IAEA-TECDOC-851
Radioactive waste management practices issuesand in developing countries
Proceedings seminara of held Beijing,in China, 10-14 October 1994
INTERNATIONAL ATOMIC ENERGY AGENCY The originating Section of this publication in the IAEA was: Waste Management Section International Atomic Energy Agency Wagramerstrasse5 0 10 x P.OBo . A-1400 Vienna, Austria
RADIOACTIVE WASTE MANAGEMENT PRACTICES AND ISSUES IN DEVELOPING COUNTRIES IAEA, VIENNA, 1995 IAEA-TECDOC-851 ISSN 1011-4289 © IAEA, 1995 Printe IAEe th AustriAn y i d b a December 1995 FOREWORD Radioactive wast generates ei d fro productioe mth nucleaf no r energe frous d e yan m th of radioactive materials in industrial applications, research and medicine. The importance of safe managemen radioactivf to e protectio e wastth r efo humaf no nenvironmene healtth d han t has long been recognized and considerable experience has been gained in this field. The managemen radioactivf o t e wast internationas eha l applications with regar dischargeo dt f so radioactive effluents into the environment and in particular to final disposal of waste. The need for the prevention of environmental contamination and the isolation of some radionuclides, especially long-lived radionuclides longer fo , r period f timso e than national boundaries have remained paststable th ,n ei requir e waste management methods basen do internationally agreed criteri d standardsan a e IAETh .s Aactivi i this h e sha ared an a introduce a RADWASd S (RADioactive WAste Safety Standards) programme aimint a g establishin promotingd an g a coheren n i , comprehensivd an t e manner e basith , c safety philosophy for radioactive waste management and the steps necessary to ensure its implementation in all Member States. While this programme is developing and various related Safety Series publication becomine ar s g available s importani t i , comparo t t e eth existing national waste management regulations, organization, technologies and methods with internationally accepted requirements and practices. In response to the growing interest in this area, the IAEA, hi co-operation with the Government of the People's Republic of China, held a Seminar on Radioactive Waste Management Practice Issued san Developinn si g Countrie Beijint sa Octobe4 1 g o frot 0 rm1 1994. It provided technical experts, mostly from developing countries of different regions, involve managemenn di f radioactivo t e wast opportunitn ea f exchanginyo g information no their regulating and operating experience and discussing the spécifie problems in every country as well as common problems which developing countries are facing in this field. Participation of developed countries which are main suppliers of waste processing equipment allowed the mlearo t n abou reae th t l technology transfer need developinf so g countriese Th . Seminar also benefited both developing countries and the IAEA through the identification of important component nationaa f so l waste management infrastructur introducee b o et r do improved. Seminae Th attendes rwa mory db e thaspecialist0 n10 s fro countrie3 m3 included san d scientifi0 4 c presentations t provideI . exchange foruda th r mfo f informatioeo wasta n no e management policy, waste management strategies, a legal framework, the responsibilities of regulatory authorities and central operating organizations, waste processing, storage and disposal technique safetd performancd san yan e assessments.
Emphasis was placed on the management of low and intermediate level waste arising from applications of radioisotopes hi medicine, research and industry and from nuclear power generation.
e SeminaTh r organizers offere technicaa d l tou f wasto r e management facilitied an s laboratorie Chine th i sah Institut Atomif eo c Energy (CIAE) includin gventilatioa n facility, levew alo l liquid waste treatment facility, waste polymerization equipment Seminae , etcTh . r was conclude a pane y b dl discussio e technicalth n o n , economic, environmentad an l institutional considerations in the establishment of a national waste management programme.
It is hoped that these Proceedings will constitute an important source of information to a wide community of scientists, engineers, regulators and decision makers dealing with the management of low and intermediate level waste. EDITORIAL NOTE
In preparing Ms publication for press, staff of the IAEA have made up the pages from the original manuscripts submittedas authors.the viewsby The expressed necessarilynot do reflect those governmentsofthe nominating ofthe Member States nominating ofthe or organizations. Throughout the text names of Member States are retained as they were when the text was compiled. Theof use particular designations countriesof territoriesor does imply judgementnot any by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies productsor (whether indicatednot or registered)as does implyintentionnot any infringeto proprietary rights, should construednor be it an as endorsement or recommendation on the pan of the IAEA. The authors responsibleare havingfor obtained necessarythe permission IAEAthe to for reproduce, translate or use material from sources already protected by copyrights. PLEASE BE AWARE THAT MISSINE TH AL F LO G PAGE THIN SI S DOCUMENT WERE ORIGINALLY BLANK CONTENTS
SUMMARY OF THE SEMINAR ...... 9
OPENING REMARKS
B. Semenov ...... 13 Li Dingfan ...... 14 Xi Zhenhua ...... 6 1 .
NATIONAL WASTE MANAGEMENT PROGRAMMES
Radioactive waste management challenge developinn si g countries ...... 1 2 . ScareD. Radioactive waste managemen Albanin i t a ...... 9 2 . K. Dollani Waste management practices and issues in developing countries - the case of Croatia . . 35 Subasic,D. S.K. Dragicevic Radioactive waste management in Cuba. Results and perspectives ...... 43 LA. Java Sed, LM. Pumarejo, H.D Nieves, N.G. Leyva Status of radioactive waste treatment and disposal in China ...... 51 Luo Shanggeng, Li Xuequn The national waste management system in Egypt ...... 59 S. Marei, KA. El-Adham Development of a national waste management infrastructure in Ghana ...... 67 KO. Darko, C. Schandorf Radioactive waste management in Kenya: Presently and the near future ...... 75 Otwoma,D. S.N. Kyoto, OnyangoSA. The Guatemalan programme of radioactive waste management ...... 81 S.R.R. Jiménez, P.O. Ordonez National programme, legal framework and experience with the management of radioactive waste in the Slovak Republic ...... 89 KonecnyL. The existing situation wit radioactive hth e waste managemen Syrin i t a ...... 9 9 . S. Takriti Statu radioactivf so e waste managemen Zambin i t a ...... 3 10 . K. Mwale Swedish waste management programme ...... 5 10 . P.-E. Ahlström Management of radioactive waste in Israel ...... 117 Brenner,S. Ne'eman,E. Shabtai,B. Garty,E. ButenkoV.
STRATEG POLICD YAN Y
National polic experiencd yan e wit managemene hth f radioactivo t e wastes from non-fuel cycle activities in the Czech Republic ...... 125 J. Holub, JanûM. Radioactive waste managemen t implementatios policit d yan Indonesin ni a ...... 3 13 . S.Yatim The fundamentals of the Russian Federation national policy in the non-nuclear fuel cycle radioactive waste management ...... 9 13 . E. Latypov, V.A. Rikunov The Hungarian radioactive waste management project and its regulatory aspects ... 143 I. Czoch Strategy for waste management in Argentina ...... 149 J. Pahissa Campa Natural decay and half-life: Two bases for the radioactive waste management policy . 155 J-C. Femique
WASTE MANAGEMENT PRACTICES
Waste managemen Nucleae th t a t r Technology Centre (CDTN) ...... 3 16 . S.T.W.Miaw Managemen f non-fueo t l cycle radioactive wast Romanin ei a ...... 1 17 . TurcanuC. Radioactive waste management at the Dalat Nuclear Research Institute ...... 175 Nguen Thi Nang Experiences in the management of radioactive wastes in Bangladesh ...... 181 MM. Rohm an
WASTE TREATMENT OPTIONS AND PRACTICES f chemicao e Us l precipitation processe liquir sfo d radioactive waste treatment ....9 .18 V. Zabrodsky, N.E. Prvkshin, A.S. Glushko Environmental impact assessment of operational practices for processing low level liquid waste Thailann si d ...... 5 19 . Yamkate,P. Sinakhom,F. SupaokitP. Treatment proces facilitied san urbar fo s n radioactive wastes ...... 1 20 . Y. Zhang, Z. Chen, J. Dca A filter stud radioactivr yfo e liquid waste treatment ...... 7 20 . Yucca,Ye Tianbao,u W Xin Guo Volume reduction of synthetic radioactive waste by the thermopress ...... 213 P. Van der Heyden, P. Debieve
WASTE CONDITIONING
Radioactive waste-mortar mixture form characterization due to its physico-chemical and mechanical properties obtaine acceleraten a n di non-accelerated dan d leaching processes ...... 223 A. Peric, I. Plecas, R. Pavlovic, S. Pavlovic Using bitumen solidificatio ILLr fo n WLLL& W ...... 9 22 . Zhang Weizheng, TingjunLi Radioactive waste forms: a review and comparison ...... 237 R.C. Ewing Development of a nuclear waste drum of concrete ...... 243 Wen Y ing Hui t chemicawe A e studth ln yo oxidatio solidificatiod nan f radioactivo n e spenn io t exchange resins ...... 9 24 . Tianbao Wu, Guichun Y un, Jiaquan W u, Yucai Yie Radiobiological wastes treatment: ashing treatment and ash immobilization with cement ...... 255 Feng,S. Wang,B. Gong,L. Wang,L. Sha L. Development of thermoplastic solidification process for urban solid radioactive wastes ...... 1 26 . Jing Weiguan
WASTE DISPOSA SAFETD LAN Y ASSESSMENT
Rock characterizatio sitn ni e selection ...... 9 26 . A.E. Osmanlioglu Competitive adsorptiof no soiSn o r l sediments, pure clay phase feldspad san r 90 minerals ...... 273 S.H. Sakuma, AhmadS. Environmental impact study for low and intermediate level radioactive waste disposal ...... 3 28 . Wang Zhiming A stud protectivn yo intermediatd an e w coverlo r sfo e level radioactive waste disposal in near-surface facilities - China's experience ...... 289 F. Zhiwen, C. Gu Screening of sorption materials for radioiodine and technetium ...... 295 J. Zeng, D. Xia, X. Su, X. Fan A systems approach for quality assurance in waste conditioning, storage and disposal ...... 1 30 . E.R. Merz
List of Participants ...... 313 SUMMAR SEMINAE TH F YO R
Wide variation developmene th nucleaf o n s i e us rd energtan evidene yar developinn ti g countries. A few have or are pursuing partial nuclear fuel cycle activities. More than ten developing countries have nuclear power plants. Becaus e increasinth f o e g demanr fo d electrical energy, more developing countries would lik havo et e nuclear power most f Bu . o t them are constrained by lack of finances and technical expertise. Some have nuclear research reactor susee productior whicb dfo y hma f radioisotopesno . Mos f developino t g countries are using nuclear energy for applications in fields of medicine, agriculture, industry, research and education. These application growine sar developinn gi g countrie fasa t sta pace. From l thesal e uses, radioactive wast produces ei d that mus managee b t d safel efficientlyd yan . Increasingl recenn yi t years, countries have turneIAEe th technica r Ao dt fo l assistancd ean waste management services to address serious problems they are facing.
In developing countries, the priority attached to radioactive waste management is not a smajorita t e caseshigshouli n I th s h. a f o ,yd be therlaca f s awarenesi eko e th f o s importanc f safeo e radioactive waste management. Consequently t tendi , receivo st w lo e priority, inadequate financing, and insufficient staffing and training support. This collectively lead littlo st e appreciatio safete th f yno implications. Waste disposal problems tendee b o dt ignored, or wastes are stored, sometimes improperly, in some remote places.
Many particular problems toda rootee yar thesn di e conditions acute th tied eo dan t , financial difficulties facing most developing countries e allocatioTh . f fundno r wastfo s e management thus is often disproportionately low in comparison with the real needs. To further complicate this situation, the most fundamental requirements for managing radioactive waste, namely policy, adequate legislation understandind an , f safetgo y issues lackine ar , g in many developing countries.
In countries dealing with management of waste generating from nuclear applications, one of the major problems concerns spent radiation sources. In many instances, information ise exten th lackinmagnitudd o an t t s ga f thio e s problem e sourceTh . s frequentle ar y negligently stored, in some cases with non-radioactive materials, and serious accidents have happened.
In countries with nuclear reactors or radioisotope production facilities, the problems of waste managemen more tar e complicated. Proper waste minimization, segregation, collection, treatment and conditioning methods must be practiced to a required international safety level.
In response to such conditions, the IAEA has put in place a number of mechanisms in suppor f efforto t s that countrie e makinar s develoo gt e necessarpth y infrastructurd an e expertis safe th er managemenefo radioactivf o t e waste. These mechanisms include technical co-operation projects, co-ordinated research programmes, training courses and some specialized activities aimed towards direct assistance to developing countries. Concerning the general area of nuclear applications, a number of technical manuals have been prepared. They address topics including minimizatio segregatiod nan wastesf no ; handling, conditioning and disposal of spent sealed sources and other solid wastes; interim storage of waste; treatmen conditionind an t radioactivf go e effluents, organi biologicad can l waste spend san t ion exchange resins; and design of a centralized waste processing and storage facility. Videos varioun o s technical aspect f wasto s e managemen e alsar to supplied often through expert mission t trainina r o s g courses. Overall e IAEA'th , s variet f service yo t enablina m ai s g developing countrie becomo st e more self-sufficien reliand managemene th an t n i t theif o t r radioactive waste.
A traditional way of dissemination of up-to-date information is organization and sponsorshi f internationapo l meeting subjecte th n f o smutua o s l interes f IAEo t A Member States. Almost every year the IAEA organizes so-called major meetings which can be either an international conference symposiua , seminara r mo meetinge t mosth A . f o t s sponsored by the IAEA or other organizations, the selection of papers is made in favour of scientists from developed countrie reporo swh interestinn o t importand gan t result f theio s r studies which, however t mucno he ar relevan, neede th f developino so t t g countries. Recognizing the deficiencie f suco s h meeting thin si s regard IAEe th , A decide organizo dt hold ean da meeting wher majoritea f speakero y s woul froe db m developing countries having similar problem radioactivn si e waste managemen lookind tan simplr gfo cost-effectivd ean t safebu e technical solution managinr sfo g similar waste streams attentioe Du give.e alss b o nowa t n to the legal framework and the responsibilities of organizations involved in various waste management activities.
e paperTh s presente e Seminath t a d r revealed thae nationath t l waste management system mosn si developinf to g countrie beine sar g establishe upgrader do e basie th th f sn o d o recent requirements and recommendations of the IAEA. It was recognized that all required components of the radioactive waste management system should be present in a national programme because only an integrated approach can assure the overall safety of radioactive wast futuree e th managemen n i . d Soman ew specifino t c factor dictatede ar s rule,a s y a b , the national situatio availabilite th n- resourcesf yo levee th ,industriaf o l l developmente th , nucleae th siz f eo r programm socio-economie th d ean politicad can l conditions.
The Seminar presented an excellent opportunity to get acquainted with the national waste management programme participatine th f o s g countrie revealed an s d that political, technical and ethical challenges are not being fully met in some developing countries. There was a general consensus that for the radioactive waste of concern in developing countries the reliable and efficient treatment, conditioning and storage technologies are available and affordabl moso t e f countrieso t . However understoos wa t i , d that proper resulte sb wilt no l achieved unless these technologies are installed, applied and controlled properly. In selecting experts for its technical assistance programme the Agency was requested to pay due attention to expertise available in some developing countries.
Regarding waste disposal present a , t near surface disposa consideres i l mose th te b o dt feasibl d intermediatan e w metholo r efo d level waste containing relatively short-lived radionuclides. Many countries are making great efforts to establish near surface repositories. Agence Th requestes ywa consideo dt developmene th r standara f o t d design packaga r efo near surface repository for low level radioactive waste. Safety assessment methodology of a near surface repository aren a ,whicn ai h intensive researc developmend han t commenced on a national level, was another area of support requested from the IAEA. Only a few countries consider deep geologic disposa realistia s a l c disposal metho ther dfo m becausf eo high cosd technicaan t l complexity n internationaA . l solutio f thino s problem muse b t considere futurn di countriey eb t hav stechnologe no tha eth o d t economid yan c resourceo st plan and construct a national deep geological repository. Consideration of regional solutions to the disposal of high level wastes was suggested as a logical cause of action to resolve this problem.
10 The technical visit to the China Institute of Atomic Energy was a successful supplement to the theoretical content of the Seminar. The visit has demonstrated the progress achieved developmene th n i f variouo t s waste management techniques within existing financiad an l technological constraints.
e SeminaTh r presente n excellena d t opportunit o brint y g together scientistd an s engineers from countries with the different level of the waste management technologies and let them exchange their views and experiences. The Seminar reviewed the status of the radioactive waste management situatio mann i y developing countries. This information will facilitate the decision making regarding the nature and extent of technical assistance to countries that participated in the Seminar and help the Agency to formulate its waste management programme.
11 OPENING REMARKS
. SemenoB v Deputy Director General, e Heath f do Department of Nuclear Energy and Safety, International Atomic Energy Agency, Vienna, Austria
title thif eTh o s Seminar shows that mor mord ean e attentio beins ni g safgivee th e o nt management of radioactive waste. While it can be noted that 75% of the IAEA 122 Member t havStateno eo d snuclea r power, mos f the o te radionuclide mus r researchfo s , medical, industria othed an l r institutional applications. Ther growina s ei g awarenes Memben si r States that in promoting the use of nuclear energy to enhance the living standards of the countries, commitmena t must als made ob safelo et y manag radioactive eth e waste tha s generatei t d from various nuclear applications. Thus, whethe countra r generatins yi g radioactive waste from small use radionuclidef so medicinn si involves i r operatioe o th n di nucleaf no r power reactor othed an s r nuclear fuel cycle activities ,establishmene therth neea r s ei dfo d an t implementatio nationaa f no l waste management programme. Suc hprogramma e must include elemente th integraten f a o f l so al d system, including law regulationsd an s , operatind gan regulating organizations, system processing/storagr sfo disposad ean effectiv n wastf a lo d ean e public acceptance out-reach programmo N . complets ei wilr f thes et o succeeno i l e e on f di components is missing. In reviewing the technical programme of the Seminar, I am pleased componente noto t th ef o tha l al t s require nationaa r dfo l waste management programme ear covered, som morn ei e detail than others opportunite .Th thereforys i e presente learo dt n from the experiences of others that will be presented here, and thus return to your countries with new ideas and practices to meet the challenge of managing radioactive waste. In concert with this thought Seminae thema , th r efo r tha supposes i t supplemeno dt meana titls it ts ea f so adequately describing wha goae thitth f lo s meetin , shoulgis ..."Sharinge db of Practicesd an Technologies for the Safe Management of Radioactive Waste". With this theme the Seminar objectives could be as follows:
1) To identify from the practices presented, those which can be useful in the implementation of your national waste management programme, and
collectivelyTo 2) dialogue theseon practices ensureto informationthe of that all on them is clearly understood.
responsibilite th s i t I scientiste engineerth e f th y o d san develoo st implemend pan t safe solutions to the nuclear waste issue. This challenge not only includes developing and implementing sound technical approache managemene th o st nucleaf to r waste translato t t ,bu e technologe th y solutions into understandabl convincind ean g statement authoritiee th r sfo d san the public. This challenge must be met and existing roadblocks will be eliminated so that full implementatio nationaf no l waste management programme achievede b n sca .
13 Li Dingfan Vice Chairma Chine th f nao Atomic Energy Authority, Beijing, China
1. STATUS AND POLICY OF NUCLEAR POWER DEVELOPMENT
Since 1955, when the China's nuclear industry was initiated, a complete system of nuclear industry, nuclear science and technology has been formed. Since early 1980's, emphasi e s developmenplaceth wa s n o d f nucleao t r power e self-designeTh . selfd an d- constructe Qinshae MW n0 Nuclead30 r Power Plan connectes t (NPPwa ) gristarteo d t dan d to generate electricit Decembeyn i intt opu s commerciawa r t 1991I . l operatio Aprin i l 1994 and now it is steadily operating under high rated power. A press conference called "Qinshan e Environmentth d an P s co-sponsorewa NP " y Chinb d a Atomic Energy Authoritd an y National Environmental Protection Agency on June 18, 1994. Environmental radiation monitoring indicated that the NPP radioactive waste management system is functioning properly and the annual release of radioactive effluents is far below the state limits, comparable with the internationally-acknowledged releases of NPPs of the same type in the world.
Unite s constructee Guangdonwa sMW Th - 0 co 90 witn P di o g h tw NP Day y Ba a operation between Chin foreigd aan n countries connectes Unie wa Th .1 t gristarteo d t dan d to generate electricit commercias Augusit n yo d , 199an 31 3t l operation bega Februarn no y connectes Unie 1,wa Th 19942 t . gristarteo d t dan generato dt e electricit Februarn yo , y7 s commerciait d an 1994 l operatio , 19946 bees n .y larges e ha bega nth ThiMa P n no tsNP project undertaken through co-operation of China with foreign countries since China's opening to the outside world. The success of NPPs in Qinshan and Daya Bay ends the history of the China's mainland without nuclear power and opens a new era of peaceful use of nuclear power and technology in China. The second phase of the Qinshan NPP with two 600 MWe Unit s beeha s n e on-sitapprove th e state d th e an y , constructiob d proceedingw no s i n . Connectio griproductioo nd t dan electricitf no f 2000o d expecte.s yen i e th y db
e worlTh d wide growt f populatioho n wil inevitable b l y followe increasinn a y b d g deman f energdo y resources. Nuclear powe s stilri l regarde maia s da n optio deao nt l with future energy demands periow ne f fas o dA .t developmen n China'i t s econome b n yca predicted in 1990s. In this context, energy industry, if it cannot meet the demands of economic development, will face a new problem. In order to meet the objective needs of social developmen reducd an tenvironmenta e eth l pollution arising from fossil-fuelled power plants believe w , e thadevelopmene th t f nucleao t rsolvo t powe e y eth Chinwa n i r a s ai energy shortage. "Simultaneous development of thermal, hydroelectric and nuclear power" is a reasonable policy for the south-east costal areas. Using the comprehensive summing-up of technical experienc constructioe th n ei operatiod n an Qinshae th f nDayo d NPPs y nan aBa , nucleae plannes 100iMW d i t r0 an powe e construco dt rMW plants0 60 t .
e scal speed Th e an developin n di g nuclear power must mee neede th t f economio s c development. Following the Qinshan and Daya Bay NPPs, the second phase of two 600 MWe Qinsha s beini P g nNP constructed . Besides that Chin s planninai construco gt secone th t d Guangdong NPP and the Liaoning NPP with two 1000 MWe units in each plant. Furthermore preparatora , y wor beins ki g mad constructior efo easterNPPe f no th n si n costal areas with well-developed economy suc Shandongs ha , Jiangs Fujiand uan believe W . e that our technology and experience can fairly satisfy the needs of developing countries. We will
14 adhere to the principle of "Equality and mutual benefit and mutual development" to promote the co-operation with countries and regions interested in the peaceful uses of nuclear energy and nuclear technology.
2. RADIOACTIVE WASTE MANAGEMEN CHINN TI A
History proves that nuclear power is a kind of clean energy resources, and the developmen f nucleao t r powe importann a s i r t solutio probleme th environmene o nt th f so t pollution. However, owing to various social and historical reasons, general public shows much concer r treatmenfo n d disposaan t f radioactivo l e wastes. Therefore, reliabl d safan ee management of radioactive wastes has become a vital subject in nuclear power development.
While developing nuclear industr nuclead yan r power, China pays great attentioo nt management of radioactive wastes. The specific policies and principles have been issued in this respect, including the following:
1. Nuclear facilities are required to reduce, as low as possible, their production of radioactive wastes; 2. Radioactive waste treatment facilities and a principal part of the project shall be all designed, constructed and put into operation simultaneously; 3. Radioactive waste shall be managed according to the classification criteria; 4. The national standards for releases of radioactive wastes to the environment must be strictly implemented; 5. The policy of regional disposal shall be implemented for low- and intermediate-level radioactive solid wastes; . 6 Regional repositorie e planne sar constructee b o dt North-Wese th n di t China, South China, East China and South-West China.After approval by the state, a preliminary design work has begun for the site, according to the design, a disposal capacity is estimated to be 60,000 m3. Site selection and feasibility study for the South China repository have been finished. Pre-selection of the sites for repositories in the East China and South-West China is now proceeding. As a pre-requisite for the whole project, the work for preparing related criteria and standardization of waste containers alss i o being carried out; 7. Deep geological disposal will be adopted for high level radioactive waste and fundamental research work is now carried out; and 8. Urban radioactive waste management. Temporary storage rooms for radioactive wastes produced in applications of scientific research, industry, agriculture and medicine have been built up in nearly 20 provinces and cities in China.
Thanks to effective policies and measures in radioactive waste management, a serious even f radioactivo t e environmental contaminatio s nevenha r occurre nuclean i d r facilities, which creates a sound basis for the sustained development of nuclear industry and nuclear power. It can be concluded that China has made great efforts in treatment and disposal of radioactive wastes with many achievement researcn si developmentd han .
15 Zhenhue Xi a Administrator, National Environmental Protection Agenc f Chinayo , Beijing, China
As is well known, China is fairly rich in coal resources. The coal supply is over 25% primare inth y energy resources. This situation predicteds a , , wilt changno l 200y eb d 0an even longer. Inappropriate techniques, scale applicationd an s s adopte used coan an di d l industry and coal-fueled industry may cause serious problems in environmental protection. The Chinese government pays great attention to environmental protection and active measures appliee ar reduco dt releas e green-house eth th f eo e cleagasese th f naddition o .I coa e us l o nt burning technolog improvo yt availabilite eth burnine th f yo g rate, great effort beine ar s g made to exploit new energy resources to meet the needs of the developing economy and reducing pollution. Among other things developmene th , nucleaf o t re poweth f o juss e ri on t option thin si s respect.
Furthermore, rich coal resource distributee sar d mostl norte th hn yi par Chinf o t a while the south part of the country and its coastal areas are short of energy resources and the transfer of coal from the north to the south is seriously restricted. For this reason, the coastal areas in short supply of energy resources and will inevitably develop nuclear power. The authorities concerne n i environmentad l managemen e supportinar t e appropriatth g e development of nuclear power in suitable places.
While developing nuclear power, management of radioactive wastes is considered as an important problem. There is no difficulty in technology to treat and dispose of low and intermediate level waste wels sa shor s a l t lived wastes. Nevertheless consequence th , f eo improper managemen radioactivf to e waste must neve underestimatede rb . Treatmen finad tan l disposa f speno l t fuel produce dissen f generao t l public towards nuclear powe n somi r e countries.
The Chinese nuclear industry concentrates a large amount of well-qualified technical personnel in respect of scientific research and engineering. A considerable amount of outstanding administrative personnel with practicable experience have also been trainede Th . National Environmental Protection AgencChine th d a yan Nationa l Nuclear Corporatioe nar co-operatin controf o y supervisind wa an l a n gi radioactivf go e waste management activities. s fulli t I y recognized, froe lonmth g term co-operation, that outstanding technicad an l administrative personnel in nuclear industry provide a fundamental guarantee for developing nuclear power, safe operatio f nucleano r reactor managemend an s f radioactivo t e wastes. Through their efforts and the increasingly strengthening international co-operation, and on the presene basith f so t record f safso e operatio nucleaf no r facilitiesassuree b n ca d t i ,tha t no t onl nucleae yth r power properl e plantb n sca y managed t alsbu o, problems associated with treatmen disposad an t radioactivf o l e solved e wastb n eca .
orden I effectivelo rt y protec environmene tth exploitation ti applicatiod nan nucleaf no r energy, the operators of nuclear facilities and their responsible departments are required to stricüy control radiation emissions into the environment. Besides that there are supervision and monitoring environmentae unitth n si l protection bodies. Therofficn a f nucleas eo i e r environmental management subordinated to National Environmental Protection Agency, and government provinciae th t sa l level also have theimonitorinn row g stations. These bodies take the responsibilities for supervision and monitoring of the environment all over the country.
16 Relevant tasks include preparation of laws, codes and standards; evaluation, monitoring and examination of the environmental impact caused by nuclear facilities. Radioactive waste managemen alss i t o included.
There is a need to supervise and control treatment of radioactive wastes produced not only by nuclear facilities but also from application of nuclear techniques. "The Prevention and Control Law of Nuclear Pollution" issued by the state has been drawn up, and it will be submitted to the National People's Congress for approval. The policy for disposal of low- and intermediate-level radioactive wastes was formulated by National Environmental Protection Agency together with China National Nuclear Corporation in 1992 stipulating requirements for regional disposa f suco l h wastes.
Radioactive waste spend san t radiation sources arising fro nucleae mth r applicatione sar accepte urbae stored th d an y ndb radioactive waste management station eacn si h provinces. So far thousands of tons of radioactive wastes and thousands of spent radiation sources have been accepted and stored.
Though China has gained preliminary experience in management of radioactive wastes, it is only an initial stage and many problems have yet to be studied and solved. Due to further developmen f nucleao t r power, increased applicatio f nucleano r technique expectind an s g decommissioning of some early-built nuclear facilities, radioactive waste amounts will be continuously increased. Investigations, studies and practices are required to solve the problems such as how to reduce the generation of radioactive wastes, how to treat and carry out safe interim storag f radioactiveo e wastescarro t t saf w you e ho ,disposa radioactivf o l e effectivelo t w wastesho d y an , contro generatee th l d radioactive waste.
17 NATIONAL WASTE MANAGEMENT PROGRAMMES RADIOACTIVE WASTE MANAGEMENT CHALLENGES IN DEVELOPING COUNTRIES
D.E. SAIRE Divisio Nucleaf no r Fuel Cycl Wastd ean e Management, International Atomic Energy Agency, Vienna
Abstract
This paper discusse challengee sth s facing Member State thes sa y pla implemend nan t a national waste management programme e challengeTh . dividee ar s d into three areas, namely, political, technica ethicald an l . These challenges have been identifie variouy db s Agency activities and contacts with senior government officials, scientists and managers in many countries. Agency programme assiso st t Member States overcom challengee eth e sar described but the paper clearly states that it is the responsibility of the Member States to plan and implement activities which will overcome the challenges and permit the establishment of a successful national waste management programme.
1. INTRODUCTION
It is indeed a great pleasure to be here in Beijing at this Seminar and I am pleased to present this opening paper. The title of this Seminar, "Radioactive Waste Management Practice Issued san Developinn si g Countries" selectes wa , challengo dt e Member Stateo st importane focuth n o s t issu f safeleo y managing radioactive wastes thageneratee ar t d from nuclear power or other applications of nuclear energy. My paper will concentrate on this challenge staro T . t this Semina believI f of rmuse ew t stat fundamentae eth l principle that link desirsa countra nucleae f eo us o yt r energy wit neee implementinr hth dfo gnationaa l waste management programme. This principle states that when a country makes the decision to use nuclear energy, it has also made the decision to safely manage the radioactive wastes that result from the use of the atom. This principle or fundamental truth can further be elaborated on into what is classified as the "Safety Fundamentals for Radioactive Waste Management". These Safety Fundamentals are contained hi the RAD W ASS1 document entitled "The Principles of Radioactive Waste Management" which is planned to be submitted to the Board of Governors of the IAEA at its December 1994 meeting. Now the primary purpose of my paper will be to consider how well countries are in fact following the fundamental principl states ea daccomplisn aboveca e W . h thi firsy sb t outlining whae ar t the basic requirements that integrate into an effective national programme for the safe management of radioactive wastes and how well countries are implementing these requirements.
2. FUNDAMENTAL PRINCIPLES OF RADIOACTIVE WASTE MANAGEMENT
Before establishing the basic requirements of a national waste management programme, we should first consider what is meant by the safe management of radioactive waste, since objective thith wasty s si an ef e o managemen t programme.
1 RADWASS is the acronym for the RADioactive WAste Safety Standards programme
21 e AgencTh s beeyha n co-ordinatin threga e year effort amon Membes git r Stateo st reach international consensus on the Safety Fundamentals/principles of radioactive waste management. These principles wile formallb l y approved sood distributean n a s a d RADWASS document principlee Th . listee sar d below:
. 1 Radioactive waste shalsecuro t managee y lb acceptabl n ea wa a i dh e leve protectiof lo n for human health. 2. Radioactive waste shall be managed in a way that provides an acceptable level of protectio environmente th f no . . 3 Radioactive waste shal manageassure o t b l s a ey sucthan di wa tha possible effectn o s human environmene healtth d han t beyond national borders wil takee lb n into account. . 4 Radioactive waste shal managee b l thay suci dh twa predicteha d impact healte th n hso of future generation greatee s b wil t lno r than relevant level impacf so t tha acceptee tar d today. 5. Radioactive waste shall be managed in a way that will not impose undue burdens on future generations. . 6 Radioactive waste shal managee b l d withi appropriatn na e national legal framework, including clear allocatio f responsibilitieno provisiod san r independennfo t regulatory functions. 7. Generation of radioactive waste shall be kept to the minimum practicable. 8. Interdependencies among all steps hi radioactive waste generation and management shall be appropriately taken into account. . 9 Safet f facilitieo y r radioactivfo s e waste management shal appropriatele b l y assured during their lifetime.
National waste management programmes should be developed and implemented hi such a manne f thesr o tha l eal t nine safety fundamental mete ar sthi n I . s way objective ,th f eo radioactive waste management, which is to deal with wastes that human health and the environmen future th i eh protectee withoud ar t an w td no placin undun ga e burde futurn no e generations can be met.
. REQUIREMENT3 WASTA F SO E MANAGEMENT SYSTEM
Since the fundamentals of radioactive waste management and the objectives have been given woulI , lik w tur o eattentiot dy no n m requirements e th o nt shoulr challengeo ,y sa dI s that mus e faceb t r overcomo d o develot e n effectiva p e national waste management programme. These challenges can be classified or grouped into three categories:
Political challenges Technical challenges Ethical challenges
3.1. Political challenges For a country to develop and implement an effective waste management programme, it must have the firm support of the senior government officials of the country. This is the fundamental requirement, for without support at the highest level of government, resources committee b wilt no l essentiad dan l legal framework developede s b wilt no l experiencr Ou . e with developing Member States has shown that it is not always easy to create an awareness e saf th e neee o r th fmanagemen dfo f radioactivo t e waste management amon e higgth h echelons of government officials. The problem can be traced to the fact that government
22 official "optico n e se ss benefit" from waste management compare othedo t r use resourcesf so . r exampleFo usualls i t i , y eas securo yt roadew schoolw fundinne ne , a r ever o s fo gna research reactor because these things have high public visibility. Waste management isn't a product that can be used but a result of using certain products. Until the fundamental truth mentioned earlier in this paper is clearly understood by government officials, it will continue a battl e o securb t e o e necessart th e y resources (fund d manpoweran se safth e r fo ) management of radioactive wastes hi some countries.
The other political challenge concerns the establishment of the necessary laws and regulations needed to set the framework for a waste management programme. These laws and regulations should provide the legal basis for operational and regulatory waste management activities. Normally issued as an Act of Government (Law on Radiation Protection or Atomic Energy), they should includ generae eth l principle wastf so e managemen implementee b o t d and responsible authorities for performing the regulatory and operational functions. Governments must recognize that any atomic energy programme cannot be exercised without the basic legal structure being hi place. Unfortunately, hi many countries, other higher priority legal requirements have ofte nationat npu l atomic energ radiatior yo n protection laws "on the back burner".
3.2. Technical challenges
case th es i wit s hA many technologie processesr so , ther mane ear y approachee th o st managemen radioactivf to efirse wasteth tt rul thiBu f .eo s discipline regardin technicae gth l aspects involved, is to use the "integrated systems approach" to manage waste streams. This "systems approach wasto "t e managemen illustrates i t reasoe System a Fig Th n dr i . nfo 1 . s Approac s quiti h e eviden s radioactiva t e waste e usuallar s y subjec a sequenc o t t f o e operations or unit steps as shown hi Fig. 1 process may lead to lower volumes and lower disposal costs. Still other considerations must be featured into the system, for example, the conditioned end-product must als compatible ob e wit disposae hth l environment with regard wasto t e leaching, corrosion, biodégradation, etc. Furthermore actuae th , l availabilita f yo disposal area must also be a part of the consideration. Where large areas of disposal space is available lese , b thersy incentivema reduco et e volumes. Thesexamplew juse fe ear a t s Systeme th y owh sf Approac wasto ht e managemen necessarys i t importans i t I .tak o t t e no t e system th e ste actiof on o p , n i nwhic y rendema h r other steps more expensivr o e technologically more difficult e necessitTh . f plannin yo e soun th r dgfo managemenf o t radioactive waste stream illustratee b n sca wha y conceptdb E calles i tIC e dth . This concept identificationstande th r sfo , characterizatio evaluatiod nan wastl al f neo stream determino st e how they should be segregated or integrated into the systems approach for their management. Figure 2 attempts to illustrate the ICE concept by showing 12 different waste streams and a resulho s f wa identificationo t , characterizatio d evaluatioan n e e streamb th n y ma s integrate r furthedfo r waste managemen processee havy b o ma et r o t d throug waste hth e management syste separatea ms a d individual stream. Onl utilizinconcepy E yb IC e s gi tth it possible to effectively plan for the management of various waste streams that may develop fro mparticulaa rmacro-basis a facilit n o r yo entire th , e country's programme. Failuro et employ the ICE concept will result hi complicating the waste management programme, as waste streams become mixed, which shoul separatede db similad an , r waste processee sar d hi entirely different ways.
Another technical challeng facine ar e classifies gi ew "overkile th s da l syndrome"t Le . me explain wha meatI thiy nb s experiencr termou n I . e with developing Member Statese w , have observed tha mann i t y instances equipmen r technologo t y selecte o perfort d ma
23 GENERATION OF WASTE
Exempted Wastes
WASTE CHARACTERIZATION INDUSTRIAL WASTE AND CLASSIFICATION DISPOSAL
WASTE SEGREGATION
WASTE TREATMENT
WASTE CONDITIONING
STORAGE
WASTE DISPOSAL
FIG. 1. Waste management - a systems approach.
FIG. 2. Identify, Characterize and Evaluate (ICE) Concept for radioactive waste management.
particular technica r proceso l s activit highls yi y sophisticate expensived dan . Thera s i e tendency among scientists and engineers in developing Member States to want to purchase a "Mercede thab jo "bicyclea t e sth Benzo d practicaa "o s "t coul A . ddo l example th f eo "overkill syndrome", I recall where a certain nuclear research facility purchased a very
24 elaborate waste incineration facility fro mwestera n country, whe totae nth l volum solif eo d waste expected to be generated was less than 100 m3/year. In this case, a simple waste compactor would have been more than sufficien cosd an t less thae quartee non th f o r incinerator. Althoug Agence hth always yha s stressed straightforwar cosw lo t solutiond dan s wasto t e management problems "overkile th , l syndrome practices "i severan di l countries. We must be able to select technical equipment and processes by their merits to do the job and not simply because developed countr sucs ha hproces a yA equipmenr so operationn i t .
lase tTh technical challeng ewanI discuso t nee e developinr th dfo s s i g Member States to perform some amoun researcf o t developmentd han . Whil businese Agence th eth i h s yi s of transferring informatio technologyd nan , Member States mus tproblem n conduco D tR& s which are of a local nature as it is not always possible to transfer technology that is completely adaptable gooA . d evaluatioe examplth neee s i r on-sitth dfo f D ef o n o eR& waste forms. Since waste forms are composed of local materials, it is not possible to directly use waste form performance data from other countries. This is particularly true in evaluating concrete or cement matrices as the chemical form and stability of the local ingredients of concrete are different. In our dealings with developing Member States, we notice an attitude which is "show us how it's done elsewhere". The Agency of course responds by providing a stream of experts to offer technical assistance and advice. However, this assistance can only go so far and it is up to the recipient country to translate the assistance into working solution r wastthein fo s ow re management problems. This entail degrea s f researcho e , development and testing. In our experience, this had been the weak link in the whole technology transfer process.
3.3. Ethical challenges
Since the fundamental principles of radioactive waste management discussed earlier provided the ethical considerations (protection of future generations, protection beyond national borders and burdens on future generations) surrounding the establishment of a national waste programme, this part of the paper will only deal with near-term aspects of the ethical challenges. In today's environment no national waste management programme strategy developee b n ca d withou firma t pla buildinr nfo g public confidenc acceptancd ean e th f eo national plan.
growte nucleae Th th f ho r power optio impedes ni mann di y countries toda publiy yb c concerns ove safete environmentarth d yan l consequence producinf so g electricit meany yb s of nuclear reactors. The main components of this public concern are the potential for nuclear reactor accidents, the day-to-day operational safety of nuclear reactors, the association in the public's mind between nuclear power and nuclear weapons, and the question of what to do with radioactive waste. Similar concern oftee sar n expresse f radioactivo e d us wit e hth e materials for non-power applications. Scientists working on the technical aspects of radioactive waste disposal have developed an international consensus that the waste can be permanently managed in a manner that protects the environment and public health. However, this view is not necessarily shared by the general public, thus the need for a public information programme.
In the public's mind, the perceived risk from radioactive waste is very high. This public's perceptio rise radioactivf th k o f no e waste differs markedly fro scientist'e mth s view because of a lack of understanding of the objective risks to the health and environment and the general mistrust that has developed over the decades since the introduction of nuclear
25 energy problee Th . gaininmf o g public trus understandind tan nationaf go l waste management programmes can be traced to the following risk concerns:
Risks involving complex technolog t welno l e understooyar ordinary db y people. Ther reluctanca s ei accepo et t risks thainvoluntare tar y (imposed) opposes a , thoso dt e about which each individual can make a free choice. The public is reluctant to accept risks from projects or technologies that are under centralized rather than local control wherd an , e local peoplt havno eo einpud t inte oth decision-making process. perceives i t I d tha failura t wasta f eo e management system could resul disastroun i t s consequences. Recognizing the above, it is our ethical duty to structure national waste management programme effectivelo st y interface wit publie hth removo ct e their natural concerns about the safety of waste management and disposal and replace myth with truth.
