. Equation (9.13) shows that for practical purposes it is important that the concentration of emitted substances at the given distance from the source in the direction of the wind depends mainly on the rate

23 1

of emission of radioactivity from the source, whereas in such a case the volume concentration of the contaminant in the emissions is not decisive. The concentration of emitted substances in the surrounding area of the source thus cannot be reduced by the simple dilution of emissions with non-active gas. The concentration of emissions in the cloud is defined by the plume rate and the volume of the cloud, which is a function of meteorological parameters and cannot be influenced by the volume of emitted gases in relation to the mass (volume) of the environment in which the emitted radioactive substances spread.

-

X,

metres

Fig. 39. Variation of emission concentration with distance from source in the direction of the wind for different types of weather; source at ground surface level

The above basic relationships for the calculation of concentrations of radioactive substances in the surrounding area of the source must often be modified to fit the actual spread of emitted radioactive substances in the given area using various corrective factors. These corrective factors quantitatively characterize the effect of changing topographic and atmospheric parameters and of the physico-chemical properties of the cloud on the travel of the radioactive cloud. These corrective factors include mainly the correction for change of emitted radioactivity in the process of the travel of the radioactive cloud owing to radioactive decay, fallout, rainout or washout from the cloud, correction for the change of the effective height of the source above the surface of the terrain owing to different effluent rate and temperature in the stack crown in comparison with the ambient atmosphere, correction for change in velocity and direction of wind with height and time,

232

-

X , mehs.

Fig. 40. Variation of relative emission concentration with distance from source in the direction of the wind for different types of weather; source at a height of 30m

-

x, metres

Fig. 41. Variation of relative emission concentration with distance from source in the direction of the wind for different types of weather; source at a height of 100m

233

correction for emission time, correction for topography (roughness of terrain), and correction for the effects of inversion layers in the atmosphere. The relationships for the calculation of the individual correction factors and the manner of their application in real situations have been analysed in detail in the literature (48, 49). The discharge or escape of radioactive substances into the atmosphere is connected with the potential danger of exposure of the population living in the surrounding area. Evaluations of the effects of radioactive emissions on the human organism usually assume that exposure of the organism occurs especially: a) by inhaling radionuclides contained in the inhaled air, b) by the ingestion of radionuclides which have entered the food chain after they were deposited on the soil or on the vegetation, c) by external /I-irradiation from a cloud, d) by external y-irradiation from a cloud. Under certain conditions, added to these so-called critical exposure paths are external /I- and y-irradiations from the deposit of long-lived radionuclides on the terrain and, in the case of an accident, also the direct radiation of the emission source. For the critical exposure paths the radiation dose exposure of the organism to ionizing radiation from emitted radionuclides may be calculated using the equation dose = source intensity x dilution factor x dose factor. The basic diffusion relation may be used for calculating inhalation, ingestion and external beta-irradiation. This will be done by expressing the ratio of the ground volume activity (concentration) of the emitted radionuclide to the radionuclide emission rate from the source or by relating the time integral of the ground concentration to the total amount of radionuclides emitted from the source in the given time interval. The dose factor includes the radiation effect of the given radionuclide on the organism with regard to the critical exposure path, its metabolism in the human body and its physical characteristics, namely the type of energy of the emitted radiation and the half-life. The numerical values of dose factors for selected, radiobiologically significant radionuclides were given by COMPER (50). In the determination of the radiation dose exposure of the organism by external y-irradiation, the calculation of the dilution factor should not merely be based on the distribution of ground concentrations in the area surrounding the source. At a given point it is also necessary to consider the dimensions of the radioactive cloud in relation to the mean linear range of the y-radiation of emitted radionuclides in the atmosphere. If the mean linear range of the y-radiation in the cloud is not negligible in comparison with the dimensions of the cloud, it is necessary to integrate in the calculation of the dilution factor the energy fluence of y-radiation in the individual volume elements ( X , X,2 ) occurring at distances r = d ( X - x ) ~+ ( Y + Z 2 from the given point

234

across the whole volume of the cloud for the respxtiv: distribution of volume activities (concentrations) in the cloud, the values of linear absorption coefficients, u, and growth factor, B, of the y-radiation in the atmosphere. The dilution factor thus obtained, defined generally by the equation B e-” exp (-

dXdYdZ (9.14) 2% may, following multiplication by the source intensity and the y-dose factor, be converted t o the value of the dose which characterizes external y-irradiation of the organism from a cloud of emitted radioactive substances. The interrelation between the radioactive emission from nuclear installations and the contamination of the environment is characterized by diffusion relations whose application for the radiation protection of the population and for environmental control purposes is based on two fundamental prerequisites. In the operation of nuclear installations it is essential on the one hand to determine dose commitments of the population living in the area surrounding the installation in relation to the known emission rates of emissions or to calculate the maximum permissible emission rates from the nuclear installation with regard to the permissible threshold doses for each inhabitant, for the critical group of the population or for larger population groups. In boths instances the basic data thus obtained may be interpreted and applied to: a) the calculation of dose rates in the area surrounding the nuclear installation from the results obtained from monitoring the effluent of the installation, b) the determination of dose rates in the area surrounding the nuclear installations in case of an accident and during normal operation of nuclear installations as criteria for decision making on their optimal sitingas well as for the preparation of safety reports on the operation of nuclear installations; c) designing safety and control systems for nuclear installations; d) comparative studies of the environmental impacts of nuclear and conventional power sources. The practical application of these models and the respective relationships for the travel of radioactive emissions to the evaluation of the environmental compact of the emissions is facilitated by the numerical processing of diffusion relationships whose results for various combinations of input parameters and correction factors may be found in the form of graphic dependences, tables, nomograms and isodose and isolethic maps in the literature (51, 54). -m

9.5.2

Waste Discharge Into Surface Waters (Rivers, Seas and Oceans)

Most countries producing large amounts of low-level liquid wastes have now accepted the British principles, methods and practical application of the discharge of wastes into coastal waters. Thanks to the active approach of most international

235

bodies and organizations, limits have been set based on maximum permissible concentrations of the individual radionuclides in all links of the food chain, starting from water. These permissible concentrations are mostly rational limits which, given the current state of scientific knowledge, constitute the required measure of safety for man and his environment. All authors of national and international legal norms are agreed that the amounts of wastes discharged should be as low as possible and that the capacity o r the surrounding area to compensate for the intake of radioactive substances contained in these wastes should be taken into consideration. In many instances the environmental impact is not restricted to the country of origin. Trans-boundary emission travel is mainly typical of gaseous emissions, yet it also applies to liquid radioactive wastes discharged into large international rivers. Coastal currents may also carry the wastes to the shores of other countries. Only international agreements which have already been concluded for certain European rivers will make it possible to solve this problem on a wide scale, i.e., in such a manner as to prevent a situation from occuring in which countries on the lower flows of large rivers would not be able to discharge into the river amounts of waste permitted under the law owing to an increased level of radioactivity in the water resulting from excessive discharges by countries on the upper flow. It is obvious that in the current state of development of nuclear power production and probably also in the coming 15-20 years such a situation is not likely to occur. Currently the radioactivity of all discharged wastes is much less than the permissible limits and the prospects of maintaining the purity of the surface waters, i.e., rivers, seas and oceans, are good. Samples of water taken in the area surrounding the US nuclear power plants Dresden-1, Yankee and Indian Point showed a level of radioactivity that could have no effect on the population dose of the local population. The discharge of liquid wastes into coastal waters has so far not caused any noticeable increase in the level of the natural background radiation in the seas. Most of the contamination of w x l d oceans comes from nuclear weapon tests, yet the levels of long-lived radionuclides 90Srand 137Csin 1963- 1969 were as follows:

... Atlantic Ocean 1.5

-5.8 Bq m-3 137Cs... Atlantic Ocean 2.8 - 10.8 Bq m-3 90Sr ... Pacific Ocean 1.5 -4.2 Bq m-3 137Cs... Pacific Ozean 1.05-5.3 Bqm-3 A comparison with the values recommended by the International Commission for Radiological Protection for maximum permissible concentrations of radionuclides in waters (30 kBq m-3 or 55 MBq m-3) shows that in this respect the situation is more than satisfactory. The situation is different in the so-called closed seas. Soviet measurements carried out in 1969 - 1971 showed a six-fold higher concentration of both radionuclides in the Baltic Sea and a ten-fold higher concentration in the Black Sea.