. IMPLEMENTATIO4 NATIONAF NO L WASTE MANAGEMENT PROGRAMMES
With the defining of the challenges or needs of a national waste management programme tims i t asseso i et ,effectiv w sho e Member States have bee meetinn ni g these challenges mighu yo s t A expect. , obtaining this typ informatiof eo easn a yt taskno t s ni bu , the Agency, through WAMAP bees 2ha n abl colleco et t some data shows A . Fign ni , 3 . WAMAP missions have visited 40 countries over the period 1987-1994. Using information
WAMAP MISSIONS AS OF MID 1994
FIG. 3. WAMAP world map
2 WAMAP (WAste Management Advisory Programme)
26 from WAMAP missions, data on the status of national programmes has been analyzed and is shown in Fig. 4. Unfortunately, this data shows that Member States visited by WAMAP missions do not have a high implementation rate for their national waste management programmes. While admittedly most of the countries visited by WAMAP represent Member States that needed advisory service, the degree of implementation shown on this figure shows thachallengee th t s mentioned t beinearlieno ge ar rfull y met. Whil degree eth f eo implementation varies, dependin regione th s interestin i n t go i , noto gt e thahighese th t t rating receive Asie Pacifid hers th di a an n ei c region s alsi t oI .worth noto yt e thae th t degree in which R&D has been implemented received the lowest rating.
REGION INFRASTRUCTURE OPERATIONS R&D TRAINED STAFF Dll (.3) (.3) (.2) (.2)
AFRICA. 35 15 5 25 21
ASIA & PACIFIC 50 30 35 50 41
LATIN AMERICA 35 25 20 25 27
MID. EAS EUROPT& E 45 30 25 25 33
AVERAGE 41 25 21 31 31
Dl! = 100 INDICATES FULL IMPLEMENTATION OF WASTE MANAGEMENT REQUIREMENTS / SAFET Y IMPLICATIONS
( ) = WEIGHT FACTOR
FIG. 4. Degree of Implementation Index (Dll)
5. WHAT CAN WE DO?
Evaluating the data shown in Fig. 4, which was developed after WAMAP had been in operation for about 4 years, the Agency has asked the question, "What can we do?" to help Member States meet the challenges that have been identified, thereby implementing effectively their national waste management programmes. Figur show5 e Agency'e th s s respons thio et s questio t listi s sna specific activities, policy and/or programmes that have been implemented or are in the development phase. As shown in this Figure, the Agency has taken actio r provideno d assistanc overcomo t e e every challeng ehavI e mentione thin di s paper. However, Member States must also develop ways to meet this challenge of implementing a national waste management programme. This is the message I leave with problee youTh . mwels i l defined Agence th , y will assis developinn i t g solutionse th t bu , Member States must implemen solutionse th t . Thi schallengee the th future s nth i f d o e an , nuclear energy dependwelw ho l thin so s challeng faceds ei .
27 RADIOACTIVE WASTE MANAGEMEN ALBANITN I A
. DOLLANK I Institut f Nucleaeo r Physics, Tirana, Albania
Abstract The policy and strategy of radioactive waste management hi Albania are described hi the Ministers Council's , 1971Decre83 . . eAccordinNo thio gt s Decre liquie eth d waste ear all contaminated liquids with concentrations 10-100 times higher than maximal permissible concentrations for ordinary water. The management of liquid waste is done through then- collectio i specianh l tanks withou treatmeny an t subsequend an t t discharg sewero et e Th . principal radioisotopes in liquid waste are 1-131 and Tc-99m. The solid waste are all materials, which contain of or are contaminated with radioisotopes up to levels greater than exempted quantities. The management of solid waste is done through its safe storage hi the premises, where radioactive decay occurs, especiall r shorfo y t lived radionuclides. Last years, many spent radiation sources were gathered hi the Institute of Nuclear Physics (INP) for conditioning and interim storage. For conditioning 200 litres standard drums with steel bars and concrete filling having a hole in the centre are used. Spent radiation sources were emplaced hi the hole until the activity of 20 GBq has been reached. Interim storage of conditioned sources is carried out hi the engineering facility near the INP with trenches of capacit cubiy5 c meters each. Last yea nationaa r l inventor f sealeyo d radiation sources begin to compile. A national programme for radioactive waste management hi the future has been developed, taking into accoun e futurth t e extensiof o f productioo n e us d an n radioisotope d radiopharmaceuticalan s e participatioth d an s f Albanio ne IAE th i Ah a Interregional Model Project on Radioactive Waste Management.
1. THE POLICY AND STRATEGY
Radioactive waste management hi Albania is aimed to protect man and his environment from undue exposur o ionizint e g radiation e legaTh . l framewor f radioactivo k e waste managemen includes principatwa e th i dh l radiation protection regulations, which were issued Decree abouth s yeara Ministerf 0 eo 2 to sag s Counci, 19783 1. [1]lNo . Accordine th o gt above-mentioned Decree, the responsibility for radiation protection rests with the Ministry of Health through the Radiation Protection Commission. A scheme of the radiation protection organizatio presentes ni thin I i Figsd h . scheme1 . s showi t i , n that radiation protection activities importan entire th r e fo tcountr developee yar d mainl Institute th i yh f Hygieneo e (IH) and in the Institute of Nuclear Physics (INP). The IH is responsible for control and inspection of nuclear facilities and for other supervising activities, while the INP is hi charge of performing technical activities such as personal dosimetric control, import and transportation of radioactive materials, the spent radiation sources management, training of user radiation so n protection, etc.
The radiation protection activities at nuclear facilities are conducted and supervised by appointed Radiation Protection Officers. Concerning radioactive waste above sth e Decree kindo tw sortf them so t sou :
. 1 Liquid waste, which include liquidl sal s wit radioactive hth e contaminatio- 10 f no time0 10 s higher than maximal permissible concentration f radionuclideso n i s ordinary water, dependin half-liven go f radionuclidesso .
29 MINISTRY OF HEALTH
RADIATION PROTECTION INSTITUT NUCLEAF EO R COMMISSION PHYSICS I LICENSING Personal control
CONTROL INSTITUTE OF Impord an t AND HYGIENE transport INSPECTION — SRS Management
Trainin— 1 g
FIG. 1. Radiation protection organization
2. Solid waste, which includes all solid materials with radioactivity contents or contamination greater than exempted quantities.
resulAe sth f modeso t t nuclear nucleae activitie lacth e f th k o d r fuesan l cycle, only low level waste and spent radiation sources are generated hi the country. The strategy for radioactive waste management is the combination of decentralized and centralized management [2]. As a general rule, radioactive waste management is the responsibility of users, especiall shorr yfo t lived daile wasteth yn I practic. liquie eth d waste managemens i t comprised of their collection hi special tanks without any treatment, and subsequent discharge to sewer. Before discharg obligators i t ei measuro yt e specific activit liquie th f dyo waste. Usually nuclear facilities hav tanko etw operatio n si n whic usee har d subsequently after their filling. Thaprovidey wa t s deca f shoryo t lived liquid wast i tankeh s befor dischargee eth . The principal radioisotopes hi liquid waste are 1-131 and Tc-99m, which constitute more than l wast95al f %eo radionuclide countrye th n si . Annual activit liquie th f dyo wast abous ei t 200-400 GBq. The management of solid waste is also the responsibility of users, especially for short lived wastes. After its collection, the solid waste is stored for a period of 10 half- live f containeo s d radioisotope thereafted san treates i r s ordinarda y waste existine Th . g legislation defines the daily limits of liquid waste discharges and exempted quantities of radioactive materials. Table I presents the above-mentioned levels for the four groups isotopes regarding then" radiotoxicity. For long lived radioactive waste the legislation requires the centralized management withou detailey an t d specification [1].
2. CONDITIONING AND STORAGE OF SPENT SEALED SOURCES.
Management of spent radiation sources (SRS) is the problem of great importance in the country, both with regard to radiation protection and nuclear safety, because during many years the gradual gathering of them in different premises have occurred. The use of sealed
30 TABLE I. LIMITS OF DISCHARGES AND EXEMPTED QUANTITIES
Nuclide Toxicity Group Liquid Activity Discharges Exempted Quantitief so (Bq/day) Activity (Bq) A (Very high) 4«104 2»103
5 B (High) 4M0 2»104
6 (ModerateC ) 4»10 2M05
7 (LowD ) 4«10 2M06 sources began at sixties for geological studies, in industrial radiography, in the army for radiometric equipment calibration, in cobalt therapy, etc. At that time any kind of local regulation safn so e handlin f radioactivgo e sources didn't exist.
During that perio occurrence dth f smaleo l incident unsafd an s e use f radioactivso e t excludedsourceno e yearn ar sTe s. late e situatioth r s changel sourcenha al d f dan o s ionizing radiation were put under control, including SRS. During the last years conditioning an centralizee dth interis a organizeP s mdIN wa storage facilitwholS e th th t SR dr a f eyefo o country firse Th t. conditionin tooS kSR placf go 1992n ei , followe othey db r conditioning activities later. For the conditioning, 200 L standard drums with steel bars and concrete filling (with defined proportions sand/cement/ /gravel/water), which hav centre holea th n eei were used [3]. The SRS with or without radiation shielding were successively placed in the hole untibees activite ha nth l q reached abouf yo GB 0 2 t . Thereafte cemene th r t mortas rwa poured over the sources. After emplacement of the lid, the drum was inspected for integrity. Finally the radiation symbol was placed on the drum. The process of SRS conditioning is presente Fig.2n di . Afte procese th r conditioninf so interie gth m storag provides ewa e th n di engineering facility with trenches of 5 cubic meters capacity, which is in the vicinity of INP [4]. Its capacity which can be increased up to 20 cubic meters makes this facility suitable for meetin countre gth y needs during many meantime yearsth n I .necessars i t ei exploro yt e the possibility of disposal for conditioned sources. Concerning the future of two sources from the cobalt therapy facilities (with initial activity 220 TBq and 110 TBq), the agreement between users and suppliers has been concluded for return of spent sources to the suppliers at the moment of their replacement.
FIG.2. Conditioning of spent radiation sources.
31 INVENTORIEE 3.TH FUTURD SAN E DEVELOPMENTS
bettee th r r knowledgFo quantitief eo activitied san sealef so d sourcese ,us whicn i e har in the country ( in medicine, industry, research, etc.) during the last year the INP in co- operation with the IH, began the compilation of the inventory for all sources. The form of this inventor intendes presentes yi wa t i identifo dy thit y Tabln dB i s wa l radiatio . yal e II n sources, whic f importancho e from radiation protectio nsame pointh t e a f vie timo td wan e to identify unknown radiation sources in the country. This inventory is also considered as the necessary measure for better management of SRS. An important issue is an inventory posseaccuracyP documentatioIN e e sth Th . manr nfo y radiation sources, whic orderee har d through regular procedures. But at the same time other ways of entering radiation sources into the country exist, for example, some foreign companies investing to Albania have brought technological lines which contain radiation sources t withoubu , notificationy an t . Another kind of registry which it is intended to compile, is that of SRS. From this registry it is hoped to be aware of the quantities and activities of SRS in the country, and therefore e need th r furthefo s r conditionin e coming th rate n I g. years foreseei t i s eventuan a n l extensio f manno y activities with radiation sources, especiall f productioe o yth e us d nan radioisotopes and radiopharmaceutical. In connection with the mentioned eventual extension a national programm s beeha e n developed aime t improvemena d d upgradinan t f o g
TABLE II. AN INVENTORY FORM FOR SEALED SOURCES
Radioactive Source Institution
No Device Isotope Initial Current and city activity activity 1 GU-3 Cs-131 300 TBq (1984)q TB 0 30 INP, Tirana
2 ALCYON II Co-60 220 TBq (1989) 115 TBq Oncological (CGR) Inst., Tirana
3 JUPITER- Co-60 110 TBq (1990) 65 TBq Oncological JUNIER F Inst., Tirana
4 WELL Am-24 1(1982q GB )0 40 q GB 0 39 Geological LOGGING Ent., Fier NEUTRON GAUGE Am/Be
5 NEUTRON Am-241 400 GBq (1984) 390 GBq INP, Tirana SOURCE Am/Be
6 CALIBRA- Cs-137 600 GBq (1988)q GB 0 52 INP, Tirana TION SO- URCE
32 radioactive waste management bot legislativn hi practicad ean l directions.Twoe yearth o sag IAEA RAPAT mission visited Albania.The recommendation thif so s mission included, inter alia, updatin radiatioe th f go n protection legislatio countre th nn i y concept basew ne n dd o san standards, which are accepted by and applied in many countries and organizations mainly those which are related to the use of radioactive materials in medicine, research and industry [2,5,6]. Now this work is headed by the Radiation Protection Commission.
We also appreciat IAEe eth A publications withi RAe nth ASDW S programme, which are considered to be a useful source of information both for the basic safety philosophy and related necessary steps for implementation into daily practice.This year Albania was involved Interregionae th n i l Model Projec Radioactivn to e Waste Managemen framewore th n i t f ko the technical co-operation and assistance with IAEA.lt is hoped that country will benefit through different activities within the project like expert services, equipment and training.
REFERENCES
[1] Principal Regulations on Ionizing Radiation Activities, Decree of Albanian Ministers Council, 197183 . . No . [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Guidance on Radioactive Waste Management Legislation for Application to Users of Radioactive Materials in Medicine, Research and Industry, TECDOC-664, IAEA, Vienna (1992). [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Nature and Magnitude of the Proble f mSpeno t Radiation Sources, TECDOC-620, IAEA, Vienna (1991). ] DOLLANI[4 , ÇUÇIK. , , TechnicaT. , l Repor Conditioninn to Interid gan m Storage of Spent Radiation Sources, INP, Tirana (1992). ] INTERNATIONA[5 L ATOMIC ENERGY AGENCY, Recommendatio Safe th er nfo Regulatiod an e Us Radiatiof no n Source Industryn si , Medicine, Researcd han Training, Safety Series No. 102, IAEA, Vienna (1990). [6] INTERNATIONAL ATOMIC ENERGY AGENCY, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Interim Edition, Safety Serie . 115-1sNo , IAEA, Vienna (1994).
33 WASTE MANAGEMENT PRACTICES AND ISSUES IN DEVELOPING COUNTRIE CASE CROATIF TH E SO - A
D. SUBASIC, S.K. DRAGICEVIC APO - Hazardous Waste Management Agency, Zagreb, Croatia Abstract Radioactive waste in Croatia is generated from various nuclear applications as well as fro e Krsk moperatioP th NP o joina ( n t ventur f Slovenio e Croatia)d an a e nationaTh . l 1 programme on radioactive waste management is aimed at establishing a new independent radiation protection and nuclear safety authority as well as the development of new legislation sitine LLW/ILa Th f .g o W repositor importane th f o i Croati yh e ton steps ai n si the whole radioactive waste management scheme. The concept of sub-regional disposal facility for the needs of the two countries is also under discussion. Faced with the needs for e establishmenth n efficiena f o t t waste management syste r botmfo h radioactivd an e hazardous waste Croatia is trying to implement a new approach by developing a joint institutional infrastructur botr efo h type wastef sactivitieso reviem e Th ma . f wo ,e rolth f eo institutions involved and some on-going projects which round up the present situation in the field of radioactive waste management in Croatia are presented.
1. INTRODUCTION
The Republic of Croatia is one of the new independent countries on the map of mid- South Europe. Besides the problems which all the countries in transition have in common, e pasyearw th fe tr sCroati fo bee s nha a faced witr aggressiohwa d basinan c survival problems. Nevertheless, Croati awarnecessits ae i th f eo establishinf yo improvinr o w ne ga g e existinth g infrastructur n almosi e t each segmen s systeit f o mt- economy , finance, production, environmental protection. As a part of these efforts, substantial changes and improvements in the field of radioactive and hazardous waste management have to be undertaken.
regulationd an w la e s coverinTh g radioactive waste managemente us n i ,e whicar w hno Republie th n i f Croatiaco adoptee ar , d fro formee mth r Yugoslav regulations f thio sl Al . legislation - from the national strategy issue to the guidelines for waste minimization - are planned to be revised in order to meet the specific needs of the new independent country. A new Croatian law in the field has already been drafted to meet the modern world/European standards.
2. QUANTITIES, TYPES AND ACTIVITIES OF RADIOACTIVE WASTE
2.1. Radioactive waste from various nuclear applications
Radioactive waste in Croatia is generated from various nuclear applications. According available toth e data, som institutions0 e50 , with more than 5,000 operating personnele ar , authorized to handle radiation sources in Croatia.In addition, some 50,000 ionizing smoke detectors are distributed in 950 buildings, and more than 600 ionizing lightning rods
(protectors) have been installed in 320 buildings. A total amount of radioactive waste
2 1 estimates generatei 3 ) m (ca w 3 6 hav.o no Croatid n t d i gros n o t ea p sau activity 2.3-10 Bq [1]. The waste contains radionuclides such as - Eu, used in ionizing lightning rods, e ionizinth n i g24 m 1smokA e detectors, 192Ir, ^Sr, 85d somKan r e others applie 152n i d154
35 measuremen processind tan g technique industryn si useo , variou n ^C 137d i d Csan s diagnostic and therapeutic method medicinen si , totae etcTh .l inventor a consist^R f yo abouf so e on t hundred needles and/or tubes. There are 31 users of the open radiation sources (in medical institutions) producing liquid waste of which only the 3H and 14C are causing disposal problems.
2.2. Radioactive waste from the Krsko NPP
Croatia is obliged to participate in finding a solution for the disposal of radioactive waste generating durin e lifetim Krske gth th f oo e NPP joina , t ventur f Slovenio e d aan Croatia, but located in Slovenia. According to the waste origin, treatment and conditioning technologies, as well as the content and form of LLW/ILW generated at the Krsko NPP, six categorie wastf so e exist [2]: spent resins, evaporator bottoms, compressible waste, super- compacted waste, spent filters and other incompressible waste. All types of waste are filled steeL int5 l o drum20 Krsk e th n soi radioactive waste process unit stored an , d on-site th n ei storage facilities. Drum providee ar s d with additional shielding, accordin waste th o egt activity. Since the Krsko NPP started operation in 1981, some 2,000 m3 of LLW/ILW with a total activity of about 3.6-1013 Bq have been generated so far. The majority of drums contain 85% of evaporator bottoms, followed by spent resins -10%, super-compacted waste - 7.5%, compactable wast - e7.0% , other waste - s1.5 spend %an t filters belo. w1% Accordin Croatiae th o gt n legislatio momenf drumse no th ove% n i f thei,80 ro t r filling belong to the category of intermediate level waste.
However, it is realistic to expect some 8,500 m3 of LLW/ILW to be generated at the NPP during its lifetime. According to a rough estimation, the total activity of the LLW and ILW generated in the lifetime of the Krsko NPP could be ca. 1.5-1014 Bq. In addition, some 11,000-12,000 m3 of decommissioning waste is expected to be produced at the Krsko NPP. HLW% e 11 Th .d an W terms n activityI it IL LLW f % projectes o si % 53 t i 36 ,e b o dt prevailing radionuclid expectes ei ^Coe b o d,t representin totae th lgf o activit som% e90 y of decommissioning waste [3].
Considering HLW totae th , l capacit spene th f tyo fue fuel8 poo82 l s assembliesi l d an , 45 f thi%o s volume (314 fuel assemblies bees ha ) n fille faro s d . Becaus recentle th f eo y introduced VANTAGE 5 fuel assembly type and the extended reactor fuel cycle time, the pool capacity will be sufficient until 2001. Afterwards, two options will enable continued spent fuel storage: either to enlarge the capacity of spent fuel pool or to introduce dry storage r fuefo l assemblies.
3. INSTITUTIONAL FRAMEWORK
3.1. Regulatory Body 3.1.1. Present structure of the Regulatory Body
Regulatore Th yRadiatioe Bodth r yfo n Protectio Radioactivd nan e Waste Management Croatin i bees aha n organize sections da three th f es o ministries Ministre th : Healthf yo e th , Ministry of Economy, and the Ministry of Civil Engineering and Environmental Protection. Unfortunately, permaneno thern s i e t bod co-ordinato yt activitiee eth ministriee th f o s s involved. It should be added that a few other ministries are responsible for licensing of some
36 radioactive waste management activities, such as transportation, import-export, release of effluent se basietcTh c. organizationa lMinistre charth f o tf Healt yo subordinatd han e institutions is given in the Fig. 1.
Sanitary Inspectorate Ministre ,th Healtf yo h section competene th s ,i t national authority for radiation protection including radioactive waste management. There is only one person in the Inspectorate dealing with all aspects of radiation protection and radioactive waste management. There is more expertise in the authorized institutions than hi the Sanitary Inspectorate itself Ministre Th . f Healtyo s authorizehha d three additional institutiono t s perform some parts of specific regulatory tasks. The Ministry of Health also suffers from very limited budget availabl regulatorr efo y tasks additionn I . , ther permanena s ei t problem preventing build-up of a competent expert structure in the ministries involved: the salaries in the ministries are not attractive enough.
3.1.2. Expected functions of the re-organized Regulatory Body
Although therdefinito n s ei Regulatore eth solutiow ho n yno Body wil organizede b l , the option in consideration (Fig. 2) is based on the Croatian needs for the forthcoming period. It also respects the recommendations given by the IAEA RAPAT Mission [4] and the experiences of some European countries.
e radiatioTh n protectio d nucleaan n r safet e supposear ye regulateb n o a t d y b d independent authority - the State Administration for Radiological and Nuclear Safety (SARNS). The SARNS could be supported by the National Nuclear Commission (NNC), whic expectes hi establishee b o dt Governmene th y db t very soon. Activities committeo dt SARNe dividee b th y Sdma into three groups radiatio) (a : n safety nuclea) (b , r) safet(c d yan common services to both sections.
MINISTR HEALTF YO H Sanitary Inspection Department
Hea f Departmendo t * Main sanitary inspector Sanitary inspectors of the state borders
Authorized Institutions t (research institutes, ECOTEC etc.) RADIATION SOURCE USERS
DIRECTOR
Officer responsible for implementation of radiation protection measures
Head of company departments
Responsible person departmentn i s s
Complete department staff
FIG. 1. Basic organizational chart of the Ministy of Health and subordinate organizations
37 NATIONAL NUCLEAR GOVERNMENT MINISTRIES COMMISSION
EXPERTS
STATE ADMINISTRATION FOR RADIOLOGICAL AND NUCLEAR SAFETY
DIVISION OF DIVISION OF DIVISIOF NO RADIATION SAFETY NUCLEAR COMMON SERVICES SAFETY AUTHORIZED EXPERT INSTITUTIONS
USERS OF RADIATION SOURCES AND NUCLEAR FACILITIES
FIG. Expected2. organizational chart re-organized ofthe Regulatory Body
On the other hand, the recently adopted Law on Health Protection foresees the creation e Nationaoth f l Agenc r Radiatioyfo n Protection (NARP) whic s supposehi perforo dt m similar activities committed to the Section of Radiation Safety of SARNS (including licensing, issuin codef go practicef so controd an , radiatiof lo n sources) consideres i t I . d that the NARP coul t withidac Ministre nth Healtf yo h only unti SARNe th l establisheds Si .
3.2. Other institutions in the radioactive waste management framework
Ther foue ear r group f institutionso s dealing with radioactive wast Croatian ei : a) National operational organization - APO - Hazardous Waste Management Agency is Governmene closth o et t (Ministr f Economyyo , Ministr f Civiyo l Engineerind gan Environmental Protection, Ministry of Health) and responsible for establishing and maintainin efficienn ga t hazardous, including radioactive waste, management system. alss i t oI authorize Governmene th y db organizo t perford ean m some specific actions like environmental restoratio humad nan n health protection.
) b National research institute "Rudee Th s- r Boskovic" institute "Institutth d r ean efo Medical Researc d Occupationaan h l Health e authorizear " o perfort d m personnel dosimetry and radiological monitoring programmes such as the monitoring of releases fro Krske mint environmentP th e oth o NP . Temporary storag radioactivr efo e wastd ean radiation sources has been established in both institutes. c) Private companies - The private company ECOTEC is authorized to import and transport radiation sources and to accomplish some other tasks. It has an important role in all in situ actions where handling of radioactive materials (and waste) is needed. d) Users of radiation sources - The users are obliged to manage their own waste by using generaln i , , three methods ) wast(a : s storeei d unti activits it l y falls beloe wth
38 prescribed level, and then is managed as non-radioactive waste; (b) waste is stored at temporaro ontw f eo y storage facilities som) (c d e spenan ; t radioactive material- re e sar t possiblno s i t ethe- i returne e f i yar use - r do d producee bacth o kt r (mostlf o t you Croatia).
4. RADIOACTIVE WASTE MANAGEMENT SYSTEM
4.1. National strategy
The radioactive waste management strategy drafted in 1992 by the APO [5] is to be approve competene th y db t ministries.The strateg recentls yha y been rearranged according to some of the latest IAEA publications [6].
These are the following main objectives outlined in the strategy:
(1) to identify in detail sources and quantities of radioactive waste hi Croatia, as well to prepare and maintain the inventory of waste; elaborato (2t ) legaea l framework, i.esystee th . responsibilitiesf mo ; establiso t ) (3 hmechanisa fundinr mnationae fo th f go l radioactive waste programme; (4) to introduce regulations on radioactive waste handling, transport, treatment and disposal; develoo t ) pLLW/IL(5 a W repository project (site selection, technical design, safety); fosteo (6t ) r public relations; givo t efula ) l (7 suppor otheo t r radiation safety related actions.
4.2. The radioactive waste repository project
Up to now the following activities have been initiated and systematically dealt with:
sita ae) selection process, b) a disposal facility preliminary design, both tunnel and surface type, preliminara ) c y safety assessmen risd kan t analysi preparee th r sfo d designs, detailea d) d characterizatio f storeno d radioactive waste, ) e creating positive climat publii eh termi ch understandinf so acceptancd gan facte th sf eo thaproblee th t f mradioactivo solvee eb thad wasto t drepositoryan a ts eha , although an unpopula i someone'h re b facility o t s s ha neighborhood, .
Siting of a radioactive waste repository in Croatia has been adjusted to the regional planning. It comprises two stages- the first is the site selection, terminating by the inclusion of candidate sites into the Regional Plan of Croatia; and the second - site evaluation stage, aimin t defininga finaa g l repository site through field investigation othed an s r necessary actions [7]. Afte vera r y slow situatioe progres th country e o t th e n i sdu , preferred sitee sar planned to be selected hi a couple of months. An interdisciplinary approach is fully applied hi the site selection (all relevant topics from geology to sociology are involved). The standard screening technique has been applied in the selection process. The stepwise approach, based on verification of every step, started with screening of the entire territory of the country, and is going to be finished with the selection of a few candidate sites. It has been based on the implementation of both exclusive and comparative criteria [8].
A system approach includes synchronous preparation of the repository project design [9], performance assessmen somd an t e other activities. Extremely high attentio gives ni o nt
39 providin a gfull , complete, constan d honesan t t informatio e publicth o t n. Necessary preparation involvemene th r sfo f locao t l communities intsite oth e selection process have already been done. Democratic efforts in the repository siting process, as a controversial facility, also includ identificatioe eth incentivef no s neede givee b o dlocano t t l communities. Being afrai f NIMTOdo O behaviour vere ar ye carefuw , preveno t l t "premature" exposure of politicians to the consequences of the NIMBY effect.
Due to the joint Croatian and Slovenian ownership of the Krsko NPP there are several possible final disposal options. A final decision, whether the LL/IL radioactive waste repository shoul boti buile h d b hradioactivp e t u countries th l themf al o r e fo ,jusr on e o ,n i t waste produced in both countries, has not been made yet. The idea of a joint radioactive waste repository construction for the Republic of Croatia and the Republic of Slovenia-as a kind of sub-regional facility, is still an issue to be discussed.
4.3. The ongoing environmental remediation projects
Amon gnumbea problemsf ro rescuine th , damagef go destroyer do d radiation sources hi the areas of Croatia affected by the war figures out as the most urgent problem [10]. In accordance with the records, the mean (per source) activity of installed smoke detectors hi the areas of Croatia affected by the war is 100 kBq, of lightning rods 7 GBq and that of sources used in industry 370 MBq-7.4 GBq. The greatest attention is paid to the sources hi destroyed buildings, sources being out of containment or out of control for any other reason. Their uncontrolled transfer could imperi onlt lno y other part thif so s countr coult ybu d have a negative transboundary effects as well.So far, the sources from roughly a half of the affected territories have been surveye removedd dan .
In 1993, Croatia joined the IAEA Technical Co-operation Project on Environmental Restoration in Central and Eastern Europe with the APO appointed as the project co- ordinator. There are four groups of radioactive contaminated sites identified hi Croatia: (1) sites containing coal slag/ash piles ) site(2 ; s containing phosphate phosphor-gypsud an s m remained from fertilizers industry ) geotherma(3 ; l spring gas/oid san l drilling) site(4 sd an ; containing natural radioactive materials.The highest priorities in the clean-up action are suppose site o givee b tw s o dcontainino t nt g coal slag/as fertilizere h pileon o wels t s a s a ls factory.
JOINE TH T. 5 APPROAC RADIOACTIVO HT HAZARDOUD EAN S WASTE MANAGEMENT
Taking into account the problems of radioactive and hazardous waste management, the point common si n become strikingly evident. Firs allf basie o t th , csamee ideath ) e a : sar reduce, reuse or recycle as much as possible, b) a fundamental requirement in final disposal, for both type f wasteo s s maximui , m protectio f presenno futurd an t e generations) c , necessit communicatiof yo n with public, opennes publicitd s an l action al projectsn yd i san .
There are a number of actions which should precede efficient radioactive and hazardous waste management such as an analysis of status and problems, review of the existing legislation, draftin legislatiow ne g n documents, communicatio co-operatiod nan n wite hth authorities wels a ,wit s la h publi locad can l community, preparatio wastf no e types/quantities data base inventory, etc.
40 The siting processe treatmenr sfo disposad an t l facilitie almose sar same th t termn ei s of public acceptance, licensin "technicad gan l process site th e f selection"o - from preparing site-selectio decision-makine th f o d n en criteri e th go at process.
In order to make a very complicated process of creating and implementing the waste management system faste mord ran e efficient approache idee th f th ,a o t leasa , somo t e parts of the required infrastructure, has been adopted in Croatia [11] .There are numerous benefits approache oth f majoe beinth ,e on r g avoidin duae gth l system establishment. Furthermore, ther certaia s ei n know-ho fiele LL/If th do w n i L radioactive waste managemen Croatian i t , which could find a good use in hazardous waste management.
f thato Op , nto associatin problee gth f radioactivmo e waste management wite hth management of other hazardous wastes, enables those problems to be compared and correlated. Makin problee gth f safmo e treatmen disposad an t f LL/Io l L radioactive waste a one-item agenda could creat impression ea i publinh c that radioactive onle wastth ys ei waste environmental proble mCroatian i , whic s definitelhi r frofa ym beine g th true n O . other hand, insisting on the problem of the radioactive waste disposal could evoke the fear opossibla f e future nuclear programme, which could hav adversn ea epublie effecth n co t acceptance of waste management activities.
Considering the joint process of site selection for disposal of radioactive and hazardous waste it is obvious that the process would be optimized due to time and money savings. Site selection is a relatively expensive, slow and time consuming process, especially owing to public sensitivity to the issue.
importane Th t benefit, especiall r smalyfo l countries wit hhiga h population density, come se countr fro th e optimaf mth o y e territoryus l t becausNo . f somo e hectaren ete s needed for the facility itself, but because of the surrounding land, the use of which is going affecte e vicinite b wast e th o t th y f db ey o facility, even whe nscientifio thern e ear c reasons for such restrictions.
6. CONCLUSION
Faced wit challenge hth f improvineo existine gth g and/o wastw r ne settin ea p gu management infrastructure for both radioactive and hazardous waste, Croatia has tried to answer that wayschallengw fe prepariny nationab a -w ne n i e e lgth radioactive waste management strategy, establishin structurw ne implementine d gth ean joine gth t approaco ht radioactive and hazardous waste management. The idea of common elements of the infrastructure enables optimizatio f limiteno d national resource significand an s t tund ean money savings. Afte e initiath r l steps have been undertake i thanh t directio e firsnth t experiences are encouraging.
REFERENCES
] [1 HAZARDOUS WASTE MANAGEMENT AGENCY, Quantitie Characterizatiod san n of Radioactive Waste Material from Institutes, Medicine and Industry in the Republic of Croatia, Zagreb (1993) KRSKP NP O ] WAST[2 E MANAGEMENT DEPARTMENT KrskP NP ,o Official Report, Krsko (1992).
41 [3] SUBASIC,D., SALER,!., SKANATA,D., "Basic Elements of Radioactive Waste Management Strateg e Republith n yi f Croatiaco , Nuclear Waste Managemend an t Environmental Remediation", (Proc.Int. Conf.Prague, 1993), Vol.2, Prague, Czechoslovakia (1993) 787. [4] RAPAT MISSIO CROATIAO NT , Travel Report Jun8 Jul2 2 , - ey 1993, IAEA, Vienna, 1993. ] HAZARDOU[5 S WASTE MANAGEMENT AGENCY, Waste Management Strategr yfo Republie th f Croatiaco , Proposal Document, Zagreb (1992). ] INTERNATIONA[6 L ATOMIC ENERGY AGENCY e PrincipleTh , f Radioactivo s e Waste Management, Safety Series No. 111-F, IAEA, Vienna (1993). [7] THE GOVERNMENT OF THE REPUBLIC OF CROATIA, Site Selection Criteria for Fossil Fuelled Power Plants and Nuclear Facilities - a regional planning background, researches and suitability assessment for the territory of the Republic of Croatia, Zagreb (1991). ] SALER,T.[8 , "General Approac Site-Selectiod han n Criteri LLW/ILr afo W Repository in Croatia", (Proc. 28th Int.Geological Congress Berkeley, 1992), University of California, Berkeley, California (1992). [9] KUCAR-DRAGICEVIC,S.,SKANATA,D., "The Tunnel Concep LL/If o t L Radwaste Repository and Results of Safety Analysis", (Proc. 1st Meeting of Nuclear Society of Slovenia, Bovec, 1992), Bovec, Slovenia (1992). [10] SUBASI , SALECD. , GUNARIRA. , NOVAKOVICM. , "ThCM. e Radioactive Waste Management in the Areas of Croatia Affected by the War", WM Symposia 93, (Proc. Tucson, 1993),Tucson, Arizona (1993) 127. [11] KUCAR-DRAGICEVIC,S., SUBASIC D., "New Aproach Towards Joint Radioactive and Hazardous Waste Managemen n Smali t l Countries", SPECTRUM'94 (Proc. Nuclea Hazardoud ran s W.M.Topical Meeting Atlanta, 1994), Vol.2, Atlanta, Georgia, (1994).
42 RADIOACIWE WASTE MANAGEMEN CUBTN I A RESULT PERSPECTIVED SAN S
L.A. JOVA SED Centre for Radiological Protection and Hygiène
L.M. PUMAREJO National Centr r Nucleaefo r Safety
H.D. NIEVES, N.G. LEYVA Centr Radiologicar efo l Protectio Hygiènd nan e
Habana, Cuba
Abstract e futurTh e safe developmen f nucleao t r energ progressivd an y e increasinf o e gus radioactive material medicinen i s , research, industr othed yan r e Republifieldth n i s f co Cuba in the past years have determined the necessity to formulate and apply a national policy to assure harmless and ecologically rational management of radioactive wastes.The ruling principles for the application of the established radioactive waste management policy in Cub e summarized.Thar a e element e infrastructurth f o s e existin e countryth n gi e th , legislative framework and the technical resources for attachment of important tasks related to the radioactive waste including spent sealed source management are further brought. Some resultstudiese th f so , which serve basia r desigs da sfo constructiod nan facilita f no r fo y treatment and conditioning of low level liquid and solid radioactive wastes, are also given.The main characteristics of the facility are described.The main ideas which govern the improvement of the safety and effectiveness of the radioactive waste management in Cuba in the coming years are finally discussed.
1. RADIOACTIVE WASTE MANAGEMENT POLICY IN CUBA
A few matters have been drawn more attention by scientists, governments, and the general e laspubli0 yearth 3 t n si c than radioactive waste issues. Nuclear techniques application programmes developed by the Republic of Cuba do not bypass the course of actions relating to this important issue. The main objective of the Cuban policy in this field o ensurt s i a harmlese ecologicalld an s y rational managemen f radioactivo t e wastey b s applying recommended and accepted methods. It lies in the following bases:
Prohibition of direct discharges of any kind and quantity of radioactive wastes into the environment. Classification and segregation of radioactive wastes and their storage for decay at the waste producer sites. Plannin sitn i uf go treatment , conditionin d interigan m storag f radioactiveo e wastes from nuclear power stations. Central collectio storagd nan f wasteeo spend san t sealed sources. Treatment, conditioning, and interim storage of wastes and spent sealed sources from institutionae th f radioisotopeo e us l waste th t esa treatment facility buil thio t t s end. Final disposal of conditioned radioactive wastes and spent sealed sources into a projected national repository.
43 2. WASTE MANAGEMENT INFRASTRUCTURE
The Cuban's integrated policy of nuclear development is entrusted to the Executive Secretariat for Nuclear Affairs, which is the regulatory body in charge of radiological protection and nuclear safety. For this purpose the Secretariat has a special implementing body the National Centre for Nuclear Safety. The National Centre for Nuclear Safety is responsibl r licensinefo supervisiod gan nucleaf no r installations.
e organizatioTh n responsibl e wastth r e fo emanagemen t e policCentrth r s i yfo e Radiological Protection and Hygiene, which is also entrusted with the main services and development f importanso tNationae partth f so l Radiological Protection Systemo t d an , whic existinn ha g treatmenfacilite th r yfo f radioactivo t e wastes belongs.
e regulatorTh y framework f legao comprise t l se documents a f so f ruline o th ,e gon which is the Decree-Law for the National Regulation of the Peaceful Use of Atomic Energy. The regulatory body, the Executive Secretariat for Nuclear Affairs (for everything concerning radiological and nuclear safety, including radioactive waste management) has been enacted in this document.
The limits which define wastes as radioactive, the prohibition of their discharge to the environment organizatioe th d an , centraf no l collectio storagd nan radioactivf eo e wasted san spent sealed source establishee sar decree th n di e which regulate wore sth k with radioactive materials [1].
regulatione Th importr sfo , transi expord an t f radioactivo t e materialse wels th a , s a l requirement issuinf so g special permit performinr sfo g those activitie establishee sar e th n i d safe decreth e r transporefo f radioactivo t e materials.The method registratioe th r sfo d nan contro f radioactivo l e waste establishee sar practica n di e guid whicn ei h more elementr sfo the classification of wastes are also given; their storage at the waste producer for a minimum of one year is established, and the requirements for their collection and segregation are stated [2].
. RADIOACTIV3 E WASTE MANAGEMENT
At present radioactive th , e wastes existin i Cubgh classifiee aar d accordin theio gt r origin hi three groups: the first and most significant one, based on the volumes as well as the number of facilities where they are generated, are the wastes from different kinds of research performe fielde th biologyf o n sdi , industry, medicin agricultured ean secone th ; d one is related to those wastes generated in medical applications; and the third group includes the wastes originated in the research hi the field of treatment of radioactive wastes, and production of radioisotopes and labeled compounds.
Table I shows the typical radionuclides of each of the above mentioned groups. As shown in Fig.l, management of radioactive wastes from any of those groups comprises collection, segregation and storage for decay at the waste producer sites, then* centralized transfee wastth o et r storag d treatmenan e t facility (WSTF) d oncan , e there, their conditioning for interim storage. Conditioning of liquids is performed by cementation, whereas conditionin f solidgo dons si compactiony eb .
During collectio wastesf no personnee th , WSTe th f lo F contro wastee th l s which have decayed, and perform their disposal. The wastes to be disposed of are those in which the
44 TABL E. TYPICAI L RADIONUCLIDE EACN SI H GROU WASTEF PO S ACCORDING TO THEIR ORIGIN
Group Typical radionuclides 1 H-3; C-14; 1-125; P-32; S-35 2 1-125; Tc-99m; 1-131; Cr-51; Fe-59; Co-57; Co-58; Ga-67 3 Co-60; P-32; Cs-137; Sr-90; Ce-144 Eu 152; H-3; C-14; 1-125; 1-131
I Medicaj l i Research Radioisotopej implications! ! Applications labeled com pj Wast e producer- 1______l product ion !
\ Collection i ! Segreg-at ion i Waste producer
'. Storagr efo 1 decay Waste producer /S^ xempt lern le«el_.X'\ ~ ' ————————s Choicp e USTF S
i Collect and transport
Contro I i — i In t er in ; ; Storage
I Treatment USTF Conditioning
j ———— WSTF -; Storage > 18 y j—————| Disposal j
ISTF - Waste Storage and Treat»ent Facility.
FIG.l. Schedule for the management of radioactive wastes
45 activity is below the exemption levels.The regulation in force states that only the WSTF personnel are authorized to perform disposal of radioactive wastes.
4. SPENT SEALED SOURCES MANAGEMENT
The planned scheme for the management of spent sealed sources is shown in Fig. 2. presente th U o pe sourcet th , s show Tabln ni I havI e e been only collecte r storagedfo . Options for the management of spent sealed sources comprise the source recycling for use in other institutions reture th , mosf no t active sources wheneve possibls i t suppliere ri th o et , and centralized interim storage at WSTF.
5. MAIN RESULTS OF THE STUDIES ON TREATMENT OF RADIOACTIVE WASTES
Studies performe mose th tn o dappropriat e method treatmene th r fo s f radioactivo t e waste Cubn si a have bee differenf o ne directeus e t th natura o dt l sorbents suc turs d ha an f zeolites; to the chemical treatment of wastes using different chemicals and to the immobilizatio wastef no cemenn si t [3-4]. Ferrocianates, carbonate phosphated san s together with the sorption by natural zeolites resulted in a very effective chemical treatment of complex radiochemica chemicad an l l wastes [5]. Modifyin nationae gth l zeolites gavn ea increase in sorptiveness for the treatment of wastes containing Cs-137, Co-60 and Sr-85 radionuclide .Th] e s[6 influenc radiatiof eo thermad nan l treatmen behavioe th n to zeolitef ro s has been also studied [7].
recycle Supplier Sources in
/ ^
H Dec it icm about no use h
Col lectio« y tesit u i ( tt Ac i Transport ! Leakage test Short lived Long lived
Storage for decay Interim storage
Conditioning Transport
hiteriro storage Evacuation
iransporx
Projected2. FIG. schedule spentfor sealed sources management
46 TABL ESTIMATEEL I D ACTIVITIE QUANTITIED SAN SPENF SO T SEALED SOURCES STORED AT THE WSTF
Radionucl ide Quantity Total of sources activity
Am-241 872 .00 9.44 GBq Am -Be 3.00 76 .22 GBq C-14 5.00 0.14 MBq Cd-107 1.00 37 .00 MBq Cd-109 1.00 0.OO3 Cf-252 1.00 0.00 Co-57 2.00 8.40 MBq Co-60 91 .00 592 .42 TBq Cs-137 90 .00 64 .09 TBq Eu-152 1.00 0.00 Fe-55 1.00 370 .00 MBq Fe-57 1.00 1.85 GBq H-3 13 .00 11 .10 GBq Ir-192 100 .00 1.88 TBq Kr-85 2.00 3 .77 GBq Na-22 4.00 74 .00 MBq Ni -63 1.00 370 .00 MBq Pm-147 3 .00 14 .80 GBq Po-Be 6.00 1.85 GBq Pu-Be 3.00 554 .00 GBq Pu-238 3 .00 2.22 GBq Pu-239 19 .00 1.27 MBq Ra-226 118 .00 58 .13 GBq Ru-106 1.00 0.00 S-35 1.00 0.07 MBq Sn-119 1.00 0.00 Sr-89 3 .00 37 .18 MBq Sr-90 148 .00 16 .91 GBq Tl-204 22 .00 1.08 GBq unknown 65 .00 370 .61 GBq
Total 1582 .00 659 .51 TBq
0.0 0- source unknowf so n activity
For conditioned wastes, studie leachinn so f matrixego f cemenso t mixed with Cuban zeolitic rock have been carried out under different conditions [8]. For the treatment of radioactive wastes containing 1-125, C-14 and H-3 radionuclides, mixed beds of organic synthetic ionites and activated Cuban charcoals have been employed [9].
6. THE WASTE TREATMENT FACILITY
The main difficulty encountered by most of the developing countries is the lack of facilitie treatmenr sfo storagd tan differenf eo t type f radioactivso e wastes resulting froe mth applications of nuclear techniques. In Cuba, it was decided to build a facility for the treatment of low level wastes and their interim storage for 10 - 15 year period of time; the storag f speneo t sealed source alss si o included.
47 As shown in Fig. 3, the technology of the WSTF comprises coagulation-flocculation process with iron and aluminum salts in a basic medium, its sedimentation and a process of ion-exchange in zeolites or organic synthetic ionites for liquid wastes; conditioning of resulting sludge spend san t sorbent cemenn si compactiod an , t f solino d wastes. Usine gth
technology of the WSTF, some 30 m3 per year of low-level liquid wastes may be processed. compactioe th r Fo f kgf/cmsolids8 no 8 WSTe n a th , s 2 press Fha volume Th . e reduction factor reache. 3 s di
Arrive and Segregation
liquids solids
non organic aqueous compressible iConpressiMe.
I i s*H£ll' c^Ki^ 4- press ing j solvent treatnent cenent jrecov-ng {supernatant lw.
sol. filtering
ion exchange 1 ; inter in \ final i conditioning / disposal ! concentrates 1 storage i
technologyThe FIG. 3. appliedWastethe in Storage Treatmentand Facility
. WAST7 SPEND EAN T SEALED SOURCES MANAGEMENT: PROSPECTS
The results obtained so far, such as training of the personnel responsible for radioactive waste management, assurance of the state supervision in radiation protection, and the existing regulatory framework allow to project the future years' work to its consolidation. r thiFo s purpose s necessari t i , r specifio applou t y n i yc condition e internationath s l recommendations relatin exemptiogo t n level wels sa intensif s la controle existine yth th f o s g practices.This way, the central collection of wastes should be directed to the relevant ones.
48 Likewise, it is necessary to develop and implement the methods for conditioning of spent sealed sources, which will enhance safety during their storage as well as foster the development of plans for the projected life of the existing sources. Special attention should informatiogivee e b th o nt n educatioe disclosurth publie o t th d f issuen co i an e s relating to the policy and practices performed in the management of radioactive wastes including spent sealed sources.
REFERENCES
[1] DECRETO NO 142 Reglamento para el trabajo con las sustancias radiactivas y otras fuentes de radiaciones ionizantes de 24 de marzo de 1988. Gaceta Ofïcial de la Repûblica (1988). [2] Guïa para el Control y el Registre de los Desechos Radiactivos, CPHR-VA-88-Ol.CIEN, CIEN, La Habana (1988). ] DOMENEC[3 , CHALEHH. , CASTILLSG. , EstudiOR. o sobr a l eposibilida l dde empleo de la turba en calidad de sorbente para el tratamiento de desechos radiactivos liquides (Resumen de los trabajos concluidos en el quinquenio 75-80). Consejo Cientifico Técnic l CAMode E par l estudis e métodoa lo e d oe tratamient d s e d o desechos radiactivo descontaminaciosy e superficiesnd , Moscû (1980). ] CHALE[4 , CASTILLSG. , EvaluatioOR. zeolita l e nd a par l e atratamient s lo e od desechos radiactivos liquides contemplando su posterior solidificaciön con cemento. Estudio Especial EE-461-06-80, CEADEN, (1981). [5] CHALES G., CASTILLO R., JOVA L., DE LA CRUZ O., Descon- lamination de desechos radiactivos liquidos mediante tratamiento quûnic osorciöy n zeolitanco s naturales, Nucleus 2 (1987). [6] NOVO , DOMINGUEAJ. , MORENZJ. , PREVAOD. . LCaracterizaciöI a l e nd zeolit yacimientl ade Piojilll oE o par empleu as gestioa l n o e desechoe nd s radiactivos, Nucleu (1987)s3 . [7] DOMINGUEZ J., FEREZ A., PREVAL I., QUInONES I., RUBIO E. Estudio de la influenci l tratamientde a o térmico, quûnic oy radiaciona sorciöa l n e le cesi nd oy cobalt zeolitan oe s naturales, Nucleu (1989)s6 . [8] CHALES G., CASTILLO R., AVILA R., Lixiviaciön de desechos radiactivos de baja actividad inmovilizados por cémentation empleando como aditivo roca zeolitica cubana, Nucleus 5 (1988). ] CHALE[9 , DOMINGUESG. , CASTILLZJ. , TratamientOR. oe inmovilizacio e nd desechos radiactivos liquidos de H-3, C-14 y 1-125, Informe final de tema, CEADEN (1990).
49 STATU RADIOACTIVF SO E WASTE TREATMENT DISPOSAD AN CHINLN I A
LUO SHANGGENG Department of Radiochemistry, China Institut f Atomieo c Energy
LI XUEQUN Burea f Safetyuo , Protectio Healthd nan , China National Nuclear Corporation
Beijing, China
Abstract
This paper outline e radioactivth s e waste management activitie Chinan i s , mainly concerned with:
) Strategie(1 regulationd san radioactivf so e waste management, ) Radioactiv(2 e waste treatmen conditioningd an t , (3) Radioactive waste disposald an , (4) Decommissionin decontaminationd gan .
Chine Th a nuclear industr histora s yha y more tha years0 n3 integran A . l systee th mf o nuclear fuel cycl bees eha n established. Wit operatioe hth variouf no s nuclear reactord san spent fuel reprocessing plants, a large amount of LLW and ILW and quite a lot of HLW have been generated.
1. STRATEGIE REGULATIOND SAN S
1.1. Polic Strategd yan y
Owine followinth o gt g measure radioactive th s e waste managemen s regardei t s da satisfactory:
The pollutant itself must manage its own radioactive wastes; Radioactive waste treatment facilities shall be set up at the same time when the main nuclear facility is constructed; Safety analysis and environment impact assessment reports shall be prepared; The principle of "Controlled generation, categorized collection, volume reduction, immobilization, reliable packaging, in-situ storage, safe transportation regionad an , l disposal" is followed in managing LLW and ILW; Release of radionuclides into the environment shall be restricted; and Radiation protection principles shal appliee b l radioactivo dt e waste management.
Environmene Th t Protectio bees ha nt promulgatednAc Atomie Th . c d Energan t yAc Nuclear Pollution Control Act are being developed. All these acts will greatly facilitate safe and efficient radioactive waste management.
51 1.2. Regulations and Standards
orden I ensuro t r e safe managemen f radioactivo t e wastes Nationae th , l Environment Protection Agency (NEPA), National Nuclear Safety Administration (NNSA) and China National Nuclear Corporation (CNNC) have paid great attention to establishing and promulgating policies, regulations and standards for radioactive waste management, particularly in the past 10 years.
At present ,standard 3 ther4 e systeme ear th i sh whic n i , hav6 h1 e been issued. These standard dividee b n sca d categoriesinto otw generas i e on : l standard radioactivr sfo e waste management; the other is those to be used in controlling or managing the various processes in radioactive waste management.
Beside generae sth l principles followine th , g specific principle followee sar draftinn di g the relevant standards:
) (1 Safety firsproteco t - t environmene tth humad an t n healt thif ho s generatio futurd nan e generations; (2) Economy - to implement the ALARA principle; ) Takin(3 g disposa core th f wast eo s a l e managementd an ; ) Adoptin(4 g internationa foreigd an t lstandard ar ne statth f eso accordin nationae th o gt l conditions.
. RADIOACTIV2 E WASTE TREATMEN CONDITIONIND TAN G
levew lo l e radioactivTh e gase liquiddischarged e b san n sca d int environmene oth t only whe permissible n th the e cleaned yar an p deu level achievede ar s . Such dischargee ar s controlle factorso tw y db : total discharge amoun specifid an t c activity.
The solid wastes are separately collected on-site according to their physical properties and specific activities.In some places volume th , e reductio carries compressiony i b t dou n or incineration. Two kinds of a compactor with unidirectional force or three directional forces have been develope t intpu o d use.Adan Chine th t a Institut f Atomieo c Energy threa , e directional forces compactor with pressur 100f eo installed1s i Qinshae th t .A n Nuclear Power Plant, an unidirectional force compactor with pressure of 30 t is used. At present, the incineratio radioactivf no extensiveleo s waste t no s si y use Chinan di .
Wet solid waste e immobilizear s cementationy db , bituminizatio vitrificationd nan . SYNROC process as the second generation HLW-solidified process is being studied.
2.1. Cementation
For solidification of LLW and ILW, cementation process is adopted at the Qinshan NPP and the Daya Bay NPP. The former uses in-drum cementation, while the latter uses the outer-drum mixing cementation Chine .Th a Institut Atomif eo c Energ developes yha dseriea s of cement formulations and the characterization methods for solidified products, and it has especially developed a planet mixer which is equipped with two twisted dragon-type mixers gnawing each other at 90°, and moving up and down in 200 L standard drum with self rotary common-rotard an m rp 2 1 rpm2 distancf e f o yo Th . e betwee mixee ndruth d ran m wals i l
52 minimue 1Th 0 nun. m distance betwee mixee ndruth d ran m . Compare bottomm 5 ms i d wit commoe hth n mixers, this mixe obvious rha s advantages suc: has (1) mixing homogeneity, (2) self clean-up without residue mixee th n rso when movin abovp gu drume eth , ) stabl(3 e operation without splashd an , (4) loading factor 90%.
2.2. Bituminization
A bituminization facilit operationn i yw witno s hi . capacitd ThimV s 2 facility1 a s yi
film-type evaporator equipped with a rotary scraper. This evaporator has a heating area of steam2a heatem scrapee MP 5 2-2.Th .y 2. db 5 r axis rotate t 650-85sa meltine 0Th rpmg. solidification products are poured into 200 L drum. The specifications of the solidified waste followss a e ar :
Sal% twt loadin 0 4 g factor % wt 1 < Water content Leaching rate (Na+) lO"4 g/cm2 Resistance against radiatio xlO5 ny G 4 Softening point 70-100°C Exothermic starting point min. 260°C Pyrophoric point min. 300°C
In order to prevent salt scale formation on the internal evaporator wall, hydrocarbonyl sulphonate having surface activatio bees nha n used, accordingly causing homogeneous mixing of the wastes and asphalt.
2.3. Vitrification MorhavW ee HL bee thaf o 3 nm produce1000 Chinan di . The safele yar y storen di stainless steel tanks. The activities in the vitrification of HLW began in 70's.The fundamental research work regarding glass formulatio glasd nan s product characterizations i carried out in the China Institute of Atomic Energy.The Beijing Institute of Nuclear Engineering (BINE) is responsible for the design. The R&D work first was focused on the batch pot process which was abandoned in 1985 because of its limited throughput. In 1988, liquid-fee th d ceramic melter process (LFCM bees )ha n chosen jouie Th .t desig fula r lnfo scale non-radioactive mock-up facility VPM (Vitrification Plant Mock-up) has been fulfilled in German 1991.Thn yi e waste oxide loadin borosilicatn gi e glas 16%.Ths si e glass frit will beadsm maximue m appliee 2 Th .b fore 1- th f mo mn di temperatur glasf eo s meltin 118gs i 0 °C.The design throughput of the VPM is 65 L/h feed.
Accordin schedulee th go t meltee th , r wil deliveree lb d from German Chino yt 199n ai 4 and put into operation hi 1995. Based on the results and experience gained from the VPM design and construction, an active vitrification plant will begin work hi the second half of the 90's. It is planned to construct an active facility by 2000.
3. RADIOACTIVE WASTE DISPOSAL
Over 50,000 m of low and intermediate level solid radioactive wastes have been accumulated hi China 3hi the past 30 years. In this decade, more or less the same amount of
53 wastes wil generatee b l d along wit nucleae hth r facilities decommissioning. Together with the NPP radioactive wastes, more than 100,000 m3 of low and intermediate level solid radioactive wastes will be stored on-site in storage facilities by 2000.
Based on the investigation of transportation risk and benefit-cost analysis, the principles of regional disposal has been established.
(1) Construction of a low and intermediate level radioactive waste repository shall be regarded as one of the prerequisites for the development of nuclear power followed by decommissionin f nucleago r facilities importann A . t conten examinatioe th s i t f no safety analysi d environmenan s t impact assessmen f nucleao t r facilitiee th y b s environmental protectio safetd nan y supervision authorities.
(2) When new nuclear power plants and nuclear facilities are put into operation, radioactive waste disposal shall be taken into consideration. Temporary storage of LLW and ILW plane th i th are limites ai years5 o dt .
(3) A regional repository shall be established and radioactive wastes shall be disposed of as close to the plant as possible. These repositories are located at favourable locations, taking into account safety, economy, technological, social factors, and the conditions of geograph communicationd yan s unde unifiee th r d national plan adjoinind an , o gt existin r plannego d large-scale nuclear enterprises. They will accept waste t onlsno y from nuclear industry and nuclear power plants, but also from nuclear applications.
(4) It is prohibited any institution to operate its own LLW/ILW repository, or use its interim storage facilit permanena s ya stipulates i t t onei d , an LLW/ILe d th tha l al t W must be concentrated and disposed of in a regional disposal repository with the state license.
e nexth t n (5I 10-2) 0 years e repositorieth , e Easth tn i sChina e Soutth , h Chinae th , Northwest China, and the Southwest China will be set up steadily.
(6) The CNNC takes the responsibility for siting, construction and operation of a regional repository for LLW/ILW, and the NEPA is responsible for reviewing and approving e environmentath l impact assessment e reportrepositoryth f o s , formulatind an g promulgating relevant standard regulationd san guidelinesd san locae .Th l environmental protection authorities are responsible for supervising environmental protection activities in disposal sites. long-tere Th ) m(7 loans shal providee lb capita e statee parth d th f y an ,o td l b construction cost of NPP shall be allocated as the initial fund which is used mainly during design construction, and initial operation. The repository provides service on the basis of compensatory approach. The income collected will be used to pay off the loans and maintai operatione nth .
3.1. Siting Activities
Since the early 80's, complying with the national standards and basing on the expert suggestion relatee th n wels do s a s IAEa l A criteria geologicae th , l selectio disposaf no l sites for LLW/IL bees W Easha e n th carrie ti Chinah t dSoute ou th , h China Northwese th , t China, anSouthwese dth t China respectively.
54 e Easth tn I China, since 198 8a programm e disposath n o e l site selectior fo n LLW/ILW has been performed by CNNC. 17 suitable areas were found suitable according to the geological map, and 21 potential sites were investigated in the field, among them 5 candidates were recommended such as one near surface disposal site, two abandoned zinc- lead mines, one abandoned uranium mine and one artificial cavity at Qinshan.
In the Northwest China, 6 preliminary disposal sites have been proposed, and two of the chosee mar candidates na characterizatione th r sfo .
e Southwesth n I t China preliminar8 3 , y disposal sites have been selectee th n o d geological mapsthef o mn te ,wer e investigate fielde th least a n d,i candidate3 t s were chosen for further investigation.
In the South China, 30 preliminary disposal sites have been selected on the geological maps, 20 of them were investigated in the field. 2 candidates, 2 km north of the Daya Bay Guangdonn i P NP g Province, were chosen.
3.2. Hydraulic Fracturing Disposal Process
hydraulie Th c fracturing proces disposaa s si l method which combines treatmend an t disposal of ILLW. Though similar to the cementation process in respect of treatment technology f cemeno s i t i t, solidificatio deen ni p stratum usiny B mature.e gth d fracturing technology and the equipment available in oil industry, the ILLW grout mixed with cement and other additive s injectei s d intundergroune oth d closed stratum with extremelw ylo permeability and solidified with shale to become an integral body. As a result, the radioactive wastes are isolated from the human environment.
Fro hydraulif o mD 1980'R& ce fracturinsth g proces initiateds survee swa th r f yo Fo . candidate sites, twelve wells were drille norte th Sichuaf hn do i n Province investigatioe Th . n shows that the shale is wide hi distribution scope and large hi thickness over 500 m; the conten claf o t y minera highs i l undergroune ;th d wate meterr0 tabl15 s ei s belo surfacee wth ; there is a impermeable stratum below the underground water table; the earthquake intensity of the area is low and the crustal stress in three-dimensional space is advantageous to the hydraulic fracturing injection. Therefore idean a s i l t sitei , .
Two runs of experimental injection at the depth of 433 m were carried out in 1989.
(1) 270 m3 of water containing 198Au 3.09 xlO11 Bq (2) 291 m3 of simulated grout containing 134Cs 3.62 xlO11 Bq
resulte followss a Th e sar :
(1) Breakdown pressure MPa6 2 : ; Corresponding injection rate: 0.13 m3/min; (2) Prolongation pressure MPa0 2 : ; Corresponding injection rate: 1-1.13 m/min3 ; maximue Th ) m(3 angl f groueo t sheetd abous an si ; 25 t (4) The maximum distance of grout sheets is 116 m from the injection well.
The observation wells for covering rock stratum are built around the injection well to observ measurd ean leakage eth e rat nakee watef eo th f dwelo re parth lt a tbotto m under pressurea MP risin e 5 Th 0. . g valu earth'f eo s surfac determines ei d after each injectionn I .
55 addition to the blowout preventer assembly, an emergency waste pool of 150 m3 was constructed to store the radioactive grout coming back to the earth's surface. Now this facility is applying for operation license.
3.3. In-situ Bulk Grouting Process Disposal
Gebe Th i Deser wese th tn i tpar f Chino t sparsela s ai y populated area with fairly ydr climate. In 80's, a comprehensive geological survey and safety analysis as well as environmental impact assessment were carried bee s outha nt . I foun t thadou t ther stabls ei e geological formation undergroune Th . d wate belom r 0 tablw4 s esurfacei . In-situ bulk grouting process as performed in Hanford, USA, is suitable for ILW disposal in this area. The liquid wastes to be disposed of can be divided into two categories, their compositions followss a e ar :
Chemical decladding wastes Concentrates
NaNO3 (g/L) 280 330 Na2CO3 (g/L) 40 50 NaOH (g/L) 80 20 NaAlO2 (g/L) 100 9oSr (GBq/L) 0.056 0.16 137Cs (GBq/L) 1.32 4.25 I06Ru-106Rh (GBq/L) 0.28 0.58 Ea (kBq/L) 26 67.5 slurry (MBq/L) 51.8
The specifications of the grouted wastes are as follows:
Matrix material Portland cement
Grout flowing 0.17m Initial setting time > 2.5 h Ending setting time <48 h Compressive strength > 10 MPa
Obasie n th researcf so formulationn ho piloa , t cold tesdemonstratior tfo carries nwa d out in 1986. The underground concrete vault was 4.24x23.5 m3. The temperature of cemented body 119°o rost p eu C afte day5 r s casting thed droppet an n,i 35°o dt day0 C6 s later.
LLW/ILW is pumped out from the collection and transfer system to the feed system, e groutine mixeth theth f o o nt r g disposa lconcrete systemth l Al e. vault e locatear s d underground, and the mixer sits at the top of the vault. The waste cement paste flows into the vaults by gravity. The vault size is 886 m3. Casting a vault will take about 26 hours. Several days later additionan a , l laye f cleao r ne surface th cementh n f o o e t t pu wil e b l solidified waste. 12 vaults form a unit, several units will be constructed as required. The vault is built by reinforced concrete with a structural cover and surrounding clay to retard the release of radionuclides from the disposal system. The HLW will be disposed in a deep geological formation.
56 . DECONTAMINATIO4 DECOMMISSIONIND NAN G
4.1. Contaminated Metal Recovery
In the nuclear fuel cycle, particularly in the case of accidents and decommissioning of nuclear facilities, a large amount of contaminated metals will be produced. Generally they ca decontaminatee nb reused dan d dependin thein go r clean-up levels orden I . reduco rt e eth radioactive waste volum recoved ean r reusable metals, melt refinin regardee b n a gca s da sound solution.
The China Institute for Radiation Protection (CIRP) has conducted decontamination of uranium-contaminated equipment coming from the diffusion plant. More than 10001 of steel, copper, and nickel were recovered. Their research results are as follows:
Melt refining can effectively remove the uranium contaminants from metal into slag; Residual uranium content ingoe th i tsh depend mainlbasicite fluxe th th n yo f ; yo When it is in the range of 1-1.3, the best decontamination efficiency can be obtained; Temperature of 200-300 °C higher than metal melting point is suitable for melt refining; The shorter the melting time, the better the decontamination efficiency. As soon as the meta meltes i l d completely t shouli , castede db ; Original contaminate deffeco n levemel n s o t ha lt refining results; Uranium distribution in slag is sufficiently homogeneous; Slag looks like ceramicsdirectle b n ca y t I dispose. f aftedo r proper packaging; Residue uranium content in ingot is 1 ppm. Metal recovery is 96%.
4.2. Uranium Mines and Mills Decommissioning
Since 1987, uraniusome th f eo m mine milld san s hav theif eo reached r servicen e dth e life. Mor ef the o tha m0 1 n wil e decommissioneb l forthcominn di g years. e Mosth f o t uranium mine milld locatee san sar densely-populaten di d areas with high moisture, abundant surface water resources and high level of underground water, which are favorable conditions for developing agriculture. Therefore, environmental remediatio urgene th s ni t need.The government and the public are also paying more attention to the environmental problem.
The major task of the decommissioning engineering is to press down the rate of radon release, keep surface water and underground water free from radiation pollution. Research work on the covers of tailings with two kinds of different materials-clay and concrete has been conducted.Though they are effective, clay is more economical and feasible in a large covering area.
57 THE NATIONAL WASTE MANAGEMENT SYSTEM IN EGYPT
S.A. MAREI Hot Laboratory and Waste Management Centre, Atomic Energy Authority
K.A. EL-ADHAM National Centr r Nucleaefo r Safet Radiatiod yan n Control Atomic Energy Authority
Cairo, Egypt
Abstract The waste management syste i Egypmh t comprises operationa d regulatoran l y capabilities. Bot f thesho e activitie performee ar s d unde legislativra e umbrella.The legal framework is well defined by both the Decree No. 288 (1957) which allowed the establishmen Egyptiae th f o t n Atomic Energy Commissio Atomie nth (nos i t cwi Energy Authority (AEA) and the Law 59 (1960) which assigned the full responsibilities for licensing, managemen radioactivf o e us contro d e an tth ef materialo l waste th d e sarisingan s AEA.The toth e operational capabilitie t Laboratorie allocatee Ho ar se th Wasto d dt an s e Management Centre (HLWMC). These capabilities include,besid operatorse eth facilitiee ,th s for treatin conditionind gan g liqui solid dan d radioactive waste liquie Th . d radioactive waste facility has been completed under an IAEA Technical Assistance Project. The facility can levew 3 /dalo ltreaf m liquiyo 0 1 t d radioactiv 3 /dae m mediuf wastyo 2 d ean m level liquid waste. The facility was commissioned in December 1993. It uses three methods for treating liquid radioactive waste: precipitation, evaporatio n exchangeio d an n . Sludged an s concentrates resulting fro treatmene mth conditionee tar cementatioy db cementatioe th n i n plan t e whic facilityth a par f e s solio i ht Th d. radioactive waste treatment includes compaction and incineration. The compactor has been supplied under an IAEA Technical Assistance Project. The building for the compactor has been completed and the compactor has been installed. The compactor is a French model (SON) with a compaction ratio ranging o 1:10t froe compacte5 1: .mTh d waste wile conditioneb l y cementatiob d e th n i n cementation plant e incineratoTh . bees ha rn supplied throug e technicahth l co-operation with Germany.lt was commissioned in 1993, for inactive solid waste and will be operated for one year before being used for low level solid radioactive waste. The regulatory activities are assigned to the National Centre for Nuclear Safety and Radiation Control (NCNSRC).These activities include issuing regulatory documents, reviewing safety analysis reports, issuing licenses, inspection and control of all the safety- related activities. 1. INTRODUCTION
Widespread use of radioisotopes in different applications is resulted in the generation of appreciable amounts of radioactive waste. Radioactive waste may also arise from the processing of raw materials that contain naturally occurring radionuclides.
Radioactive waste needsafele b o st y managed becaus potentialls i t ei y hazardouo st human health and the environment. Safe radioactive waste management requires the application of technology and resources in a regulated manner, in accordance with internationally agreed principles [1] so that the exposure of the public and workers to ionizing radiatio controlles n i environmene th d an d protecteds i t . Basic requirementr sfo such safe managemen providee ar tSafete th n ydi Standard Prédisposan o s i [2], Disposal
59 [3,4], Uraniu Thoriud man m Mining and Milling Waste [5] and Decommissioning of Nuclear Facilities [6].
objective Th radioactive th f eo e waste managemen Egypn ti dea o t s li t with radioactive waste in a manner that protects human health and the environment now and in the future without imposing undue burdens on future generation. A waste management system is established in Egypt for the management of waste in accordance with the objective and principles as set out in the RADWASS Safety Fundamentals [1].
2. THE NATIONAL WASTE MANAGEMENT SYSTEM IN EGYPT
The waste management system in Egypt comprises operational capability for dealing with radioactive waste and the regulatory capability for controlling the way in which it is dealt with. Bot thesf ho e activitie performee sar d unde legislativra e umbrella schematiA . c presentatio Componente th r nfo Nationae th f so l Waste Management Syste mshows i Fign i . 1.
A legal framework is well defined by both the Decree No.288 (1957) which allowed the establishment of the Egyptian Atomic Energy Commission (now it is the Atomic Energy Authority (AEA) and the Law-59(1960) which assigned the full responsibilities for licensing, management and control of the use of radioactive materials and the waste arisings to the AEA. Fig. 2 shows the AEA structure.
2.1. Operational Capabilities
operationae Th l capabilitie Laboratort assignee Ho s ar e th Wast d o dt y an e Management Centre (HLWMC). These capabilities include, beside the operators, the facilities for treating and conditioning liqui solid dan d radioactive liquiwastea therw e deNo ar . radioactive waste facility compactoa , incineraton a d an r r treatinrfo solie gth d radioactive waste.
RADIOACTIVE WASTE UANASEMEKT SYSTEM
OPERATIONAL CAPAaUTY REGULATORY CAPASILTTY
FAOones opsvaons LEGAL REGULATORY FRAMEWORK BODY
FIG. Components1. nationalofthe radioactive waste management system.
60 National Center For Radiation Researcd han Technolog • (NCRRT)
ATOMIC ENERGY AtrrHORrTY(AEA)
National Center For Nuclear Safer» And Radiation Control (SCNSRQ
FIG. Structure2. Atomicthe of Energy Authority (AEA).
2.1.1. Liquid Radioactive Waste Facility
The liquid radioactive waste facility has been constructed under an IAEA Technical Assistance Project [7]facilite 3leve /daTh .w tream n lo 0 y ca lyf 1 tradioactiv o e wastd ean 3 /dam yf mediu2 o m level waste facilite commissiones Th . ywa Decemben di r 1993t I . uses the three methods for treating liquid radioactive waste: precipitation, evaporation and ion exchange.
e schemTh e adopte intermediatd r treatmenan dfo w lo f o te level radioactive liquid wastes was planned according to the IAEA guides, so as to prevent any release from the environmene planth o t conformitn i tt whicno e har y wit ICRe hth P recommendationd san the national regulations [10].
Figure 3 shows the different treatment processes applicable in the Egyptian Radioactive Waste Treatment Plant, as briefly described in the following:
Low level radioactive liquid waste collected from the research laboratories and institutes (a total volume up to 10 m /d with the salt content of 700 g/m and specific activity of IE-6 Ci/L), wil subjectee b l hydroxida o dt e coagulatio3 n proces alloo st w efficient separatio3 f no several contaminating radionuclides and suspended matter. Coagulants and settling sludge (0.12 m /d with salt content of 5 kg/m and specific activity of 4E-4 Ci/L) are collected and the décantâte3 s are passed through sand3 filters to reduce suspended matter within a ratio of 2-3.
The décantâtes are freed from salts and radionuclides by two successive ion exchange demineralizee stepsth d an , d water with radionuclide sald tsan content belo permissible wth e level drainee sar d int normae oth l sewage system liquide Th . s wit hhighea r salt contend tan radionuclide concentrations will be subjected to additional decontamination through ion exchangers. Afte firse th rt decontamination cycle sale th ,t conten demineralizee th n ti d water does not exceed 100 g/m3, with specific activity of IE-8 Ci/L.
61 SUJOCtSJLliCnW
F/G . Schematic3 . flow sheet of liquid radioactive waste treatment conditioning.d an
The intermediate level radioactive liquid waste with an average salt content of 25 kg/m3 and specific activit IE-o t 4p y u Ci/L, also includes residues resulting fro regeneratioe mth n of use exchangersn dio . Processin thif go s typ wastef eo baseds si principlen i , volumn o , e reduction through evaporation to the liquid concentrate with the salt content not exceeding 250-300 kg/m3. Condensâtes with salt conten g/m0 specifid 70 t 3an c activit f IE-yo 7 Ci/L, will be cooled and directed to low level liquid waste treatment flowsheet.
The sludges and the concentrates resulting from the treatment are conditioned by cementatio cementatioe th n ni n plant whicfacilitye parth a f s hi o t . Figur schematia s i e3 c diagram for the flowsheet of the processes in the liquid radioactive waste facility.
2.1.2. Solid waste treatment
Treatment of solid waste includes compaction and incineration. The compactor has been supplied under the IAEA Technical Co-operation Project [8]. A building for the compactor has been completed and the compactor has been installed. The compactor is a French model (SON) wit compactioha n ratio rangin 1:10o gt fro5 .1: m
It is 160 kN bailing press for compaction of low level solid waste into 200 L drums. This pres bees sha n designe compaco dt t radioactive wast draeL insid0 m 20 eithe ea buln ri k containersL 0 10 on boti rn I . h case finasa l compactio obtaines ni successivy db e stepf so waste feeding and pressing. The necessary time for one motion of the piston down and back uppen i r positio approximatels ni minutes.Thy2 e compacted waste wil conditionee b l y db cementation in the cementation plant.
Activw Lo ee WastTh e Incinerator (LAW! s beeha ) n supplied through technical co-operation with German ys commissione [9]wa t I . 1993n di inactiv r fo , e solid wastd ean will be operated for one year before being used for low level solid radioactive waste. The incinerator is the research tool available to the Nuclear Research Centre, Atomic Energy Authority, where scientific investigation f contaminateso d wastes incineratio flud s nan ega
62 cleaning may be conducted. The primary goal of these research activities is to develop an ecological solid waste incinerator process combine controe th o dproblemf t o l s arising from pollutants presen l streams al concentrat o t n ti d an , e pollutant residues from flucleanins ega g to make them suitable for disposal.
The test facility has been designed for a nominal capacity of 15 kg per hour. Typical low active waste to be incinerated is paper, textiles, plastic items and HEPA filters. The incinerator is schematically presented in Fig.4. The main parts are:
Glove box and loading station. Gas reactor, Combustion chamber, Air mixing chamber, Cyclone separator, Filter group which includes: Bag filter, HEAP filter, Valves, fittings, and piping system, feedins Burnega d gran system, Process contro instrumentationd an l , Stack.
To assur gooea d performance wels a ,accurats a l e thermo-hydraulic data collection, LAW equippes i I d with process control instrumentatio safetd nan y systems operatine Th . g switche controd san r varioufo l s electro-mechanical components (valves, burner, exhaust blower and fresh air blower) are located on the control panel, as the entire installation can be operated easily by only 2 employees. Construction allows for very fast start up and shut down periods of 0.5 hour. All process variables (temperature, pressures and flow rates) are recorded, displayed and printed by a special data acquisition system.
Stcdt
FIG. 4. Simplified flow sheet of the incinerator.
63 2.2. Regulatory Activities
Regulatory activities are assigned to the National Centre for Nuclear Safety and Radiation control (NCNSRC). These activities include legislativ regulatord ean y parts.
2.2.1. Legislative part
This includes establishing national regulations and/or adopting other relevant regulations. In this issue, National Regulations concerning the radioactive wastes resulting from the users of radioisotopes were prepared. The IAEA guidelines and the experience of other countries were considered in preparing this document which is available in two versions (Arabic and English). It comprises two parts:
1. Responsibilities of the users of radioisotopes (waste producers) such as:
Segregation and collection of radioactive wastes: The different categories into which radioactive waste classifiee sar d were defined differene Th . t methodr sfo collectin differene gth t categories were described. Controlled releases: This determines the amounts of different radioisotopes which can be released into the sewage. The release conditions were identified. The released amount Minimue basee th sar n do m Annual Limi Intakf o t e (ALIs ^a determine ICRe adopteth d IAEAe Py an db th y db . Interim storage: This include regulatione sth s relevan storage th o t e room design, the storage conditions physicae th , l protectio recore th d dn an keeping system. This first part also illustrate licensine sth g procedure licensind san g conditions.
. 2 Responsibilitie t LaboratorHo e th Wast d f so yan e Management Centre (HLWMC). transportation; treatmen conditioningd an t d an ; shallow ground disposal.
2.2.2. Regulatory part
This part deals with assuring compliance wit requiremente hth regulationse th f so , with particular reference to:
- radiation protection - assessment and approval - emergency planning and preparedness - quality assurance - inspection and enforcement.
In this issue, the IAEA basic radiation protection standard (Safety Series No.9) was adopted safetA . y analysis report (SAR s prepare e liqui)wa th r dfo dradioactiv e waste treatmen d solidificatioan t n plant repore Th . t deals wit site e buildingse hth th , procese th , s description auxiliariee th , radiatiod san n plant repore Th . t deals wit site e buildingse hth th , , the process description auxiliariee th , radiatiod san n protectio monitorind nan g systems with emphasi n safeto s y aspects. Safet hazard an y d analyse e dealar s t a separatwit n i h e chapter.This chapter includes the risk evaluation due to releases resulting from normal operation or accidental conditions. This report was reviewed and assessed by the staff of the
64 NCNSRC and approved by the Chairman of the AEA. The welds and the pipes in the plant were checke inspected experte dan NCNSRCe th th f y o sdb preliminarA s . wa R ySA prepared for the facility. This report was reviewed and assessed by the NCNSRC and a final SAR is nearly terminated. The waste stores in the hospitals, laboratories and other sites are periodically inspected and checked.
The co-operation between different organization is constructive and the assistance of the IAEA is considerable.
REFERENCES
[1] INTERNATIONAL ATOMIC ENERGY AGENCY, The Principles of Radioactive Waste Management, Safety Series No. 111-F, IAEA, Vienna (1995). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Prédisposai Management of Radioactive Waste, Safety Series No. lll-S-2, IAEA,Vienna (in preparation). [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Near Surface Disposal of Radioactive Waste, Safety Serie . lll-S-3sNo , IAEA, Vienn preparation)n a(i . ] INTERNATIONA[4 L ATOMIC ENERGY AGENCY, Geological Disposaf o l Radioactive Waste, Safety Series No. lll-S-4, IAEA, Vienna (in preparation). ] INTERNATIONA[5 L ATOMIC ENERGY AGENCY, Managemen Wastf o t e from Minin Millind gan f Uraniugo Thoriud man m Ores, Safety Serie . lll-S-5sNo , IAEA, Vienna (in preparation). ] INTERNATIONA[6 L ATOMIC ENERGY AGENCY, Decommissionin f Nucleago r Facilities, Safety Serie . lll-S-6sNo , IAEA, Vienn preparation)n a(i . [7] IAEA Technical Assistance Project EGY/9/007. [8] IAEA Technical Assistance Project EGY/9/012. ] KRUNG[9 ; SCHMITZW. , , HJ.; ABDEL-RAZEK THONED AN . L , , I.D., Incineratio Activw n PlanLo r e Wastfo t Inshast ea , June 1993. [10] INTERNATIONAL ATOMIC ENERGY AGENCY, Treatment of Low- and Intermediate-Level Radioactive Wastes, Technical Report Serie . 223sNo , IAEA, Vienna (1993).
65 DEVELOPMENT OF A NATIONAL WASTE MANAGEMENT INFRASTRUCTURE IN GHANA
E.O. DARKO . SCHANDORC , F Radiation Protection Board, Ghana Atomic Energy Commission, Accra, Legon, Ghana
Abstract
Radioisotope applications in medicine, research and industry in Ghana is on the increase. Existing waste management infrastructur inadequats ei copo et e wit probleme hth s of radioactive wastes. With expande nucleaa f o e rd us researc h reactor, gamma irradiation facility, radiotherapy and nuclear medicine, waste management practices are being reorganized wit requisite hth e trained manpower, equipmen supportind tan g facilities. Under a new programme, a radioactive waste management committee has been set up to advise the Ghana Atomic Energy Commission on the establishment of a National Waste Management Infrastructure. Wit expern ha t advice under RAF/9/007, AFRA-I project drafa , t regulation s beeha n submitte r stud d promulgatiofo d an y e Commissionth y e nproposeb th n I . d legislation, a radioactive waste management centre will be established which shall be capable of managing all radioactive wastes in the country. Regulatory control of waste management activities will be the primary responsibility of the Radiation Protection Board (RPB). The waste management infrastructure envisage developee b o dt effectivr dfo e waste management contro discusseds i l .
1. INTRODUCTION
Radioactive materials and other sources of ionizing radiations including X-rays generators have been varieta use r d fo f application yo morr sfo e than three decades. Most of these sources might have outlived their usefulness and left in a state of neglect without appropriate storage and/or disposal.
This stat attributee f affairb eo y numbea ma s o dt f reasonso r :
(a) policy makers are unaware of the presence and dangers of radioactive sources used in countrye th , (b) ignorance of the general public about the presence and inherent dangers associated with radioactive materials, ) (c absenc larga f eo e numbe radiatiof o r n sources nuclead an , r installation facilitiesd san , (d) lac f welko l trained manpowe fiele radioactivf th do i h r e waste management, (e) lack of adequate or basic equipment to handle the wastes generated and (f) allocation of state resources to more pressing and important priorities.
Increasing evidence of the existence of spent radiation sources in medicine, industry and research, coupled wit constructioe hth gamma f no a irradiator, nuclear research reactod ran radiotherapy units, necessitates effective control measures to deal with the problems of wastes. In furtherance to the provisions of the Atomic Energy Act (Act 204) [1], the Radiation Protection Instrument (L.I. 1559) was promulgated in January 1993 [2], to regulate and control the use and management of radioactive materials in all national endeavors. The legislative instrument (L.I. 1559), however provided minimal legal basis for regulatory
67 control of waste management in its totality. A request for an expert mission under RAF/9/007 AFRA-I [3] to advice on the establishment of a Waste Management Infrastructure enabls ha draftine eth Wasta f go e Management Regulatio complimeno nt existine th t g L.I. 1559 on matters of radioactive wastes. The regulation which has been studied by the Radiation Protection Board is receiving attention and subsequently promulgation by the Ghana Atomic Energy Commission (GAEC).
To enforce the regulations, the Radiation Protection Board is first instituting certain action retrievae planth r sfo f spen o l t radiation sources. These include:
(i) announcement in the news media for identification and registration of active and spent radiation sources, (ii) visit premises suspecte possessinf do g radioactive materiald san (iii) send out questionnaires to institutions and organizations to find out whether they possess any radioactive material.
thin I s paper essentiae th , l component establishinn si effectivn ga e waste management regim discussede ear .
2. PRESENT STATUS IN RADIOACTIVE WASTE MANAGEMENT PRACTICES
promulgatioe Prioth o t r Radiatioe th f no n Protection Instrument [2] Nationae ,th l Nuclear Research Institute (NNRIonle th ys Institut)wa e established unde Ghane rth a Atomic Energy Commission (GAEC) responsible e NNRth s Th . wa I e organizatio chargn ni f o e radioactive waste management. Activities in this area were executed by its Health Physics section under the Department of Physics and Reactor Technology. Due to the absence of legal support sectioe th , n operated under very limited jurisdiction, withi framewore nth f ko the Atomic Energy Act [1]. The functions of the section was to advise, and provide services to end-users upon request.
2.1. Waste management committee
Cognisanc evee th rf eo increasin f radioactivo e gus e material othed an s r sourcef o s ionizing radiations in the country, with the future attendant problems of waste envisaged by Commissione th wasta , e management followinp committeu t se appointmene s gth ewa f o t an AFRA- I co-ordinator, to address this issue.The committee consisted of staff from the various Department GAECf so havo ewh , some knowledg radioactivn ei e waste management.
The waste management committee, though not formally inaugurated, meet regularly to advise on safe management of radioactive wastes. The committee requested expert service from the International Atomic Energy Agency (IAEA) under the AFRA-I project. This led to the drafting of the Waste Management Regulations which are now receiving attention.
2.2. Waste management regulation
In pursuance of the powers given in Section 10 of Act 204 and the L.I. 1559, the waste management regulations were drafted with assistanc expern a f eo t fro IAEAe mth brieA . f summar revisee th f yo d regulation provides si followine th n di g paragraphs.