236

This is probably caused by the relatively small depth of these two seas and by the relatively high concentration of these radionuclides in the soil (600 - 700 Bq m-') in the coastal areas which are washed by several river flows, thus contributing to the permanently high level of both radionuclides in the waters of these two seas. Legal norms and permissible maximum amounts of liquid radioactive wastes which may be discharged into surface waters should be adopted and implemented only after thorough hydrological and hydroecological surveys of the given locality. Such surveys should include: - the study of the dispersiveness of the medium and its dependence o n river or sea currents, the tide, the wind, and the distance of the point of discharge from the shore. Many of these data may be known from existing hydrological reports. It is recommended, however, that they be specified, using conventional or radioactive tracers; - the study of the biological concentration of important radionuclides in the aquatic organisms characteristic of the given area; - the study of the sorption capacity of the sediments for the individual radionuclides and the study of their effects on establishing an equilibrium, - the study of the effects of the composition of the discharged wastes on the dispersion and sorption processes in the given medium. On the basis of the critical paths by which radioactive substances may reach man and threaten him by doses and dose commitments, it is then possible t o determine the critical radionuclide. The permissible amounts will then be calculated from its proportional representation and with regard to the other radionuclides present in the discharged wastes. In most instances the critical path will be the consumption of those aquatic organisms in which are reconcentrated long-lived toxic radionuclides. The exposure of the population during fishing and swimming cannot be excluded. At Windscale the following maximum permissible concentrations were set for discharging liquid wastes into the sea: coastal sand coastal silts sea bed fish meat

other foodstuffs

1 MBq kg-' 4 MBq kg-' 4 MBq kg-l cca. 1 kBq kg-' 74 .Bq kg-' 37 kBq kg-' 300 Bq kg-' 0.3 kBq kg-' 25 kBq kg-' 9 kBq kg-' 100 Bq kg-'

(25 nCi/g) (100 nCi/g) (100 nCi/g) (30 nCi/g) (2 nCi/g) (1 nCi/g) (8 pCi/g) (10 pCi/g) (670 pCi/g) (330 pCi/g) (2.7 pCi/g)

all beta emitters all beta emitters all beta emitters all alpha emitters all beta emitters '06Ru

90Sr all alpha emitters all beta emitters lo6Ru 90~r

237

10 Transport of Radioactive Wastes Radioactive waste disposal is not possible at source and such wastes must therefore be transported not only on the site of the waste-generating installation but also along public roads. Handling and transportation on the site of the installation are relatively simple and do not require expensive containers. Any transport of radioactive wastes on public roads must be carried out observing national and international regulations. These regulations haw mostly been drawn up for the transport of emitters and small-sized radioactive materials or for the transport of spent fuel elements, and it therefore sometimes becomes very complicated to observe them for the transport of radioactive wastes.

10.1 Transport of Radioactive Wastes on Installation Site The transport of radioactive wastes on the site of the installation is relatively simple as it is fully within the responsibility of the operator, and persons not acquainted with handling such materials are not allowed access. The basic requirements of workplace safety for processing and handling radioactive material must be observed and the exposure of employees during transport must be maximally limited. The most frequent occurrence is the transport of solid materials. These wastes are classified and collected separately into containers with plastic-bag liners which, when full (or after every shift) are sealed and put into a larger bin and placed in a temporary storage area. Depending on the handling technology these bins are either used for transport only and following decontamination are re-used or are deposited with the container in which the unprocessed waste is deposited. Depending on their number and the distance over which they are transported, the bins are either carted or hauled by trucks either separately or on palettes. For the disposal of liquid radioactive wastes, experience has shown that the use of pipelines for their transport is only ecDnomic with sizeable amounts of wastes.

238

Many installations generate smaller amounts of such wastes, which must also be collected and transported separately. Small amounts of liquid wastes are therefore often collected in laboratories in bottles or in other suitable plastic containers in which they are also transported. If the liquid wastes are to b: transported to a greater distance they are pumped into drums in which they are transported by special trucks or hauled on special trailers.

10.2 Organization of Waste Collection The centralized collection of radioactive wastes (1 1) has been introduced in almost all socialist countries and in a number of other countries. In the USSR the Central Station for Radiation Protection is responsible for the transport of radioactive wastes from the generating installation and its secure disposal at the burial site. Under current hygiene regulations all solid and liquid wastes containing radionuclides with a half-life of more than 15 days must be disposed of on these central burial sites. Wastes containing radionuclides with a shorter half-life are stored on the site of the waste-generating installation and after decay are removed as ordinary waste. The Central Station removes unlimited amounts of liquid wastes with a specific activity of more than 3.76 Bq/m3, while lower level liquid wastes are removed by the Station only if their amount does not exceed 0.2 m3 (200 1) per day. An installation which generates more than 200 I of low-level liquid wastes per day is responsible for their processing. All liquid wastes, regardless of their chemical composition, are removed only in a neutral state. Solid radioactive wastes are removed in suitable containers which cannot be disturbed by handling or transport and from which the radioactive material cannot spill. Paper and plastic bags must not be used for the transport of sharp or pointed metal objects, broken glass, etc. Also, the surface of the containers should not be contaminated after filling. The various installations conclude agreements with the Central Station on the average amount and type of wastes to be disposed of and cost of their removal. The route to be used for the transport of the wastes must be approved by the hygiene authorities. The wastes are prepared for transport by the installation itself and are removed by the Central Station, whose surveillance officers check each batch before transportation, especially for suitability of the bags and containers and their surface contamination, the p H of liquid wastes, the intensity of radiation on the surface of the containers, etc. In the USSR radioactive wastes may be transported only by special trucks whose design has been approved by the hygiene bodies and by the Ministry of Hzalth. Several types of vehicles are used for this purpose.

239

The OT-2 truck is designed for the transport of solid wastes and has a tilting body of stainless steel with a volume of 3.3 m3. The individual parts of the body are joined so as to prevent any leakage of material that might spill from the containers as well as the infiltration of atmospheric precipitation into the load. The wastes are loaded on to the truck through six openings (0.6 xO.6 m) on the sides of the body. The back wall is fitted with a hinged door, which is secured against opening during loading and transportation. The driver is protected by a lead door 0.015 m thick. The 0 2 - 2 tanker is used for the mechanical filling, transport and discharge of liquid wastes with a specific activity of up to 3.76 Bq/m3 (0.1 Ci/m3) 6oCo.Tanks with a volume of 1 m3 are made of stainless steel and placed in a cast-iron jacket. Vacuum filling of the tank is carried out using the engine of the vehicle - the air evacuated from the tank is first run through an aerosol filter. From storage vessels the wastes are pumped into the tanker by special portable pipes with special sheathing, The tank is fitted with special control equipment to prevent spilling. The height of the level is shown on the control panel in the driver’s cockpit. The wastes are discharged by gravity and the tank is flushed with water from a revolving jet. The cast-iron shielding has a thickness of 0.12 m in the direction of the driver and 0.05 m on the other three sides. The 02-1 tanker is designed for transport of liquid radioactive wastes with a specific activity of up to 37 MBq/m3 Ci/m3). It is made of stainless steel and has a volume of 1.7m3. The cockpit is separated from the tank by a lead shield 0.15 m thick. The technology of filling and discharge and the design and equipment of the tanker are the same as with the 02-2. Solid radioactive wastes deposited in large containers are transported by the OKG-I truck, which has a tank with a volume of 9 m3 made of 2 x m thick stainless steel. The wastes are loaded and unloaded by a special crane mounted on the truck. Each of these special vehicles is equipped with a radio transmitter and receiver that enable the driver to keep in contact with the control room, an emergency signal system and cabinets for equipment, protective clothing and aids and an emergency fence. The wastes are always transported in truck convoys with a police escort. In Czechoslovakia the collection of solid wastes and small batches of solidified liquid radioactive wastes from radioisotope workplaces is carried out by the Institute for the Research, Production and Application of Radioisotopes in Prague. At the workplace the solid wastes containing artifical radionuclides are deposited into special transport and storage containers of zinc-plated sheet with a volume of 0.05 m3. Incinerable and non-incinerable wastes are deposited separately and small batches of liquid wastes are solidified with cement in the containers. Wastes with natural radionuclides are deposited into 0.1 m3 (100 1) metal drums whose maximum weight must not exceed 400 kg.

240

The containers are transported by a standard SKODA RT 706 truck equipped with a crane. The containers should be placed on the truck in such a manner as to ensure that the radiation intensity in the driver’s cockpit does not exceed 0.15 Sv/s (50 mrem/h) and 0.6 Sv/s (200 mrem/h) on the surface of the truck. In practice only fractions of these values are reached. The vehicle is marked with a warning sign indicating radioactive radiation and written instructions for the driver are displayed in the cockpit. No approval from the hygiene service or an escort are required for indidivual transport of radioactive wastes.