68 2.2.1. General provisions Radioactive Th e Waste Management Regulations requir activity e thaan n i ty that will generatioresule th n ti subsequend nan t managemen radioactivf to e wastes valia , d licensn eo the detailed design, operation and decommissioning of the proposed facility and/or equipment shal obtainee b l d fro Radiatioe mth n Protection Board.
e wastTh e produce s fulli r y responsibl e wastth r e efo generate d includin s finagit l disposal waste Th .e management compliancn shali e b l e wit Nationae hth l Regulationd san any other legal requirements impose Boarde th y f db I waste. Radioactive senth e sar o t t e Waste Managemen e GAEtth Centry Cb p undeeu e NNRIt (RWMCth rse e e th b , o t ) responsibilit transferres yi Centree th o dt .
envisagee Th d organizational Commissioe charth f to n wit positioe hRWMth e th f no C is shown in Fig.l.
2.2.2. Return to supplier
buyee sealea Th f o r d radioactive sourc epurchase shalth n i l e contract, requese th r fo t return of the source after its useful lifetime and not later than 15 years after the purchase, provided the activity of the source does not exceed 100 MBq, 10 years after the purchase.
J
Biotech-ANuelear Agnc, Institue« i NNRI I RPB Director Director j .. --' J
J j
Dcpt of Food Socnc« &. Dcp Nucleaf o t r Food & Envinxunenu! R*d Proceisuig Moiutortng Lab
DejX of Plant Occupational Ridttlopcal — Dep Chemattf co y «cSoilScicocc J Protectiob nLi j
<— SupportSav*« Gunm* Imdtauon Centre
lectronics & iatturoe Onu*
e Wtste Management Centre
1—— Support Servie
FIG. Organizational1. Ghanachartthe of Atomic Energy Commission
69 A contrace copth f yo t wit reture hth n clause shal submittee b l Boarde r somth fo o dt f e I . reasons, the buyer cannot fulfil this obligation, the source may be sent to the RWMC at a prescribed fee.
2.2.5. Handling, treatment, conditioning storaged an of radioactive waste
Waste generated shal collectede b l , characterize segregated dan d on-sit accordancn ei e with the guidance provided in the IAEA Safety Standard on Prédisposai Management of Radioactive Waste from Medicine, Researc Industrd han y [4], Thifacilitato t s si e subsequent management of the waste.
Release of waste into the environment shall be done under authorization from the Board after segregatio decad nan y storag meeo e t requirements e th t .
Wastes to be stored for more than 5-10 years shall be sent to the RWMC in suitable containers e containerTh . s shal properle b l y labelled labele Th . s shal e e legiblb lth r efo whole period of storage and bear the following information:
i) dat f storageeo , ii) name and place of waste producer, iii) radiation symbol, iv) activity and dominating radionuclide(s), v) maximum surface dose rate, ) vi nam f sendereo , vii) category of wastes, viii) name and signature of responsible person and ix) identification number or batch code.
2.2.4. Transportation
Apart from licensing, the RPB is responsible for control of transportation of radioactive wastes in the country. The Board shall administer the transport packaging requirements for waste transportation in accordance with the National Regulations based upon the IAEA Regulation Safe th er Transporsfo Radioactivf o t e Materials, Safety Series No.6 [5].
On-site procedure r transfesfo f wasteo r s shal undee b l r supervisio designatea f no d Radiation Safety Office qualifiea r o r d perso consultation i n wit approved h an Boar e th y db in the licensing process. Appropriate training should be provided to the designated officer with emphasis on accident scenarios.
2.2.5. Disposal
e CommissioTh n shal responsible b l establishmene th r efo operatiod an t neaa f no r surface repository for waste meeting the requirements arising from a safety and performance assessmen repository.The th f to e design, construction, operatio closurd repositore nan th f eo y shal approvee b l Boarde th y db . Waste acceptance requirements shall als developee ob d dan approved by the Board.
Wastes not meeting the requirements for exemption, controlled release or disposal in a near surface repository shall be stored until the waste can be safely disposed of in a deep underground repository.
70 2.2.6. Records
updaten A d inventor f wasteyo storagn si thosd ean eRWMe senth o t C wil kepe b l t waste byth e produce RWMe th d Cran respectively currene Th . t inventor spenf yo t sources which are in RWMC storage is presented in Table I.
Table I: Spent source inventory
Nuclide Type Form Activity Quantity [mCi]
Sr-90 sealed solid 4.0 5 Sr-90 sealed solid 5.0 2 Sr-90 seated solid 2.0 2 Cd-109 sealed solid 3.0 1 lr-192 sealed solid 5.0 1 Fe-59 sealed solid 100 1 t -129 sealed solid 125 1 Co-60 sealed solid 1.0 1 Co-57 sealed solid 1.0 1 Sr-89 sealed solid 129 1 TI-204 sealed solid 1.0 2 Cs-137 sealed solid 1.0 1 H-3 unsealed liquid 10.0 10 Fe-59 sealed solid 200 1 Sr-90 sealed solid 10.0 10 P -32 unsealed liquid 10.0 1 fn-113m unsealed liquid 50,0 12 Cs-137 sealed solid 9.0 5 Tc- 99m unsealed liquid 135 6 Cf-252 sealed solid - 6
A e wastreporth f eo t management activitie e previouth r fo s s fiscal year shale b l submitte Boare th boty o dwastt b e hth e produce RWMCe th e d reporTh an r . t should contain the following information:
(i) Exempted wastes disposed of in the municipal landfill and/or discharged into the sewer system; (ii) Sealed radiation sources returne suppliere th o dt ; (iii) Wastes transferred to and/or received by the RWMC for treatment, conditioning and storage.
2.2.7. Exemption from regulatory control
In addition to licensing, certain radioactive materials may be exempted from regulatory control. These material municipae disposee th b n n i ca sf do l landfill, released inte oth
71 atmosphere or discharged into the sewer system provided they satisfy requirements for the disposal of solid, liquid and gaseous wastes into the environment [3,6]. Records of disposed exempted waste shall be kept by the producer for at least three years. The records shall be opened for inspection by the Board.
2.3. Manpower development
Manpower development is an essential prerequisite in enforcing a waste management regulation. To develop the manpower resources and to strengthen the staff for operational waste management purposes, specific training and experience are necessary to ensure safety in the performance of duties.
Suitable graduate or diploma qualifications from the University is essential with some years of experience.The Board will in future also introduce a system of written and oral examinations to ensure that personnel working in various areas are competent and abreast with current trends hi handling of radioactive materials. Staff will also be encouraged to participate actively in the IAEA regular training programmes to upgrade then" skills. Periodic national training course alse sar o envisione augmeno dt IAEe tth A regiona interregionad an l l programmes.
Educational programmes are not yet available in the Country's Universities.The Board is however seeking collaboration hi this area for such courses to be offered at the postgraduate levels.
2.4. Equipmen othed tan r resources
Under RAF/9/007 project a gamma spectrometer with a sodium iodide detector has been acquired for identification of unlabelled radioactive sources kept m a storage facility. The NNRI has also a decontamination unit, two concrete vaults and a number of concrete wells which were constructed in the early sixties for interim storage of wastes and spent fuel respectively. These facilities were constructe researci anticipatiodh W M 2 ha reactof no r whic s nevehwa r installe unforeseeo t e ddu n circumstances. Wit e constructiohth a f no research reactor in progress and the increase hi radioisotope applications in the country, radioactive waste expectee ar s increaso dt i future.Thereh plane ar erehabilitat o t s e these facilities and convert them to a waste management centre and a near surface disposal site. This will be done after safety and performance assessment to be carried out and approved Boarde byth .
Considerin expectee gth d leve nucleaf lo r applications inevitabls i t i , e that facilitier sfo processin conditionind gan wastef go s will als needede ob firse Th t. step wil acquisitioe b l n of a cementation plant to condition the wastes already hi storage and those expected.
3. CONCLUSION
The continuous expansion of nuclear applications hi Ghana needs adequate and commensurate infrastructure to address the problems of wastes. The commissioning of the research reactor (GHARR-1 latn i ) e 1994 will inevitably contribute significantle th o t y problems of wastes. Introduction of methods for treatment and conditioning of the wastes are necessary for immobilization of the spent ion exchange resins and spent sealed radiation sources.The establishment of a practical guidance incorporating rules and procedures to control and regulate the management of wastes in all aspects of nuclear applications is a
72 necessary prerequisite. The promulgation of the L.I. 1559 and the issuance of the Waste Management Regulation are timely to strengthen the legal basis for effective waste management contro countrye th i h l .
REFERENCES
[1] REPUBLIC OF GHANA, Atomic Energy Commission Act, Ghana (1963). [2] REPUBLIC OF GHANA, Radiation Protection Instrument, Ghana (19930. [3] TSYPLENKOV V., Report of Expert Mission, IAEA, Vienna (1993). [4] INTERNATIONAL ATOMIC ENERGY AGENCY, Pre-disposal Management of Radioactive Waste, IAEA, Vienn preparation)n a(i . [5] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport of Radioactive Materials, Safety Series No. 6, IAEA, Vienna (1990). [6] INTERNATIONAL ATOMIC ENERGY AGENCY, Principles for the Exemption of Radioactive Source Practiced san s from Regulatory Control, Safety Series No.89, IAEA, Vienna (1988).
73 RADIOACTIVE WASTE MANAGEMENT IN KENYA: PRESENTL NEAE TH R D FUTURYAN E
. OTWOMD A National Radiation Protection Laboratory,
S.N. KYALO Radiation Protection Board, Ministr Healthf yo ,
Nairobi
S.A. ONYANGO Port Health Office, Mombasa
Kenya
Abstract The expériences and plans of the Radiation Protection Board (RPB) in thé management of radioactive waste ss establishe (RWwa e Radiatioe describedB th )ar y RP b d e nTh . Protectio t whic environmene enactes nAc th hwa d proteco an dt n tma t fro harmfue mth l effects of ionizing radiation. The secretariat of the RPB basically implements the policies formulated. The interplay between the RPB, RW producers and proposed RW managers is covered. Results of a two year service of a RWM programme is elaborated. Recommendations by experts from the International Atomic Energy Agency (IAEA) have not yet been followed and hence while the RPB regulates importation, generation, ownership and use of radioactive materials, it has not started licensing disposal of RW, although some RW is release e environmenth o dt a incinerationvi t , sewerag municipad an e l landfills. Also unlicensed pits have been use buro dt y uncombustibl . FailureRW havo et operationan ea l organisatio folloo nt w throug plann ho develoo st exotin pa c waste management programme prompted a number of institutions using unsealed sources to build then" own waste management facilities. Experience gaine demonstrates dha d that:
(a) Mor f in-houseo tha% n80 e generated treated e wastb n eca . (b) Reliance on the somewhat uncertain future of external waste management programmes resultes ha hasün di y availed storage facilit untreater yfo conditioned dan d waste. waste Th e managemen) (c t programm pland ean t develope effective dar e working models whicn o h other institution basn sca e theiwastn row e management programmes.
operatinn a righte d responsibilitied Th an san gB bodRP handle o yt th f so e generated RW and implement RWM should be clearly defined and surbodination should be separated. This will enabl operatoe eth l operationa carral o r t t you l functions require complo dt y with the national regulatory requirements and the RPB will exercise full control of RW arising in Kenya and their safe management.
1. RADIOLOGICAL AND NUCLEAR ACTIVITIES
Ther mainle ear y sealed a smalsource o t lt extentsbu , unsealed radionuclides use medicaln di , research and industrial institutions. There is no isotope production, nuclear power or fuel facilities
75 although one learning insitution is striving to obtain a research reactor. The stock of sealed sources comprise about 740 Terabecquerels (TBq) Co-60,222 TBq Cs-137,35 Tbq Am-241/Be, 55 5 GigaBecquerel (GBq) Ra-226 and 74 GBq Th-232. More than 37 TBq Ir-192 is imported annually. Unsealed sources imported annually range from tens to thousands of MegaBecquerel (MBq)ofH-3, C-14, P-32,P-33, S-35, Ca-45, Cr-51, Fe-59, Tc-99m, 1-125, 1-131 and In- Ill. Ther informatios ei n availabl radioactive th n eo e substances imported since 198 word 9an k procesn i s i compilo st enationaa stocchartr n li registe disposeba r kW o e sR (figurf Th ro . dof e 1.) show yearle sth y import radioisotopef so s since 198 1998o t somr 3fo e seale unsealed dan d sources. The y-axis denotes activities of the radioisotopes in becquerels while the x-axis denotes the years.
2. REGULATORY STATU EXPERIENCED SAN S
The Radiation Protection Act regulates the import, manufacturing, possession, handling, export and disposal of any radioactive material including radioactive waste [ 1 ]. Two secondary legislations have been produced One covers standards of radiation protection as contained in the International Commission on Radiological Protection Publication No. 26, while the other covers building and structural requirement alsd soan schedule . Aftelicencef 3] o issuinse d th r feean d f 2 sgan s[ o ICRP 60 of 1990 the present standards as contained in Legal Notice No. 54 have been superseded [4]. Specific regulations on the management of radioactive wastes are being prepared and are schedule releasr dfo e late 1994n ri . They inco-operate managemen , limit conditiond RW san f to s for their exemption from regulatory control. Technical guidelinegeneratory b e W us r R f sfo s o have been prepare afted dan r publication they wil sole b lconcerne o dt d parties. Quantitief so radionuclides which can be treated as exempt are included in the guidelines. A questionnaire which when completed provides the necessary information for inclusion in the national register of RW is in use. Since 1990 it is mandatory to include into the purchasing contracts of sealed source claussa e which assure suppliereture e sth th o nt r oncs it esourc a f o reaches d eha en e dth useful lifetime or it is no longer in use.
The secretariat submits to the RPB, for approval, licences for handling, interim storage and disposal of radioactive wastes as and when applied. Licences for handling and interim storage radioactivf o e waste beine sar g issued annually. Licence disposar sfo l started being approven di 1993 but they have been withheld as negotiations between the proposed central facility for storage continueB RP e th . d Futuran W eR licencf o e application unsealer sfo d sources will have b o et accompanie environmentan a y db l impact assessment repor disposan to l (release) l licenceal f I . s applied for are issued, then more experience will be acquired in waste management especially with the implementation of the conditions fixed in the licences.
Inspections of institutions which produce and handle RW and which use radioactive sources have started. Safet ypractiset conceptye t systematia no n de i sar c manner exampln A .deliberat e th s ei e mixing of liquid waste from H-3 and Cr-51 applications in one container despite the institution having been advise havdo t e separate container shorr sfo t live lond dan g lived radioisotopes. Also mandatore th y requirement tha radiatiol tal n worker monitoree sb concludee b flouteds i n ca s da d from analysi survef so y result Nairobn si i Province nee e [Tabl safetr Th dfo . eyI] analysid san bees assessmentha n M recognisedRW r sfo . Currently three member secretariae th f so e ar t involved in a RWM project assisted by the IAEA. A database of RW generated and related information has been started. This will provide information on wastes and projection of future wastes arising. Most private institutions inform the RPB but the same is not done by most public institutions. Hence the data of what is received by all users is not complete, and the conclusion follows that the RWM in Kenya cannot be stated as safe. With one research institution an
76 ot tlie impact oi" R W released lo tne environment is being uone. The goal u> to compare the quantities released with those given in ICRP publications as secondary limits e.g. annual limits of intake (ALIs) and derived air concentrations (DACs) [5].
No central facilit treatmentr yfo , conditionin interid gan m storag sealef eo d source radioactivd san e waste exists. Since 1989, when an IAEA, Waste Management Advisory Programme, team visited Kenya, the JRPB has been collecting spent sealed sources, mainly from learning and medical institutions. Between 1989 and 1991 several spent sealed sources were conditioned into four 200 litre drums. This exercises wer e drumMTRe e donth th kep e t d e sa ar theiLt an a t r premises. Storage conditions of these drums is also being monitored. Due to a blurred line on who is responsible for interim storage and eventual disposal, the drums are kept in the open and are subjec ravagel al weathe e o t th f so r i.e. raisunshined nan unpublishen .A d thesis concludes (after using film badge from Januar Aprio yt l 1994 dos)a e rat 0.2f eo 5 mSv/ 0.0d han 5 mSv/e th t ha surfac metr1 d eean distance, respectivel drume recommendee on r Th . fo ] y[6 d metho interif do m storage has not been implemented. Scientists from both the RPB and MTRL recommend separatio bodieo tw dutief e implementinnse o th wit th lattey e e sb h b th o rt g organizatio chargn i e of a central storage facility for treatment and conditioning RW.
Lack of certain equipments and also limited experience and training of the secretariat has resulted in postponement of certain duties which oft to be done. It has been recognised that the absence implementinn a unresolvee f th o d an dM gproblem bodRW r yfo s involving spent sealed dan unsealed sources which accumulate are the consequence of the weakness of both the regulatory and legal basis and infrastructure. Experts from IAEA have recommended that MTRL be appointe agencn a s dwastr a yfo e treatment, conditionin interid gan m storage since they have eth relevant experienc equipmentsd an e . Fro longeo mn s Juli r B yinvolve 199RP e e th 3th n di collection of spent sealed sources and other RW. The RPB only keeps track of the radioactive material producedW R d san .
3. WASTE MANAGEMENT PRACTICES
The waste management practices vary from one place to another. Low and intermediate level waste handling strategies include treated/untreated, packed and stored and some incineration. Below are listed some of the WM practices currently employed in some institutions.
In hospital solie sth d waste storee disposese ar decar ar dfo inactivs d a yan f do e waste after monitoring. The excretions of patients (1-131 from thyroid cancer diagnosis) are discharged to the sewage system numbe.A researcf ro h institutions atemp meeo tt t specific conditions stipulaten di the licensees institutione Th . s submit monthly or/and quarterly report radioisotopef so s imported, quantities of RW generated and the methods used to dispose them. In some institutions the liquid wastes are kept in storage awaiting the RPB's decision. Two of these institutions receive their radioisotopes from international organisations, which are not subject to normal customs checks. Henc secretariae eth t only discovered later, afte wastese r th the d yha , that the radioisotopee yus s generatd an e wastes. Effort rectifo st y this situatio ongoinge nar . IAE Ahelpins i requiriny gb g tha l institutiontal assistt si beforB s notifRP ee beginninyth g work with radionuclides.
researce On h institution when moving fropremisee mon anotheo st r left theiB rRP wastee th r sfo to decide on the action to take. The principal researchers were not even Kenyans and they used receivo t e their radioisotopes through their embassy researcherse Th . , while shiftind g ol fro e mth premises, abandoned two 1001 drums which-when the secretariat confronted them, they claimed contained tritiated liqui keepins di momenwastee B drumth e opet e RP g th . A e th ntth n s i skies. IAE witsupplies B A hha liqui a RP e ddth scintillation counter, unde Technicara l Co-operation
77 R WM Project activite Th . drumthif o yo stw s gave 426,03 287,74d 2an 4 count minutr spe e (cpm) in August, 1994 and hence plans to release this waste to the sewerage system have been pu abbeyancn ti e (backgroun NRPe cpm)2 th 3 t dLa s i .
4, WASTE DISPOSAL
Waste disposal practices vary from plac researc e placeo et on t A . h institution, using unsealed sources liquie ,th d effluent pass through three consecutive lagoons wit htotaa l residence timf eo abou thee month3 t ar n d releasesan d int osmala l researco riverTw . h institutions have incinerators which the foe ryus burnin g both inactiv decayed ean mixes i dh wastesdas wite Th h. incombustible solid waste disposed san clan i yf d o pits within their compounds bees ha nt I . recognised that environmental impact assessment and monitoring of the radionuclide content in groundwatee th r aroun disposae dth l pits need performede b o st insitutioe disposas On .it s nha l pit lined with impermeable concrete and the initial pits (now full) have concrete covers on top. Another institution cla g useydi do pitt s (not linedhare th d o )bedroct bottoe th t ka m (dept- 7 f ho 12 metres). This, it has been found, is contrary to one of the IAEA requirements for RW repositories (that the separatee yb d from fractured bedrock) [7].
. 5 CONCLUSION
The volume of RW will grow in the years ahead. The time it will take to prepare, implement, and demonstrate safe, permanent solution disposaW R o st l reflect politicale th s , economid can environmental importance to public health and safety of the task at hand and its technical challenges [8]. However, it is not the technical challenges and issues that will stand in the way of progress. It is the institutional and socio-political issues that need to be resolved for progress to continue. National programmes are being re-defined and practices developed to eliminate negative environmental impact LLf so W disposed futurn .I generatore eth wil requireW e lb R f so launcdo t h and implement a RWM programme under the oversight of the R PB. Important preconditions suc s keepinha f recordsgo , monitorin safet B presentind an RP gy e assessmenth o gt d an t environmental impact assessment reports will be emphasised. IAEA has given and continues to advic providd ean e technical assistanc develoo et expertise p th managemene th r efo wastesf to . For reasons of safety, surveyability, security, monitoring and economics it would be desirable to hav facilitiel eal s (surface storage, deep geological disposal, incineration locatioe ,on etct a ) n [9]. problemn solvo ow t r s u eweaou f so o t p ku legaIs i t l basi lacd f infrastructursan ko safelo et y handle RW..
REFERENCES
1. [ACT 1982] Radiation Protection f 1982o Act 0 ,)2 198o Chapte(N 2 Law3 24 rf so Kenya. Government Printers.
. 2 [REGULATIONS 1986] Radiation Protection (Standards) Regulations. Legal Notice No. 54 of 1986. Laws of Kenya. Government Printers,
. 3 [REGULATIONS 1986] Radiation Protection (Building Standards) Regulations. Legal f 1986o Notic5 . 5 . Lawe NO Kenyaf so . Governemnt Printers.
78 4. [ICRP 1991] 1990 Recommendations of the International Commission on Radiological Protection. ICRP Publicatio . Annal ICRPne 60 th f so . Pergamon Press, Oxfordw ne , York, Frankfurt, Seoul, Sydney, Tokyo.
5. [ICRP 1981] International Commission on Radiological Protection, Limits for intake of radionuclide workersy sb , ICRPAnale th f so , 6,2/3 (1981).
. 6 [Otwoma 1994] Monitorin environmentf go s modifie man-mady db e radiation, MSc. thesis, unpublished data.
. 7 [IAEA 1984] International Atomic Energy Agency, Site investigations, design, construction, operation, shutdown and surveillance of repositories for low- and intermediate-level radioactive wast rocn ei k cavities, Safet y, IAEA serie62 . ,s No Vienn a (1984).
8. [IAEA 1988] Radioactive Waste. 1989 International IAEA Bulletin Vol. 31, No. 4 1989 Vienna, Austria.
9. [IAEA 1990] IAEA-TECDOC-562, Low level radioactive waste disposal: An evaluation of reports comparing ocean and land based disposal options, IAEA, Vienna, 1990.
79 GUATEMALAE TH N PROGRAMM RADIOACTIVF EO E WASTE MANAGEMENT
S.R.R. JIMÉNEZ, P.G. ORDÔNEZ Nuclear Energy General Directorate, Guatemala City, Guatemala
Abstract Guatemala aims at ensuring safety of present and future generations as well as the environment, this is to be achieved by preventing the release of radioactive substances containe n radioactivi d e wastes inte environmentth o e mai.Th n activities that produce radioactive wastes in Guatemala are medical practices (radiodiagnostic and radiotheraphy), wastes are also generated in industry and research, but to lesser extent. The most frequently used radioisotopes are cesium-137, cobalt-60, iodine-131, technetium-99m.Some spent source radium-226e ar s , cobalt-6 contaminated 0an d material generate medicinn di d ean research.
radioactive Th e wastes generate basicalle intermediatd dar an w ylo e level wastese Th . collectio wastee th f ndons o si e periodically usere th , s must deliver them correctly packed and marked. When the radioactive wastes are short lived the user must manage them himself, as in the case of technetium-99m. We collect only Chromatographie columns. When the decay period is longer, the Nuclear Energy General Directorate (DGEN), as the regulatory authority chargn i s i ,f supervisineo controllind gan managemene gth f wasteo t s considerin l radiologicagal l protection principles.
Presently, Guatemal tryins ai achievo gt meany eb f Nationaso l Centr f Radioactiveo e Wastes (CENDRA) the adequate practices in managing, storing and subsequent disposal of radioactive wastes. Three facilities for storage and final disposal of radioactive wastes are planned in Guatemala. The CENDRA installations located in Guatemala City, will consist of: storageA . / levelw facilitylo wastes:r fo
With an area of 28 square meters with its corresponding security systems.
2. A storage facility for high level wastes:
With an area of 11.2 square meters.
3. An area for inmobilization and storage of spent sources:
This will hav aren ea f 85.1 ao 2 square meters.
e desigTh n diagram e facilitieth f o st Nationaa s l Centr f Radioactivo e e Wastef o s Guatemal includede aar .
. INTRODUCTIO1 N
n "ControlNucleae Applicatiod o Th an w e La r Us , f Radioisotopeno Ionizind an s g Radiation" (Decree-Law 11-86) made the provision that the Nuclear Energy General Directorate wil responsible lb e bod licensinr yfo operatione gth s with radioisotope othed san r
81 source f ionizinso g radiation wels a ,controllins a l handlinge gth , transport, treatmend an t final disposa radioactivf lo e waste. This Law, since 198 creates countrr 6ha ou n dlegai e yth l Governmene basith r sfo privatd an t e institution operato st controd ean l nuclear activities.
With the Governmental Accordance 989-92 put in force in December 1992 the Regulatio Licenseer nfo fiele Radioisotopef th do n si Ionizind san g Radiation rulee th , s were established for issuing the licenses.
At present, collectio radioactivf no e waste t i lackcarrie s complesi n d i t san dy ou xwa of appropriate resources and infrastructure. Guatemala is looking for the assurance of protectio f presenno futurd an t e generations environmene th s wel a ,s a l e adequatth y b t e managemen radioactivf o t e wastes. Owin RAPAe th o gWAMAt d Tan P mission have w se decided to create in the Nuclear Energy General Directorate a National Centre of Radioactive Wastes (CENDRA), which will have three primary facilitie r storagfo s f radioactivo e e sources of low and medium activity and for immobilization and storage of sources of intermediate life.
secone Th d phasCENDRe th f eo A project coul finae db l disposa radioactivf lo e sources in an installation specially designed for this function to be located outside the Nuclear Energy General Directorate building.
Fro othee mth r hand wile w ,l continue locate radioactive source properld san y manage them, makin througt gi hprogramma e whic incorporates hi s ha dt i int d Nucleae oan th w rLa the Regulation for each of this activities.
In Guatemala, the waste management begins simultaneously with the design of radioactive installations. Collection, treatment, storage and/or discharge into the environment is considered as a complex work, because this has different disciplines that take the engineering aspects and the evaluation of the radiological impact of radioactive discharge to the environment.
Guatemala is developing a project on waste management with limited resources, a principal e sourcinfrastructurth s i e e that formerly belonge e fueth l o t dGul f plant. Considering the real situation it would very important to provide the adequate utilization of the resources available.
. SITE2 S
The site is a physical space, meteorogically and geologically stable, in which authorized installations for treatment or storage of waste are located.Our installation is located in the respectivs it 1s 2 ha distric cite d eth yan f securito t y measures.
The evaluation of the installations and layout of the site considered several factors, such as:
naturae Th l environment: weather, hydrology, radiation exposure fro backgrounde mth , flora and fauna. The geological stability of the site and its resistance to the climatic variations.
82 Socio-economi technicad can l considerations: Guatemal realizins ai g this project with limited economic resources; the infrastructure has been object of several remodelling requiree th adjuso t o t t di t needs exampler fo , :
Activity of the wastes; Treatment methods to be applied; Types of waste storage.
These facilities are considered as controlled zones, all accesses to the zones are clearly identified.The radioactive waste, in accordance with its activity, is classified as low and intermediate level othed an , r following parameters shoul takee accounto db t n ni :
Type of radiation Half-life Dose rate Physical state Radiotoxicity
. COLLECTIO3 N
Collection of wastes consists in transfer them from the place where those were generated to the place where it will be treated and/or stored. Collection is made periodically and transportation is carried out in accordance with the actual Regulation. Guatemala is implementin police gth returf source ycountro e th origith s f it no o et f yo n afte usages it r .
The waste management must comply with the Regulations. It is the responsibility of the Nuclear Energy General Directorate competens a , t Authority controo t , l throug differens hit t Departments fulfillmene th , Lawe Regulations.th d f san o t .
4. FACILITIES
There are the following facilities at the CENDRA:
THE FACILITY FOR STORAGE OF LOW LEVEL WASTES:
An area of 28 m has been assigned for the storage of short-lived wastes containing such radionuclides as: technetium-99m2 , phosphorus-32, iodine-131, etc. This facility includes the area for personnel monitoring, shower facilities to be needed in case of contamination and security system (see Fig. .1)
THE STORAGE FACILITY FOR INTERMEDIATE LEVEL WASTES:
An beeares f 11.ha ao n 2 2assignem r storagdfo f speneo t sources. This facilits yha adopte desige dth undergrounf no d storage concep thir fo ts kin wastef do , (see Fig. .2)
THE FACILITY FOR IMMOBILIZATION AND STORAGE OF SPENT SOURCES:
In the area of 85.12 m2 the sources are immobilized by packaging them in barrels (see Fig. 3).
83 5. SPECIFICATIO CONSTRUCTIOF NO N MATERIALS
Storage of radiation sources
Technical specifications:
Reinforcement Steel Structures concretr Fo : e foundation, column ceilingd san , reinforcement stee degref lo e 40 and diameter 1/4", 3/8" and 1/2" was used. Walls or concrete blocks utilized mesh electrowelded with rib higf so h resistance steel, type 6x6-9/9.
Concrete foundationr Fo utilizes wa t si dmixe a portlanf ro d cement typ wite1 h added river sand and gravel of 1/2" and/or 3/8", with a compression resistance of 3,000 pounds (210 kg/cm2).
Walls a) Concrete blocks wit followine hth g dimensions 19.x 5 539.x :9. 14. , 5 519.cm x 5 x 39.5 cm. and 19.5 x 19.5 x 39 cm. with resistance of 50 kg/cm2. b) Concrete with resistance of 3,000 psi (210 kg/cm2)
Plastering Wall ceilingd san s used whitewashin havd g an finishine a g cove f varniso r d han epoxy components o smeltw f to , formulated with epoxy resin 100%, wit agenn ha curedf to , pigments, solvent additived san f higso h quality. Floor a) Marble-granite bricks were used with dimensions of 30 x 30 cm of a thickness no less of 2.5 cm, manufactured in press totally automatic, oleodynamic system and vibro-press, submitted in a minimal pressure of 150 kg/cm2. f thicknesso m m 2 .d an m ) c b Viny0 3 x l 0 floo3 f o r
Installation: a) Potable water, all of the tubes and accessories are of PVC and installed underground. ) b Laboratory tubewatere accessoried th l san al , made sar f concreteo e with union case on the direction changes, which are manufactured of baked bricks. All the tube undergrounde sar .
6. CONCLUSIONS
e facilitieTh s develope managemene th r dfo finad an t l disposa f radioactivo l e wastes in CENDRA were realized studying the radioactive wastes generated in our country and considering the growth of wastes, which is expected by the end of the century due to the quantity and quality of generated wastes which are of short-lived radionuclides. The facilities with compactors and incinerators not are included. Guatemala is not planning the installation of researc powed han r nuclear reactors neithe thin ri beginin e s th centur nexe n i th r tf go y no one. Guatemala is looking for an adequate area to built a facility for immobilization and storage of spent sources.
Guatemal interestes ai co-operation di n wit countriee hth Centrae th f so l Americd aan the Caribbean Region having the same level of nuclear applications.
84 m
.?o_i ao•__!. s o o, __ [ -i. oo
. F/GStorage1 . levelw facilitylo r radioactivefo waste
85 \
<&
F/G. 2. Storage facility for intermediate level radioactive waste
86 L >o a § o - o v o : o o (N ^ i in F er
j l / J
=TVU
F/G. 3. Facility for immobilization and storage of spent radiation sources
87 NATIONAL PROGRAMME, LEGAL FRAMEWORK EXPERIENCD AN E WIT MANAGEMENE HTH F TO RADIOACTIVE WASTE IN THE SLOVAK REPUBLIC
L. KONECNY Nuclear Regulatory Authorit Slovae th f yo k Republic, Trnava, Slovakia
Abstract The syste radioactivmf o e waste managemen Slovae th n ti k Republi describeds ci . This syste formee mth followy b r t Czechoslovapolice ou sth t yse k Atomic Energy Commission. Radioactive waste produce nuclean di r power plant plane stored th an stn do site s wile b l solidifie packed dan fibre-concretn di e containers thin i sd foran , mt wili dispose e b l n i f do neaa r surface repository. Such near surface radioactive waste repository located close th eo t Mochovc phase th e commissioningf n esito i s ei e radioactivTh . e waste that wilt complno l y wit acceptance hth e criteri r neaafo r surface disposal, wil plane storee b th l tn do site d san wil e a disposegeologicab l n i f o d l repositor e constructedb o t y e amountTh . d an s characteristics of the most important types of radioactive waste generated in nuclear power plants n industryi s wela ,s a l , medicin d researcan e e given.Thar h e technologier fo s processing this radioactive wast outlinede eseconar e th n I paperde . parth activitiee f o tth , s Nucleae oth f r Regulatory Authorit Slovae th f yo k Republi ) whiccentrae th cSR s h(UJi Dl government authorit e are f th nucleao a n i y r regulations e presente ar e fiel, th f o dr fo d radioactive waste management.The fundamental legal documents dealing with radioactive waste managemen describee ar t d briefly.
1. INTRODUCTION
The development of nuclear power in the Slovak Republic has proceeded together with Czece th h Republic withi Czece n th Slova d han k Federal Republic (CSFR) till 1993 when the CSFR separated into two independent states. Czechoslovakia was a country with insufficient natural energy resources, however, with a developed industry and scientific and technica developmenle basisth d an , t programm nucleae th f eo r powe thus wa rs established already in sixties.This programme focused on the construction of units designed in the former Soviet Union.
l BohunicA- e Ith n 1972e ,plan commissiones twa d (HWGCR MWe)0 . 15 , This reactor type was chosen mainly because it did not require enriched fuel for operation. The plant was operation i n till 197 7s shutdow whewa t ni n ultimately.Currentl e phasth f n o ei s i t i y decommissioning.
In late seventies and in early eighties, four PWR nuclear units with the electric capacity of 440 MWe each were commissioned gradually. Two of these units at the V-l plant are of earle th y V-23 newee th 0 f designplano r2 V-21e V- 3 ar t e , th anothet designa o e tw r .Th operation of these units is expected to continue till 2000 (V-l) or 2015 (V-2). At the Mochovce site, anothe runit R fouundee ar sPW r r construction wit capacite 0 hth 44 f yo MWe each. The commissioning of the Mochovce Unit 1 is expected in 1997.
89 2. NATIONAL STRATEGY AND EXPERIENCE IN THE FIELD OF RADIOACTIVE WASTE MANAGEMENT
Strategy in the area of radioactive waste management established by the former Czechoslovak Atomic Energy Commissio approved n an governmen e th y db Czece th d f hto an Slovak Federal Republi s i followece Slova th y kb d Republic.This strategy requires conditioning of radioactive waste into a form suitable for disposal in a near surface repository and constructio f thino s repository. The following radioactive waste have to be considered for disposal in the Slovak Republic at present and in the near future:
a) operational waste from nuclea i Jaslovskrh powe2 V- d re plantBohunican l sV- e b) the waste from the nuclear power plant A-l hi Jaslovske Bohunice which is under decommissioning c) the waste from the Mochovce nuclear power plant d) radioactive waste from medicine, industry and research.
2.1. Radioactive waste volume generat2 V- d followine ean th Bohunic e l Th V- a g)P volumeeNP radioactivf so e waste (as to December 31, 1993) liquid waste (active concentrate) 6603 0m sorbents 300 m3 solid radioactive waste 3003 0m
The radioactive waste production has been stabilized. The radioactive waste is stored in storage planttanke th t sa s
radioactive Th ) b e waste wit higha h activit bees yha n producel A- a e resul s th da f o t operation, plant accident and subsequent activities. Amount of this waste is shown in the Table below:
Radioactive waste type Amount Total activity [m3] [Bq] Chrompik (water solution 20 10x 2 142. of potassium bichromate) 18 10x 6 133.
Dowtherm 50 2.5 x 1012 Storage pond water 500 10x 5 143. Biological shielding water 360 7.9 x 109 Evaporated concentrate 290 10x 1 11. Sludges 150 4.7 x 1012 Metal scrap 1 400 t 2.5 x 10U Soft solid RAW 850 4.0 x 108 Contaminated soil 1400 2.0 x 1010
90 totae Th l amoun radioactivf o t e wast l planepresent a A- fro s e i t mth t approximately f liqui o f soli 300d 3 o d3 m an d 0 m 0 waste150 , however, their total activity exceede th y sb factor of more then 10 the total activity of all radioactive waste at the V-1 and V-2 plants. l plant'A- e s Th radioactive wast stores ei varioun di s storage systems which representsa relatively high risk, because of the possible negative impact on the environment.
commissionine Th ) c Mochovce th f go e Uniexpectes i 1 t i 1997 dh productioe Th . f no radioactiv plants2 eV- wastd . an consideres ei 1 V- same e th th t er da ratfo s ea
amoune Th dactivitd ) tan radioactivf yo e waste from medicine, industr researcd yan e har given in the following table:
Type Amounts Total activity [Bq] Sealed sources 1945 pieces (Co-60, 1.2 x 1015 Cs-137) Others 1500 1 liquids 4.0 x 1010 3000 kg solid 2.5 x 1010
2.2. Radioactive waste composition
compositioe Th mose th f tn o importan t type f radioactivso e2 wastV- d ean fro1 mV- plants wels a ,fro s l a planl mA- shows followini te th n ni g tables:
91 Radioactive waste compositio nV-1- 2 V- ,
Active concentrates from V-12 V- ,
Nuclide Activity [Bq/1] V-1 V-2
Cs-136 7 10 x 9 3. 5 10 x 9 1. Cs-134 9.5 x 105 2.1 x 104 Ag-1104 m10 x 5 4. xlO7 1. 3 Co-583 10 x 2 6. 1.4xl03 Co-60 1.3 xlO5 4.1 x 104 Mn-54 xlO8 1. 4 3.4 x 104 Sr-903 10 x 2 2. 92.3 C-14 1.8 x 104 4 10 x 2 3. 1-129 27.3 123.8 Tc-99 103.0 <24.0 H-3 SxlO5 SxlO5 Pu-239, 8.35 0.642 Pu-240
Sorbents from V-12 V- ,
Nuclide Activity [Bq/1] V-1 V-2 Cs-136 7 10 x 1 7. 1.0 x 105 Cs-135 4 10 x 9 2. l.OxlO3 Co-60 Ix 103 3 10 x 0 1. Sr-90 2xl03 100 C-14 2xl04 3xl04 1-129 30 100 Tc-99 100 0 H-3 3 xlO6 2xl06 Pu-239, 10 1 Pu-240
92 Radioactive waste composition - A-1
Active concentrate
Nuclide Activity [Bq/1] Cs-137 3.8 x 107 Co-60 6 10 x 2 3. Sr-90 5 10 x 4 3. 1-129 5.3 x 102 C-14 1.7 xlO3 H-3 8.0 x 106 Pu-239, 3.5 Pu-240 SUM alpha 60
Chrompik from can lonn si g term storage
Nuclide Activity [Bq/1] Cs-137 1.4 x 1010 Sr-90 4.3 x 106 1-129 2 10 x 5 6. C-14 5 10 x 8 2. H-3 5.0 x 107 Pu-239, 3 10 x 6 2. Pu-240
93 Chrompik from short term storage
Nuclide Activity [Bq/1] Cs-137 1.4 x 109 Sr-90 2.0 x 105 1-129 6.0 x 102 C-14 6.4 x 104 Tc-99 xlO2 1. 4 H-3 1.5x 107 Pu-239, 1.4 xlO3 Pu-240
Water from long term storage pond
Nuclide Activity [Bq/1] Cs-137 2.1xl08 Sr-90 3.1xl04 1-129 5.3X102 C-14 1.7 x 104 Tc-99 1.2x 104 H-3 4.6 x 106 Pu-239, 49 Pu-240 Am-241 30 alfM a SU 110
2.3. Conditionin f radioactivgo e waste into form suitabl disposar efo l For the solidification (i.e. bituminization) of evaporator concentrates, a rotary film evaporato bees rha n develope full-scala d dan e facilit currentls yi y being commissioned. Solid waste compactee sar d within drumsexchangn plaio e r nTh .fo e resin solidifo t s si y them into a cement matrix.
A vitrification process has been developed for the solidification of the most active liquid wastes of the A-l NPP. That facility is currently in a non-active test phase.
94 Waste conditioning centre was constructed in Bohunice and is currently planned in Mochovce. Cementation line, incinerator and supercompactor are planned in Bohunice. A supplie technologe th f o Germae r th s yi n company NUKEM cementatioA . na lind ean sorting box will be installed hi Mochovce. A supplier of the technology is the French company SON.
Radioactive waste from medicine, industr researcd yan h wil conditionee b l produco dt e fora m acceptabl r finaefo l disposa usiny b l g standard technologies adapte r radioactivdfo e waste from nuclear power plants.
A repository for low and medium acitivity radioactive wastes is close to commissioning. locatefros r i Mochovc e fa t mI t th dno e NPP. This near-surface type repository will provide ultimat mediud e disposaan w mlo f levelo l wastes generated fro plane mth t operation wels a , l as waste l decommissionins A- fro e mth frod gman nuclear applications wastee e th b l o st Al . disposeMochovce th n i f do e Radioactive Waste Disposal Facilit plannee y ar package e b o dt d i cubih c fibre-concrete container backfilled san d with concrete fabricatioe planth A .r fo t n of these container undes si r construction (unde licence Frence rth th f eo h company Sogefibre) at the Mochovce site.
Radioactive waste which doe complt sno y wit criterie hth disposar afo neaa i lh r surface disposal facility wil aftee lb r solidification temporary store subsequentld dan y dispose intf do o a deep underground repository.
. LEGA3 L WASTE MANAGEMENT FRAMEWORK
fundamentae Th l objectiv radioactivf eo e waste managemen deao t s li t with radioactive wast mannea i eh r that protects huma future th n environmene i healteh th d d han an w no t without imposing undue burden futurn so e generations meeo T . t this goal, each nation using nuclear energy shall develop a structure of legislation for this field and establish an independent authority which assures the compliance of nuclear facilities with state regulations nuclean o r safety.
Following the separation of CSFR into two independent states, the Slovak Nuclear Regulatory Authorit ycharges (UJwa ) Dd SR with thi scentra e taskth s sa l authorite th f yo state executiv e e Slova s thui bodth i sR h yS kresponsibl RepublicD e UJ th e r fo eTh . regulation of all nuclear facilities including radioactive waste management, spent fuel managemen wels a t fissios a l n materials.
The other tasks of the regulatory body are to review the peaceful use of nuclear energy an ensuro dt participatioe eth f Slovakino i nucleaah r safeguards regime. Among thise th , regulation of radioactive waste management is very important task. The recently established Nuclear Regulatory Authority deals wit followine hth g basic problem i ordesh : rto
ensure compatibility of all utility's approach to radioactive waste treatment, conditionin d disposaan g l with approved general radioactive waste management conception; facilitate the minimization of radioactive waste production; ensure that condition providee sar r long-terdfo m storag subsequend ean t disposaf o l spent fuel;
95 secure controlled disposal of radioactive waste meeting acceptance criteria;
support and direct R&D works on radioactive waste management and decommissioning participate activel developmene th n yi regulationf o t fiele radioactivf th do n si e waste management and decommissioning. With regarpresene th o dt t situatio radioactivn i e waste management objectivee th , f so regulation are as follows:
storage of liquid radioactive waste storage of solid radioactive waste technology of treatment and conditioning of radioactive waste transport of radioactive waste disposal of radioactive waste releases into atmosphere and hydrosphere radioactive waste from non-nuclear facilities.
The activities of the UJD SR are focused mainly on review of concepts, on compliance with regulations qualitn o d yan , assuranc radioactivn i e e waste management.The basic documen radioactivn i t e waste managemen Decree th . 67/198s i t CSKAEe No th f 7o t I . specifies basic technica organizationad an l l requirement eliminatioe th r sfo f releaseno f so radioactive materials into the environment. It also specifies mandatory procedures for authorities, organizations and their staff which design, construct, commission, operate or decommission radioactive waste management facilities.
The regulation specifies further the basic safety requirements for radioactive waste management:
collection, sorting and storage of radioactive waste treatment and conditioning of radioactive waste final disposa f radioactivo l e waste.
The regulation finally specifies the requirements for safety documentation presented with application licencesr sfo sitingr fo : , construction, commissionin operationd gan . Spent t considereno fue s i l radioactivs da e wast i thieh s regard.
Another regulatio r radioactivnfo e waste managemen e Decree Healtth th s f i to eh Ministry No. 65/1972 on the protection of public health against ionizing radiation. Conditions for the release of air-borne and liquid radioactive waste into the environment are specified in the regulation.
. CONCLUSION4 S
The policy in the area of radioactive waste management established by the former Czechoslovak Atomic Energy Commissio approved nan governmene th y db formee th f o tr Czech and Slovak Federal Republic has been followed by the Slovak Republic.
This policy determines:
Radioactive waste produce plann di t operatio stored nan d provisionall plann yo t sites will be processed into a form suitable for disposal in near surface repositories.
96 Radioactive waste generated from the decommissioning of the A-l Bohunice plant stored on the plant site will be solidified. The solidified radioactive waste which comply wit acceptance hth e criteri near fo ar surface disposal wil storee b lnea a n di r surface repository. The solidified radioactive waste which does not comply with the acceptance criteria will be stored on plant sites and later disposed of hi a geological repository.
Radioactive waste originated in industry, medicine and research will be solidified in processing facilities in nuclear power plants. The solidified radioactive waste will be dispose f eithedo nean i r r surface repositor geologicaa n i r yo l one, accordine th o gt acceptance criteria.
Cementation, bituminization and vitrification technologies will be used for the processing of liquid radioactive waste into a form suitable for disposal or long term storag incinerationd ean , supercompactio cementatiod nan n technologies wil usee b l d processine foth r f soligo d radioactive waste int ofora m suitabl disposalr efo .
For disposal of low level and medium level radioactive waste, the near surface repositor s beinyi g built. This repositorphase th f commissioninn o ei s i y s i d an g located close to the Mochovce site.
r radioactivFo e waste which wilt complno l y wit acceptance hth e criteri r neaafo r surface repository, another typ f repositoreo y wil providee b l d whic geologicaa s hi l repository. Activities for finding a suitable site for such repository started currently.
fiele legislationf th do initiatn R I S D eactivitiee froUJ th , same e mth th f eso coded san other legal document activitiee th formee s sa th f so r CSKAE preparatioe Th . legaw ne lf no documents in this field is going on nowadays. Principally it is a law on the peaceful use of nuclear energy developmene Th . f suco t bilha f expectes i o l d completee b en o de t th y db 1995. Among other legal documents relate fiele f radioactivth do o dt e waste management, the law on a state fund for the decommissioning of nuclear power facilities, spent nuclear fuel and radioactive waste was approved by the Parliament, and a regulation on exemption metal materials from radiation contro beins i l g prepared.
97 EXISTINE TB G SITUATION WIT RADIOACTIVE HTH E WASTE MANAGEMENT IN SYRIA
S. TAKRITI Radiation and Radioprotection Department, Atomic Energy Commission, Damascus, Syrian Arab Republic
Abstract
existine Th g radioactive waste management infrastructur presentes i e d includine gth legal framework, responsibilitie regulatore th f so y bod wastd yan e management practices. amoune Th radioactivf to e wast expectes ei increaso dt e dramatically whe nucleana r research centre with a research reactor and radioisotope production are established.
I. INTRODUCTION
Safe management of radioactive waste in Syria is presently based on the Atomic Energy Commission's Law No. 12 dated 5 April 1976, which established the Syrian Atomic Energy Commission, SAEC. The SAEC has the full responsibility for all nuclear energy matters, it has to set up the procedures required for protection of personnel and the members of public from radiation exposure, suggest legislation, control its implementation and issue subsequent instructions, guides and safety standards.
According to the Law No. 12, the SAEC has both controlling and promoting functions for atomic energ regulatore y th matters ha t Syriai n yi s d bod r radiatioan , yfo n protection and waste management within the SAEC. The Department of Radiation Protection and Nuclear Safety (RPNSe responsibilitth s ha ) r reviefo y f applicationwo d controan s f o l nuclear activities.
To implement its regulatory function, the SAEC established the Syrian Nuclear Safety Committee (SNSC) in 1985. It is composed of 16 permanent members (representatives from ministrie thred san e member Commission)e th f so .
SNSe approves Th Cha followine dth regulationsg9 :
1. Basic safety standard for radiation protection; . 2 Regulation r safsfo e managemen f radioactivo t e waste; 3. Regulations for medical supervision and examination; . 4 Regulation industriaf o r safe sfo e us l radiography; . 5 Instruction radiopharmaceuticar sfo l material diagnostir sfo c purpose i hospitalssh ; . 6 Regulation conditiond san f licensinso wore gth k with radiation; 7. Regulations and standards for safe operation of reactors; 8. Regulations for safe transport of radioactive material; 9. Emergency preparedness programme.
99 2. WASTE EXISTING AND ARISING IN SYRIA
Presently radioactive wast generates ei medicaln di , researc industriad han l fields:
(a) medicae Inth l field, only short-lived radionuclide usee sopear s da n sources (e.g. 131I, therapr fo "Tcd yan ) seale 192dusedd e Isourcerar an .o ^C s (b) In the research field, there are many radioactive liquids which are used in different departments of SAEC such as 137Cs, ^Sr, "Tc, 210Po, ^Co and 14C. These radioisotopes after the use are stored in bottles in the interim store, and there are also many radioisotopes which are used for calibration of instruments. (c) In the industrial field, the largest producer of radioactive waste in Syria is the phosphate industry, but oil and gas industry also generates wastes with enhanced radiation levels.
3. SEALED SOURCES
Sealed sources used in Syria are returned to the supplier when no longer in use. Since 1987 there has been established a comprehensive system of registration and notification of all sources entering and leaving the country.
Table 1 and 2 list the spent sealed sources which are stored in the interim store.
TABL E. SPENI T SEALED SOURCES BEING STORE SAECT DA .
Nuclide Origin Activity Notes 60Co Russia High ? &OCo Russia Unknown Salt in a glass tube 60Co Russia Unknown Wir glasa n ei s tube
60Co Russia Low Salt in a glass tube 137Cs Russii a mC 9 1. 3 Sources in original contaier,used for calibration
137Cs Russia Low metalli a d use ro n ci d Cement industry 226Ra ? ? 33small containers fron General Organization o Geology
137Cs Russia i 10mC Density/moisture detector 24lAm-Be U.S.A 60 mCi Ministr f Irrigatioyo n ? 137Cs Russia Low 226Ra ? 200 mR/h Lead box with sources detecte n scrai d p meta from Lebanon
60Co UK ICi d sourceol 8 s fro< th m Syrial oi industrn y delivered to SAEC
100 TABLE H. INVENTORY OF SOURCES AT SAEC. Ser. Type Activity Origin Entrence Export Users No:
1 137Cs i 100mC German 27-10-89 11-10-90 Oil Comp 2 192i IC r 1 0. France 12-11-89 10-03-9 l Com3 Oi p 3 228Th 10/^Ci U.S. A 14-09-90 No Oil Comp 4 241 Am 0.9/Ci U.K 01-06-91 no Cooling company 5 131i ! mG 0 10 France 16-03-92 Local Cement 6 99Tc 400 mCi France 23-02-92 Local Cement 7 201-ri 5 mCi France 04-06-92 Local Cement
. FUTUR4 E WASTE ARISINGS
The largest individual waste generator in the nuclear application field, will be the Nuclear Research Centre together with a 30 kW research reactor to be commissioned at Damascu years2 n si . Radioisotope production will als establishee ob centree e th th t t da A . Centre, irradiatiothern a s ei n facilitcommissionine th n i y g stage which will initialle yb loade^Cof o i . dkC Spen wit0 h10 t sources fro facilite mth expectee yar returnee b o dt o dt the supplier. Waste management activities are planned near the research centre.These include chemical treatmen othed an t r processe r treatmen slevefo w lo l f radioactivo t e waste.
5. RESEARCH
There are many studies in progress related to radioactive waste management such as:
diffusioe Th 137f o nC s and^S locan ri l rocks (Limestone, Dolomit,....) The cinatique studies on the radioisotopes migration in local rocks The radioactive pollution studies in phosphogypsum in dumping aria.
101 STATUS OF RADIOACTIVE WASTE MANAGEMENT IN ZAMBIA
K. MWALE Radiation Protection Service, Ministr f Healthyo , Lusaka, Zambia.
Abstract Zambia being part of the world community clearly understands that careless handling of radioactive waste would cause problem - sworldwid r humafo - e n e healthth r fo , environmen naturad an t l resources management r thifo s s i reaso t I . n thaRadiatioe th t n Protection Board has initiated a Radioactive Waste Management Programme covering the following areas:
. i Legislatio Radioactivn no e Waste Management, ii. Immobilization of spent sealed radioactive sources, and iii. Sitin constructiod gan interin a f no m storage facility.
1. LEGISLATION ON RADIOACTIVE WASTE MANAGEMENT
The Radiation Protection Board, the competent authority in Zambia is about to pass the controlling regulation radioactivr sfo e countrywaste th n ei . importanThin a s si te steth n pi right direction as the regulations will put more specific emphasis on the requirements that will followee hav b handlin e o et th y db g authority.
Handling of radioactive materials starts from the time of importation. The executive Boarde th ar Radiatiof e mo th , n Protection Service needinformee b o st writinn di e th f go intention to import before an import license can be granted after all the requirements are satisfied. One condition among others is that the user must pay a license fee. The user must also have adequately trained personnel and that the facility where the radioactive materials will be used must satisfy the physical and other requirements. This includes radiation safety precaution of persons involved in case of a radiological accident.
Apart from the above the safety in transportation must also be supervised by the competent authority through the Radiation Protection Service. All these steps are initial and necessar radioactivn yi e waste management becaus onls i t ei y then that registratioe th f no sources is done. The expected useful life of the source is followed and at decommissioning the Radiation Protection Service is involved.
. IMMOBILIZATIO2 SPENF NO T RADIOACTIVE SOURCES
Most of the radiation sources in Zambia being used and those that are spent are sealed solid sources. These include caesium-137, plutonium-238, americium-241, iridium-19d 2an cobalt-60 which are or have been used in industry and mining. Other sources which are being use researcn di h institution iron-55e ar s , cadmium-109, americium-241, cobalt-57, strontium-90, europium-152, etc.
All the sources in the country are registered with the Radiation Protection Service. These include both those in use and those that are spent and are awaiting immobilization,
103 interim storag disposald ean . Short lived source decayee sar disposed dan aftef do r they attain acceptable level of activity. Initial work leading to immobilization of sources has advanced so well. This wil done b l t useea r institutions whils interin a sita tr efo m storage facilits yi being decided upon. Immobilization of a few sources has already been done under the supervision of an IAEA expert and the country now has the expertise to proceed with this project. However onle th , y setbac constructioe th s ki storagf no site th e e n storag sheo r dfo e whils constructioe th t centralizea f no d interim storage facilit beins yi g considered.
3. SITING AND CONSTRUCTION OF AN INTERIM STORAGE FACILITY
Zambia does not have the resources and high level expertise to construct a deep underground disposal facility. Other surfac shallor eo w level method beine sar g considered. However, this does not mean that necessary requirements for such a facility will be overlooked.Technical considerations would include the necessary requirements that the conditioned sourcet contaminatno o sd environmente eth .
. SUITABILIT4 INTERIE TH F YMO STORAGE FACILITY
suitabilite Th f sucyo h facility will hav tako et e into accoun amoune sourcee th t th f o t s already spent both in terms of volume and activity levels. There should also be an element of future requirements in terms of the facilities that will be decommissioned in medium term of about 15-30 years.
othee Th r requiremen geographicae th s i t l fror locationfa m o site to ee Th . b mus t no t e areth a wher e countrth e e i source h th moso avoit y f e o ar tds riskd costan sn i s transportation. The country will constantly seek IAEA and other member countries' advice and support hi achieving this goal.
104 SWEDISH WASTE MANAGEMENT PROGRAMME
PER-ERIC AHLSTRÖM Swedish Nuclear Fuel and Waste Management Co., Stockholm, Sweden
(Presented Gustaffson)B. by
Abstract Sweden has developed a comprehensive system for the manage- wastel al men f so t arising fronucleas it m r power production. An interim storage for spent nuclear fuel is in operation since 1985. A repository for low and medium level waste has been constructed and is in operation since 1988. Transpor- tation of the fuel and other radioactive wastes is made by a sea transport system e existin.Th g facilities will with some moderate expansio e sufficiennb handlo t radioactivl eal e waste lona r g fo stime . An encapsulation plant for spent nuclear fuel and a reposi- torr finafo y l disposa a limite f lo d amoun f speno t t fues i l planne buile b o dt til le repositor 2008th n I . e fueyth l wil e isolatelb multiply db e engineere geologicad an d l barriers. The ongoing waste management RDScD-programme is mainly concerned with questions related to the encapsulation of fued constructioan l f sucno hrepositora e granitith n i y c bedrock in Sweden. During the 1990s the emphasize will be on finalising the development and the design of the needed facilities and on the characterization of candidate repository sites. The cost for spent fuel management including final disposal has been calculated to 4800 SEK/kg U.
Introduction Swede s twelvnha e nuclear power reactors wit totaha l capacit producind an 1000f o y e annuae 0MW th gf lo abou % 50 t electric demand in the country. These reactors are located at four different sites. The production of electricity creates annually about 250 tonnes of spent nuclear fuel and about 3000 m3 of other radioactive wastes. Up to the year 2010 the accumulated amount wil e somlb e 8000 tonne f speno s t fued an l f otheo 3 m r 0 wastes900 0 e decommissionin.Th d gan dismantling of the plants will create another 100 000 m3 of radioactive wastes. The power plants have storage ponds for spent fuel at each unit with capacities varying from one up to eight years of output of used fuel. They also have capacit interir fo y mediud man storagw mlo levef eo l wastes. Accordin plantSwediso e ownere t gth th e respon w f sar o h la - e safth sibl er handlinfo e d disposa an gl radioactiv al f o l e wastes arising from the plants. In order to meet these requi- rements the four utilities which own the nuclear power plants have formed the Swedish Nuclear Fuel and Waste Management Co. - SKB. Planning, construction and operation of, as well as researc d developmenhan l facilitieal r fo t se needeth r fo d
105 safe handling and disposal of all spent nuclear fuel and radioactive waste thue maie sar th s n task SKBf o s . managemene Th radioactivf o t e waste Sweden i s bases i n n o d some firmly established guidelines. All wastes will be taken care of in Sweden and disposed of inside the country. Decision parliamenn i s t afte referendura nuclean mo r power in 1980 limit the power programme to the existing twelve nuclear reactors. With these constraints o thern s i e incentiv reprocessinr fo e spene th g t nuclear fuel. s SKdevelopeBha safe systea dth e r managemenfo m l al f o t radioactive waste Swedenn i s . Majo re systepartth e f o msar alread operationn i y . This paper describes briefly some experiences gained from the development of the system and from the operation of the existing facilities. An outline of the plan r encapsulatiofo s d finaan n l disposa f spenlo t fuel is also given.
Short-lived low and medium level wastes The low and medium level wastes from the nuclear power plants as wel s frola m hospitals, industr d researcan y e seno har t the repositor shorr fo y mediutd an live w mlo d level waste- Forsmarkt a SF - R . This0 10 facilit o t 0 s buil5 ywa t a t meters depth in the bedrock, one kilometre off shore, below the BalticForsmare th t a , k nuclear power plant starting 1982 and completed 1988 e repositorTh . y consist rocf so k caverns of different design accordin wasto t g e Figur e typse - e. e1 The present capacity could be expanded to contain all low and medium level wastes including the decommissioning waste from the Swedish programme. The experiences from almost six years of operation are very f 199o d 3en aboue goodth ty B .1300wastf o s bee3 m 0eha n disposed in SFR. Doses to workers are minimal and most of the radiation exposure is due to the common radon exposures in bedrock caverns.
Interim storag f speno e t nuclear fuel e spenTh t nuclear fuel wil storee lb abour fo dyear0 4 t n si the central interim storage facility - CLAB - which is nucleae th f o r locate e poweon t ra d plant Oskarshamn- s e .Th interim storage period allow residuae sth e lth head an t radioactivity to decay by a factor of about ten. Thereby the handling and final disposal will be considerably easier. CLAB was constructed in the early 1980s and taken in oper- ation in 1985. It consists of an above-ground receiving and handling building and an underground storage complex in rock - see Figure 2. The fuel is handled and stored under water. e capacit Th w abouno ts i y 5000 tonne f spenso t fue n foui l r storage pools o stor T .l 800 eal 0 tonnes from twelve reactors some additional pools wil needee b l d around 2000.
106 KeR y datSF r afo 1. Rock vault for intermediate-level waste in 4. Silo for intermediate-level waste In metal 1 Disposal 3 capacitym 0 00 0 :6 concrete tanks tanke handlee Th . sar y db drum mouldsr so waste Th . handleds ei Planned expansion: 30 000 m3 forkiift truck. by a special remote-controlled handling Receiving capacity: 2. Rock vault for low-level waste in freight machine. m'/yea0 600 r containers containere .Th handlee sar y db . Operatin5 g building with operations centre Operating personnel: forkiift truck. and personnel quarters. abou person0 t2 s 3. Rock vault with pits for intermediate-level Constructio millio0 n 74 cost K n :SE waste in metal drums or moulds. The waste (through 1992) handles i remote-controlley db d overhead Operatin milliog7 ' 2 costr K npe :SE crane. year (1992)
FIG. 1. SFR disposal facility for low and intermediate level radioactive waste.
CLA n receivBca e abou tonne0 30 t f spenso t fueyearr pe l t .A present abou 0 tonne25 t arrivins i s g each yea d abouran t 1800 tonne poolse e storeth sar n .i d Besides spent fuel also used core component d reactosan r internal e storeb n CLABt dsa ca . e plan Th operates i t d aroun cloce th dunloadint kbu s i g performed only during the daytime shift. The operating staff number persons0 s7 . Additional service alse sar o obtained fronearbe mth y power plant staff e constructioTh . n costs were about 1750 MSEK.
Transportation All present major nuclear facilities in Sweden are located at the coast. This a transportatiomakese naturat e i s us o t l n heavl foal r y transport d froan mo t sthes e plants. a Thuse sa transportatioe th r fo n e systeus bees n mi ha ns i buil d an t
107 J) Terminal vehicle i~ FIG. 2. Central interim storage facility for spent nuclear fuel. - CLAB 108 shipmen f speno t t r othefue fo wels a lrs la radioactiv e wastes e syste.Th m consist followine th f o s g components: a specially designed ship, the M/S Sigyn, n speciate l transport cask r spenfo s t e Figurfuese - le 3, two transport cask corr fo se components, four diesel-powered terminal vehicle locar fo s l road transpor e reactoth t a t r CLABt a site d ,an s additional transport containers and a terminal vehicle for the transportation of medium level waste to SFR. The M/S Sigyn was taken in operation in 1982 and has since then shipped about 2000 tonnes of spent fuel plus some 13000 m3 of medium level waste. The spent fuel casks meet the ~he spent nuclear fvel fiansporîedis m ery sturdy "casks" that provide radi- ation shielding protectionand the in event of accidents The casks ai e made of stainless steel tilth copper finsheatfor -Dissipation caskA containing fyel hs 80 tonnes A transport vehicle usedis moveto casks to and from M/S Sigyn The casks are anchored earnera on fiame, v,hichis secured to the ship's cargo deck FIG. 3. Transportation of spent fuel 109 stringent IAEA requirement radiation so n shieldind gan ability to withstand external stresses and fire. Each casks can take somewhat more than three tonnes of fuel and has a transport weight of 76 tonnes. Ten spent fuel casks or medium level waste containers can be carried by Sigyn at the same time. The investments in the total transportation system so far are annuae abouth MSE 0 d l25 t Kan operatin maintenancd gan e costs are abou MSEK5 1 t . Plans for final disposal of spent fuel and other long-lived wastes The work carried out during a period of about fifteen years in Sweden, and similar work in other countries, has led to a broad agreement among international experts that methods exis implementinr fo t g final disposa high-levef o l l wastd ean spent nuclear fuel. Proposals and alternative options for the final disposal of spent nuclear fuel have been reviewed and studied by both regulatory authorities and industry in extensive R&D projects during the 1980s. Thus, the important issues relating to encapsulation and final disposal of spent nuclear fuel in Swedish bedrock have been thoroughly elucidated. Spent nuclear fuel contains large quantitie f radioactivso e materials. Final disposal shal arrangee lb thao waste s d th t e is kept isolated in a safe manner while it has a higher radiotoxicity than otherwise found in nature, i e over a perio years0 arounf 00 o d brino 0 .T 10 d g about this isolation a fina, l repositor spenr fo y t fue designes i l d according to the multi-barrier principle. Safety assessments show tha y usinb t g stable material e engineereth n i s d barriers radioactive materials can be kept isolated for one million year morer so . After having examined safety, technical feasibility and other aspects for a number of different alternatives, work in w reacheSwedeno s poinna dha t wher shoult i e cone db - centrated to a main line. SKB has concluded that the present knowledge is sufficient to select a preferred system design, to designate candidate sites for siting a repository, to characterize thes eadapo t sitee repositor d th t an s o locat y l conditions. RD&D-ProgrammB SK Thue th s call2 9 ecompletior fo s e th f no research, developmen demonstratiod an t n wor buildiny kb a g final repository done b stagen .i eo t Thi s si s starting with above th ef minoa o give % r10 n quantit o totat 5 lf o yamount . maie Th n reaso thir fo ns stage-wise approace th s i h possibility to demonstrate in the first stage: The siting process with all its technical, administra- tive and political decisions; 110 e step-wisTh e investigatio d characterizationan e th f no repository site; e systeTh m desig constructiond an n ; The encapsulation of spent nuclear fuel; handline Th g chai f spenno t nuclear fuel from CLAo Bt deposition in the repository; e operatioTh deea f pno repository; The licensing of handling, encapsulation and deep dis- posal, including the assessment of long-term safety: (Retrievability of the waste packages); e timth eo t period e Du s involve pose th dt closure safetf yo the final repository cannot be demonstrated through field tests. The longterm safety must be demonstrated by a tech- nical-scientific assessmen repositore th f o t y performanc. e When the first stage has been completed, the results will be evaluated before decidin expango t whethe e t dth no r ro facility to accommodate all the waste. This makes it also possibl consideo t e r whethe depositee rth d waste shoule db retrieve alternativr fo d e treatment s 8KB'i t sI . opinion that suc hstepwisa e approac disposao ht spenf o l t nuclear fuel wit ha freedo o future t f choic gooa th mo y s r dwa i e fo e enlist broad support for the method of disposing of the nuclear waste. Additional facilities and systems that will be needed are: Encapsulation plant for spent nuclear fuel, including a buffer stor encapsulater fo e d fuel. Deep repositor encapsulater fo y d spent nuclear fuel. Transportation system betwee encapsulatioe nth CLAd an B n plant for spent fuel and between the latter and the site of the deep repository. 8KB believes that the first stage including construction of the encapsulation plant and the deep repository and also depositioe spenth e tcompletef b o fue% n f 5-1no ca 0l d within about 20 years. The SKB RD&D-programme 92 has been reviewed by a comprehensiv f experto t universitiest se esa , research competene instituteth r fo tc Swediset s h authoritiese .Th authorities - in particular the Swedish Nuclear Power Inspectorate e scientifith ,d an SKI, c advisory committeo t e the ministry, KASA hav- M e summarized their conclusions from the review in reports to the government. In general they find thaprogramme th t e complies wit e requirementhth e law.th f so They endorse the main direction and the start of practical work towards final disposal of the spent nuclear fuel. e inspectoratTh e stresse necessite sth concentratinf o y n go one method for disposal in order to make it possible for the generation using the benefits of nuclear power to also take full responsibility for the waste. If decisions are postponed burdens will unnecessaril n futuro t epu generationse yb e .Th inspectorate also points out that landbased geologic disposal of encapsulated fuel is the only realistic alternative for Sweden. Ill The authorities also give several critical comments e g about o optimistito a c timeschedule, about lac systematif o k c approac e sitth en hi selectio n proces d aboue san th t comprehensiveness of the safety analyses at early stages of the process. Deep repositor r spenfo y t fued othean l r long- lived wastes In the long-range perspective the safe isolation of spent fuel is achieved in a deep geologic repository. The technical solutions that have been studie Sweden i de basee th n ar n do following principles: Final disposa Swedise th n i lh bedrock. A multibarrier system with mutually independent natural and engineered barriers. Natural materials in the engineered barriers. Limited temperatures, radiation dose and other impact on the rock. The implementation of these principles can of course be made bmultitudya differenf o e t s designs ha 1980e B th SK sn I . evaluated several such designs e result.Th s were reporten i d the R&D-programm RD&D-programmn i d an e 9 conclusio8 eTh . 92 e n continuo t s wa e wit reference hth e repository design selected alreade latth e n Figure i y1970t consist I se . - sa e4 f so system of tunnels at about 500 m depth in the crystalline bedrock. From the floors of the tunnel deposition holes are drilled bout 7.5 meters deep and 1.5 meters diameter. Other long-lived wastes mainly from research activitiet sa the national laboratory in Studsvik would be disposed of in a specially designed part of the deep repository separated from the spent fuel. A primary role of the bedrock around a repository is to provid mechanicallea chemicalld an y y stable environmenr fo t the engineered barriers protectin wastee th g . Studied san investigation bedroce th f Sweden so i k n durin pase 5 th g1 t years indicate that there are many sites possessing the properties and stability needed for constructing a safe repository. The work on siting and construction of the deep repository is planned to proceed in the stages shown in Figure 5. e selectioTh f candidatno repositore th site r fo s y wile lb based on the qualities necessary from safety-related, technical, societal and legal viewpoints. It must be demonstrate e selecteth r fo dd sitd selecteean d repository system thae safetth t y requirements imposee th y b d authoritie mete sar must . I possible b t builo et e th d repository and carry out deposition as intended. The siting process investigatione th , constructioe th d an s n work shall be carrie thao l relevans al tt dou t lega plannind an l g 112 Schematic drawing of a deep repos- itory. systemA of tunnels with vertical deposition holts will be bttilt At a. depth of about metres.0 50 spente Th fuel - c>f Multiple barriers protect the spent deep fuel:he in repository. 1. Copper canister. The canister isolates fuelthe gro from the luater. The fuel itself ii in solid form and has very hu: solubility. Blocks2. of benionite day.clayThe prevents ground'izaitr flow around canisterthe while protecting against tn;nar movements in the rock. mixtureA 3. of ber,:&>ii:e claysandand fills :!:-eup :;:>:i:ch. 4. The rock o/ffr; n stable environment, both n:ccl'.:ri:'c.;iH .;»c chemically. /: .;/js .-c:< ,-j .-7J }!:erfcr :he ^•o;iKr':;.;:cr FIG. 4. Deep repository for spent nuclear fuel 113 1995 -1995 Regulatory review concernint undeAc e rth e gth Management of Natural Resources (NRL) 2000 -2000 Regulatory review under NRL and the Act on Nuclear Activities (KTL) 2005 -2005 Regulatory review undeL rKT 2010 -2010 The repository can be expanded to full scale after evaluation FIG. 5. Timescale of the work on siting and construction of a deep repository in Sweden related requirement mete t least sar no d last. An t t i ,bu , shall be possible to carry out the project in harmony with the host communit locae th ld populationyan . The ongoing work is mainly concentrated on pre-studies in interested municipalitie generan o d san l sitin desigd gan n studies. A pre-study is a preliminary investigation based on existing information and data on the impacts and pos- sibilitie sitinf o s repositora g municipalitye th n i y e .Th study is made as a cooperative effort in order to provide the municipality and SKB at an early stage with all available facts to give a base for decision on further work. It is a clear understanding thastude th t y doe t implsno y thae th t municipalit s committei y accepo t d t future site investigations .A forma l agreemen pre-studa r fo t bees yha n signed with one municipality in northern Sweden and discussions are under way with a few more. SKB would like to make such pre-studies for five to ten municipalities. The pre-studies will be followed by site investigations on o sitestw e purpos.Th provido t thesf o e e basie ar er eth fo s selectin e sitdetailer on gfo e d characterization e latte.Th r will require a permit from the government according to the AcManagemenn o t Naturaf o t l Resources e plannin.Th g foresees that such a permit should be obtained before the end of this decade. Spent fuel encapsulation e encapsulatioth r Fo f speno n t nuclear fuel planB SK ,o t s expand the central interim storage facility for spent fuel 114 (CLAB). The spent fuel is already being stored at CLAB, and SKB believes that expansion of CLAB with an encapsulation plant for spent fuel has clear advantages in terms of logis- tics, resource utilizatio d environmentanan l impact. Several alternative designs were canistee studieth r fo o drt encapsulate the spent fuel. A copper-steel canister for 12 BWR fuel assemblies or 4 PWR fuel assemblies - see Figure 6 - was chosen as the reference alternative and is the basis for the ongoing design work. Final desig plannes i ne b o t d selected in 1995. The canister consists of a steel container providing mechanical protection and an outer copper container providing long-term corrosion protection e empt.Th y space between the fuel rods will be filled with some suitable inert Cutaway view o fuelwithrod pellets of uranium dioxide. Schematic drawing of canister. The canister aboutis longm 5 and has a diameter of 88 cm. The can- ister wall consists steelof Scm of of cm copper. 5 and With fuelthe canister will weigh about IS tonnes. FIG.Encapsulation6. of spent nuclear fuel disposalfor 115 material like glass bead d inersan t s importangasi t I . t that the moisture conten minimuma keps i to t t e coppe.Th d li r will be sealed by electron beam welding. The waste package will hav totaea l weigh almosf o t tonne5 1 t d contaisan n about 2 tonnes of spent fuel (uranium weight). Cost wastr fo s e management e existincoste Th th r som fo sf eo g facilitie d systeman s s have already been mentioned majoe .Th e rspen th par tf to fuel management costs will howeve e incurre rb e future th n i d. These costs will be covered by funds which are built up from nucleae th n o r fee t powespu r production e feee .Th ar s revised annuall governmente th y yb e revision.Th basee sar d on detailed cost calculations whic reportee har d each yeay rb SKB and reviewed by the Nuclear Power Inspectorate - SKI. The fee has been 0.019 SEK/kWh on the average for the last ten years and includes not only costs for spent fuel management but also cost other fo s r waste handlin disposad r gan fo d lan decommissioning of nuclear facilities. e calculateTh d cost spenr fo s t fuel managemen e summarizear t d e followinith n g table. Please note thacoste r th t fo s researc d developmenhan t includ costl r findineal fo s d gan characterizin repositore th g y site. Calculated costs (price level 1993). Average for 7700 tonnes 4800 SEK/kg U Marginal for additional quantity 2000 SEK/kg U Relative distributio thesf o n e costs. Average Marginal Transportation 4% 3% Interim storage 25% 20% Encapsulation 21% 35% % 42 % 34 Final disposal Researc% 15 developmend han t 116 MANAGEMENT OF RADIOACTIVE WASTE IN ISRAËL S. BRENNER . NE'EMANE , . SHABTAIB , . GARTYE , . BUTENKV , O Ministry of the Environment, Tel-Aviv University, Sackler Medical School, Tel-Aviv, Israel Abstract Radioactive materials are used extensively in Israel in labeled chemical n i hospitalss , research laboratories, industriad an l agricultural premises and for environmental studies. In many instances they provide scientists and technicians with unique methods' of monitoring processes and measuring reactions. A by products of many of these methods is radioactive waste (R¥). e primarTh y concere Ministre Environmenth th f o f no y n wasti t e managemen o implement s i t n effectiva t e contro d disposaan l l system that ensures the safety of people and protection of the environment. e responsiblTh e authorit r Elfo ?y managemen Chiee n Israei tth fs i l Radiation Executive (CRE) who is nominated by the Minister of the Environment according to the "Pharmacists Regulation - Radioactive Element Productd san s Thereof". These regulationso t authoriz E CR e th e issu elicensa wastr efo e disposal services, after consulting wite th h Israeli Atomic Energy Commission (IÀEC). Each R¥ producing institute in Israel has to acquire a license for its operation. This license limits the amount of radioactive materials purchased by the institute and approves the nomination of a radiation officer. The radiation officer is responsible for the appropriate handling of R¥ inside the institute. Hospital d researcan s h institutes pos a unique ¥ problemR e . They produce a large amount of R¥ and the adequate segregation and disposal of this waste by these institutions deserves special attention. However e maith ,n y sorrequiremenan t f o wil e ¥ b s R lthati t o N : disposed off through the ordinary waste system or through the general sewage. Radiation waste disposal service offeree I&EC'e sar th y sb d Nuclear Research Center—Negev (NRCN) which operates and monitors a National Radiactive Waste Disposal Site (NR¥DS). The NRTOS which is the only one in Israel is located in the Negev Desert in the southern part of Israel. Official E scontro CR unde floe e f th radioactivlth ro w e materials in Israel aided by a computerized Data Base Management System (DBMS). This software comprises of the following modules: licensing module, impor d distributioan t n module d wastan , e disposa d transportatioan l n module. At the present time only the first and second modules were completed. The waste disposal module of the DBMS described above, will includ a etheoretica e estimatiol th modee volum¥ r R th fo lf f o eo n production by large institutions. This model will provide a "first guess" that can be used to validate the information given by the waste disposal agency. 117 l. SURVEILLANCE OF RADIOACTIVE MATERIALS IN ISRAEL Radioactive materials are used extensively in Israel in many area d applicationsan s e.g. medical diagnosi d therapyan s , industry, agriculture, research and development and related subjects. À radiation protection infrastructur f o regulationse , educational facilities, licensing and supervision arrangements was developed in Israel including the formulation of radioactive vaste (R¥) disposal rules. The system of sharing responsibilities for radiation protection among the Ministry of Environment (MOE), Ministry of Health (MOH), the Ministr f Labo o yd Sociaan r l Affairs e Israe(MOLSÀth d l an )Atomi c Energy Commission (IÀEC) was developed, especially during the last 15 years. The main two sets of regulations relating to radiation protection concerning radioisotopes are: 1.Pharmacists' regulations - radioactive elements and products thereof. E togetheMO e Th r H wit share MO responsibilithth e e th r fo y enforcement of these regulations. 2. Safety-at-work regulations (person engage n i ionizind g radiation). These regulation undee supervisiose ar rth MOLSÂf no : It should be noted that the Ministry of Transportation in coordination -with MOE s responsibli ,e transportatioth r fo e f o n radioactive materials includin. gOT w (1994no Ther consumer4 e )30 ear radioactivf so e materiale th n i s country according to the different groups described in Table I. Table I. List of radioactive materials consumers in Israel GROUP TYPE NUMBER OF INSTITUTES Industrial nuclear measurement devices 138 Medical laboratories 37 Researc educatiod han n institutes 36 Hospitals 29 Impor distributiot& n companies 27 Smoke & fire detectors 16 Non-destructive test— radiography 11 H-3 light emitters 7 Radioisotopes producers 3 Total 304 Since each institute may contain and handle more than one radioisotope, the computerized screening program enables the authorities to review the overall situation at any given time. Table II present breakdowe sth 15.08.94r nfo . Table II. List of Radioactive Installation and Sources institutef o o N s (Tabl) eI 304 f installationo o N s 2169 No of "sealed sources" 2147 No of "unsealed sources" 5819 118 1.1. Paaraaci&ts ' regulations - radioactive elements and produces thereof (1992 Ed.) These regulations d MOE,an ,undeH prohibiauthorite MO th re th tf yo the purchase, distribution, transportation and any application of radioactive materials unless a special license has been issued to the user. Licenses are issued by the Chief Radiation Executive (CEE), appointed by the Minister of Environment. These are the main regulations dealing with the issue of radioactive vaste. The principle behind the regulations is supervision over radioisotopes "from cradle to grave" whenever this is feasible, usuall e tendenco th t folloy s i wy international guidelinef o s recognized bodies such as IAEA, ¥HO and ILO. e regulationTh s specif e generath y l conditions under whica h license wil grantede b l , including shieldin d storaggan e arrangements. regulatione Baseth n o d s ther e specifiar e c guideline r releasinfo s g radioactive materials from customs and it is also one of the main duties of the CRE to prevent the introduction of radioactive vaste intcountre th o y which sometimes occurre pase th t n i unded r different names. e regulationTh s also specify practice d activitiean s s exempt from licensing, and list services and practices which may not be undertaken without special permit from the CRE (such as dosimetry, vaste disposal, radiotoxicology services, etc.). e regulationTh s requir e appointmenth e a Nationa f o t l Advisory Committee for Radiation Protection whose members are professionals in various field f scienco s d technologyan e d expertan , n radiatioi s n protection. Finally e regulationth , s requir E consuleCR e IÀEe thath th tCt prior to licensing certain practices (e.g. offering vaste disposal services) r whe,o n relatively high activit f radioactivo y e materials (beyond specified limits) is concerned. 1.2. Safety-at—vork regulations (persons engagea in ionising- radiation) These regulations are enforced under the authority of the MOLSA. Thet fortse y h guideline r controfo s n facilitiei l s where employers handle radioactive material r radiatioso n equipment.The main elements regulatione oth f e necessarth e ar s y requirement e protectioth r fo s n of the workers. The limits for the annual radiation doses to the whole body and to single organs of radiation workers are based on the recommendations of the International Commissio Radiologican no l Protection. 2. RADIOACTIVE ¥ASTE MANAGEMENT The responsible authority for Rff management in Israel is the Chief Radiation Executive (CRE) accordin e "Pharmacistth o t g s Regulatio- n Radioactive Elements and Products Thereof". These regulations o issut authorize E a elicens CR r waste fo th es e disposal services, after consulting with the IAEC. 119 The requirements for handling RW in each institute possessing radioactive material s integratei s e specifith n i d c licens f thao e t institute e licensTh . e limit e amounth sf radioactivo t e materials purchase institute th y db d approve an e nominatioe sth a radiatio f no n protection officer e radiatioTh . n e office th s i liabl r rfo e appropriate handling of RW inside the institute. Hospital d researcsan h institutes pos a euniqu problemW R e . They produc ea larg d theie an w leveamounr lo W radiatioR lf to n offices i r generall physicisa y perforo wh t m this part-tima dut s a y e joba s A . consequence adequate ,th e segregatio cannoW R f assuree no b t y thesdb e institutions. À safety assessment of the above actual situation dictate scleaa instructiot rcu disposalW R r nfo : y soran t f o wil e dispose? Nb R lo f througof d e ordinarth h y waste system or through the general sewage unless a special permit was granteCREe th .y db RW disposal services are offered by the lÀEC's Nuclear Research Cente — rNege v (NRCN) which operate d monitoran s a sNationa l Radioactive Waste Disposal Sitee currenTh . t report contain4 9 s institutes disposin e Nationath g o theit lW R rSit e (Table III)e Th . National Waste Disposal Sitn Israei e e onls e i th lwhicon y s i h locateNegee th vn i dsouthere Deser th n i t n par f Israelo t . However, most of the hospitals and institutions generating radioactive waste are located far away from the above site. The cost of transporting the radioactive wast s usualli e ye larg th ver eo y t distance hige du h s involved n ordeI .o solvt r e this proble e w intenm o instalt d l intermediate waste concentration site n i severas le regionth f o s country. Table III. Radwaste disposal from institutes (1994) (except sealed sources) Medical centers 21 Medical laboratories 24 Research institutes 21 Other (fire detectors. distributors, producers) 28 Total 94 The CRE control over the flow of radioactive materials in Israel is a computeriseaide y b d d Data Base Management System (DBMS). This software comprises of the following modules: licensing module, import and distribution module, and waste disposal and transportation module. At the present time only the first and second modules were completed. The CRE gets monthly reports from the national agency for waste disposa e NRCth N f o lwhic h specif e quantitieth y f drumo s s arrived from each hospital and research center in Israel (e.g. Fig. 1-2). As can be seen from these figures, the volume of radwaste produced by large institutions often decreases with time. Howeve suco rn h decrease s foun wa n theii d r consumptio f radioactivo n e materials n ordeI . o t r achieve safe and efficient waste disposal control, the CRE should have the ability to estimate whether the reported waste quantities are reasonable. The waste disposal module of the DBMS described above, will include a theoretica e estimatiol th modee volum r th fo lf radwastf o eo n e 120 production by large institutions. This model will provide a "first guess" that can be used to validate the information given by the ^faste disposal agency. 