10.3 Transport of Radioactive Materials National and international regulations for road, maritime, rail and air transport are based on the Regulations for the Safe Transport of Radioactive Materials published by the International Atomic Energy Agency ( 5 5 ) . Low-level solid and solidified liquid radioactive wastes may in most instances be classified as low-level radioactive materials, which are defined by the following properties: - activity under normal transport conditions is scatterred throughout the whole volume of the solid material or in a greater number of solid objects or is evenly distributed in the solid matter of the solidification agent (cement, bitumen, ceramic materials). - activity is and remains in an insoluble form; even after the destruction of the container the dispersion of the radioactive material from the container owing to the effects of wind, water or other factors or as a result of the total immersion of the material in water less than one tenth of the maximum activity which may be transported under the regulations will be released in 1 week, - items of non-radioactive materials which are contaminated with radioisotopes that are difficult to remove and where the average level of contamination per square metre (or on the whole surface if it is smaller than 1 mZ)does not exceed 7.46 Bq/m2 (20 pCi/cm2) for beta and gamma emitters and for low-toxicity alpha emitters and 740 MBq/mZ (2 pCi/cmz) for other alpha emitters. Some low-level radioactive wastes may be classified as materials with low specific activity. In addition to uranium and thorium ores, their physical and chemical concentrates, non-irradiated natural or depleted uranium, non-irradiated thorium and aqueous solutions of T,O with a concentration of up to 370 TBq/m3 (10 Ci/l) it is possible to include in this group: - materials in which under normal transport conditions the activity is evenly scattered and whose average specific activity does not exceed 1/10 000th of the maximum activity permitted under the transport regulations, - materials in which the activity is evenly scatterred provided that owing to conditions which may occur during transport, e.g., dissolution in water with

24 1

subsequent recrystallization, precipitation and evaporation or burning, etc., the product of such a process will not have a specific activity higher than 1/10 000th of the maximum activity which may be transported under the regulations, - items of non-radioactive materials contaminated with radioactive materials provided that the free surface contamination at any point of the surface with an area of 300 cm2 does not exceed by 10-fold the following values: natural and depleted uranium, natural thorium /I and y emitters and low-toxicity a-emitters (23sU, 238U,232Th,228Th,230Th)diluted to the concentration of natural uranium and thorium and radionuclides with a half-life of up to 10 day other a emitters

370 kBq/mZ

pCi/cmz)

37 kBq/m2 (lod4 pCi/cmz) 3.7 kBq/mZ (loe5 pCi/cmz)

A further condition is that the concentration of the contamination into a minimum volume as a result of the conditions which may occur during transport (e.g., solution in water with subsequent precipitation and recrystallization or burning) should not raise the specific activity of the concentrate to more than 1/10 000th of the maximum activity permissible under the regulations, - items of non-radioactive materials contaminated with radioactive materials, where the radioactive contamination is difficult to remove and where the average contamination per square metre (or the whole surface if it is smaller than 1 mZ) does not exceed 370 MBq/m2 (1 pCi/cm2) for /I and y emitters and low-toxicity a emitters and 37 MBq/m2 (0.1 pCi/cmz) for other a emitters. Table 35 shows the specific activity (Bq/kg) and maximum activity of some radionuclides permitted for transport. These data apply only to the individual radionuclides; regulations for determining the maximum activity for a mixture of radionuclides are specified in detail in the IAEA Regulations. If radioactive wastes may be classified as low-level radioactive materials or as materials with a low specific activity, and if they moreover meet the following requirements: - the radiation intensity at no point of the surface of the container exceeds 1.5 x 10-3Sv/s (0.5 mrem/h), - the minimal outer dimensions of the containers containing 232Uare 0.1 m and the amount of 235Udoes not exceed 0.015 kg, then it will be possible to establish conditions permitting the transport of such materials. These materials may be transported on condition that: a) the total amount of materials loaded on to a truck or sea-going ship does not exceed 1.85 PBq (50 000 Ci) of tritiated water or with liquids 100 times the amount which under the regulations may be transported in one container (the amount of solid materials is not restricted). For river transport the total volume of solid and liquid materials is limited to 100-fold the maximum amount permitted for one

242

Table 35: Maximum Activities of Some Radionuclides Permitted for Transport

Radionuclide

41Ac 13'Ba '4C 45Ca 1 3 9 a

249Cf 6OCO "Cr I37CS 1.311

42K 140La fission products 24Na 95Nb 237Np Z39NP 32P 239Pu 224Ra Io6Ru 35s

90Sr 65Zn 95Zr

Maximum activity permitted for transport

740 1480 8700 1480 8700 0.074 259 22 200 333 370 370 1110 14.8 185 740 0.183 7400 1110 0.074 18.5 259 11 100 14.8 1110 740

Specific activity (GBq/kg)

159.1 X loL o 321.9 x lo' 170.2 70.8 x 107 240.5 x lo6 114.7 40.7 x lo6 340.4 x 107 3,626 44.4 x 108 222 x 109 207.2 x lo8 321.9 x 144.3 x 255.8 x 85.1 x 107.8 x 229.4 x 59.2 x 125.8 x 159.1 x 55.5 x 296.0 x 77.7 x

lo9 lo7 lo-' lo8 lo8

lo1 lo8

lo6 lo7 10s

lo8 107

consignment and to 185 TBq (5000 Ci) of tritiated water. The cargo must be loaded in such a manner as to preclude, under normal transport conditions, the dispersion of the cargo owing to the leakage of the vehicle or ship, b) the containers meet the following requirements: - the container should ensure easy handling and reliable containment of the transported material, - containers weighing 10-50 kg must be fitted with equipment for manual hand ling, - containers weighing more than 50 kg must be fitted with equipment for safe mechanical handling, - the outer surface of the container should not retain water and should be easy to decontaminate,

243

- the minimal outer dimensions of the container must not be less than 0.1 m, - the contamination of the container must be as low as possible and in no instance should it exceed the maximal permissible values, - where radiation intensity does not at any point of the surface exceed Sv/s (0.5 mrem/h) the containers are painted white, 1.15 x - yellow containers are used when radiation intensity on the surface exceeds 1.5 x Sv/s (0.5 mrem/h) but does not exceed 0.6 Sv/s (200 mrem/h) at any point of the surface, - each container must be marked with at least two radiation warning signs. These signs are placed on the outer opposite walls of the container or on all four walls of the container. The regulations are much stricter for the transport of high-level radioactive wastes. In addition to the above requirements the containers: - must be designed in such a manner as to preclude the initial radiation intensity of 3 x (10 mrem/h) of Ig2Irto exceed 3 Sv/s (1 rem/h) to a distance of 1 m from the container. The fulfilment of this conditions may be tested by gradually submerging the container in water to a depth of 0.9 m for at least 8 h, a free fall on to a solid surface from a height of 9 m and 1 m and exposure to a temperature of 800 "C for a period of 30 min, - must be designed in such a manner as to ensure that the loss of activity 1 h after the control checks will not be more than 1/6th of the maximal activity permissible for the transport of such a consignment. Such checks will include: a) even spraying with water at an angle of 45" from the vertical axis in a n amount corresponding to a level of precipitation of 0.05 m/h for at least 1 h the container must not be submerged in water, b) free fall on a flat solid surface from different heights depending on the weight of the container (5000- 10 000 kg, 0.9 m; 10 000- 15 000 kg, 0.6 m; more than 15 000 kg, 0.3 m), c) compression load test which corresponds to a minimally five-fold weight of the container and a load to 1300 kg per square metre of the upper surface. The container is tested in the position in which it is usually transported, d) penetration test by the free fall of a bar 6 kg in weight, 0.032 m in diameter with a hemispherical point from a height of l m, - the total loss of the radioactive content in 1 week must not exceed 1/1000 th of the maximum permissible transported amount. The following tests - free fall of the container from a height of 9 m on to a flat solid surface and the free fall of the container from a height of 1 m on to the flat upper surface of a 0.2 m long rounded steel bar, 0.15 in diameter with a maximum radius of 6 x lo-' m; b) temperature endurance tests follow the mechanical tests. The container is left at an ambient temperature of 800 "C for 30 min, c) finally, the container is submerged in 15 m of water for at least 8 h;

244

- the container must be dcsigned in such a manner as to ensure that not even a t an ambient temperature of 38 "C will decay heat affict the transport conditions, i.e., to ensure that changes will not occur in the configuration, shape and state of the radioactive material, that the properties will not deteriorate owing to the cracking or melting of the shielding material and that the temperature on the accessible parts of the surface of the container will not exceed 50 "C, - the maximum operating pressure in the container must not exceed 7 kg/cm* continuous ventilation of the container during transport is not permitted, - continuously operating cooling or mixing equipment should not be part of the container, - the container must not contain a pressure safety valve which would allow the escape of radioactive material in trial conditions.