160- 140- K xstfftïftïtS? K^äOQpj^.ey- '- ; ^ A .-»x-XvXvi*«*« A? ICHILLOV SOROKA RAMBAM KAPLAN BEILINSON SHIVA CARMEL FIG. l. Disposal of radioactive waste (drums) from hospitals (1989-93). 7001 .,./ • C/Î " • '/SSSSlSi^.'^'.L -y*./1-*•!/y £==^7^^^àrÀ^// vvv 7 7 \ .<" »U;———/.•• £1. • ~~~7J-r WEIZMAN T-A-UNIV .HAIFD ME A. B-GUMV. MED. JER. HEB.UNIV. AGR.RHVT. FIG. Disposal2. of radioactive waste (drums) from research (1989-93). 121 The amount of R¥ produced by each laboratory in a hospital depends on the various kinds of tests performed there, the work load, and the specific technique e personnels th use y b d e assumed¥ . a firs s a ,t order approximation, that this dependenc s lineari y . Since most laboratories in Israel exercise similar techniques for the same kind of test, it is possible to consider a limited number of hospitals in calibratin e coefficientth g a linea f o s r simplified model. Typical e numbevalueth f testr o r fo s s neede a o produc¥ drut dR s a me on e function of labs' discipline is listed in table IV. Tabl Typica. eIV l numbee valueth f r testro s fo s neede produc o dru¥ t dR me eon DISCIPLINE i a^ 0 40 i Gastroentherologv Bacteriology 2,200 Nuclear Cardioloqry 200 Hematology 800 Genetics 2,400 Endocrinologv B,800 e drume numbeth Th N sproducef o r d each a hospitamont y b h l according to such a model will be given by , HZnj/am ( -dk b = N+ i) where the summation is carried over all the laboratories of the institute, and n - is the quantity of tests done per month in each laboratory; m - is the number of licensed radioactive facilities in the hospital; e totath ls i activit - k n milicurii y f radioactivo e e materials purchased by the hospital each year, not including Tl-201 and Mo-99 - Tc-99m. constante th modele e th ar f s. o d coef o d ficientan t f b se , e sai Th e constantTh e takear ± n sa werd fro d e me valuean tablTh b . f so IV e estimated to be 0.08 and 0.0035 respectively. There are few cases in which the CEE has decided to grant permits to dispos d ß-emitterse an liqui - sewage a ¥ (noth R r d o fo te)t system. This practice is performed only when the regular procedure is not practical and comprehensive calculations have demonstrated that the material dilution in the sewage water of each institute will result in radionuclides concentrations below national drinking water levelst A . the momen tinstitute9 1 ther e ear s with such permits. FinallyE MO e th , is engaged in the legislation process of a new law on "Disposal of R¥" based on the approach allowing disposal of short lived R¥ as regular solid waste provideprecautione th l dal s were taken thawaste s th t ewa segregated and stored the required time ensuring complete decay. 3. CONCLUSIONS We have describe disposa¥ R w dho manages li d supervisedan e th y db CRE, the responsible authority in Israel. An important part of the enforcement capabilities of the CRE is his ability to independently validat reporte th e s concernin amoune th g f wasto t e produce y largb d e user f radioactivso e materials .partialle b Thi n sca y achievee th y db modee th lf o presente e us d here. 122 STRATEGY AND POLICY NATIONAL POLIC EXPERIENCD YAN E WITE HTH MANAGEMEN RADIOACTIVF TO E WASTES FROM NON-FUEL CYCLE ACTIVITIES IN THE CZECH REPUBLIC J. HOLUB, M. JANÛ Department of Ecology, NYCOM, Prague, Czech Republic Abstract Research, production, and application of radioisotopes in many fields of science, industry, agriculture, medicine, education, etc. proceeded in the former Czechoslovak Republic (CSFR) sinc mid-fiftiese eth . These activities resulte greaa n di t accumulatio f relativelno y large volumes and activities of radioactive wastes.Therefore, in 1959 the Czechoslovak government appointed the Institute for Research, Production, and Application of Radioisotopes (IRPAR) NYCOMe centraw th no , e b l o authoritt , r collectiofo y disposad an n f theso l e radioactive wastes 197n I . 2 these responsibilities were define morn di e decree detaith y b le of the Ministry of Health of the Czech Republic No. 59/1972 on the protection of public health agains effecte th t ionizinf o s g radiation. From the very beginning the services for collection, transport, and disposal provided by IRPAR (NYCOM) were based on the concept of waste concentration and their safe disposal in well-controlled facilities. The aim of disposal is to guarantee that man and his environment wilt sufferno l , neithe futuree th t presen ra n i , r frono t m these wastes .achieve s i Thim sai d by isolation of radioactive wastes from the human environment by a system of multiple barriers for a sufficiently long period of time to allow activity to decay below acceptable limits disposae Th . radioactivf o l e waste centrae th n si l repositories starte 1959n di , whee nth first repository located nea village rth e Hosti Beroue th mn operationi n i n t Districpu s .wa t operationae Th l perio thif do s repositor closes endes wa y wa t 1965n di 196n i d i d 3.an At present, there are other two repositories in operation. The repository Richard serves for disposa f wasteo l s containing artificial radionuclides, i.e., nuclides with induced radioactivity and fission products. The repository Bratrstvi serves for disposal of naturally occurring radionuclides, i.e., nuclide uraniuf so thoriud man theid man r daughter products. 1. INTRODUCTION The NYCOM was given the responsibility for collection and disposal of institutional radioactive wastes which mean the wastes from applications of radionuclides in various branches of industry, medicine, agriculture, etc. The effective protection of public and biosphere from the potential hazards arising from these wastes is the main objective of radioactive wastes management. Many investigations and efforts in this field have led to the general agreement that underground disposal, wite wasteth h s suitably immobilized an d isolated, can provide adequate protection for man and his environment for a sufficiently long perio f timer do disposa Ou . l strateg compliancn i s yi e wit e internationallhth y accepted standards and requirements. The characteristics of the waste types led to the choice of disposal 125 in rock cavities. The most convenient and economical options were abandoned mines. The philosoph f separato y e disposa f radioactivo l e wastes containing artificia d naturaan l l radionuclides has been applied. Transport and disposal rules are given by the legislative regulations of the radiation hygien safd eean transport repositoriee th s A . s were buil abandonen i t d mines respective th , e mining regulations had been applied as well. In 1991-94, a safety assessment for the individual disposal sites have been started. The evaluation of present conditions of the disposed wastes, the technical state of the repositories, present qualities of the natural and engineered barriers, and the evaluation of the possible impac disposee th f o t d wasteenvironmene th n smaio e th ne objectar t thesf so e studies. 2. CONDITIONING AND TRANSPORT OF RADIOACTIVE WASTES Radioactive wastes are transported and disposed of by the NYCOM staff. The wastes mus preparee b t r transportatiodfo disposad nan l accordin e "Acceptancth o gt e Criterir afo Radioactive Wastes" endorsed by the General Hygiene Office of the Czech Ministry of Health.The present acceptance criteria define technical requirement wastr sfo e conditioning and organizational and legislative relations between NYCOM and the institutions producing radioactive wastes.The document define classificatioe sth characterizatiod nan wastesf no e ,th method theif so r treatment, conditioning preparatiod ,an transporr nfo disposald tan , dutied san responsibilities of the waste producers and NYCOM, and general and legislative regulations. At present double-conditionine th , concretn gi principae th s ei l requirement firse th t n I . step the wastes, packed usually in plastic bags, are put in a 100 L inner barrel and fixed with concrete. Soft solid wastes are pressed in 100 L barrels prior to cementation. Then the inner 10 barre0L l wit fixee hth d waste places si externaL d0 int20 ola barrel whic provides hi d with a 5-cm thick concrete layer at its bottom. The free space between the two barrels is fille concretef do witm c barrele h5 Th .mad e sar steef eo l plated from both sides wit hzina c laye sealed ran d wit hhermetia c lidsurface .Th barrelf eo paintes si d with bitume epoxyr no - bitumen qualit e concrete Th th . f yo e should conform wit requiremente hth technicae th f so l standard CSN 731201-86 B 12.5. The upper part of the concrete layer should be painted with bitumen after 14 days of concrete hardening. Wastes in the inner barrel should be distributed and provided with shielding so that, at a distance of 5 cm from the surface of the external barrel, the dose rate equivalent would not exceed 1 mSv/h. The contamination on the surface externae th f o l barrel shoul t exceedno kBq/md3 naturar 2fo alpha-toxid an l c radionuclides, 0 kBq/m30 r explicitlfo 2 y specified low-toxic radionuclide 0 kBq/m3 d r othean fo 2s r radionuclides. Liquid wastes, that cannot be easily solidified, are transported in a special tank to a treatment facility for evaporation and fixation of concentrates in cement in 200 L drums. Organic solutions, especially those containing 14C and 3H, are fixed in cement containing synthetic resin VAPE absorXo t b organic compounds. Special attentio gives ni biologicao nt l radioactive wastes. They are incinerated and the ash is fixed in cement. When this procedure t possibleno s i , the treatee day0 3 ar y formaldehyden r i s dfo t intpu ,o plastic bags with chlorinated lime and fixed with concrete in steel barrels. The handling of spent radiation sources used in various applications, e.g., in oncology, industrial gauges, well logging, fire detection devices, etc., is another special problem. These 126 sources contain mostly 60Co, 137Cs, 241Am and 226Ra. In CR a special system of safety supervision on the movement of these sources was adopted. Each organization intending to sealee us d radiation sources must obtai nlicensa hygienissuee e b th o ey t d b e authoritied san by the State Office for Nuclear Safety. Each sealed source is provided with a certificate of the producer. This certificate is transferred to NYCOM together with each discarded spent radiation source senr disposalfo t . Afte activite th r decreases yha d belo e givewth n level, radiotherapeutical 60Co source reusee sar other dfo r purposes, e.g. technologican ,i l irradiators, etc. For such radiation sources, a special tube bunker was built in one of the chambers of the repository Richard. Radiation sources destined for ultimate disposal are fixed with their shielding containers in concrete in 100 L barrels. If necessary, barytconcrete, instead of normal concrete, could be used in order to reduce the dose rate on the surface. Then the 100 L barrel spacbarrele L th place e 0 d esar 20 betwees an n di n their wall filles si d with concrete or barytconcrete.The barrels containing spent sources with artificial radionuclides, such as Co, Cs, Pu, Am, Kr and others are disposed of in the repository Richard and the barrel60 s137 containin 239 g 241source s85 with natural radionuclides, such as 226Ra, 210Po, etc., are disposed repositore th n i f o y Bratrstvi. presene th U o pt t time about 4.10sealef o q dB sources have beene disposeth n I . dof 16 period 1977-199 followine 3th g isotope f differenso t activities were collected: 60Co - l,2.1015Bq I37Cs - 3,6.1014Bq - 3H target3,2.10s13 Bq - l,8.10I2B85qKr 192Ir - l,0.10I2Bq 204T1 - 6,0.10"Bq 239Pu - 2,8.10nBq - 2,4.10nB90qSr 241Am - 2,3.10" Bq I47Pm - 2,2.10uBq 226Ra - I,0.10"Bq 238Pu - 3,2.10IOBq The activit f otheyo r radionuclides, originally used mostl s dosimetriya c standards, ranges from 2,4.10 Bq to 8.10 Bq. The following radionuclides as such standards were disposed of: C, Se3 , Pb, Pb9 , Ru, Ce, Fe, Zn, Cf, Mn, Ba, Po and Ni. 14 75 210 109 106 144 55 65 252 54 133 2IO 63 A special truck is used for transport of conditioned wastes. Liquid wastes are transported in a special tank for liquids. . CHARACTERIZATIO3 REPOSITORIEE TH F NO S centrae Inth l repositories requiremene th , f wasto t e isolation fro environmene mth s i t realized through a system of multiple barriers. The basic barrier is the immobilized wastes themselves. Another barrie provides ri packaginy db barrelo tw concrete d n gi san . However, the geological formation hosting the repository and its isolation characteristics are the most important barriers from the viewpoint of the long-term safety. According to the waste characteristics and its disposal procedure, there is no threat of any significant release of radioactivity to the environment. The natural barriers provide reliable shielding against the increased radiation level. The repository operation requires careful checking of any 127 contaminatio drinkine d sitth an e f o n g water sources y imaginablAn . e accidene th n i t repository should not constitute a significant threat to the population living in the vicinity of the repository. REPOSITORE TH . 4 Y HOSTIM repositore Th y Hosti establishes m wa t int pu o d operatiodan 1953n i . Constructiof no this repositor d beginnin an e ywaste-disposa th f o g l service s e initiateth wa s y b d government.The repositor situates yi galleriee th abandonee n dth i f so d limestone minee Th . repository was closed in 1965 by the decision of the local Hygiene Office of the Central Bohemia Region. Before the repository was closed, the containers (barrels, packages) with higher activities were transferre repositore th o dt y Richard. The characteristics of the repository Hostim (gallery B) are as follows: waste th - e3 volumm 0 11 gallern ei : yB - numbe packagef o r : s 2000 somd an , e unpacked wastes - total remaining activity: 0.1 TBq - predominant radionuclides: 3H, I4C, 60Co, 90Sr, 137Cs totae th l- volum gallere th f : eo 120yB 3 0m Simila disposere datth n ao d activit valie ytotaar e gallere d Th th lals r volum. oyfo A e of this gallery is about 400 m3. This repository constitutes no real danger to the environment.This conclusion will be verified by the on site monitoring system. The proposal of the final solution for the repository Hostim assumes that the remaining wastes will be left in place. After all these activities are finished, the entire volume of the repository will be filled with inert material consisting of clarepositorcemene d th yan d an t y wil sealede b l detaiA . l study evaluatin presene gth t state and future solutio f thino s repositor beins yi g worked out. 5. REPOSITORY RICHARD, LITOMËRICE The repository Richard near Litomëfice has been in operation since 1964. It was built as a relatively large-capacity repository for low and intermediate level radioactive wastes and spent sealed abandonesourcee aree th f th a o n si d limestone mine Richar . DurindII g World undergrounn Waa , II r d factory workin militarr gfo y purpose situates swa thin di s minee Th . cost for the adaptation amounted to more than 10 mil. Czech crowns. However, only a part of the mine Richard II was used as the repository. Up to the present time the total volume of the disposed radioactive wastes amounts to about 2700 m3. The underground water level is approximately 50 m below the disposal modules in a sandston repositore e th laye d an r continuousls yi y monitore possiblr dfo e contaminatiof no water, land and air. e repositorTh y Richar I acceptdI s only wastes with artificial radionuclides s totaIt . l volume th s i thi f 3 o e st m availabl volumfigur2 Ou . 61 3 e 8 disposam r ee 4 fo capacit 68 6 l1 s yi and 8 072 m3 is used for communications (gangways and corridors). By 1993 about 5 200 m3 were filled with wastes so that about 2 800 m stil3 l remain free for disposal. As the filling facto s aboui r t 40% remaim , 0 aboun 12 stil 1 t l frer emplacemenfo e f futuro t e packed 3 wastes. 128 The total disposed activity is estimated in the range of 10 Bq. Out of this amount about 16 31% of the activity is Sr, about 31% Am, 30% Co, 3% C, 2% H, 1% Cs, the rest being small amountf so Pu 90, T1, C1, Se 24I, Co, Kr, 60 Ce, Na 14, Ba, 3 Y, Mn 137, Ca, Fe, Zn, Pu, 239Cd, Cs204 . Abou 36 t 7595% o57f the 85 total 144activit 22y is i n133 the for 88 m 54 of seale 45 d sources. 55 65 238 109 I34 Fro remainine mth g unseale activite , d1.7th H f source3 o %s yi 137 % Css96 , 1.4%14 C, 0.7% Pu, 0.6% Ce, 0.3% Co and 0.2% Am. From these figures it is evident that tritium 239 is one of th e144 most importan 60 t radionuclide 241s from the point of view of radiation hygiene, especially because of its volatility and difficulty with immobilization. Other critical nuclides are 241Am, 239Pu, 137Co, 14C and 90Sr. A project has been worked out for the reconstruction of another 2 800 m3 capacity but beet ye nt approveno s locae iha tth ly db authorities . Fro recene mth t hydrological studiest i , follows that the isolation characteristics of this site are relatively good. The underlying geological bed of the site is formed partly by marl. Small amounts (several litres per day) of mine water flow out during the whole year. The repository has not yet been equipped with a full monitoring system. Such a system is now being developed. However, the packages are in a relatively good condition. While recent studies revealed that there is no immediate threat to the surrounding environment, it is necessary to take the following measures in order to be abl guaranteo et e full radiation safety: revisioe statith f co n condition e repositorth n i sd makine necessarth an y l al g y adaptations; revision and repair of the drainage system; building a central isolated retention basin for the accumulation of mine water; establishin f systematigo c monitorin repositore th mine n th i ef r d g o ai wate yan d an r outside; completing of deep monitoring system in agreement with the hydrogeological studies presene th mado t tp etimeu ; supplyin repositore gth y wit modere hth n equipmen r saffo te handlin wastese th f go ; Some of these measures have been already taken. A detailed preliminary safety study has been worked out that will be used for the enhancement of radiation safety of the radioactive wastes disposal in this site. 6. THE REPOSITORY BRATRSTVÎ, JÂCHYMOV repositore Th y Bratrstvï Jâchymo gallere buils abandonen th a vwa n f i t y o d uranium mine with five chamber r disposal.lfo s s i appointet r wastefo d s containing natural radionuclides, predominantly Ra, Po, Pb and uranium and thorium isotopes. The wastes 226 210 210 also contain spent sealed sources and neutron sources, mostly with 226Ra and 210Po.The main reasoseparatioe th r nfo thesf no e wastes from other waste radoe th s sni emanation that would cause serious problems in the repository Richard. basie th cose f cTh o tadaptatio formee th f no r uranium mine amounte abouo dt 2 1, t million Czech crowns repositore Th . bees yha operation n i n since 1974. characteristice Th repositore th f followss so a e yar : the volume of disposed wastes is about 250 m3; the free capacity is still about 40 m3; 129 repositore th y wil completele b l y fille aboun di years3 t ; disposee th d activities are: about 1012 B226f qo Ra , 109 B232f qabouo d T han t 109Bf qo other radionuclides. possibilite Th repositore th f yo y expansio bees nha n rejecte locae th ly db authoritie r sfo the time being license Th controllable . bases ei th n do epackagese statth f eo safete Th . y analysis shows that this mine canno consideree tb stabla e b eo d t syste m wit time hth e horizon t leas yearsa 4 f perio10 e to th , d necessar substantiaa r yfo l decreas 226e Rth af e o activit y with the half-lif f 160o e 0 years acceptance Th . repositore th f eo face bases yi th t n thado t only a small part of the activity that was originally mined from this locality has been returned to the repository. followine Th g provision increasinr sfo repositore safete th gth f yo beine yar g made: a tracer examination of the underground water movement using artificial radionuclides; hydrochemical determination of the mine waters, mineral waters and surface waters from poine th vief o t hydrochemistryf wo ; estimatio engineerine th f no geologicad gan l stabilit repositorye th f yo . . ECONOMICA7 SAFETD LAN Y CONSIDERATIONS pase Inth t transpor disposad an t f wasteo l provides swa d fre f chargeo e coste f Th . so this service were fully covered by the state. The average costs amounted to about 3 million Czech crown yeare r yearth spe n s I .1991-9 3 onl inevitable yth e costs connected wite hth operation and maintenance of the repositories in the amount of 1,5 million Czech crowns were covered by the state. The costs connected with the transport, treatment and conditioning of wastes wer waste paith y ed b barre producersL 0 wastef o l20 charge e son Th . r amountefo s to 10 000 Czech crowns. It is evident that the charge-free service led to the uneconomic operation. However, abnormal increas complete th f feer eo fo s e operatio maintenancd nan e e repositorieth f o requestee th n o s d level woul e dincreasn th a e learis f th o o kdt f o e uncontrolled disposa f radioactivo l e waste environmente th n si . Considerin rathee gth r imminent exhaustio presene th f no t capacit repositoriee th f yo s necessars ii t wore speeo yt th othe n kp o d u r sites desigo t , construcd nan repositoriesw ne t . t woulI f greao e tdb advantag f disposaei f institutionao l l radioactive waste s solvei s n di connection with the wastes from the nuclear fuel cycle because both are faced similar problems and the combined solution can result in economic savings.This is true not only for treatment and conditioning of wastes but also for final disposal. It is necessary to take up the following strategy in the operation of these repositories: guaranteo t e operatio mine th f en o systems including their safety; to maintain the packages in a good and transportable state; operato t d builo an t e p reliabldu e monitoring systems; to establis maintaid an h n good relations wit e locahth l authoritief o m s ai wit e hth achieving the atmosphere of trust and constructive cooperation; fino t d way provido st locae eth l municipalities with reimbursemen operatioe th r fo t n of the repositories on their respective territories. 130 FIG. 1. The mobile compactor for volume reduction of solid radioactive wastes. FIG. 2. The view into one gallery of the repository Richard with stored drums containing radio- active wastes drume . prepae th Som ar sf o e - red for final loading. 131 REFERENCES [1] HOLUB , JANÛJ. , , MARSALM. , , "ThJ. , e Institutional Wastes Managemenn i t Czech Republic", Proceeding e 199th f 3so International Conferenc Nuclean eo r Waste Managemen Environmentad an t l Remediation, Prague, Vol (1993)337.3 . [2] JANÛ , MARSALM. , , HOLUBJ. , , "SecurinJ. , f Safetgo Maintenancd yan e th f eo Repository Richard", IRPAR Repor . DE/1/9No t 3 (1993). 132 RADIOACTIVE WASTE MANAGEMENT POLICD YAN ITS IMPLEMENTATION IN INDONESIA S. YATIM Radioactive Waste Management Technology Centre, National Atomic Energy Agency, Kawasan Puspiptek, Serpong, Indonesia Abstract Sinc establishmene eth Nationae th f to l Atomic Energy Agenc 1958n yi , several nuclear research centres have been establishe o carr t dt researc ou y d developmenan h e th r fo t promotion of nuclear energy in supporting the national development programme. To realize such objective firse sth t research reacto operation i Trigf t o r pu a s 196n Marni d wa 5an I k I flowed by the establishment of Pasar Jumat and Yogyakarta research centres. To support further developmen nucleaf o t r energy programme several nuclear facilitieW M 0 s 3 suca s ha research reactor, radioisotope production, fuel element researc fabricatiod han centralized nan d radioactive waste treatment facility are put in operation at Serpong site in 1988.Through out developmene effore th th n o t promotiod tan nucleaf no r energy programme, healt safetd han y of the workers, population and the environment have been the primary concern as it is stated in National Act No.31 of 1964 on Basic Provision on Atomic Energy. Based on the act, the radioactive wastes generated from nuclear programme should be treated to minimize its harmful effect to population and environment. To meet such a requirements, many works and efforts have been directed toward formulating the national policy of radioactive waste managemen implementations it d an t papee Th . r present broasa d vie nationaf wo l policd yan programme of radioactive waste management and its implementation to support the safety aspect of the present and future development of nuclear energy programme in Indonesia. 1. INTRODUCTION Sinc operatioe eth firse th tf nnucleao r research reacto 1965n ri , several nuclear research facilities have been established to develop and promote nuclear programme in supporting different area f nationaso l development programme t presentA . , nuclear programm stils ei l limited to research and applications activities, and in the near future will be extended for electricity generation. e efforth o promott n tI a nucleae r programme, healt safetd an hf worker o y d an s populatio protectiod nan environmene th f no t have bee nprimara y consideratio states i t i ds na in the national Act No.31 (1965) on Basic Provision of Atomic Energy and Act No.4 (1982) on Basic Provision of Environmental Management [1,2]. Based on the acts, radioactive waste generated from nuclear activities should be treated to minimize its radiation effect to the population and the environment. To accomplish such an objective, several steps have been taken in establishing a national structure for radioactive waste management. 2. POLIC STRUCTURD YAN E The basic polic radioactivf yo e waste managemen Nationae th n t i p beeAc ls u t tha nse No. 31 (1965) on Basic Provision of Atomic Energy and Act No. 4 year 1982 on Basic Provision of Environmental Management, and implemented in several regulations such as the Governmental Regulatio n Operationao n l Safety f Radioactivo e , LicensinUs e th e f o g 133 Material Radiatiod san n Sources, Transportatio Radioactivf no e Material Préparatioe Th d san n of Environmental Impact Analysis [3-6] basie Th c. policy radioactive wast Indonesin ei s ai as the following: 1. Generation of radioactive waste from nuclear activities should be minimized. 2. Discharge liquif s o gaseou d dan s radioactive wastenvironmene th o et w t lo shoul s a e db as possible. 3. Environmental aspects of handling, treatment and disposal of radioactive wastes should takee b n into account. 4. Solidified and solid waste should be emplaced in a facility specially constructed for that purpose. . 5 Radioactive waste management problems shoul takee db n into account befor largey ean r nuclear programme is carried out. 6. Research and development on radioactive waste management should be carried out to support the safety of present and future nuclear programmes. e presenTh t structur radioactive th f eo e waste managemen s showi t Fig.ln ni e Th . structure comprises promotio regulatiod nan n sideMinistre .Th Sciencf yo Technologyd ean , the Atomic Energy Council and the National Atomic Energy Agency set up research and development programme to be implemented. On the regulatory side, the Atomic Energy Control Burea Nationae th f uo l Atomic Energy Agency Ministre th , f Healtyo Statd han e Ministr Environmenf yo t provide regulations, guideline criteri d safete san th n yao aspectd san contro f theio l r implementation addition I . theso nt e organizations Radioactive th , e Waste Management Technology Centre acts as a supporting organization and carries out research and development in the radioactive waste management field. Several steps are being carried out to separate the Atomic Energy Control Bureau from the National Atomic Energy Agency and put it as an independent organization under the Ministry of Science and Technology. ATOMIC ENERGY COUNCIL MINISTRY OF HEALTH MINISTRY OF SCIENCE & TECHNOLOGY STATE MINISTRY F ENVIRONMENO T NATIONAL ATOMIC ENERGY AGENCY CONTROL SI DE PROMOTION SIDE ATOMIN ENERGY NUCLEAR RESEARCH CONTROL BUREAU AND APLICATION SUPPORTING ORGANIZATION RADWASTE MANAGEMENT TECHNOLOGY FIG. 1. RADWASTE MANAGEMENT STRUCTURE 134 3. CURRENT PRACTICES 3.1. Waste source generatiod san n At present, mos f radioactivo t e wast generatee ear d from nuclear research activities carried out by the National Atomic Energy Agency in several research centres located in Bandung, Yogyakarta, Jakart Serpongd aan . Small amount f radioactivo s e wast e alsear o generated from application of nuclear techniques in medicine and industry. The total amounts of radioactive waste generate 199n di show3s i Tabln i . eI TABLE I. WASTE SOURCES AND WASTE ARISINGS IN 1993 Nuclear Research Waste Types Amount Liquid (low level) 520m3 Solid - Compactable (low level) 197 drums (10) 0L Burnabl- e (low level) 39 drums (100 L) - Other (lo mediud wan m 28 drums level) Applications Solid (low level) 6 drums (100 L) Spent sources - Cs-137 2i lea8h d containers - Sr-90 10 in lead containers levecontai d wastew e Mosan lo th l f e no tsar short-lived radionuclides. These wastes consist of contaminated process equipments, used filters, protective devices and concentrates and sludges from the liquid waste treatment. Small amount of intermediate and high level wastes are also generated from radioisotopes production and radiometallurgy facility. From application activitie wastee sth mainle sar y spent radiation sources radioactive .Th e wastee sar treated at a centralized waste treatment facility by different techniques and then solidified in a cement matrix. 3.2. Centralized waste treatment facility Before the establishment of the Serpong nuclear industrial research centre the quantity of radioactive waste generated from nuclear researc relativels hwa y small, consistin lowf go - level activit mostld yan y contain short half-life radionuclides.The treatmen thesf o t e wastes is simple through deca d dela d needan y an y s simple process e equipmentth o t e Du . developmen nucleaa f to r programm orden i d givo rt e an e better service radioactivn si e waste management centralizea , d treatmen operation ti facilite Serpont th pu n s i nyo g site.This facilit operates yi d unde Radioactive rth e Waste Management Technology Centre whic alss hi o responsible for research and development needed in the radioactive waste technology field. The centralized radioactive waste treatment facility consists of three buildings: the waste treatment, interim storag powed ean r supply buildings procese Th . s building accommodates liqui spend dan t resins storage tanks capacityh evaporatoL/ n a ,0 incineraton 75 a , a f o r f o r 135 200 kg/h capacity cementatioa , compactoa , kN 0 60 nf o runi t witL 0 hcapacita 20 6 f yo drum laundrdayr a sd pe ,an y syste cleaninr mfo decontaminatiod gan protectivf no e devices. The process system is also equipped with a transportation unit for solid and liquid wastes and spent resins. Interim storage buildin spaca s accommodatn geha ca ared f 150aan o 2 0m 0 e50 e powe Th d 150 . an rL 0 supplL 0 drum 0 20 concret 35 f y o s d buildinan e 0 shell95 g f so provides steam, compressed air, electrical powe auxiliard an r y support. 3.3. Radioactive waste management programme The radioactive waste management technology scheme used for different waste forms and categories is shown in Fig. 2 and summarized in this section. Low and medium level liquid wastes. Low level liquid wastes are treated by chemical methods and evaporation. Precipitation technique is mostly used as a chemical method to treat low-level carries wastadditioi y b d t ean d ou chemicalf no s after adjustinsludge Th . s ei gpH filtere solidified dan d with cemen drumL t 0 mixtur10 . Steaa n ei m evaporatio uses ni o dt concentrat liquie eth dconcentrate wastth d ean solidifies ei d with cement slurr concreta n yi e capacityL 0 sheleffluent95 e f Th .o l s fro liquie mth d treatmen collectee ar t controlled dan d before release dichargee Th . s shoul beloe db authorizee wth d discharge limits. Low and medium level solid wastes. Solid waste consists of different materials. These waste e classifiear s s compactablea d , non-compactabl d burnablan e e materialse Th . compactable material drumL collecte e compacte0 d sar 10 san a compacton a di y db 0 60 f ro drumL the0 d s20 an nk Nn solidifiei d with cement burnable Th . e materia x packes i l bo a dn i of 30 x 30 x 60 cm and incinerated at a temperature of about 950°C.The ash is then collected in a 100 L drum and solidified with cement. Non-compacted material is placed into a concrete capacitL 0 solidified 35 shel yan f o l d with cement. Spent resins, after pretreatmente ar , FUEL FABRICATION n & FUEL RESEARCH STORAGE TANK ACTIVITIES LL & ML V»S TE RESEARCH REACTOR r SPENT FUEL OPERATION EVAPORATION COMPACTION RADIOISOTOPES VITRIVICATION 3) INCINERATION __ CHEMICAL PRODUCTION PROCESS APPLICATION NUCLEAR CEMENTATION RESEARCH INTERIM 2) MEDICAL STORAGE — INTERI—— ' 1 M & INDUSTRIAL STORAGE APPLICATION 1 SHALLOW 1) LAND 1) DESIGN REPOSITORY ) UNDE2 R CONSTRUCTION ) PLANNIN3 G FIG . RADWAST2 . E MANAGEMENT PROGRAMME 136 solidified with cement in a concrete shell of 350 L capacity. Spent radiation sources from nuclear medicin industrd ean solidifiee yconcretar L 0 drum35 L d n ei 0 eithe shellr s20 o n ri . Some radiation sources, especially from industrial radiography t treate no use d e ds an ar d,a radiation sources for different research activities. High level waste spentd an fuel. Small amoun f higo t h level waste s alssi o generated from productio Mo-9f no radiometallurgd 9an y laboratory.The wastes contain fission productd san transuranium (TRU) radionuclides.The waste containing TRU radionuclides is solidified in borosilicate glass in high integrity stainless steel containers and stored in steel lined underground vaults specially designe thar dfo t purposes.Immobilizatio thif no s wast inern ei t materials is necessary for its long storage. High level waste which contains different fission products is placed in a temporary storage facility for decay and after the activity has decreased certaia o t n level solidifies i t i , d with cemen concreta n ti capacityL e0 shel35 f o l. Spent fuel is stil centralizea l d storereactoe an th y n di rba d spent fuel storage wate underw poono rs i l construction. Engineered Storage. solidifiee Th d waste interie storee th ar s n di m storage buildinge .Th buildin constructes gi brickf do concretd san e with thic wallm c provid o kt 0 s4 e shielding.The buildin divides gi d intareaso aree otw uses ai On . storo dt e solidified low-level wasted san has a surface area of 500 m to accommodate 1500 200 L drums.The other area is used to store medium and high level2 solidified waste and it has a surface area of 1000 m2 and can accommodat concret0 e50 e shells.The buildin designes gi year0 3 r se d fo b operatio n ca d nan extended to allow additional storage capacity. . FUTUR4 E PROGRAMME Several effort increaso t s capabilite eth radioactivn yi e waste management have been undertaken for the present and near future needs. Several techniques are being adopted and evaluated to immobilize alpha-bearing waste in glass matrix. Several steps are also being carried out to increase the safety in handling of high level waste from radioisotopes productio radiometallurge th d nan y facility establishmen e studA th . n yo demonstratioa f o t n shallow land repository on the Serpong site is also being carried out. The study is necessary to gam the experience in different aspects of shallow land disposal. Another programme whic s alsi h o bein e techno-economicagth carries i t ou d l stud f radioactivo y e waste management for a nuclear power plant. Training of personnel in different aspects of radioactive waste management is also scheduled through bilateral or international co-operation. A long term programme is also directed towards the demonstration required to establish a deep geological disposa higr lfo h level waste. Several steps towards strengthenin nationae gth l infrastructure of radioactive waste management and enhancing the safety of radioactive waste managemen recommendes ta IAEe th Ay db RAD WA SProgrammS e [7,8] wil anticipatee lb d in the near future. 5. CONCLUSIONS Radioactive waste management has been introduced in the beginning of the nuclear activitie becomd san integran ea nucleae l parth f o t r programme. Several effort differenn i s t aspect f radioactivo s e waste management have been undertake d implementenan e th n i d nuclear programme. In terms of the magnitude of the radioactive waste management problems, the action taken is adequate to ensure the safety of radioactive waste management fro presene mth t nuclear activities futuree th n I ., many works have been planne increaso dt e 137 the national capability in different aspects of radioactive waste management. To achieve such a goal international co-operation under the IAEA programme is one of the important aspect to be considered. REFFERENCES f Republio t Ac f Indonesico ] [1 a No.31 year 196 Basin 4o c Provisio Atomie th n no c Energy. [2] Act of Republic Indonesia No.4 year 1982 on Basic Provision on Environmental Management. [3] Government Regulation No. 11 year 1975 on Working Safety Provision Against Radiation. [4] Government Regulation No. 12 year 1975 on the Licensing of the Use of Radioactive Material Radiatiod san n Sources. ] Governmen[5 t Regulatio yea3 1 . Transportatioe r nth 197No n o 5 f Radioactivno e Materials [6] Government Regulation No.29 year 1986 on Environmental Impact Analysis. [7] INTERNATIONAL ATOMIC ENERGY AGENCY, The Principles of Radioactive Waste Management, Safety Series, No. 111-F, IAEA, Vienna (1994). [8] INTERNATIONAL ATOMIC ENERGY AGENCY, Establishing A National Legal System for Radioactive Waste Management, Safety Series No. 111-S, IAEA, Vienna (1994). 138 THE FUNDAMENTAL RUSSIAE TH F SO N FEDERATION NATIONAL POLICF TH N YI NON-NUCLEAR FUEL CYCLE RADIOACTIVE WASTE MANAGEMENT E.M. LATYPOV, V.A. RIKUNOV Federal Radiation and Nuclear Safety Authority, Moscow, Russian Federation Abstract Extensive manufacture and use of sources of ionizing radiation result inevitably in the generation of a considerable amount of radioactive waste. The crucial objective within the contex generae th f o t l proble radioactivf mo e waste management involve safe sth e isolation of radioactive waste fro environmene mth entire th r efo tperio e existencth f do f theieo r potential hazardous impacts upon it. The complex nature of the problem requires substantial efforts to be placed for the establishment of an integrated radioactive waste management system providing a national control in medicine, industry and science. To this end, the fundamentals of the national policy safe foth re managemen radioactivf to e waste from non-nuclear fuel cycle activitie beine sar g develope Russiae th n di n Federation.The essential component nationae th f so l policy are: development of a scientifically sound concept of radioactive waste management; adoptio f legislativno e documents suc standards ha actsd san , relevan thio t s area; implementation and enforcement of state regulations and supervision of the relevant activities; development of a national programme on radioactive waste management; provision and maintaining of a national radioactive waste inventory; radiation monitoring. safe eTh radioactive waste management concept elaborate Russiae th n di n Federatios ni based upon a multi-barrier defense in depth system and is targeted at the substantiation of such control methods whic eliminatn hca e negative impact human so n healte wels th ha s a l environmen future th n ei takind tan botgw inthno o accoun sociaf o t economid an l c factors. For example e concep e isolatioth , th f o tf hig o n h level radioactive wast n deei e p geological formations has been considered as safe in the scientific community. The concepts intermediatd ofan disposaw lo f o le level radioactive wast nean ei r surfac intermediatd ean e depth repositories have been developed. While elaboratin above gth e concepts e IAEth , A recommendations shall be taken into account, in particular, the disposal method should be governe radionuclide th y db e compositio specifid nan c activit radioactivf yo e waste. 1. LEGISLATIVE DOCUMENT STANDARDD SAN RADIOACTIVR SFO E WASTE MANAGEMENT Unfortunately o ,specia n ther e ar le legislative acts stipulatin e rightth d g an s responsibilities of the Federal executive agencies and operators regarding radioactive waste managemen Russiae th n i t n Federation. Nevertheless, some section followine th f so g actf so e Russiath n Federation Environmene th n "O : t Preservation" Sanitare th n d "O yan , 139 Epidemiological Welfar Generaf eo l Public" Entrailsn usee "O b , executdo n t "ca e legislation in the area under consideration. All in all, the legal framework in the radioactive management area include lawse sth Presidente , actth f so Russiae th , n Federation State Agenciee th d san Russian Federation Subjects wels a ,nationa s la l standards, special norm rulesd san , normative technical documentation of various branches of industry. 2. NATIONAL CONTROL AND SUPERVISION OF THE RADIOACTIVE WASTE MANAGEMENT In compliance with the Law, the Russian Federation State Control Agencies have been created within the executive power system to deal with the issues of immediate execution of the functions on the provision of public and environmental radiation safety. Among various state regulatory and supervision bodies in this field are the State Committee for Sanitary and Epidemiological Supervision, the Federal Radiation and Nuclear Safety Authority, the Ministry of Preservation of Environment and Natural Resources. The State Committee for Sanitary and Epidemiological Supervision executes the national standard-based regulation speciae th s lwels ha s a authorizatio s a l controd nan l functiono t s provid sanitare th r epidemiologicad efo yan l well bein Russiaf go n Federation citizens. The Ministry of Preservation of Environment and Natural Resources is the central agenc federae th f yo l executive authority accomplishing contro fiele preservatiof th do n i l n of the environment and natural resources. The Federal Radiation and Nuclear Safety Authority is involved in all efforts for the organization and implementation of the state regulations and supervision over safety in the area of nuclear power generation, nuclear applications and waste management aimed at assurin personnee safete gth th f yo nucleaf o l radiatiod ran n hazardous facilitie generad san l public, at protection of the environment and safeguarding security of the Russian Federation. operatinn A g organization (operator facilitya r )o , which perform kiny activitiesf sdan o , pertaining to radioactive waste management, bears the responsibility for the radiation safety, personnel health protectio environmend nan t preservation. NATIONAE 3TH . L PROGRAMME RADIOACTIVN SO E WASTE MANAGEMENT Currently nationae th , l programmes dealing wit radioactive hth e waste management issues, namely "The Russian Federal Special Purpose Programm Radioactivn eo e Wastd ean Spent Nuclear Material Management, Utilizatio Disposad nan f theso l e Substancee th r sfo Period of 1993-1995 and up till the Year of 2005", "Radiation Rehabilitation of the Ural Region Territor Measured y an Assistanc e th n Populatiose o th o et n Suffered fro Effecte mth s of Nuclear Accidents", etc. have been adopted or under approval. The first of the above programmes envisages particularn ,i modificatioe ,th enlargemend nan existine th f to g regional radioactive waste disposal sites desige th ,constructio d nan onesw improvemene ne th ,f no t and developmen wastf o t e processing methods developmene th , manufacturd an t f wasteo e processing facilitie equipmentd san . However presene th , t economic situatio Russiae th n i n Federation makes it difficult to successfully implement the above programmes in the near future. 140 4. NATIONAL RADIOACTIVE WASTE INVENTORY inventore Th y constitute systesa appropriatelmf o y arranged dat data a( a base) regarding location and status of the radioactive waste management facilities and sites on the territory Russiae th f o n Federation inventore Th . intendes yi prompr dfo t provisio state th eno t contro l and supervision authorities and agencies of the information on the location and status of the facilities and sites with any activities in radioactive waste management, as well as the information on deterioration of radiation situation, caused by these activities. 