245

11 Economic Problems of Radioactive Wastes Disposal

The protection of the health of the population and of the environment are the most important criteria for any considerations on processing and discarding radioactive wastes. It is therefore generally accepted that only such amounts of radionuclides as may be detected using currently available technologies may be discarded into the environment. In the development and application of these technologies the factor of economy will be a major criterion. The total elimination of radioactive waste discharges would be extremely costly and is hardly ever necessary. In the past, unnecessarily strict hygiene regulations in some instances greatly increased the costs of handling radioactive wastes while on the other hand attempts to reduce the costs of wastes processing will sometimes increase the costs of measures that will have to be taken to protect the existing environment or to reclaim an environment that has been disturbed, damaged or devastated. A reasonable balance should therefore be maintained between the capacity of the environment to receive wastes safely and the costs of processing these wastes so as to preclude any health or environmental hazards. Waste handling includes the following operations, the costs of which will have to be included in any economic consideration: - waste collection, - transport of wastes to the processing plant, - monitoring before processing, - processing, - monitoring of processed wastes before discharge, - discharge of processed wastes, - processing of radioactive concentrates prior to disposal or removal, - storage or disposal of concentrates, - environmental control. Other costs to be included in the economic consideration will include the costs of research and development, the development and manufacture of new equipments, the development of new technologies, the updating and upgrading of existing equipment and technologies and the solution of current problems of operation. The

246

last category also includes the costs of an environmental survey, which should precede any concept of waste processing. The costs of processing radioactive wastes vary from country to country and may differ from workplace to workplace even within one country. This is caused by many factors, namely: - different activity levels and volumes of processed wastes, - intensity of the use of equipment. In some instances the capacity of the wasteprocessing installation is determined by considering envisaged future needs, while elsewhere it will meet the current requirements of the nuclear installation. In the former instance the transport capacity will be used very inefficiently and sporadically, which in turn will result in a high depreciation of the equipment and will raise the costs of the whole operation, - the decontamination factor required. This factor does not depend solely on the activity level of the radioactive wastes but also on the permitted level of radionuclides discharged into the environment. The higher the decontamination factor the more complex will be the technology, the higher the depreciation and the higher the costs of operation and maintenance, - different chemical and physical composition of the wastes, - different costs of operation, the effects of climate, availability of energy, etc., - overhead costs, - different capital costs and amortization. The economy of radioactive wastes processing is affected by a number of factors which cannot be influenced by the nuclear installation itself. The most common case is a change occurring in the composition of the wastes owing to changes in the research and production programmes of the installation. Liquid wastes may contain various amounts of salts and suspended solids, and they may be contaminated with only one radionuclide or a mixture of many radionuclides having different physical and chemical properties. Solid wastes may also consist of various materials having widely differing properties. The costs of waste processing are greatly affected by the siting of the nuclear installation. Geological and geographical conditions may necessitate costly processing technologies, while the high capacity of the environment in another area will allow simple and inexpensive methods. In many countries demands on processing technologies are strongly influenced by public opinion. On the other hand, there are problems whose solutions is entirely the responsibility of the nuclear installation. On the site of the installation the project designer will separate the area in which he will lay out buildings where radioactive materials will be handled. This will make it easier to prevent the dilution of radioactive materials in non-active material and the contamination of ncn-radioactive wastes with radionuclide wastes which in both instances would immensely increase the costs of processing. The sorting of wastes inside the individual buildings on the site of the nuclear installation is carried out along the same lines.

247

Another example is the monitoring of the contamination of the environment with radioactive wastes. Under certain conditions strict surveillance inside the nuclear installation together with costly waste processing may considerably reduce the costs of environmental control, while low-cost waste processing will enforce costly environmental control measures. Both alternatives show that although waste disposal will not be an environmental hazard the choice of the feasible variant will depend on local conditions. Most published data on the costs of waste processing are random and incomplete and the individual items are not accurately defined, and it is therefore impossible to make a comparison in terms of absolute costs. Data on the costs of the individual processing operations may prove to be misleading unless the costs of the whole process are known. The proportion of capital and running costs of the nuclear installation which goes to the processing and disposal of radioactive wastes may serve as a rough guideline. In research nuclear installations this is 1- 6 % (mean approximately 2.5 %) of capital costs and operating and maintenance costs will be in the region of 1-6 % (mean 2 %). In nuclear power plants, capital costs for waste processing will be 1.5-6 % of the total. The actual value depends on the installed output of the reactors and will decrease with increasing unit output of the reactors and with the number of reactors installed in one power plant. Running and maintenance costs will make up 5 - 10 % of the total. The first international attempt made at evaluating the overall costs of the processing and disposal of radioactive wastes and at assessingthe cost ratio of the individual operations was the comparison conducted in 12 countries by the International Atomic Energy Agency (56). The results were published in 1968, but the unified method of economic evaluation proved to be extremely efficient and is still used. As for the data themselves, it is obvious that inflation has increased the costs of all operations, and of the entire fuel cycle, but the cost ratio has remained almost unchanged. A survey of costs is given in Table 36. At six nuclear installations some low-level waste is collected by a pipeline into tanks where it is checked for activity level and then discharged without processing. The larger is the amount of wastes the lower are the costs of dischargz, because the discharge pipes are also used for other wastes which makes the system very efficient and considerably reduces the cost of depreciation. On the other hand, long and costly pipelines will increase the costs of depreciation. Of the running costs wages and salaries make up 16 %, material and services 11 %, maintenance 5,5 %, overheads 25 % and depreciation 42 %. The annual maintenance costs will be within the region of 1 - 5 % of the capital costs. In four nuclear installations, low-level waste was collected by pipeline and, following activity checks, the waste was processed using a simple processing technology. This was usually a simple chemical precipitation with a decontamination

248

factor of 10- 100 used for removing the critical radionuclides before discharge. The generated sludge was then solidified using cement or bitumen and stored, dumped or buried in the sea. The given costs include the processing and removal of concentrates. The lowest costs were achieved with highly automated equipments processing large volumes of waste. A low decontamination factor was attained by processing and the radioactive concentrates were discarded into the ground on the site without further processing. Of the total costs wages and salaries make up 14 %, material and services 13 %, maintenance 11.5 %, overheads 31 % and depreciation 31 %. Annual maintenance costs will be within the region of 1-9 % of the capital costs. Intermediate-level liquid wastes in four instances were collected by pipeline or in flasks and, following activity checks, were transported to the processing plant. The costs include solidification and disposal of the concentrates and discharge of treated waste waters. The minimum decontamination factor was lo4. The lowest costs were attained with the largest volume of processed wastes and with equipment that was most efficiently used. The highest costs were recorded in installations with large evaporation equipment, numerous operating personnel and high depreciation. Of the running costs wages accounted for 24.5 %, materials and services 15 %, maintenance 7,5 %, overheads 22 % and depreciation 31 %. Annual costs of maintenance were within the region of 3 - 5 % of the capital costs. The characteristic feature was the irregular and inadequate use of the evaporators, i.e., within the region of 10-13 % of their theoretical capacity. This increased unit costs and caused a widening of the cost range. In order to be able to make a more accurate comparison a conversion was made to a uniform 24 % use of evaporators of the same theoretical capacity, i.e., 8 h per a day for 5 days per week. This conversion narrowed the range of unit costs from US $ 2 9 -207 per m3 to US $ 3 7 - 116 per m3The lowest and highest specific running costs were recorded by the same workplace as before the conversion. The share of the costs of chemical processing was investigated in four instances where the system included the collection of wastes, their chemicalprocessing, solidification and removal of concentrates, the discharge of treated water and monitoring of the environment. It was found that of the total costs of processing and waste disposal (US$ 1.65 per m3) the cost of chemical processing accounted for roughly 62 % of the total, i.e., US $ 1.02 per m3. Of the running costs of chemical processing wages and salaries made up 13 %, materials and services 14 %, maintenance 15 %* overheads 22.6 % and depreciation 36 %. The annual costs of maintenance will be in the region of 1-9 % of the capital costs. Average costs of evaporation were US $ 4 2 per m3. A comparison with the total costs of waste processing (US $ 120 per m3) will show that with regard to the higher specific activity of the wastes, the waste collection and environmxtal monitoring will prove very costly. Of the costs of evaporation, wages and salaries accounted for 16 %, materials, and services 13 %, maintenance 5 %, overhead%

249

Table 36: Analysis of Costs of Processing Radioactive Wastes Total running

Unit operation

Number of operating personnel

Capacity m3/year

1

2

3

4

Liquid wastes 1. Collection by pipeline activity checking, no processing, discharge

6

2. Collection, activity check, simple processing,discharge

4

3. Collection, check, evaporation, discharge

4

4. Chemical processing

4

5. Evaporation

4

900-3.4

6. Collection by pipeline

5

3 x lo4-8.4

7. Collection by drums

6

8. Water discharge

7

%lid

8 x 104-8.2 x lo5

6.3 x 104-5

x lo5 x 103

900-3.4

4.3 x 104-2.7

x lo5

1.5 x 105-4 x lo5 8 x 104-2.5 x lo5 7.4 x 104-2.3

x lo5

lo3

1.3

X

x lo5

1.5 x 10'--1.8

x lo5

950-6.4 x lo3

1.2 x 10'-9.4

x lo4

1.4 x 103-1.7

x lo4

1.5 x 10'-1.4

x lo5

6.3 x 104-5

2 x 104-5

X

X

X

lo5

10'

X

10'--1.3

lo5

WMW

1. Total costs

5

250-4 x 103

2. Collection, activity check, transport

7

300-4 x 103

5 x lo3-6.6

3. Compacting

4

270-3.5 x lo3

7 x lo3--1.6 x 10'

4. Incineration

120-1.5 x lo3

8.6 x 10 -5.5

X

lo4

x 10'

19 % and depreciation 42 %. Annual costs of maintenance were in the region of 1-4 % of the capital costs.