5. RADIATION MONITORING radiatioe Th n monitoring system implemente Russiae th n di n Federation surveye sth parameter radiatiof so n situatio followine th t na g levels site-specifi:a c level regionaa , l level and a departmental level. In perspective, it is intended to set up a Unified State Automated Radiation Situation Survey Syste entire th n meo territor Russiae th f yo n Federation. 6. CONCLUSION Proceeding from the above, one can arrive at the conclusion, that the national policy of Russiae th n Federatio radioactive th n ni e waste management field cover widsa e rangf eo issue focused solutioe san problemth e t dth a f no s relate radioactivo dt e waste including their collection, storage and disposal. But, taking into account the current economic situation, the lac financiaf ko industriad an l l resource forces i e believo dt son e tha accomplishmene tth f o t e nationa e th objective th y b lt radioactivse s e waste management policy will present considerable difficulties. 141 THE HUNGARIAN RADIOACTIVE WASTE MANAGEMENT PROJECT AND ITS REGULATORY ASPECTS . CZOCI H Hungarian Atomic Energy Commission, Budapest, Hungary Abstract In 1993, a National Radioactive Waste Management Project was launched to handle, treat and dispose of LLW/ILW from the nuclear power plant. Within the framework of the projec complea t x strateg bees yha n elaborate managemene th r dfo l typeal radioactivf f sto o e wastes from the NPP, including HLW, spent fuel and wastes from the decommissioning of the nuclear power plant. The first phase of the project will be realized between 1993-1996 and thif o sd perioen possible be ydth th e sitsitesr e(o accepte)- publie th y dcwil- b selectee b l d for the LLW/ILW wastes. To this date the funding problem of radioactive waste management solvede b o t legas a , ha l framework shoul available db regulatord ean y requiremento t e sar be clarified. In 1980 a law on nuclear energy was promulgated in Hungary based on former governmental level decrees regulating the application of nuclear energy. It reflects the system o centrallfa y planned economy, facilitiewhere th l eal s were stat nuclean o e ownedw la r e Th . energy prescribes the solution of safe storage of radioactive waste as a prerequisite for licensino stepn t s bu gwer e take o providt n e special fundin r radioactivfo g e waste management. Now there is a new law on nuclear energy in preparation, and one of its most important chapters is dealing with radioactive waste management, the responsibilities for it and its funding. In the framework of the National Radioactive Waste Management Project a programme was initiated to elaborate the technical basis for the detailed legal regulations.The firs programmte steth f po definitioe th s ei exemptiof no n level wastd san e acceptance criteria. An analysis will be prepared to compare the Hungarian regulations with the internationally accepted requirement defino st e those areas where further regulations shoul e issuedb r do wher existine eth g ones shoul amendede db . 1. INTRODUCTION As soon as nuclear energy started to be used in Hungary, the relevant legal regulations came into force, first at the level of standards and medical norms, later followed by a comprehensive governmental decre handlinn eo radioactivf go e materials including radioactive wastes. A repository for final disposal of radioactive wastes has been in operation in Solymàr since the fifties. Later it turned out that safety of the repository is not sufficient and in 1976 the Hungarian Atomic Energy Commission opene repositorw ne da Püspökszilagn yi y where the wastes from Solymar have also been transported. Since that time this facility is used for final disposa e wasteth f o sl from application f nucleao s r energ n industryi y , medicine, agriculture, researc developmentd han t alsI . o accepted some solid wastes fro e Pakmth s Nuclear Powe r capacits thie Planit decidet us s facilito bu t t wa limite s yt i t i r dno d yfo dan the wastes from the nuclear power plant. 2. THE LAW ON NUCLEAR ENERGY AND ITS EXECUTIVE ORDERS The original concept of the Soviet design nuclear power plants was that radioactive waste remain facilite site s lonoperatioth n th ea i s preparationo n gs a i n yo s d nan s were 143 foreseen for the decommissioning. Therefore when four units of the Hungarian WER-440 Pakn i typP s were NP t int epu o operation between 1982-87 sitn o e n storaga , e capacits ywa provided for radioactive waste, but no final repository was looked for. In accordance with this situation the act on nuclear energy promulgated in 1980 requires only that a license for constructio operatiod nan nucleaf no r facilitie sgrantee shalb t lno d unless sufficient measures are taken for the safe storage of radioactive waste produced by it. executive Th definet ac e responsibilite ordee th sth f ro variouf yo s Minister fiele th dn si of application f nucleao s r e Ministeenergyth d f Publian ,o r c Welfar s authorizewa e o dt regulate disposa f radioactivo l e wastes Ministee ordee th f Th . o rf Publio r c Welfars wa e issue 1988n di t regulate.I s radiation safety, licensing procedur applicationr efo f nucleaso r energy other than nuclear facilities (research reactor nuclead san r power plants amon- d )an g others - the special requirements for final disposal of radioactive wastes. The ministerial order specified also those other authorities, who should be involved in the licensing process. In Hungary it is the general rule of administration that a licensing authority invites in its licensing procedure all those other regulatory organizations who are authorize mako dt e decisions relate subjec e procedure th th o df t o t e from some special point of view. The license can be granted only if all the involved authorities gave their consent to casradioactive a th f eo itn I . e waste repositor followine yth g authorities have responsibilities: Licensing authority is the Public Health and Medical Officer Service (on behalf of the Ministe f Publio r c Welfare). Other authorities participating in licensing procedures of the licensing authority rGeneral Inspectorate of Transport, National Headquarters of Fire Service and Civil Defense, Municipal administration, National Police Headquarters, Inspectorat f Environmeneo t Protectio Wated nan r Management, Veterinar Food yan d Control Servic Hungariad ean n Geological Survey. Some other authorities have also regulatory tasks in connection with the radioactive waste management, such as the Nuclear Safety Inspectorate of HAEC for waste collection, handling and conditioning on the site of the NPP and the Institute of Isotopes in international transportation, packaging and recording of radioactive materials. A facilit r finayfo l disposa f radioactivo l e waste s- otheliky ean r facilitie s alsi - os subject to the conventional licensing procedure. The relevant authorities and organs among others are the following: Municipal administration (utilization of land and construction of buildings), Mining Bureau of Hungary, National Agency for Nature Conversation, National Water Management Directorate, National Agenc Historir yfo c Monuments, etc. nuclean o w la r s executive energit Th d yan e t creatorderno o cleaea sd r connection between nuclea conventionad an r l licensing procedures sequence Th . takenf stepe eo b o st , co-operation of the authorities are issues that should be solved on a case by case basis. E NATIONATH . 3 L RADIOACTIVE WASTE MANAGEMENT PROJECT Though the amount of generated waste is much lower than the designed value, in accordance with international practic Paktriee P e th fin o dst NP disposada l site alreade th n yi eighties for its radioactive waste. This effort failed mainly because of the lack of public acceptanc n 199i d 3a nationaan e l projec s launchewa t o solvt d e problee th e th f mo 144 managemen radioactivP NP disposad e an t th f eo l waste decisioe Th . s takenwa n wite hth understanding that: the plant's operation must not be adversely affected by handling and storage of radioactive waste durin whols git e lifetime, quantite th volumd yan wastef eo s temporarily storenucleae th i dh r power plant should be as low as technically possible taking into account safety conditionin wastee th disposaf r disposage o s fo th d an l l itself should take plac soos ea n as possible, 3.1. Organizatio e projecth f no t Safwideld an e y accepted management, including final disposa f radioactivo l e waste requires co-ordinatio f scientifico n , economical, technical, social, legal, financiad an l international activities inter-ministerian a thio d T sen . l projec establishes wa t d undee th r leadership of competent ministries and authorities, such as the Ministry of Industry and Trade, Ministry of Environment and Regional Policy, Ministry of Public Welfare, National Committe r Technicafo e l Development, Hungarian Atomic Energy Commissioe th d an n Hungarian Power Company Ltd. A Project Governing Board was established with high level representatives of the competent ministrie Boare organsd th san assisted s di Advisorn an , a y db y Committee.The chairman of the Board is the vice-president of HAEC, the main contractor is the Paks NPP. Late independenn ra t institution will possibl establishede yb , responsibl constructioe th r efo n and operatio e repositoryth f o n o enhanc.T e solutioth e f regulatoro n y tasks a specia, l Regulatory Workin co-ordinato t p u g t Grou se activitiee s eth authoritiese pwa th f so . 3.2. Goals and results of the first phase of the project The first phase of the project (1993-1996) is aimed to determine the outline of the complex strateg r managemenyfo disposad l kinan t al radioactivf do f o l e wastes, including spent fuel and wastes from future decommissioning of the NPP and to select one or more sites for disposal of LLW/ILW. The activities and results in the main areas of the project are the following: The complex strategy for radioactive waste management was elaborated and accepted Projece byth t Governing Board subjec updatino t t finae th ln gi projectphase th f eo . The selection of procedures and equipment for treatment and volume reduction of radioactive wast completeds ewa . Solid wastes wil super-compactee lb d applying mobile service of relevant companies. For liquid wastes the Finnish technology was selected with boron recovery and Cs removal. Quick screenin countre th f gfino o t undes y i d y potentiawa r l region r LLW/ILsfo W disposal, near surfac undergrounr . eo m) 0 30 o d(t A Public Relation company was selected to elaborate a programme for enhancement of public acceptance and to assist the project management and the Paks NPP in their public relation activity. The possibilities to assure funding waste management (with the pricing of electric energ r otheyo r methods) undee ,ar etc.r consideration. 145 3.3. Development regulatioe th n si radioactivf no e waste management nuclean o w la r energw ne Ae y th drafdefine f o tresponsibilitiee th s e wastth f eo s producer tako st e car wastf eo e disposal t clearlI . y outline responsibilitiee sth waste th f eo s producers, the authorities and the Government. The Government shall create a radioactive waste managemen e constructiotth fundd an a repositor, f o n y shoul e decideb d n o d governmental level. In the division of responsibilities among the authorities no major changes will take place but their detailed licensing procedure requirementd san e stilar s l missing enhanco .T e eth elaboration of the complete licensing procedure the Regulatory Working Group made a proposal for a practically applicable procedure that avoids overlapping or omission. The Regulatory Working Group is compiling now all the regulations that the various authorities have relating to the licensing documentation and its acceptance criteria.The authoritie requestee sar enumerato dt e their written regulations, guidelines, directivef i r so case-law is applied the relevant precedents. In 1988, when the first attempt was made to select a site for LLW/ILW disposal-in accordance wit internationalle hth y accepted view-i s supposewa t d tharepositore th t r yfo LLW/ILW a shallo wile b lw ground repository n e i existinth , e juss a on gt Püspökszilagy.Therefore the ministerial order of the Minister of Public Welfare specified requirements, related to this type of facility - among others - such as the following: A shallow land disposal facility can be sited only in a geological environment acceptable fro poine m th vief o t tectonicsf wo , seismology distancm k t leas1 a , n etcd eo t froan . m larger living areas, recreational districts, surface waters (river, lake), dams, mined san factories producing dangerou explosivd san e goods. If natural parameters of the site are not quite adequate, the selected site should be improve engineerey db d structures. disposae acceptee Th b n finaonl a e lastt lca i s d a f on ylleast i s a t twenty time halfe sth - life of the longest lived dominant radio nuclide. post-sealine th n I g perio operatoe dprovidth o t supervisioe s th r rha efo facilite th f no y for monitorin radiatiof go environmene th n i preventiod tan intrusioe th f no personf no s animald t leasan a r t sfo fift y year afted san r that datauthorite lons th ea s ga y requires it. These randomly selected examples show that some basic requirements are defined, however furthe done r b woreo t wits ki h respec classificatioe th o t f radioactivno e wastes, the definition of exemption levels, waste acceptance and site selection criteria. In this effort we are supported by the RADWASS programme of the IAEA and - hopefully in the near future - by the relevant project in the PHARE programme (CASSIOPEE) where a CEC study was proposed to support the Hungarian authorities in the selection of a disposal option and candidate disposal sitLLW/ILr efo W radioactive waste. 4. CONCLUSION Hungary attaches great importanc establishmene th o et appropriatn a f o t e legal system to assure thainternationalle th t y agreed principle formulates sa RADWASe th n di S Safety Fundamentals are applied in all stages of radioactive waste management. The legal 146 instruments now in force provide already a framework for it, and these principles were nuclean o w la r energyconsiderew lowene e e drafTh e th . rf th leveo tdn i l regulations, guides, criteria, etc. hav supporo et attainmene th t thif o t s goal. Their elaboratio vers ni y important for the success of our National Waste Management Project, therefore we are looking forward for all kind of international co-operation and exchange of experience in this field. 147 STRATEG WASTR YFO E MANAGEMEN ARGENTINN TI A J. PAfflSSA CAMPA Waste Management Programme, National Atomic Energy Commission, Buenos Aires, Argentina Abstract The National Atomic Energy Commission (CNEA) of the Argentina Republic was established in 1950. It has in operation two nuclear power plants; a third one is under construction (70 % completed undes fourta i e d rh)an on study orden I . supplo rt fuee yth l element nucleae th o st r power plants mentioned before, CNEA has implemented the front part of the Fuel Cycle. Regarding the back- end, the actual policy concerning to the spent fuel elements, is to storage them waiting further decision. Together with this activities, the CNEA has developed, in practice, all the peaceful applications of nuclear energy. Mentioned activities, generate important volumes of radioactive wastes of different characteristic overale th d lsan strateg Argentine th f yo e Progra plano t ms i , develod pan implement the technology and provide the facilities for the permanent isolation of the generated wastes, with the aim that not compromise the health and safety of general public. To implement and coordinat thesl eal e activities CNE establiss Aha Radioactivha e Management Program thin I . s paper an outline is given concerning the policy, treatment, characterization, storage, transport and final disposal of radioactive wastes in our country. . INTRODUCTIO1 N Argentin nucleao tw s raha power plant operationn si MW(e0 38 e )th : PHWR, Atuchd an , aI the 640 MW(e) Candu, Embalse. Both use natural uranium and heavy water and represent 8% of the country's electric power capacity, but frequently produce more than 17% of the total electricity generated. A third plant is under construction (70% completed), the 720 MW(e) PHWR, Atucha II, and a fourth is being studied. In orde o supplt r e fueth yl element e nucleath o t s r power plants mentioned beforee th , Comision Naciona Energie d l a Atomica (CNEA implementes )ha frone dth Fuee t th par lf o tCycle , which include prospection, exploration, minin millind gan g ores, refinin standare th f go d concentrates; conversion into uranium dioxide; sinterin pellets2 UO f ;go productio Zircalof no y tubes; fuel elements fabrication and a high pressure testing loop. Complementin nucleae gth r plants requirements industrian ,a l plan heavr tfo y water production with capacit Mg/yea0 25 operationn f i yo s ri . For the back-end of the cycle, the current policy for spent fuel elements is to store them in pool concretr so e silos while considering further action. In addition to power generation, CNEA has designed and constructed several research and radioisotope production reactors (it produces over 90% of the radionuclides used in the country and is one of the main Co-60 producers), developed and commissioned a uranium enrichment plant in order to supply the fuel to be used in experimental and research reactors (including those exported) and to perform light enrichment in the fuels to be used in domestic power plants to improve "burn- up". Other sources of waste include research centres, universities, hospitals and industries. Alf theso l e activities generate important volume f radioactivo s e waste f differeno s t characteristics that must be treated and conditioned. For that purpose, the CNEA has established since 1986 the "Radioactive Waste Management Program". Although a "wait and see" policy for spent fuel was adopted, immobilisation of high level liquid wastes by vitrification is being studied. Two methods are being tested, employing borosilicate encapsulatioglasn a s sa n matrix: fusio sinterind nan g (hot pressing). Cement-based matrices are used to immobilise medium and low level wastes and formulation 149 development studies to produce acceptable wastes-forms composition are being made. A polymeric matrix (Alkatene) has sometimes been used in the past. With the aim of guaranteeing the behaviour of the waste-form, which is the first barrier, in the short and in the long term, modelling and experimental studies were made. The second type of barrier is formed by repository, wherein a series of complex engineering and geological elements mus consideree b t ensuro t s a eo ds isolation stratege Th . bases y i neaa n dro surface disposal of medium and low level wastes. A trench system is being used for low level wastes. A monolithic type concrete repositor mediur yfo m level waste undes i s r design deeA . p geological repository is being studied for high level wastes. 2. THE SIZE OF THE PROBLEM Since the beginning of the nuclear programme in Argentina, the following quantities of waste have accumulated: level2.1w Lo .. levew lo lf o waste 3 m s 2 treateTher96 conditioned e an dar e d int drumo3 m 506 2 s00. accordin predefineo gt d specification proceduresd an s . From these drums, 3500 have been placen di remainie TrencTh . 1 ° gh N 156 hundree 0 drumTrencOn n i . 2 e d° sar cubih N c metre treatef so d dan cemented biological wastes have been placed in a concrete pit ad hoc. From Atucha I, 15 m3 of evaporator concentrate are being immobilised by cementation, equivalent to about 750 drums. 2.2. Medium level. mediuf o Abou3 m m6 5 tleve l wastes, mainly sealed source structurad san l materials froe mth re-design of an experimental reactor, are immobilised by a cement grout in a concrete pit. Both power plants under operation have spenf generate o exchangn 3 tio m 8 d12 e resins, which once treated and conditioned will produce 3200 0.2 m3 drums. These drums will be placed in interme- diate storage until a final repository has been constructed. Medium level wastes also include filters fro operatine mth g power plants. The storee yar n di concrete pit awaiting a further decision. In Atucha I there are about 400 filters in fourth concrete pits. In Embalse there are 135. The principal sources of activity in both are Co-60 and, to a lesser extent, Cs-137. annuae Th druml0 yiel21 f sdo arising from Co-60 productio aboud drum0 nan 16 t s arising fro Mo-9e md 9 production mus takee b t n into account. . Alpha-contaminate23 d wastes. At present, 66.8 m3 of alpha-contaminated wastes treated and conditioned are intermediate storage awaiting final disposal. These waste arisine sar g fro experimentan ma l mixed oxide facility. 2.4. High level. These wastes basically comprise spent fuel elements: - Atucha I: A total of 6305 spent fuel elements equivalent to 1103 Mg of uranium are stored in pools at the reactor site. Pool building N° 1 has capacity of 3500 elements. The remaining elements are in pool building N° 2, which has at present 4158 free positions, enough capacity for the life of the plant. This second pool also contain 1 coolan4 s t channels froe reactomth r core, which were replaced. - Embalse: Spent fuel elements total 44181 at the time of writing, which is equivalent to 936 Mg of uranium e storagTh . e capacit a poo s lha 5690f yo 0 elements bees ha nt I . decide adopo dt interin a t m dry storage concept with modular concrete silos in order to increase the storage capacity. Actually ther 3554e ear 1 spent fuel element poosilose e 864d th th t an ln 0s.a i 150 - Spent fuel elements from experimental and production reactors: There are now 232 elements with 1418 uraniuf o 8g m (U-235 element,7 20%20 d )an s with 2998 uraniuf o 8g m (U-235, 90%). 2.5. Wastes from uranium mining and milling. These consist of: Uraniu- m tailings estimates i t :I d that about 5.182.00 wastf o g e0M have been generatede froth l mal uranium mines. All radioactive wastes produce managee dar d directl Wasty yb e Management Programmr eo under its supervision and control. 3. STORAGE, TREATMENT AND DISPOSAL In addition to the installations from power plants for waste management, the following facilitie treatinr fo s conditionind gan g waste availablee sar . 3.1. Radiochemical facility. This facility has; three hot cells to deal with high level wastes; radiochemical laboratories for medium and low level wastes; high integrity gloveboxes and a laboratory for complex radioactivity measurements. Different method f treatmeno s d conditioninan t f radioactivo g e waste s wela s a l characterisatio f matriceno waste-formd an s controld an s s relateacceptance th o dt perfomee ear n di this facility. levew 3.2lLo . solid waste treatment plant Wastes from CNEA atomic centres wels a , fros a l m medical centre industriad san l activities are processed in this plant. In a classification room wastes are divided into incinerable and non-inci- nerable, and the latter into compactable and non-compactable. r incinerablFo e waste incineraton sa availabls i r capacity3 e m wit 1 ha incineration a , n rate m3/hworkino4 a f3 d ,an g temperatur 970-107f eo ashee incorporatee Th . sar 0K d them bitumene Th . mixtur s loadeei d into drums , LLSe whicth carriee o Wt har t trenchdou . Compactable waste ear compacted with a hydraulic press into drums (3-6 Mg). The drums are loaded with classified and pressed wastes and are carried to the trenches for final disposal. Non-compactable t directlwastepu e sar y into drums wit cemenha t grou thed an nt disposed as before immobiliso T . e levemediuw lo l d wasteman cementationy b s , remotely operated mixer equipmen useds i t . 33. Trenches for low level radioactive wastes. There are two trenches. The first is completely full with 3500 drums and has been closed. The second is 120 m long 20 m wide and 1.20 m deep. It is enclosed by a perimeter barrier supported by walls of concrete. Its base is a 0.60 m thick bed of compacted caliche, plus a 0.10 m thick layer of soil, concret uppen a d re an laye granitif ro c stone. slope 5% Th .fros eo i t m2 This trenc accomodatn hca 560o t each3 p 0eu m drum2 . Whe0. f so n about 1000 drume sar accumulated, they are covered with the local clayish earth. Over the compacted soil a hot asphaltic layer is spread to a density of 2 kg/m2 achieving a 2 mm pore-free layer. Over the asphaltic, fine sand placeds si . Afterward blaca s k fil polyethylenf mo eplaceds i (25 ) m 0 . Finall toscya a layes i r placed ove polyethylenee rth , free from lumps; followe blaca y db k laye f soiro l 0.1 5thicm k with grass over it. The drumtrence th n shi contain radionuclides with half-lives shorter than five years, although limited amounts of nuclides with longer half-lives are also accepted. Drums with contact exposures higher tha t admittedmSv/0 n1 no e har samplinA . g statio monitorinr nfo aquifee gth operationn i s i r . Each trench has a database in which the characteristics and the history of each drum are recorded. 151 3.4. Low activity liquid waste decay and evacuation plant This installatio designes ni receivo dt e active effluents fro radioisotope mth e plant, store them for decay and, later, after monitoring, either discharge them to trench designed for liquids, or to the environmental. The plant is underground, divided into three main cubicles, two of which contain one 15 m3 tank each, while the third contains the pump and valves. 3.5. Semi-containment trench system for low level activity liquid wastes. This installation was designed for the disposal in selected ground sections of radioactive liquid waste containing radionuclides with short half-lives and low activities. The trenches were constructed according to a conventional design, after a study of the retention capacity of the soil by ion exchange Ezeize ath t a Atomic Centre f loessia soie o s i lTh . n slimes wit higha h conten f clayo t . Trenchee sar deepm 3 wide lon.m d m Eac10 g0 an 2 , h trenc samplina s hha g statio monitorinr nfo aquifee gth r and a database where the characteristics and origin of the pumped liquid are recorded. 3.6. Concrete cubicles. Two concrete pits are available. Each is 4 m in diameter, 10m deep and has a wall thickness of 0.30 m. They are for final disposal of difficult-to-handle structural parts, such as contaminated experimental reactor components, graphite, irradiation boxe higd san h sources o activitC n a r y.I These pits are cemented periodically to mantain an acceptable dose at their mouth. 3.7. Temporary store forspent fuel elements (Material Testing Reactor(MTR) type) and control roads. This is a building 85 m long, 12m wide and 4 m high. It consists of six longitudinal batteries pitf o s wit totaha l capacit pits8 19 , f eacyo inne whicn f ha o s r hlininha stainlesf go s steel tubes, 0.15 diametemn i 2.1 d longm 0ran , capabl receivinf eo maximuga fueo tw l f melemento controe on r so l eacn i t h pi batter e rod interconnectee Th . yar d with each othe stainlesy rb s steel tubes provided with manual valves to regulate the flow of demineralised water. 3.8. Installations under construction and yet to be builts. intermediatn A - e stor conditioner efo d medium level completedwaste% 0 (3 . ) - Laboratories coverin completed% 5 aren (2 g a f . 150 ao m2 0) piloo tTw cementatio r plants- fo e othee on vitrification,th r rd fo nan completed% 0 (7 . ) - A facility for the decontamination of large components. (95 % completed) - An additional temporary store for spent fuel elements (MTR type) and control rods. - Three more trenches for conditioned low level waste (one at Ezeiza Atomic Centre, one at Atucha and other at Embalse). - An intermediate dry store for spent fuel elements from power stations. 4. REPOSITORIES Studies are being made into a shallow monolithic concrete repository, for final disposal of level wastes majorite Th . countrr MLWf yo ou n i ,y come fro operatioe mth nucleaf no r power plants (spent resins, filters, irradiation channels replaced, liquids from decontamination, etc), as well as, in minor proportion from radioisotopes production (Mo-99, Cs-137,etc) reserce th productiod d han an , n of reactors remodelling. Total volume of treated and conditioned medium level waste until year 2020 was considered taking into account the simultaneous operation of four nuclear power plants and the activities mentioned before (including decommissioning). drum3 timee m numbeth e t 2 th ,s A 0. containin f ro g conditioned radioactive wastes wile b l u50.000o pt . 152 These drums will be inserted into a concrete container. After drums are inserted (12 drums per container) residuae th , l volume wil fillee b l d with cement grout closed han d wit concretha e cap. The containers holdin drume gth s wil placee b lspecia a n di l concrete module paraa f o t wil-I . e b l llelepiped shape, partl (totay m s dimension burieit 7 l d x volum dan m 2 2 se x wil aboue m b l 3 2 t 3500 m3). This is constructed to accept a (18.4 Mg/m2 total load over its floor. Each module will hav ecapacita concret0 40 f yo e containers, tha 480s i t 0 drums. Inspections wil performee lb t eacda h module. Durin operatione gth mobila , e structure includin neccesare gth y inspection equipment will cover the module and when it is completed and closed, it will be moved to the following module. The procedur sitr eefo selectio vers ni y important requiremene Th . establiso t hdisposaa l site are that it should satisfy all relevant peformance criteria taking into consideration technical, environmental, socia economid an l c considerations. From a geological point of view the following requirements are involved: Wast- e packages placeselectee th n di d site must ensure tha radiologicao n t l effects will occur before a yearperio0 elapseds 30 sha f do . - The design requires the possibility on construction down to 6 m deep. - Candidate sites will be located in a /one between Atucha and Embalse. Only low-use areas will be considered. - Candidates sites would hav aren ea smalleo an r than 500.00. 0m2 - Density of population must be less than one inhabitant per km2. - Neither permanent nor temporary water courses must be in the area. - The underground water table would be below 20 m. - The annual rainfall would be less than 200 mm per year. selectee Th - d site mus seismologicalle b t y stable. At present, the geological studies on the candidate disposal sites are in progress. Ten probable sites have been already studied. These site locatee sar d nea bordee rth r between Cordob Sand -aan tiagodel Estero provinces, with five site eacn si h one. The vere yar y stable formations. Eigh thef to m are plutonic outcrops and two are sedimentary sandstones. Also at an advanced stage are studies for siting a geological repository for high-level waste. pétrographie Th structurad ean l feature granitif o s c rocks, dimension deptd san rockf ho y formation, seismic and hydrogeological conditions mining and oil potential, as well as population and human activities have all been analysed. Sierra del Medio (Province of Chubut, is one of the sites under consideration. Further studies have included photointerpretation, alignment, statistical analysis, geological and geophysical recognitio rocke di f nyo formation, drilling geomorphologica, m s dow0 20 no t hydrod lan - geological analysis of the area and deep drilling down to 800 m. Conditioned wastes (borosilicate glasses or encapsulated spent fuel elements) will be placed deptn i i n diameten i m holeh 9 borem d 1 s galleriesn ran di , seale turn di n wit mixturha f saneo d and bentonite wit higha h ion-retention capacity. Each vitreous matrix containing 10 % by weight of oxides from fission products and transuranic elements, will generate a thermal power of 500 W after decaying over 20 years. In order to prevent exceeding the adopted maximum temperature of 333 K in the rock , the minimum distance between containers will be 5 m with a thermal power of 5 W/m2 on a horizontal plane. The distance between repositore gallerieTh . m s0 y 2 wil wilt operationae b l no l l until after 2010. 5. QUALITY ASSURANCE - QUALIT Y CONTROL SYSTEM To guarantee that activitie service d able b san en satisfactorca s provideP WM y ydb fulfilled, a Quality Assurance Progra mbeins i g developed. Ah essential requirement is that the radioactive waste management must be adressed so that safete th y objective guaranteee ar s d durin requiree gth d reasonabltime unded th ean l al r y foreseen circumstance proteco t generam e s ai th twit e h th le environmen publith d an c t from unacceptable radiological risks. The Quality Assurance Program involv a Qualite y Policy n Organizationaa , l Structure, Definitio Responsbilitief no Qualitd san y Control System. 153 The Quality Assurance System Involves: - The general requirements and acceptation conditions which have to met by a waste package to qualify for transport, long-term storage and/or disposal. procedurese Th - , spécifications, instruction inspectionprograme th d san , related wit materialse hth , the productprocese th d ordesan n i , obtaio rt requiree nth d quality. controe Th - l syste processe th f m o plante ,th operator e produce ,th th d san t durin differene gth t steps waste ofth e management. defineP e requireWM th s e Th d quality taking into accoun e limitth t s establishee th y db Regulatory Body; achieves and mantains the required quality by controls on waste treatment, condi- tionin characterizationd gan ; verify this qualit continuouy yb s inspectio recordd l casenan al e n sth i data, to demnostrate that specified quality of the end product meets the established requirements. All the facts here exposed show that the Argentina has decided, within certain limitations, to face with emphasis all about radioactive waste management. alwaye W s remark, without fea mistakef o r , thasuccèe th t failurr so nucleaa f eo r power plan wil resolvo t l depen w nucleae eho th n do r waste issue. 154 NATURAL DECA HALF-LIFED YAN BASEO R :TW SFO THE RADIOACTIVE WASTE MANAGEMENT POLICY J-C. FERNIQUE ANDRA, France Abstract n environmentaca w Ho l protection imperative d technicaan s l requiremente b s reconcile radioactivn di e waste disposa Francen I o kin? lf facilitietw o ,d s illustratw ho e radioactive waste disposa mergn ca l e scientific, regulator politicad yan l concerns, e baseth n do natural decay property of radioactive material. - Andra's near-surface disposal facilitie r short-livesfo d wast operatee generatioe ear on r dfo n (30 years) and monitored for ten generations (300 years), with the radioactivity of the waste declinin naturally-occurrino gt g levels throug procese hth f radioactivso f o d en e e decath y yb that time. The waste to be disposed of in such facilities contains nuclides with half-life below sais yeari 0 3 dd time-degradablsan humat ea n scale -•The challenge differene sar r long-livefo t d waste, whic alse har o time-degradablet no t bu , t humaa n scale. Risk assessment r disposasfo f suco l h waste, relatively straightforwarr dfo the first few thousand years, must also demonstrate that levels decline to naturally-occurring levels, even though this may occur in tens of thousands of years, when it is predicted that climatic change, new glacial activity, and a drop in sea level will occur, and when civilizations wil doubo n l t have change wells da . This demonstratio verf no y long-term safetexpresn a s yi s requirement for radioactive waste disposal. The paper briefly describes the criteria used in the French regulation to determine what waste can be accepted for near-surface disposal and the recent significant steps taken to resume field e sitinworth f undergrounr go kfo d laboratorie possiblyd an s , much later a repositor, r yfo waste non acceptable for near-surface disposal. The conclusion focuses in demonstrating how a consistent National or International Waste Management Program based on clear ethical, societal, scientific and technological choices has to be prepared and presented to the e PublicAuthoritieth o t , d allowinan s e wastth g e management Organizatio o gait ne th n necessary Public Confidenc Acceptanced ean . 1. RADIOACTIVE WASTE PROPERTIES Radioactivit spontaneoue th s yi s disintegratio nucleue th f n o f certai so n unstable atoms s thea y transform r decayo , , into stable atoms. This transformatio s accompaniei n e th y b d emissio energyf no radiationr o , . Unstable atoms decay into stable atom randoa n si m manner. For many atoms of a given radioélément, a statistically constant number of transformations will occur over a certain period of time, called the decay constant. Radioactive half-life is the time it take halr originae sfo fth l numbe atomf o r decayo st . Each radioélémen intrinsin ow s it c s ha t half-life. Radioactive half-life is an important notion in long-term radioactive waste management becaus t determineei lonw t wilgsho i l tak radioactivr efo e material decao st o yt harmles f comparisono y s wa levels y B , . non-radioactive toxic materials never lose their toxicity, and can therefore be said to have an infinite half-life. The longer the half-life of a radionuclide e loweth ,s radioactivityit r , since more tim s needei e r hals atomfo dit f o t s disintegrate.fi] 155 basie Th c waste properties use creato dt wastea e management syste mhalfe th wil -e b l life and the initial level of radioactivity. Natural radioactivity is often contrasted with artificial radioactivity, but in reality there is no difference between the two, except that one originates from natural sources and the other is manmade. However, natural sources are not usually highly radioactive. We all live in a radioactive environment, with radiation emanating from several natural sources. The comparison between natural and artificial level of radioactivity can also be used to determin considereee b wha n ca t negligibls da acceptabler eo . 2. USE OF THE " TIME DEGRADATION " PROPERTY Comparing radioactive waste with toxic waste makes clear that the decay property of radioactivit existence th d y an acceptabl f eo e level negligiblr so e use e levelb determino dn t sca e waste ith f e will become harmles a perio n i sf tim o d e tha s manageabli t humay b e n made barrier notr so . It will the possible nb categoriesoro wasteo e t th ttw n ei s: * Waste that will decay to an acceptable level at human scale, or short-lived waste, that can be safely managed by human actions and can be disposed of in near-surface facilities. * Waste that will not decay to an acceptable level at human scale, or non-short-lived waste, that has to be protected by non human-made barriers from the biosphere. Such waste will generally be disposed of in deep geological disposal in order to be protected by the geological barrie durabilitra thas ha t accordancn yi e wit lone hth g half-life nuclideth f o e s contained.[2] implementatioe Th f sucno wastha e management system will nee anweo dt maio tw rn questions : consideree Whab n ca t acceptabls da e? consideree Whab n humae ca t th s da n scale? 3. WHAT CAN BE CONSIDERED AS ACCEPTABLE ? e answeTh thio t r s question belong e Societyth o t s . When establishing acceptable levels of protection, Authorities typically take account, among other things, the recommendation Internationae th f so l Commissio Radiologican no l Protection (ICRPd an ] [3 ) the IAE specificalld Aan three yth e concept justificationf so , optimizatio dosd nan e limitation. Bu remaint i t Society'e sth s responsibilit determino yt e suc threshholdha . Elements to answer this question can also be taken out of the Annual Limit of Intake concept [4], through ingestion and inhalation scenarios developped by the nation Safety Authority. s alsi t oI possibl o compart e e potentiath e l maximum exposure wite locath h l backgroun naturaf do l radioactivity. 156 4. CONSIDEREE WHAB N TCA HUMAS DA N SCALE? The answe thio rt s question belongs Societye agaith o nt , taking into account historical culturad an l factors. In France it is considered (Fundamental Safety Rule 1.2.)that human actions can be relied t leasupoyeara 0 r therd 30 t n fo existine s an ear g example thar sfo tentite sucth s n yhi a charge of keeping the memory of the quarries located under the region of Paris, France. In the U.S. it is considered (10 CFR 61 §61.59.a) that "the period of institutional control will be determined by the Commission, but institutionnal controls may not be relied upon for more than 100 years following transfer of control of the disposal site to the owner." The duration of the institutional control period is also depending of the waste generated in each countr examplen a s y a year 0 and ,convenien e 30 , sar safe th e r managemenfo t f moso t t waste coming from the normal operation of PWR reactors as it allows for a decay often half- lives or by a factor of more than 1,000 for 137 Caesium and 90 Strontium. 5. DETERMINATION OF ACCEPTANCE CRITERIA maie Th n basi r establishinfo s g acceptance criteri r short-livefo a d waste eth wile b l potential exposure of people living on the site after the end of the institutional control period. This value wil connectee b l specifie th o dt c activit half-life th r nuclido t ype ed thaean t gives decae th y factor durin institutionae gth l control period. 11001 A 1000 c 900 t ' 800 v 60 Cobalt » 700 t y 600 r 500 e m 400 a i 300 n i 200 n 100 238 Uranium 30 60 90 120 ISO 180 210 240 270 300 Institutional Control Period (Years) Uraniu8 23 m» Cobal0 6 tCaesiu— 7 - 13 — m Specific Activity Acceptance Criteria as a function of acceptable level 157 r eacFo h nuclid possibls i t ei e through selected scenario givo st maximue eth m valuf eo the specific activity at the end of the institutional period. The decay calculation taking into account the duration of this period and the half-life allows an easy determination of the maximum specific activit operationae th nuclidr f ype o d acceptee een b thale n periodth ca t t da . Other limitations can be added for reasons connected to transport regulations or to operational safety, specially for very short-lived waste. As there is always a small amount of long-lived nuclides in the short-lived waste it is threshola necessarp u t thir se d fo s o y t amount Frence Th . hBq/average0 n limia 37 s gs a i t . FRENCE TH H . RADIOACTIV6 E WASTE MANAGEMENT SYSTEM Some countries, particularly those that have electesoro t t t wastdno e into long-lived and short-lived categories, like German Switzerlandd yan , pla disposo nt l wastal f deeen o ei p underground repositories. This approach is sometimes a matter of convenience, as is the case for countries like Swede d Finlandan n , which have built repositorie e Scandinaviath n i s n granite shiel t nucleada r power plant sites. France, Spain, United States, Japan otherd an , s dispos short-livef eo d wast near-surfacn ei e facilities. France has elected to dispose of short-lived solid radioactive waste with low and medium activity level near-surfacn si e facilities using multiple-barrier concept accordancn si e with national safety regulations. [5],[6] Near-surface disposal methods have gradually evolved since 1969, whe e firsth n t French radioactive waste disposal facility at the Centre de la Manche began operating, and have reached maturity wit desige hth constructio d nan Centre th f l'Aubene o e d ] .[7 The underlying principle of near-surface radioactive waste disposal is to protect the waste from human intrusion and from exposure to water for as long as it takes for its radioactivity to decline, through the natural process of decay, to levels that are no longer harmful to the environment. The disposal facility will be monitored for the duration of the institutional control period, which should not exceed the 300 years mandated by the regulatory authorities. The institutional control period requiremen s accompaniei t site-specifiy db c acceptance limitr sfo mas totad san l activitie botr sfo h long short-lived -an d emitters. However, high-leve long-lived an l d waste e anothear s re mus storyOn t ] resolv.[8 e difficult technical f affectivproblemo t lo a s ewel a s s e onesWasta l t Th passe.Ac e n o d December 30,1991, stipulated that research on long-lived radioactive waste management was carrie e thren b i t o t ed ou mai n areas: - enhanced actinide separatio transmutationd nan , - waste solidification processes for interim storage, and - retrievabl r non-retrievablo e e disposa deen i l p geologic formations, particularly through the creation of two underground laboratories. 158 The last of these subject areas, which falls more directly within Andra's scope, meant that the 1990 moratorium had to be ended, which was not an easy thing to do. To accomplish this, Christian Bataille a membe, e Parliamentar f Parliamenth o r f o e d authoon an t f yo r assessment reports appointes wa governmen,e th y db Mediatoa s a t r "negotiator(o r e th n i " U.S.) charged with identifyin undergrouno g tw site r sfo d laboratories s confirmewa e H y . db the new French government elected in March 1993, showing that the waste problem is an apolitica d nationaan l e transcendinon l g party politics. Through informal discussione th , Mediator and his small team worked to persuade local communities, cities and Departments (France is divided into 95 territorial administrative departments) to volunteer to host an underground laboratory, armed with the knowledge that a laboratory would mean 1.5 million French francs of investment, 150 jobs and a 60 million French franc per year contribution to the economic development of the region. In the end, Mr. Bataille selected four Departments, all of which had voted in favor of hosting a laboratory. The Mediator submitted his report to the governmen latn i t e December 1993 earln .I y January 1994 governmene ,th t gave Andre ath green ligh begio t t n detailed foue worth r n ki arease 199 th 0o ,t moratoriumputtind en n ga . The phased program established by the 1991 Waste Act will now be as follows : - one to two years for geologic reconnaissance (seismic reflection, drilling) of the four areas and selection of two sites to host the labs ; - one year for the license application to construct and operate the two underground laboratories ; twelvo t n te econstructiob year- la r sfo operationd nan . Afte yeae th r r 2006 decisioa ,mad e o b whethen n tw eo ne ca th convero f t ro e on t laboratories into a real waste repository, subject to positive laboratory testing results. In accordance wit Waste hth e Act, conversio nParliamentf woulo t ac contingene db w n ne I . a n o t addition, the law provides guarantees to ensure that all research, including research on incineration, has been taken into consideration before final decisions are made. Needless to say, the Waste Act also requires continuing oversight of the program and statue th f researcf so o independenn a y hb t National Review Board, whic s appointehwa d very recently. 