At some installations, intermediate- and high-level liquid waste is transported in containers or drums. Smaller volumes of such wastes (0.27 m3/d; 100 m'lyear) are transported in flasks (with a volume of 0.025-0.4 m3; 25 to 400 1) while tanks are used for the transport of larger volumes. The cost of transporting wastes in flasks

costs

Percentage of total running costs labour

material and services

maintenance

overheads

depreciation

5

6

7

8

9

10

0.1-0.7 av. 0.36

5-28 av. 16

3-21 av. 11

2-19 av. 5.5

15-38 av. 25

21-60 av. 42

0.50-2.65 av. 1.65

7-25 av. 14

3-25 av. 13

3-17 av. 11.5

9-58 av. 31

1945 av. 31

29-207 av. 120

1040 av. 24.5

5-24 av. 15

1-13 av. 7.5

10-31 av. 22

19-44 av. 31

0.2-1.50 av. 1.2

4-1 8 av. 13

7-20 av. 14

4-25 av. 15

9-38 av. 22.5

2847 av. 36

14-112 av. 43

14-18 av. 16

3-20 av. 13

1-9 av. 5

7-35 av. 19

18-50 av. 42

0.2-1.6 av. 0.7

6-25

av. 17

2-16 av. 7.5

2-1 8 av. 6

10-50 av. 22

20-63 av. 46

1-28 av. 15

1663 av. 27

1-19 av. 11

1-22 av. 7.6

15-35 av. 22

19-56 av. 33

0.03-0.1 3

av. 0.06

6-40 av. 16

2-32 av. 10

1-20 av. 6

14-43 av. 25

16-88 av. 51

35-84 av. 60

7-32 av. 20

9-1 8 av. 12

2-13 av. 6

9-34 av. 25

12-60 av. 28

9-58

av. 22

13-56 av. 31

8-37 av. 22

4-1 1 av. 7

17-53 av. 31

3-37 av. 20

2649 av. 40

3-37 av. 16

9-59 av. 34

1-31 av. 11

5-1 3 av. 11

14-66 av. 28

20-72

10-28 av. 17

2-8 av. 5

7-19 av. 10

19-28 av. 24

21-61 av. 36

av. 50

is US $ 110 per m3 while the costs of transporting wastes in drums or tanks will be US $ 15 per m3.The cost of container transport depends on the activity of the wastes. For higher level activity special containers are used with shieldings of different thicknesses, while wastes with a lower activity may be transported in simple containers. The costs of container transport of liquid wastes will be distributed as follows: wages and salaries 14 %, materials and service 3-4 %,

maintenance 5 - 6 %, overheads 30 %, depreciation 45 %. The annual costs of maintenance will be in the region of 1-2 % of the capital costs. As for the total costs of tank transport, 27 % will go on wages and salaries, I 1 % on materials and services, 7 % on maintenance, 22 % on overheads and 33 % on depreciation. The most widespread technique for collecting and transporting liquid wastes is by pipeline. The c x t s are given by the activity and amount of transported wastes. The average cost of the transport of liquid wastes with an activity of up to 370 MBq (lo-’ Ci/m3) was US $0.7 per m3. The lowzst cost (US S 0.2 for m3) was for the largzst volume of transported waste (2301 m3/d; 8.4 x lo5 m3/yzar), while the highest cost was for the smallest volume of transported waste waters (82 m3/day; 3 x lo4 m3/year). The costs of transport by pipeline will increase at least ten-fold with an increase of the activity of the waste waters to 370 GBq/m3 (10 Ci/m3).This increase in costs is given by higher demands on safety, higher costs of materials used for the pipeline and by the necessity to employ more personnel t o operate the pipeline. Of the total costs for the transport of waste water by pipeline, wages and salaries account for 17 %, materials and services 7 %, maintenance 6 %, overheads 22 % and depreciation 46%. Annual costs of maintenance will be 2-4 % of the capital costs. A comparison of the transport of liquid radioactive wastes in tanks and by pipeline has shown the pipeline to be less costly when the volume of liquid wastes does not exceed 3.3 m3/d (1200 m3 per year). The pipeline must, however, be used to capacity. The average costs of the discharge of liquid wastes are US S 0.06/m3 and are considerably influenced by the capacity and complexity of the discharge system. The highest unit cost was for the lowest discharged volume (55 m3/day; 2 x lo4 m3/year) and a pipeline that was not used to full capacity. Of the total costs, depreciation accounted for 51 %, wagas and salaries 16 %, materials and services 10 %, overheads 25 % and maintenance 6 %. The annual cost of maintenance was less than 1 % of the capital costs. The costs of concentrate processing varied considerably, mainly owing to the wide variety of processing techniques that have been developed and applied for processing concentrates whose physico-chemical properties differ widely. No analysis has been made of the costs but it was found that the costs of concentrate processing are within the region of 10-40 % of the total costs of processing the respective liquid wastes. The costs of the storage of concentrates vary widely, depending on the type of wastes and the place and technique of disposal. The c x t s of the burial of solid or solidified liquid wastes are 5 - 10 % of the total costs, the permanent disposal of concentrates and sludges in tanks at source 3 % of the total costs, and transport to and disposal on a central off-side disposal area up to 30 % of the total costs. In countries where solidified liquid wastes have been buried at sea the average costs for dumping 1 ton of material were 20 % of the total costs.

252

Compared with the costs of the disposal of liquid wastes, which varied considerably, those for the disposal of solid wastes varied very little, i.e., within the region of 35 -84 US $/m3, with an average of US $ 6 0 per m3. The least costly was the disposal of solid wastes at source without processing-here the costs were 5 - 10-fold lower than the costs of any other disposal technique. Of the total costs, wages and salaries accounted for 20 %, materials and services 12 %, maintenance 6 % , overheads 25 % and depreciation 28 %. The annual costs of maintenance will be between 1 - 7 % of the capital costs. The average running costs of collection, activity monitoring and transport of liquid wastes were US $ 22/m3, with 170-1500 m3 of wastes per operator per annum. Of the total costs, wages and salaries made up 31 %, materials and services 22 %, maintenance 7 %, overheads 31 % and depreciation 20 %. Annual costs of maintenance will be 1-4 % of the capital costs. Compacting solid wastes into metal tanks using a simple hydraulic press with a reduction factor of 5 : 1 cost US $ 40/m3 on average, of which wages and salaries made up 16 %, materials and services 34 %, maintenance 11 % and depreciation 28 %. The average annual costs of maintenance varied widely within the region of 1-22 % of the value of the capital costs. Solid wastes are usually incinerated in commercial incinerating plants with stack gas control equipment. These incinerators have a high output and hardly ever operate to capacity. The average costs were US $ 50/m3 and the capital equipment costs US $ 2 0 000-80 000. Of the total costs, wages and salaries accounted for I7 %, materials and services 5 %, overheads 24 %, depreciation 36 % and maintenance 10 %. Annual costs of maintenance were 2-6 % of the value of capital costs. The incineration of high-level and toxic materials requires complex handling of the incinerated material and a highly efficient stack gas monitoring and control system. In such plants the costs of incinerating 1 m3 of solid wastes reached us $200 - 220. The costs of processing radioactive wastes were analyzed using the same method in the countries of the Council of Mutual Economic Assistance. The results are given in Table 37. The specific activity of the processed wastes was between 0.37 and 370 MBq/m3 and lo-’ Ci/m3) and the salt content was between 0.2 and 30 kg/m3 (0.2 and 30 g/l). The wide variation in costs for processing radioactive wastes is given by the great. differences in volumes of processed wastes and in their radiochemical and chemical composition. Moreover, the Moscow plant did not include solidification and concentrate removal. A comparative study was made at the Leningrad processing plant made of the costs of the storage of liquid wastes in tanks and the storage of solidified wastes. Cementation was carried out using the technique described in Section 6.1 and bituminization was carried out at elevated temperature, similar to that used at Mol (see Section 6.2). The results confirmed the advantages of the solidification process:

- storage in tanks

220 roubles (Rb)/m3

- cementation and disposal in tanks with solid wastes 39 Rb/m3 94 Rb/m3 - cementation and disposal in separate tank 75 Rb/m3 - bituminization and ground disposal - bituminization and disposal in concrete disposal 83 Rb/m3 pit The results obtained at the Leningrad plant provide a good guideline for assessing the costs of waste processing in socialist countries. The plant has 58 employees and 98.6 m3/day (36 000 m3/year) of waste waters in continuous operation, which corresponds to the capacity of the plant. The waste waters have an intermediate specific activity of 370 MBq/m3 Ci/m3) and have a I3’Cs content with a specific activity of 185 -296 MBq/m3(5 x -8 x Ci/m3), -2 x a 90Srcontent with a specific activity of 1 1 . 1 -74 MBq/m3 (3 x Ci/m3) and a content of alpha emitters with a specific activity of 0.74-0.074 MBq/m3 (2 x - 2 x lo-’ Ci/m3). Waste waters are treated by chemical precipitation and evaporation and the condensate is further processed on ionexchange columns. The liquid waste treatment process generates 720 m3 of sludges and concentrates per annum with a salt content of 400 - 500 kg/m3 (400 - 500 g/l) which is cemented. The solidification process will double the volume of concentrates. The capital costs for constructing the plant were 817 600 US $, of which capital costs for the building amounted to 400 000 US $. The most cotly equipment is the evaporator, costing 224 000 US $, of the other processes, 12 800 US $ goes on waste collecting, chemical processing and ion exchange will cost 48 000 US $ each, cementation 30 400 US $ and environmental control 40 000 US $. Annual running costs reach 662 400 US $ and are broken down as follows: 8640 US $ waste collection chemical processing 51 200 US $ evaporation 397 440 US $ ion exchange 49 920 US $ cementation 110 560 US $ environmental control 44 640 US $ The costs of processing I m3 of waste waters will be 14.9 US $ for processing. 3.10 US $ for cementation and 1.20 US $ for environmental control, i.e., a total of 19.2 US $/m3. In Leningrad the collection and disposal of solid wastes from radioisotope workplace is also organized. At the workplace the waste is put into a plastic bag which is placed in a 0.03 m3 (30 1) container. The waste is then transported by a special truck which carries 16 containers. At the disposal area the plastic bags are put in reinforced concrete tanks and the containers are re-used following decontamination. About 224 m3 of solid wastes are discarded in this way every year.

254

The capital costs of the collection and disposal of solid wastes amounted to a total of 342 400 US $, of which 273 600 US $ went to building costs. Of the total capital costs the collection and transport of wastes was 91 200 US $, removal from containers to tanks 19 200 US $, decontamination of containers and trucks 169 600 US $ and environmental control 62,400 US $. Annual running costs were 130 100 US $. The total costs of removing 1 m3 of solid wastes from radioisotope workplace amounted to 580 US $ of which collection and transport amounted to 225 US $, removal from containers to tanks and disposal 121,6 US $, decontamination of containers and trucks 130 US 96 and environmental control 104 US $. The lowest costs for liquid waste processing were achieved at the Moscow plant (31), which processes large volumes of low-level wastes. The plant processes an annual 150 000 m3 of waste waters with an intermediate specific activity of 370-3.7 MBq/m3 - lod4 Ci/m3), namely 90Sr, I3’Cs, 95Zr, 9JNb, lo6Ru, I 3 l I etc., a salt content of 0.5-0.6 kg/m3 (0.5-0.6 g/l). The plant has an output of 20 m3/h. The processing of waste waters includes chemical precipitation, sludge sedimentation, sludge filtration and two-stage ion exchange and the regeneration solutions are thickened in an evaporator. The concentrates and sludges with a salt content of 400 kg/m3 (400 g/l) are transported to the disposal area in tanks and are solidified with cement at the disposal area. The costs of the transport and processing of the concentrates by solidification have not been stated in the report of the Moscow plant. The capital costs for the construction of the plant were 1.9 x lo6 US $, of which building costs amounted to 536 000 US $; capital equipment costs were 1 344 000 US $, of which the costs of waste collection are 227 200 US $, chemical processing 324 000 US $, ion exchange 384 000 US $, evaporation of regeneration solutions 316 800 US $ and sludge processing 160 000 US $. Annual running costs were 528 000 US $ and the total cost of processing 1 m3 of waste waters was 3.50 US $ (not including transport, solidification and removal of concentrate). Of the total unit costs, 0.51 US $/m3 went on waste collection, 0.67 US $/m3 on chemical precipitation, ion exchange cost 0.86 US $/m3, the processing of 11.3 m3/ day (4125 m3/year) of regeneration solutions cost 1.07 US $/m3 and the thickening of 4.11 m3/day (1500 m3/year) of sludges amounted cost 0.38 US $/m3. So far no detailed analysis of the cost of processing and disposal of radioactive wastes from nuclear power plants has been published. The only known data concern the share of the total capital and running costs remarked at certain nuclear power plants for waste processing. The breakdown of the costs was as follows: - with pressurized water reactors 2 x350 MW(e) 1.5 % Beznau (Switzerland) Mihama I1 (Japan) 500MW(e) 2 % IlOOMW(e) 5 % USA average

255

- with boiling water reactors

Muhlenberg (Switzerland) 300 MW(e) 2 - 3 % 380MW(e) 2 % Tarapur (India) Fukushima I (Japan) 460MW(e) 3 % 11OOMW(e) 5 % USA average - with heavy water reactors Rajastan (India) 220MW(e) 2 % The situation is similar with nuclear power plants of the WWER type built in Czechoslovakia. It is generally accepted that equipment for processing radioactive wastes requires 2-5 % of the total capital costs of a nuclear power plant. Regarding running costs, the US estimate is that with pressurized-water and boilingwater reactors radioactive waste processing will account for 2 %. In other countries the percentage is lower. The total costs for processing and disposal of radioactive wastes from nuclear power plants are within the range of 2-5 % of the total costs expended in a nuclear power plant for the production of electric power. Detailed data are known only from the Byeloyarsk boiling-water reactor nuclear power plant. Evaporators with a decontamination factor of lo3 - lo5 are used for waste water treatment. The concentrate from the evaporators is stored in tanks, further processed using charcoal filters and then re-used or discharged. The on-site waste water treatment plant processes annually 150 000 m3 of waste waters and the costs of processing 1 m3 of waste water amount to 6.13 US $. The total costs may be broken down as follows: wages and salaries 15 % depreciation 69 % power 3% 5% steam chemicals 1% overheads 7% In Czechoslovakia solid wastes from radioisotope workplaces are transported to one central off-site disposal area in Northern Bohemia. The disposal area is part of a disused limestone quarry reconstructed for this purpose at a cost of 1.0 x lo6 US $. The disposal area has a volume of 9200 m3, which corresponds to unit capital costs of 140 US $ per m3. The amount of solid wastes discarded annually in the disposal area is 70-90 m3. The wastes are placed in 0.05 m3 (50 1) metal containers. The annual costs of waste disposal ranged between 30 000 and 60 000 US $, with a mean of 47 000 US $. Running costs were broken down as follows: transport of wastes 32 % material, wages and salaries 32 % power 5% maintenance 9% overheads 22 %

256

The disposal area is used to 75 % of capacity. At this rate the running costs per m3 of discarded wastes are 474 US $. The costs of the collection and disposal of 1 m3 amount to 615 US $, including depreciation. The costs may be reduced by more efficient operation of the disposal area and more efficient use of the capacity of the containers. Liquid wastes have been processed for many years at the Institute for Nuclear Research. The deactivation unit was built for processing intermediate-level liquid wastes and is equipped for the complex processing of such wastes, including equipment for chemical precipitation, evaporation and ion-exchange columns, and equipment for the cementation of sludges and concentrates. The annual capacity of the unit is 900 m3 of waste waters with a specific activity of 370-3.7 MBq/m3 Ci/m3) with a salt content of up to 3 kg/m3 (3 g/l). Radionuclides present in the liquid wastes include "Sr, I3'Cs, 144Ce, lo6Ru and "Zr. The unit produces an annual amount of 1.5 m3 of sludges and 4.5 m3 of concentrates, which after cementation have a total volume of 17.9 m3. The solidified concentrates are transported to an off-site disposal area. The number of personnel employed at the unit is 13. The total capital costs of the deactivation unit were roughly 2.4 x lo6 US $, of which building costs amounted to 1.0 x lo6 and equipment 1.4 x lo6 US $. Of the total capital costs, the processing of solid wastes accounts for 6 %; for liquid wastes, collection of waste waters accounts for 5 %, activity checks 8 %, storage 9 %, chemical processing 35 %, evaporation 20 %, ion exchange 6 %, sludge filtration 5 %, sludge and concentrate cementation 4 % and discharge 2 %. The total annual costs (including depreciation, overheads for the whole Institute. research and development, transport and disposal) per m3 of processed waste waters were 517.48 US $6. The costs are broken down as follows: collection at source transport to deactivation unit activity check storage at deactivation unit chemical processing evaporation sludge filtration sludge cementation discharge of purified water transport to disposal area final disposal environmental control

2.83 US $/m3 18.68 US $/m3 99.61 US $/m3 29.17 US $/m3 141.54 US $/m3 93.07 US $/m3 28.02 US $/m3 8.56 US $/m3 27.69 US $/m3 9.42 US $/m3 0.73 US $/m3 27.46 US $/m3 30.70 US $/m3

The annual production of solid wastes is 30 m3, of which 4 m3 cannot be

257

compacted. Compactable wastes are compacted into drums which reduces their volume to 0.03 m3/day (10.7 m3/year). Non-compactable wastes are placed in drums without processing and the total volume of 0.04 m3/day (14.4 m3/year), is transported to the central disposal area. The total costs of processing and disposal of solid wastes amount to 21 12 US $ per m3 and the costs of the individual operations involved are as follows: 79.1 1 US $/m3 150.77 US S/m3 297.07 US $/m3 355.70 US $/m3 210.35 US $/m3 14.78 US $/m3 685.46 US $/m3 319.26 US $/m3

collection at source transport to processing unit activity check compacting packaging transport to disposal facility final disposal environmental monitoring

The relatively high unit costs of processing liquid and especially solid wastes result from the extremely low volumes of these wastes and from the very inefficient use of the individual equipment. Table 37: Costs of Waste Processing in the CMEA Countries Country

Amount (m3/4

:apital costs (103

us $1

Running costs ~

(103

us $1

Processing costs (US $/m3:

Liquid wastes

Poland GDR (low-level) GDR (intermediate-level) Czechoslovakia USSR (Leningrad) USSR (Moscow)

4.54 0.88 0.24 3.19 98.63 410.96

2,629 2,629 2.926 4.365 802 1,878

822 326 402 435 608 507

512 1,021 4,514 373 33 3.5

0.29 0.27 0.05

888 730

330 140 79

3,077 1,405 4,021

Solid wastes

Poland GDR Czechoslovakia

--___

258

References

1. 2. 3. 4. 5.

Standardization of Radioactive Waste Categories, IAEA, Technical Report Series No. 101, 1971 Marsily G., et al.: Science 197 (1977), No. 4303 Martin C. J.: Health Phys. 21 (1972) 235 Orlova E. I.: Gigiena i sanituriya 12 (1973), 65 (in Russian) Management of Low and Intermediate Level Radioactive Waste, IAEA Symp. Proc. Ser., STI/PUB/264, 1970 6. Techniques for the Solidification of High Level Wastes, IAEA, Technical Report Series No. 176, 1977 7. Status Report on Radioactive Waste, Swedish AKA, 1974 8. Aurand K., Kernergie und Umwelt, Erich Schmidt Verlag, 1976 9. Radioactive Waste Management in Western Europe, OECD, 1971 10. Waste Management Technique and Programmes in Czechoslovakia, Poland and the Soviet Union, IAEA, 1971 11. Management of Wastes from the Mining and Milling of Uranium and Thorium Ores, IAEA, Safety Series No. 44, 1976 12. Robinsky E. I.: Canadian Mining and Metallurgical Bull., December 1975 13. Chemical Treatment of Radioactive Wastes, IAEA, Technical Report Series No. 89, 1968 14. Operation and Control of Ion Exchange Process for Treatment of Radioactive Wastes, IAEA, Technical Report Series No. 78, 1967 15. Design and Operation of Evaporators for Radioactive Wastes, IAEA, Technical Report Series No. 87, 1968 16. Management of Radioactive Wastes at Nuclear Power Plants, IAEA, Safety Series No. 28, 1968 17. The Volume Reduction of Low Activity Solid Wastes, IAEA, Technical Report Series No. 106, 1970 18. Warner B. F., et al.: Proc. Symp. Management of Radioactive Wastes from Fuel Reprocessing, OECD, 1973 19. Lennemann W. L.: ibid. 20. Salom L., et al.: Trans. Am. Nucl. Sec. (1975), 21 (April) 21. Guber W., et al.: Report KFK 1789 (1973) 22. Wilhelm J. G.: Report KK 1065 (1969) 23. Fuchs N. A.: The Mechanics of Aerosols, Macmillan N. Y.,1961 24. Management of Low and Intermediate Level Radioactive Wastes, IAEA Symp. Proc., 297, Vienna 1970 25. Management of Radioactive Wastes from the Nuclear Fuel Cycle, IAEA, Symp., Prac., Vienna, 1976

259

26. Bituminization of Low and Medium Level Radioactive Wastes, OECD Seminar, Antwerp, May 1976 27. Bituminization of Radioactive Wastes, IAEA, Technical Report Series, No. 116, 1970 28. Bituminization of Low and Medium Level Radioactive Wastes, OECD Seminar, Antwerp. May 1976 29. Technique for the Solidification of High Level Wastes, IAEA, Technical Report Series No. 176, 1977 30. Grover J. R.,et al.: Proc. Symp. Richland 1966, Conf. 660208, 1966 31. Morris J. B., Chidley B. E.: Proc. Symp. Management of Radioactive Wastes from the Nuclear Fuel Cycle, IAEA, Vienna 1976 32. Bonniaud R., et al.: Proc. Symp. Management of Radioactive Wastes from Fuel Reprocessing, OECD, Paris 1973 33. Eschrich H.: ibid. 34. Gee1 J. V. et al.: Proc. Symp. Management of Radioactive Wastes from the Nuclear Fuel Cycle, IAEA, Vienna 1976 35. Krause H., Randl R.:Proc. Symp. Management of Radioactive Wastes from Fuel Reprocessing, OECD, Paris 1973 36. Lynch R. W., et a].: Proc. Symp. Management of Radioactive Wastes from the Nuclear Fuel Cycle, IAEA, Vienna 1976 37. Kubo S., Rose D. J.: Science 182 (1973), 1205 38. Kouiim V., Vojtech 0.:Atomic Energy Review 12 (1974), 215 39. McKay H. A. C.: Chem. Ind. (1972), 275 40. Wilfert G. L.: Report BNWL 389 (1967) 41. Glueckauf E. (Ed.): Atomic Energy Waste, Interscience, New York, Butterworths, London, 1961 42. Lennemann W. L., et a].: ZAEA Bulletin 17 (4) (1975) 2 43. Beard S. J., Moore R. L.: Progr. Nucl. Energy, Ser. 111. Process Chem. 4 (1970) 645 44. S t a j J., KyrS M.,Marhol M.: Separation Methods in Radiochemistry, Academia, Prague, Czechoslovakia, 1975 (in Czech) 45. Meade P. J.: WMO No. 97, Techn. Note 33, T. P. 41 (1960) 46. Pasquill F.: Atmospheric Diffusion, Van Nostraiid, 1962 47. Hilameier W. L., Gifford R.: Report USAEC - O R 0 545 (1962) 48. Gifford F. A,, Pack D. H.: Nucl. Sufery 3 (4) (1962) 76 49. Chamberlain A. C.: Int. J. Air Pollution 3 (1960) 50. Comper W.: Report KFK 1615 (1972) 51. Beattie 1. R.,Bryant P. M.: Report ASHB/S/R 135 (1970) 52. Bryant P. M.: Report AHSP/RP/R 42 (1964) 53. Fritelli L. Report CNEN RT/PROT (72) 7 (1972) 54. Doury A,: Report CEA-R-4280 (1976) 55. Regulations for the Safe Transport of Radioactive Materials, AEA, 1973 Revised Edition, Safety Series No. 6, 1973 56. Economics in Managing Radioactive Wastes, IAEA, Technical Report Series No. 83, 1968

2 60

Index

A Active solution, 14 Activity, beta, 10 Activity, gamma, 10 Aerosol filters, 126 Alpha emitters, 10, 23,49, 105, 183, 237, 242 Aluminothermal process, 169, 172 Americium, 193

B Beta emitter, 185, 237, 242 Biological treatment, 83 Biotoxicity, 98 Bitumen, 140 -, blocks, 144 -, categories and properties, 140 -,types, 141, 142 Bituminization, 140, 145-146 -, Czechoslovakia, 147, 152 -, evaluation, 153 -, installations, 146-152 -, USSR,147, 148, 151 C

Calcination, 167, 170 Cement blocks, leaching, 133 Cementation, 130 -, new variants, 138, 139 -,process and technology, 131, 133-136, 138 -, technology USSR, I37 Central Station for Radiation Protection, 239 Cerium, 190 Cesium, 191

Coal-fired power plants, 19 Cold bituminization, 150 Compacting, 92 Concrete trench, 21 1 Containers, 219 Contamination, 33, 36, 37 -, atmospheric, 57 -, monitoring, 174, 248 Corrosion products, 12, 40 Critical exposure paths, 234 Curium, 193 D DAREX technique, 52 Death hazard, 27 Decay constant, 1 13 Decontamination factor, 56, 75, 79, 80, 112,119, 246, 249 Delay line, 112, 114 Denitration, 166 Dilution factor, 234 Dilution of wastes, 20-21 Disposal area, diagram, 214 Disposal in deep geological formations, 102-103, 220 Disposal trenches, 21 3 Dressing plants wastes, 28, 33 Dynamic adsorption coefficient, 112, 11 3

E Electrodialysis, 84 Emitters, 60 -, open, 62 -, sealed, 61

261

Emulsified bitumen, 147, 153 EUROWAT process, 104 Evaporation, 80

F Fast breeder reactor, 14 Fast reactors, 46 Fission products, 36, 48, 53,98, 111, 181 -, application, 184-186 Fixation, technologies, 163-1 68 Fixation process, testing methods, 174-1 79 Fixation products, 155 -, chemical composition, 176 -, chemical resistance, 176, 177 -, form, 157-162 -, macrostructural properties, 175, 176 -, microstructural analysis, 174 -, physical properties, 177-179 -, requirements, 156 Fluidized-bed calciner, 168 Fluoride reprocessing wastes, 106 Fossil fuels wastes, 19, 29 Fragmentation, 91 Freezing process, 74, 193 G Gamma emitter, 185, 242 Gaseous fission products, 111-120 Gaseous radionuclides, 55, 57, 110 Gaseous wastes, 10, 11, 30, 31, 38, 54, 55 -, processing, 107-129 -, storage, 224 H Half-life, 48, 49, 113 HARVEST process, 170 HEPA filters, 108, 126, 128 High-level wastes, 10,48, 51, 163 -, disposal, 100-103, 219 -, solidification, 155-1 79 -, solutions, 53 High-temperature process, 168-169 Hot bituminization, 147

I Incineration, 93, 253 Intermediate-level wastes, 55, 104

International Atomic Energy Agency, 11, 241, 248 International Commission for Radiological Protection, 19, 226,236 Iodine radionuclides, 120-125 Iodine 129 world production, 59 Ion exchange, 78, 173, 188, 190, 194 Ionizing radiation, 184, 234 Irradiation rate, 10 Isohyets, 201

K Krypton, 23, 2 6 2 8 , 119, 193

-, world production, 57

L Land burial, 208-219 Liquid extraction, 47 Liquid waste processing, 65-68, 72-90 -, biological treatment, 83 -, comparison, 83 -, chemical treatment, 75-78 -, electrodialysis, 84 -, evaporation, 80-82 -, filtration, 73-75 -, ion exchange, 78-80 -, solar evaporation, 85 Liquid wastes, 10, 11, 13, 30, 31, 38, 50 -, from mining, 32 -, transport, 240, 249 Long-lived radionuclides, 107, 191,220, 236 LOTES process, 161, 171 Low-level wastes, 10,26, 63-64, 78, 133, 241 Low-temperature distillation. 117

M Medium-level wastes, 140 Mine waters, 31, 32, 65 -, treating, 65-68 Mining of radioactive raw materials, 31-36

N Neutron scavenger, 52 Niobium, 192 Non-fission products, 53 Nuclear fuel cycle, 17-18

Nuclear power plants, 39, 40, 41, 42, 43-45, 236 -, accident, 28, 36 -, Czechoslovakia, 40, 88 -, liquid wastes, 86-90 -, safety, 17 -, solid wastes, 97 -, radioactive wastes, 26, 3 6 - 4 7 USSR, 40,43,46,90

-.

0 On-site incorporation, I03

P Palladium, 192 PAMELA process, 171 Permissible concentrations, 237 Permselective membranes, 11 1, 120 Plutonium, 47, 49, 133, 155, 180, 183, 187 Population doze, 27, 28 Portland cement, composition, 131 Pressure filters, 74 Processing of radioactive raw materials, 31-36 Production of radioisotopes, wastes, 58-64 Promethium, 190 PUREX system, 47, 51, 104, 189, 193 PUREX technique, 51

R Radioactive aerosols, 125-1 29 -, separation methods, 127 Radioactive concentrates, fixation, 130-1 54 Radioactive emmissions, 228 Radioactive materials, transport, 241-245 Radioactive substances, migration, 196 Radioactive tracers, 206 Radioactive wastes, 10, 25 -, as raw-materials, 180-194 -, atmosphere, 227-235 -, bituminization, 140-154 -, categories, 10, 11 -, cementation, 130-1 39 -, characteristics, 3 L 6 4 -, chem-ical ore processing, 33, 34 -, CMEA countries, 258 -, collection, 239 -, Czechoslovakia, 240, 256, 258

-, -, -, -, -, -, -, -, -,

disposal, 16, 23, 195-237 EEC countries, 36,46, 64 geology, 203 hydrology, hydrogeology, 202 long-term burial, 140 meteorological factors, 201 origination, 11-15 sources, 30, 37 storage, 16, 23 -, storage in seas and oceans, 225-227 -, surfaces waters, 235-237 -, transport, 238, 252 -, USSR, 137, 147, 151, 239,253,254,258 Radiolysis wastes, 52 Radiolytic decomposition, 52 Radionuclides, 10, 2 6 2 8 , 36,40, 63, 110, 185 -, categories, I 1 -, maximum activity, 243 -, of rare gases, 12 -, separation methods, 188-189 -, use and wastes, 60 Radiotoxic materials, 48, 183 Radiotoxicity, 10, 177 Radium, 33 Radon, 33 Reactors, 18 -, BWR type, 20-21, 26-28,42-43 -, CANDU heavy-water type, 4 3 4 5 -, gas-cooled, 12, 44-46 -, PWR type, 20-21,2628, 37-39 -, water cooled, 13 -, WWER type, 12,40 Redox technique, 52 Rhodium, 192 Ruthenium, 192

S Salt deposits, 220 Sand filters, 73 SANDIA process, 173 Semi-sunken hall, 212 Separation of elements, 189-194 Silver reactors, 123 Simple trench, 209 Soil, migration of radionuclides, 197 Solar evaporation, 85 Solid wastes, 10, 11, 30, 31, 38 Solid wastes processing, 65, 69-72, 91-97 -, compacting, 92-93

263

Solid, fragmentation, 91

-, incineration, 93-96 -, other methods, 96 Solidification processes, 98-102 -, continuous French process, 170 -, survey, 164 Specific activity, 10, 48, 49 Spent fuel reprocessing, 26-28 -, dry technologies, 54 -, liquid extraction, 47-53 -,wastes, 47-58, 98-106 Spray calciner, 167 Storage areas, waterproofing, 218 Storage pits, 213 Stored wastes, properties, 215-217 Strontium, 189, 190 Sulphides, chemical oxidation, 33 -, oxidation by bacteria, 33 Surface deposit, 209

T

Uranium ore dressing, 33, 35

v Vacuum cementation, 139 Vacuum filters, 74 Volatile fission products, 51 Volatile fluorides, 54 Volume activity, 10, 11, 38 W

Waste burial forms, 99, 100 Waste cementation equipment, 137 Waste conversion into silicates, 158 Waste disposal, 195 -, economy, 246-258 -, salt deposits, 220-223 -, sites, 199 -, sites, basic units, 208 -, sites, evaluation, 203-206 Waste fixation into ceramic systems, 161 Waste fixation into glass systems, 159, 160,164, 170, 172 Waste handling, 246 Waste processing, economy, 247 Waste waters, 31, 34 Waste rock, 31, 69 Wastes from research centres, 58-64

Tailings disposal, 69-71 THOREX technique, 5 2 Thorium, 31 Technetium, 192 Time delay, 112 Tracer activation method, 207 Transuranium elements, 49, 105, 184, 186, 187 Tritium, 23, 26, 44 -, world production, 59

X

U

Xenon. 193

Uranium, 31,49

Z

-, concentrate, 33 Uranium mines, wastes, 28

264

Zirconium, 192