7. CONCLUSION The basic propert f radioactivo y e waste when comparing with toxic wast s thai e t radioactive waste will loose its potential hazardous effect as it will decay according to its half-life. This property frencalowe th f disposo hso t wast% e90 e volum surfacn ei e disposal facilities subjec institutionan a o t t l control perio humae th n di n scal timef eo . A waste management system based on possible human remedial actions is far more understandable and acceptable by the public than a system based on demonstration by calculatio modelingd nan . It is obvious that such a surface disposal facility will not accept all the waste volume and will in fact receive only a small amount of the total activity ; but this will allow to create, d maintaian n everyday e Publith , c Confidence (instead Public Acceptance e entitth n n yi i ) charge of waste management. 159 Later ,e credibilit baseth n o dd confidenc an y e gaine thiy b d s entity, thisa wil e b l positive factor in the process of solving the difficult problem of safe final disposal of long-lived waste, possibl emplaciny yb g the deemn i p geological disposal. REFERENCES [1] FAUSSAT, A.,"La gestion et le stockage des déchets radioactifs", La Technique Moderne, n° 4/5, 1992 [2] FERNIQUE, J.C., "Radioactive waste Management in France", (Proc. POWER-GEN Europe'93),Paris, (1993) ] [3 INTERNATIONAL ATOMIC ENERGY AGENCY, Basic safety standard radiatior sfo n protection, 1982 Edition, jointly sponsored by IAEA, ILO, NEA(OECD), WHO, Safety Series N°. 9, IAEA, Vienna (1982) [4] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION. 1990 Recommendations of the International Commission on Radiological Protection. ICRP Publication 60, Annals of the ICRP Vol.21 . 1-3,N° , Pergamon Press. Oxford. (1991) ] [5 BAZOT . ,VOIZARDG , , "Disposa short-livef o l d radioactive wast near-surfacn ei e facilities", Revue Générale Nucléaire. June 1993 [6] MARQUE, Y.,"La gestion des déchets radioactifs", Arts et Métiers Magazine, n° 165, 1991 [7] FERNIQUE, J.C.,"The new low-level waste disposal site in France, a ten-year new experience", (Proc. Nuclear Industry China 92), Beijing, (1992) [8] ALLEGRE, M.,"Radioactive waste management ; the new french way, an approach to the future" (Proc. WM94 Conf), Tucson, (1994) 160 WASTE MANAGEMENT PRACTICES WASTE MANAGEMEN NUCLEAE TH T TA R TECHNOLOGY DEVELOPMENT CENTRE (CDTN) S.T.W. MIAW Nuclear Technology Development Centre - CDTN, Belo Horizonte, Brazil Abstract In several laboratories and pilot plants at the Nuclear Technology Development Centre (Centro de Desenvolvimento da Tecnologia Nuclear - CDTN), low level solid and liquid radioactive wastes are generated from nuclear fuel cycle activities, radioisotopes application, R&D and routine works. The CDTN also receives spent sources from radioisotopes users. These wastes are to be managed to avoid contamination risks and minimize the costs of treatmen d furthean t r storage o systematizT . e wastth e e control a Wast, e Management Programm s beeha e n implemented since 1983, which strateg e nationa s basei yth n o d l regulations and the available infrastructure at the Centre. This paper presents an overview of the waste management at the CDTN, some results of research and development and the technical support communitye giveth o nt wels a , dealins a l g wit emergence hth y causey db radiologicae th l accident that occurre Goiânian di . 1. THE CDTN WASTE MANAGEMENT STRATEGY Low level solid and liquid wastes are generated in several laboratories and pilot plants at the Nuclear Technology Development Centre (Centro de Desenvolvimento da Tecnologia Nuclear - CDTN). As support to the community, CDTN also collects spent sources, smoke detectors, lightning rods and Ra-needles for further treatment. All these wastes are to be managed to avoid contamination risks and to minimize the cost due to their production. To systematiz waste eth e control Wasta , e Management Programm bees eha n implemented since 1983 and regularly revised. The strategy is based on the Brazilian standards and the infrastructure availabl e Centrth t a ee suc s resulta h s from R&D, developed treatment processes, equipment and installations and supporting laboratories. All contaminated materia segregates li origis it t da n accordin physical-chemicae th go t l and radiological characteristics. To minimize the waste volume and therefore the cost of treatment and storage, it is verified if the contaminated material is reusable, before release for treatment. The strategy adopted for the waste management at CDTN is shown in Figure 1. The non-sealed sources, except ore, are treated if their half-life is greater than 60 days and the activity is higher than 74 Bq/g. These types of wastes are mainly contaminated liquid solution solid san d materials. Liquid wast collectes ei d separately aqueous a , organid san c wastes, either in polyethylene flasks or in glass bottles. Solid waste is segregated as compactable and non-compactable and is collected in small containers, protected by plastic n metallibagsi r o , c drums l thesAl . e package e labelear s d with radiation symbod an l identified. After collection l datal , a applicabl e wasteth o et , lik s origineit , composition, volume, weight, chemica radiologicad an l l contaminant exposurd san e rate recordede ar s . This informatio s importanni waste th r e fo tinventor y .After monitorin classificationd gan , radioactive wast s transferrei e r storagfo d d furthean e r treatment accordin o safett g y requirements. Non-compactable wastes suc s rubbisha scrad immobilizee han par cemena n di d an t bentonite matrix. Damaged contaminated polyethylene flask smald san l irradiation flaske sar 163 reduced in volume by a 130 kg/h throughput shredder. Compactable waste (waste paper, plastic, clothes, gloves etc.) is directly pressed in 200 L drums using a 16,000 kgf compactor. Non-reusable spent sources are stored for further conditioning. They will be packed in qualified container immobilized san cemena n di t matrix accordin kine f sourceth do o gt s it , activity and physical condition. Special additives can be added to absorb the radionuclide in cas f leakageeo . Smoke detector lightnind san g rods wil dismantlee b l sourcee th d dan s will be conditioned.Liquid aqueous waste is treated by chemical precipitation.The radionuclides are concentrated in an insoluble form, reducing greatly the activity of the overflow that is released accordin standards.The th go t e sludg cementeds ei organie Th . c wast absorbes ei n di vermiculit alsd eoan cemented cementee th r Fo . d waste product, quality control samplee sar taken. The process control and the product evaluation are done through viscosity, setting time, after 28 days compressive strength assays and leaching tests. Table I shows some data applicabl CDTe th o eNt cemented wastes. Liquid effluents fro laboratoriee mth collectee sar d in tanks, analyzed and discharged according to the release limits, otherwise the dilution is carried out. TABLE I. CDTN CEMENTED WASTE DATA Contaminants , nat , nat Ra , daughterTh .U . s Miscellaneous radionuclides Matrix Cement / bentonite Cemen / tbentonit e Waste/product (wt%) 36-43 10-40 Viscosity (Pa.s) 13 - 150 88 - 250 Setting tim) e(h 3-9 3-5 Density (g/cm3) 1.6-1.8 1.6-1.8 Compressive strength (MPa) 6- 10 11 -22 Activity alpha (Bq) 8 10 10 x x 1 71 -1. beta (Bq) 1 x 107- 1.3 x 108 . RESEARC2 DEVELOPMEND HAN T PROGRAMME 2.1. Immobilization The immobilization process is based on the technology where the waste is fixed in a matrix and the obtained final product is solid with low leachability and good resistance characteristics, minimizing the contamination risks and is suitable for disposal. Research and development works for cementation and bituminization of wastes generated from the nuclear plants operation and radioisotopes applications are carried out at the Centre. The results from both studies are helpful to choose a more suitable process for a specific waste.They are also useful to give support to the nuclear power plants Angra I and Angra II or to the regulatory bodies on the establishment of waste acceptance criteria. Cementation - the cementation R&D includes investigation of different kinds of matrix, equipment and methodologies in order to find a more efficient process and high quality products, establishmen procesf o t s contro procedured an l testd immobilizer san s fo d product characterization, evaluation of the cement-waste compatibility and establishment of cementation parameter reae th l n scalesi supporo T . t these works ther cementatioa s ei n batcL laborator0 h 20 cementatio a d yan n plant constructioA .batc L 0 2 h a syste f no ms i 164 planned in order to improve some operational and product parameters such as to evaluate different liquid-solid mixer configurations and product homogeneity, to set the reaction time and to establish decontamination procedures. Storage Transport I Treatnent/ ImnoMlzatlon Iirterln Storage 1 Transport Disposal 1 © Yes No Fig 1 - Waste Management Strategy. 165 Several Brazilian natural clays are evaluated, specially bentonite, vermiculite, kaolin and serpentine. Experiments were carried out with active and inactive wastes. It was observed that the use of amounts up to 10 % of bentonite increases the cesium retention, about 99 %, without reducing the mechanical resistance. Several samples are under leaching tests over almost nine years and are in a good shape. These results are applied to solve the CDTN waste cementation problems. Additionally, experiments wit 3 typeh f chemicao s l additives (retarding, accelerato fluidizatiod ran n agents from different manufacturers progressn i e ar ) . The preliminary results show that these kinds of additives can improve the cementation process and obtained products. The contaminants retention is under evaluation. The investigation of different parameters that influence the final cemented products was done specially with the aim to solve Goiânia's accident waste problems. Because of the open air storage, corrosion was detected in 200 L drums. They have to be reconditioned and immobilize specian di l packaging orden si ensuro rt e safe interim storag disposald ean . Based resultD R& cement/bentonits a e onth e matri specifies xwa immobilizo dt remaie eth n source and the 200 L drums. The mixture of water/cement = 0.70 and cement/bentonite = 0.15 was used. The achieved viscosity was in the range of 21 Pa.s, initial and final setting time of 2 respectivelh 6 d han measureds ywa average Th . e densitmatrie th arouns f o xywa d 1.79 g/cm3. Several samples were taken to evaluate the mechanical stability (compressive strength of 30 MPa) and absorption rate of the waste form. The developed methodology is also applied to treat hazardous wastes. Experiments were performe evaluato t d e heaveth y metals retentio cement/claa n ni y matrix. Some results showed that, using clays retentioe th , leachee higheth s n i d dan ramoun tha% 9 n9 lowes i t r than those recommende environmentay db l standards. Bituminization - The investigation of bitumen product acceptance criteria is based on Brazilia t climatnho e conditions firse th t r approachFo . , Brazilian bitume softenine th f no g range pointh f 80-100°n eo i t penetratiod Can f 10-2no 0 (L/1 ) wer0emm selected. Three different simulated evaporator concentrates were investigated. Experiments with solid content incorporation that varied from 27-4 wer% 0wt e performed. Product parameters sucs ha softening point, flash point, penetration, water content, grain size/homogeneity, swelling and leaching were investigated. Both leachin swellind gan g tests showed that products containing < 30 wt% of solids incorporated presented lower teachability and swelling than those products containing > 30 wt% solids. The leaching rates of borate ions, for a period of one year varied from l.OxlO" 5.OxlO"o t s 11m/ 12 leachine m/sTh . g rate r chloridsfo e ions were around 8.0x10' m/s. The product softening point varied from 96-102°C, flash point > 320°C 12 and penetration from 0.4-1.1 mm. Experiments with simulated spent resins immobilized in bitumen were carried out after the resins were loaded with lithium and boric acid to the breakthrough capacity of the resin. range Th f incorporateeo d resins varied from 33-48 wt% produce ,th t softening point from 113-124°C, flash point from 205-224° penetratiod Can resin-bitumee Th n. arounmm 4 dn0. products presented lower penetration than tha f puro t e bitumen watee Th .r contene th n i t product was less than 2%. The leaching rate of borate and lithium ions was around 5.OxlO"9 m/s and l.OxlO"10 m/s respectively for the anion and cation exchange resins incorporated in the bitumen. The swelling of the product was in the range of 10-30%. The products appear to be in good physical condition after a year of leaching tests. The distribution of mixed e bitumeresinth n i s n phas d micro-structuran e f resin-bitumeo e n products have been investigated by microscope. The homogeneity of resin in the bitumen phase and resin size ground have been determined resine Th . s were fragmented into different size thed san y were 166 uniformly distributed in the bitumen phase. These experiments were performed by a pilot plant extruder wit kg/h1 h kg/produc4 3- h d condensatan t e throughput. 2.2. Packagin radioactivr gfo e material. Since 1982 the CDTN has designed, tested and qualified Type A packagings for radioactive materials. These packagings are used for the transport of radioisotopes, spent sealed source wasted san s fro nucleae mth r fuel cycle facilities, radioisotopes user thosd san e generated during the radiological accident in Goiânia. For this purpose, facilities and equipment were developed, qualifyin Centre Braziliae gth th s ea n official packaging testing institute. Studie done sar orde n ei creato t r e capabilit qualifyinn yi g intermediat higd ean h level radioactive materials, using finite element computer code thermal-structurar sfo l analysis and shielding calculations. This programme has the aim to give support to the storage of spent fuels from Angra INP solvd Pprobleme an eth s fro conditionine mth highef go r activity spent sources.All the gained experiences can be also applied to deal with problems related to the transportation of hazardous/chemical materials that is very critical in Brazil. In orde evaluato rt durabilite eth commerciaf yo l intermediatdrumd an w s lo use r dfo e level wastes conditioning CDTe ,th carries Nprogramma ha t dou e since firse 1983th t n parI . t of this evaluation, unsectioned drums from two different manufacturers, containing both compacted and cemented simulated wastes, were stored inside a storage area and in the open. To study the corrosion phenomena, from a quantitative perspective, samples were submitted to metallurgical tests for later comparison. A total of sixteen drums were tested. After the predicted duratio f fivno e verifies yearswa t i , d thadrume th t s stored inside wer goon ei d condition and those outside presented large corrosion areas, mainly at the lid surface, because its design allowed rain water collection. Three of them were opened and metallurgical tests were performed, presenting similar results to those from the beginning. In a visual inspection performed in 1992 the existence of holes at the lid surface was detected. This part of the corrosion programme was executed. The results obtained show that the tested drums are not suitable for open storage, since they failed after an 8-year storage period. In the second phase studye th f o , onltype f druy on eo tested s mpurpose wa evaluat o t Th . s elone wa eth g term influenc botf eo environmente hth , externally waste th d e internallyan , , upo drumse nth e Th . drums were sectione representativd dan e sample theif so r body contact n weri t epu , internally, with pure mortar or simulated cemented borate waste and stored in a simulating storage condition. This programm duratioa s eha f samplef fivo no t e se years so havTw . e already been take t (199 n1993)ou d 2resultan e Th . s show thainternae tth l painting (epoxy-phenolic) is not suitable as a cemented waste packaging coating due to its poor resistance to this kind of waste. It is planned to test a more resistant commercial drum, with a thicker steel plate and havin electrostaticalln ga y applied epoxy paintin internan a s ga l coating. 2.3. Repository safety assessment Since November 1993 multi-disciplinara , y grouCDTe th beet s pa N ha n forme deao dt l wit repositorha y safety assessment. This grou composes pi f expertdo s from different areas such as those with knowledge hi source term, ground water flow, hydrology and dosimetry. The studies are conducted in an interactive manner by starting simple calculations, and followed by calculations using computer codes. Experimental works will be determined along the studies purpose Th . f thieo s worcreato t s k i nationaa e l capability, withi three nth e different Brazilian institutions in safety assessment analysis. All the results applicable to waste characterization, packaging backfild appliee b san n thir ca dlfo s purpose. 167 2.4. Hazardous waste e CentrAth t a elarg e amoun f chemical/hazardouo t s materia s generatedi l e Th . experience gaineradioactive th n di e waste managemen uses i t manago dt e these hazardous wastes. After the waste qualification and quantification, a hazardous waste management programme will be implemented. The strategy includes waste collection, recycling through stoca k exchange data bas treatmend ean safa r e fo tstorage.Th e developed methodologn yca also be extended to other institutions. 3. CONCLUSION Based on experiences accumulated in the R&D and CDTN waste management, the Centre was able to provide assistance and technical support to the fuel cycle industries, radioisotope users, hazardous waste management, as well as with the emergency caused by radiologicae th l accident that occurre Goiânian di . Studies were don characterizo et e eth effluent stream fro mmonazita e processing industr establisd yan hliquia d waste treatment process for the uranium enrichment plant. Establishment of the waste management programme for the nuclear fuel element manufacturing plant, evaluation of leaching resistance of cemented wastes from Angr , establishmenaI proceduref o t cementatioe th f so n procesd san obtained product qualification generate t cellevaluatiod ho an sn di f packagingno e th r sfo transpor f non-irradiatedo t - fuel elements were carried out. Medica d industriaan l l radioisotopes packagings were also teste qualifiedd dan . maie Th n important featur thas gainei e th t d experience allowed participatioe th n no Goiânia radiological accident waste managemen n 1987i t e staf.Th f e worketh n o d establishment of the general planning, and the strategy adopted for these wastes management, definition of specific procedures and the identification of the short term available infrastructure suc packagingss ha , treatment processes, definitio operatiod nan f interino m storag d alsan eo decontamination works. Untitherw s no participatioi le e th n o n establishmen f strategieo t r fo safs e disposal, definitio d desigan n f reconditionino n g packagings and the composition of the backfill for the drums immobilization and operation of the first Goiânia's waste repository. Support has also been given to several incidental situations with unsealed sources. BIBLIOGRAPHY AWWAL, M. A., GUZELLA, M. F. R., SILVA, T. V., "Research and development work on bituminization of low level radioactive wastes", SPECTRUM' 94 (Proc. Nuclear and Hazardous Waste Mangement Tropical Meeting, Atlanta, 1994), vol , Atlanta.1 , Georgia (1994) 313. MI AW, S. T. W., "The Goiânia accident waste management - strategy for a safe storage and disposal", SPECTRUM' 94 (Proc. Nuclear and Hazardous Waste Mangement Tropical Meeting, Atlanta, 1994), vol. 3, Atlanta, Georgia (1994) 2184. MOURÄO, R. P., MIAW, S. T. W., Packaging design and qualification: the experience of CDTN/CNEN, RAMTRANS, vol. 4, Nuclear Technology Publishing, England (1993) 22. , SILVA A. , REIS. SILVA P. C , Contrôl . . F. , L M , . E rejeito,e ed s radioativo CDTo sn N- 168 participaçâo da Supervisâo de Rejeitos, Rep. CT3-NI-03/93, Centro de Desenvolvimento da Tecnologia Nuclear (1993). f Braziliao e Us , n TELLOretentio e O. clay th . n C so . contaminantsf C n,o , SPECTRUM' 94 (Proc. Nuclea Hazardoud ran s Waste Mangement Tropical Meeting, Atlanta, 1994), vol. , Atlanta2 , Georgia (1994) 1253. 169 MANAGEMENT OF NON-FUEL CYCLE RADIOACTIVE WASTE IN ROMANIA C.N. TURCANU Institute of Atomic Physics, Bucharest-Magurele, Romania Abstract managemene Th non-nucleaf to r fuel cycle radioactive wastes fro l ove mcountral e rth y is resolved in a centralized way by the Institute of Atomic Physics. For waste treatment, batch co-precipitation, evaporatio incineratiod nan mais na n processe usede ar s . Conditionins gi performed by cementation and final disposal is assured in an old uranium mine. A R&D programm objectivee th s eha upgrado t s radioactive eth e waste managemen meeo t te th t international standard recommendationsd san . In Romania, the nuclear activities are regulated by the Nuclear Energy Act, Law No. 61/1974 and Quality Assurance Law No. 6/1981. The regulatory body is National Commission for Nuclear Activities Control, within the Ministry of Waters, Forests and Environment Protection. Considerin nationae gth l nuclear energy power programme th d ean necessity of a better protection of operational staff, population and the environment, the regulatory body initiated the revision of the National Energy Act, with special references to: a) development and establishing an overall strategy for radioactive waste management, and b) development of regulations for decommissioning. Accordin presene th o gt t legislation managemene th , non-nucleae th f o t r fuel cycle radioactive wastes from over the country is the responsibility of the Institute of Atomic Physics (IAP), Bucharest-Magurele Radioactive th , e Waste Treatment Plant (RWTPd an ) National Repositor Radioactivf yo e Wastes (NRRW) thren i , e technical steps: primary segregation and storage at the waste producer site waste transfe centralizey b r d collectio treatmentr nfo , conditionin interid gan m storage (RWTP) long term disposal in geological formation (NRRW) 1. SOURCE RADIOACTIVF SO E WASTE Radioactive materials are extensively used in Romania since 1957 for research and applications, after the construction of the WR-S research and isotope production reactor. The radioactive wastes, generate thesy db e nuclear activitie divide e levesw ar lo l n dwasto i t p e(u 10~3 Ci/m3) and medium level waste (up to 103 Ci/m3). These wastes contain mainly short and medium half-lived radionuclides (with exception of 14C) e wasteTh s result fro d mfro an researcmP nucleaIA t ha r applications mainln i y medicine, biology, agriculture, geological exploration, quality contro constructionn i l d an s metal processing industries. Wastes containing short lived radionuclides as I do not require 131 any special treatment, except temporary storage at the producer site, before the transfer as normal non-radioactive waste takes place wastee Th . s containing longer lived radionuclides are properly collected, treated and conditioned before final disposal. 171 Special cases of radioactive wastes are the wastes collected before the implementation of Nuclear Energy Act, which wer t appropriateleno y treated and/or conditionedw no d an , canno transferree b t r finadfo l disposal without remedial actions. sourcw A ne radioactivf eo e waste decommissionine s th wil e b l WR-f go S research reactor of IAP. The reactor commissioned in 1957, has a nominal power of 2 MW and has been operated nearly 38 years with its original equipment and systems without any major incident. In the next two years the fuel reserve will be consumed and the reactor life will be considered accomplished.To prepare for the decommissioning, in 1992 the institute commence a dspecia l programme concernin e radioactivitth g y inventory, radioactive decontamination and waste treatment. 2. TREATMEN CONDITIONIND TAN G All non-nuclear fuel cycle radioactive wastes generated in Romania are collected and treate t RWTP-Bucharest-Magureleda annuae Th . l designed capacit treatmene th f yo t plans i t levew lo lf 150aqueouo 3 0m s wast levew lo el f soli(LLAW)o 3 dm wast0 10 , e (LLSWd )an 100 shielded drum f mediuo s m level waste (MLW) normae Th . l present statu f RWTo s P concernin buffee gth r storage capacities, radioactivit annuad yan l arisin presentee gar n di Tabl. eI Low level aqueous waste o stepse treatear tw s :n i dfirstl y triplb y e chemical co-precipitation (iron hydroxide, calcium phosphate, copper ferrocyanide) operated batch-wise, secondld an evaporationy yb concentrate Th .sludg e th conditionee d ear e an cementatioy db n in 200 L standard drums, and the distillate is released after checking radioactivity and the chemical control. Low level solid wastes are treated according to the waste form, final conditioning is performed also by cementation. Medium level wastes, spent sources and wastes containing 3H, 14C and 129I are conditioned in shielded drums, without any treatmentTreatment and Table I. Present Status of the Radioactive Waste Treatment Plant Storage Storage Activity Annual arising Type of waste capacity present status LLAW 2 x 300 m3 ful% l80 up to 10-3Ci/m3 up to 103m3 LLSW 20m3 5m3 up to 10-3 Ci/m3 3m 0 1 uo pt MLW 200 shielded 30 shielded in limif o t up to 70 shielded drums 200 mRem/h drums Spent sources 3000 1000 i C 4 10 uo pt up to 400 Conditioned 3000 drums 1200 in limit of up to 200 drums waste 200 mRem/h 172 Tabl I eI Treatmen Conditionind tan g Capability Typ f Wasto e e Capability segregation incineration shredding LLSW compacting (200 kgf/cm2) drum1 0 cementatioy sb 20 conditionind an 1 0 n10 (fina n gi l : 200 1 drums) reconditionin drum1 drum0 1 s24 0 n si 20 f go MLW fire detectors direct conditioning without processing by cementation in 200 damaged sources 1 drums chemical treatment by precipitation (first step) using iron chloride, sodium LLAW phosphate, potassium ferrocyanid FD=3; e 0 (approx.) evaporation (second step) FD=1000 (approx.) conditioning capabilitie RWTf so summarizee Par Tabln di . RWT et licenseII e no th s P i r dfo treatmen r conditionino t f alphgo a wastes, excep smokr fo t e detectors. From November 1974, when RWTP became operational, to December 1993, 20,000 m3 of LLAW, 1,500 m3 of LLSW and 2,700 spent radiation sources were treated, resulting in 5100 conditioned drums.The transfer of conditioned wastes from RWTP to NRRW for final disposal starte 1987n di , wit annuan ha l rat f nearleo drum0 y50 s (8-10 rail transportsd an ) in the interim storage are still 1200 drums (1000 drums damaged by corrosion) some of which with an age up to 20 years. There are 1000 spent radiation sources without a known history, waiting for treatment and/or conditioning in the buffer storage of RWTP. 3. FINAL DISPOSAL A national repository for low level and medium level wastes (NRRW) is situated in the uranium min t Baita-Bihor.Thea e sit situates ei compaca n di t formatio crystallinf no e rock providing a solid geology and shielding, with low porosity and good chemical homogeneity. Because of the positioning at a mountain, no shallow underground water occurs and no risk f floodino g with surface expected e wateb n rca . Minin dons gensurwa o e t e acces ventilation.Thd san e present capacit galleriee th n yi s is 20,000 standard drums.The capacity can be enlarged up to 200,000 standard drums, providing additional galleries. 173 4. R&D ACTIVITIES r improvemenFo non-nucleae th f o t r fuel cycle radioactive waste managementD R& a , programme starte t RWTPda , wit hshora t mediuranga d ean m range planning. The short range period will cover: up-datin technica e procedureA th Q f go d standardlan C meeo s t IAE e EE tth d d Aan san recommendations. improvemen RWTe th f o tP general maintenance afte year0 2 r f operationso . reconditionin f damagego d waste drum transfed san r the NRRWmo t . definition of spent radiation sources by gamma spectrometry, treatment and conditioning. mediue Th m range period will cover: diversificatio planf no t processes accordin decommissioningo t VVR-e th f go S research reactor. study of radioactive decontamination methods. new treatment processes based on chemical precipitation, extraction chromatography, reverse osmosis and dialysis. improvemen conditionine th f o t g technologies. These R&D activities are based on the bi- and multilateral co-operation with the International Atomic Energy Agency and nuclear developed countries through research contracts and technical assistance. In the last years RWTP-Bucharest-Magurele benefited by the IAEA with a WAMAP mission, training courses, grants to international meetings and participate Co-ordinatee th n di d Research Programme entitled "Treatment Technologier sfo Intermediatd an w Lo e Level Wastes Generated from Nuclear Applications". e nexth tr year Fo areaw ne ss concernin e radioactivgth e decontamination, reactor decommissioning and waste treatment and conditioning are defined. 174 RADIOACTIVE WASTE MANAGEMENT AT THE DALAT NUCLEAR RESEARCH INSTITUTE NGUEN THI NANG Nuclear Research Institute, Dalat, Viem Na t Abstract The Dalat Nuclear Research Institute (DNRI) uses the reactor for iso6tope production, activation analyses basicand research neutronin solid-stateand physics. mainthe is producerIt of radioactive waste in Vietnam. The whole Institute generates about 100 m3 - 150 m3 of wet dryof liquidand radioactive 3 m 5 - waste about3 m and 3 waste year.per The system for radioactive waste management at the DNRI consists of two main parts: - Radioactive liquid waste treatment station; Storage- disposaland unit. The treatment station locateddesignedis basement the The building at the 2. of No capacity of the treatment station is 5 m3 per day. The station is collects radioactive wastes from the reactor operation and from radioisotope production and other laboratories. DNRI, the treatment the At methods currently used liquidfor wastes coagulationare and precipitation, mechanical filtration and ion - exchange. After being treated, the beta and gamma activities solutionthe of reache levels lower than 0.01 nCi/J. radioactivewet Dryand wastes collectedare stored disposala and in building unitthe in No 5. In this building, six concrete pits (100 m3 each) have been constructed for disposal and solidification of radioactive wastes. Up to now, cementation has been performed for 10 m3 of sludge waste with the proportion of waste to cement (W/C-ratio) of 0.47 - 0.5. Although the existing treatment systems for radioactive waste management at DNRI can in principle meet needsnuclearthe the of center, there stillis needa reevaluateto some majorits of components and also to optimize the operation of the system. There is an urgent need to reevaluate renovateand cementationthe volumeand reduction facilities sludgefor solid and wastes. 1. INTRODUCTION The Dalat Nuclear Research Reactor (DNRR reconstructes wa ) 198dn i 2 wit hele hf th p o the USSR and put into operation in March 1984. A short time later, the DNRI became the most important Nuclear Research Center of Vietnam. For the time being, it is also the main radioactive waste producer in Vietnam. The systems for radioactive waste management were newly designed and put into operation in 1984. Thank radioactive th so t e waste management system Nuclear ou , r Research Institute could do safely various researche applicationd an s nuclean i s r fields. Ther medicale ear , industrial, agricultura researcd an l h usersworkine people ar o Th . radioactivn gi ewh e waste management area have gained some practical experiences. In this report generae th , l problem f radioactivso e waste managemen DNRIe th t s it a ,t syste treatmend man t methods currently use radioactivr dfo e liquid wast e introducedear . Besides, results of the treatment and some difficulties met in radioactive waste management are presented. Last but not least, the suggestion about the ways of how to improve the present situation with radioactive waste management at the DNRI in both aspects, safety and economy. 175 2. WASTE ARISINGS AND WASTE CHARACTERISTICS Accordin IAEe th Ao g t TECDOC 656, Vietnam belong groue th contriespo C st a s i t I . country which has multi-use of radioisotopes and nuclear research center which is capable of indigenous production of several radioisotopes[1]. We use the reactor for training, isotope production, activation analyses and research. Many types of waste are generated in this nuclear center. Mos thef to m belon levelw lo ,go t shor t lived wastes. 2.1. Liquid radioactive wastes Mos radioactivf to e wastes generate DNRdn i liquis i I d waste whole .Th e Institute generates about 5 m -15 m liquid wastes per month. There are 0.5 m -1 m from the reactor operation, the 3 3 3 3 remaining from radioisotopes production and other laboratories. The quantity and chemical and radiochemical compositio waste th f neo greatly depen activitiee th n do f reactoso radioisotopd ran e departments. As nuclear medicine departments in Vietnam increase year after year, the DNRI has to produce more and more radioisotopes{2]. In this nuclear center, I -131, P-32, Tc-99m, Cr - 51 are produced for medical use resulting in solid and liquid wastes. The liquid wastes from collecting tanks contain mainly I -131, Cr - 51, Co - 60, Ce -139, Cs -134, Mn - 54 with beta activity less than 10-6 Ci/l and gamma activity less than 10-^Ci/l. chemicao Thertw e radiochemicaed ar an l l composition liquid e collectoe th th f so sn i r tank. The first case, "simple" compositio reactoresule e th th f ns o i t r operation, whil secone eth d one, "difficult1 composition is resulted from laboratory activities. These compositions are summarized in table I. TABUE I CHARACTERISTIC DALAF SO I TNR LIQUID WASTE BEFORE TREATMENT Parameter "Simple" composition "Difficult" composition pH 6-8 2-11 Conductivity (pS/cm) 0 2030 0- 1000 - 50000 Oxygen Deman Og 9/ld(m ) 3-5 5-25 Tota Activitß l y (nCi/1) 1-100 1-100 Co-60 (nCi/1) 10-50 10-150 Mn-54 (nCi/1) 1-5 1-10 Cr-51 (nCi/1) 10-50 30 - 300 1-131 (nCi/1) 1-5 10 Ce-139 (nCi/1) - 20-50 Cs-134 (nCi/1) - 20-50 solit we d d radioactivan 2.2y Dr . e wastes About 5 m3 of dry solid and wet wastes are generated per year. The compactible and combustible solids are paper, swabs, plastic, rubber (gloves), ion exchange resins, carcasses and excreta. The non-com partible and non-combustible solids are glass, scrap, brick work, sealed sources. Accordin IAEe th Ago t TECDO C 655 ,soli mose th df to radioactiv e wastes generatee th t da DNRI belon categorgo t 2[3]d an . y1 . EXISTIN3 G SYSTE RADIOACTIVF MO E WASTE MANAGEMEN DNRE TH IT TA The DNR equippes i I d wit systeha wastr mfo e management formee baseth n do r USSR regulations valid at the beginning of the eighties. The system includes the following: 176 1- Radioactive liquid treatment station, 2- Storag disposad ean l unit, 3- Control room for treatment station, 4- Laboratory room 3Liqui1 d treatment station The radioactive liquid treatment station is located at the basement of the building No 2 The desig consistt nI capacity statioda e f r sth o rr)f 5 pe y 3o n s i - 4 storage tanks (5 m3 each) for collection and precipitation wastes, ion-exchang-8 mechanicao tw d ean l filters, -16 pumps for solution and sludge, sludg-4 e reservoirs, - 4 storage tanks containing alkaline-acid solution Nowadays, at the DNRI, about 100 m3 - 150 rt|3 of radioactive liquid waste are treated by the treatment station annually Some problems encountered with the control- operation of the treatment station are due to absence th f segregatioeo f differenno t waste streams generated Thereb unable e ar us e o et yw specific treatment condition differene th so t t batches 3 2 Storage and disposal DNRe Atth I radioactive wast storeethis buildini n e I s th 5 buildin dn i o gN concretx gsi e pits (100 m3 each) are constructed for disposal and solidification of radioactive waste A "yellow tank" with volume of 8 rr>3 collects radioactive sludge from the treatment station for subsequent cementation Recently, annual 5 m3 of solid waste is collected in this disposal There are some facilities for cementation Storage and disposal of solid and solidified waste depend on the design and is directiy done into six concrete pits Transportation inside the building is done by an overhead crane, capable of lifting a concrete slab which covers the pits There are no any equipment for waste volume reduction The existing facilities for cementation are not operated at optimal capacity With the help and recommendation fro WAMAe mth P mission[2] cementatioe th , s beenha n don in-liny eb e mixing cementation process with containers in the form of drum of 200 I 4 TREATMENT METHODS OF LIQUID WASTES AND DECONTAMINATION FACTOR DNRIe A th ttreatmen e th , t methods currently use r liquidfo d wast coagulatioe ear d nan precipitation, mechanical filtration and ion-exchange 4 1 Chemical precipitation Liquid waste streams in the reactor building are collected in a tank of 5 m3 (See figl) From the collected tank, liquid wast pumpees i d into precipitation tank swast w wherra ee esolutioth ns i precipitated by chemicals Some chemicals and precipitation processes have been tested and used in DNRI The hydroxidee yar , phosphate, barium sulphate precipitatio combined nan d processes According to the test results in our laboratory, the best precipitation process is hydroxide M" nOH*+ M(OH)> = - ni where Mn+ is Fe3+, AI3+ etc Because ferric ions may already be present in the effluent, their floes are easy settled in the bottom tank[4e o solutiofth e Th j treatens i addiny db g FeSO valusolutioe H NaOth p d f ee o 4an HnTh must be more tha5 Sometime n8 e phosphat us e hav w so et e precipitatio r solutionfo n including radioactive strontiu decontaminatioe mTh n facto (ser0 (DF10 e frotableo )s t i m0 5 , III sII ) 177 o-Jo RAD. LieuiO MASTES LEAKAGES «NO DECONTAMINATION SOLUTION AFTER REGENERRTION WASTE STORRGE TRNK ST ST ST ST CEMENTRTION TO REACJOR HATER SUPPLY SYSTEM PT - PRECIPITATION TRNK ST - STORRGE TRNK MECHRNICR- M L FILTER M M I. EX I. EX I. EX I. EX I. EX I. EX - ION EXCHRNGE FILTER C I R I C I' R I' C I] - PUMP u c FIG, Principal1. diagram radioactivethe of waste treatment system NRI-Dalatat 4Ion-exchang2 e process Ion-exchange method has been used to remove soluble radionuchdes from liquid wastes After precipitation, the solution is pumped to the ion-exchange units (see fig 1) It is done m stepso tw mechanicay b , l fitter stagfolloweo tw ea y ion-exchangdb e colum activite Th n f betyo a and gamma emitting isotopes reaches the level lower than 0 01 nCi/l The DF is more than 1000 (see table III) The cleaning solution may be released into the environment or supplied to reactor water preparing system after activit chemicad yan l control TABL EH CHARACTERISTIC DALAF SO I TNR LIQUID WASTE AFTER PRECIPITATION Parameter "Simple" composition "Difficult" composition pH 8-9 8-10 Conductivity (uS/cm) 0 3040 0- 500 - 5000 Oxygen Demand (mg O?/!) 2-4 3-10 Total ß Activity (nCi/l) 1-10 1-10 Co-60 (nCi/1) 1-7 6 Mn-54 (nCi/l) <0.1 <1 Cr-51 (nCi/l) <15 <1 1-131 (nCi/l) 2 5 Ce-139 (nCi/I) - 1-2 Cs-134 (nCi/l) - 20-50 TABLE IE CHARACTERISTICS OF DALAT NRI LIQUID WASTE DURING TREATMENT PROCESS Type of pH Conductivity Oxygen Total ß Co-60 Cr-51 Mn-54 1-131 Ce-139 Cs-134 DF Solution (US/cm) Demand Activity (0/1) (0/1) (0/1) (CM) (0/1) (Ci/l) (Decont (mgOyl) (Ci/1) Factor) Raw 10-8 . 3 10-8 . 8 2-11 1000 - 50000 5-25 10-'7 - 10-'- lo-s 210- - 210-8- Solution io- 810-8 3 10-7 10-8 510-8 510-8 After 10-'- 8-10 500 - 5000 3-10 610-' 510-' 10-' 210-'- 10-9 210' 10- precipitation 10 8 510-' 100 After 10'- 10 8-10 50 0500- 0 3-10 8 IG"' 10-1° 10-10 4 3 Conditioning of sludge Both sludges from the precipitation tank and the first part regeneration solution (from ion- exchange filters collectee ) ar volum 3 tankm 4 1 d m f e so eac interir hfo m storage The sludge nth es i pumped along pipeline to the waste storage serving as the feeding tank for the cementation process Until now, cementation has been performed for 10m3 waste sludge with total activity of 10 ' -10 6 densitCi/e Th l sludgf yo mores i e than 10g/ cementatioe lTh bees - nin ha n e donth y eb line mixing cementation process with a W/C-ratio of 0 47-0 5 The cement is fed with a screw feeder while the waste is fed with mono-pump From the mixer the cement waste mixture is directly released into the storage container (drum of 200I) 179 CONCLUSIO. 5 SUGGESTIOD NAN N The DNR equippes i I d with system r wastsfo e management formee baseth n do r USSR regulation beginnin e valith t eightiesde a th f go considere.e Thougb n ca ht i d adequat principaen i l for need thif so s center, ther stiles i l nee reevaluatdo t e som t majoi f eo r components alsd o an ,t optimize the operation of the system. First, facilities for cementation and volume reduction of sludge and solid wastes mus equippede tb . Ther urgenn a es i t nee assisdo t upgradinn i t waste gth e management syste DNRe th mt Ia including safety, economy and environmental assessment for radioactive waste management REFERENCES ] [1 INTERNATIONAL ATOMIC ENERGY AGENCY, Treatmen d Conditioninan t f o g Radioavtive Organic Liquids, IAEA-TECDOC-656 [2] Bergman, C., Detilleux, E., Berci, K., Report of WAMAP Mission to Vietnam, 30 November - 5 December 1992 [3] INTERNATIONAL ATOMIC ENERGY AGENCY, Treatment and Conditioning of Radioactive Solid Wastes, IAEA -TECDOC-655 ] [4 INTERNATIONAL ATOMIC ENERGY AGENCY, Chemical Precipitation Processer fo s e th Treatmen Aqueouf to s Radioactive Wastes, Technical Reports Serie 337o sN , IAEA, Vienna, 1992 180 EXPERIENCE IN THE MANAGEMENT OF RADIOACTIVE WASTE BANGLADESN SI H M.M. RAHMAN Bangladesh Atomic Energy Commission, Dhaka, Bangladesh Abstract Banglades bees hha n firmly committe peacefue th o dt l application f ionisinso g radiations in agriculture, medicine, industry and research in order to achieve socio- economic developmen diversn i t e sector shore th tn s i ter wels ma lons a l g term perspective since its emergence in 1971. Consequently f radioactivo e us e ,th e material radiatiod san n source alwayy sma s produce radioactive wastes warranting safe, planned and proper management so as to protect man and the related environment (at present and in the future) from the undue risks of ionizing radiation. Bangladesh Atomic Energy Commission(BAEC principlen i , is ) , responsiblr efo developin implementind gan gnationaa l strateg necessard yan y infra-structure th r efo collection, handling, treatment, conditioning, transportation, storage and disposal of radioactive wastes including their regulatory control, duly considering the local condition socio-economid san c patter countrye th f no . A considerable progress has already been made in this regard. The Nuclear Safety and Radiation Contro t (NSRAc l C Act, 1993 promulgates )wa n 199di d 3an necessary regulations(sub-ordinate legislation) are under preparation in the light of the act, government policy and the up-to-date internationally accepted waste managemen disposad tan l recommendation standardse th d san . Special emphasis has been attached to the Basic Safety Standards for Protection Against Ionizing SafetRadiatio e e Radiatioth th r f yo fo d nn an Sources . Source f radioactivo s e wastes: The main sources presently include - a) operatio maintenancd nan TRIGW 3M A f eo Mar researcI I k- h reactor b) production of mostly short-lived radioisotopes : 99mTc, 131 Jroutine basis n nuclean i e forus r medicine o c) production of ^Sc for sedimentological studies at the Chittagong Port d) production of 32P, 35S, 5ICr, 192Ir »^A^etc. in future subject to the availability of appropriate facilities. e) extensive usages of 125I, 51Cr, 57Co and sealed 90Sr (for opthalmic purposes), in medicine and 65 Zn, 32p, 54Mn, 35s, 3/H and *4C in researcn i o teachinC d 0 han 6 d gan s agriculturC 7 13 d ean 181 0 a significant number of sealed 192fr sources (around 100 Ci each) constantly used forNDT purposes in the industrial sectors ) g sealed 22<>R previouslas neediegm 2 nearlf so o t y 5 usey 1. brachytherapyr dfo . h) several sealed radiation sources of various strengths ranging from 500 Ci to 120 kCi of ^Co and 137Cs are used for radiotherapy, medical sterilization and industrial processing of food and R&D purposes which will produce used/spent sources needing special managemen futurn ti e i) a large amount of imported thorium nitrate used in several Gas Mantle industries j) a large amount of monazite tailings (ore of thorium) following separation of heavy mineral beacf so h sand f Cox'so s Baza off-shord ran e islande th n si Bay of Bengal In addition to the above, Bangladesh is contemplating to build up a nuclear power plan futurn ti whicr efo h separate waste management infrastructure wil needede lb . Radioactive waste from Research Reactor (3MW TRIGA) a) Solid waste: s Mostly short h'ved having low specific activity, e. g., contaminated mops, tissue and absorbent papers, stack and ventilation filters, trash papers, gloves, clothings, foot wears, corrosion and activated products, mixed bed non-regeneable ion exchange resins, etc. b) Liquid wastes : Mostly of low specific activity , e.g., spent-ion exchange resins, ion-exchange regenerates, filter-transfer liquors, leakage froe systemsth m , plant decontaminations, washings, etc. c) Gaseous waste: s 1 131 Noble activated ga Ars4 , fission gases (Kr, Xe, etc) HalogeI fron m irradiated - targetTe , ventilatio stacd nan k air, allied discharges, etc. Probable sources of radioactive wastes due to prolonged operations, maintenance and allied activities of the research reactor taking into future consideration of rupture of fuel elements, fuel claddin follows a e b g s: y failure ma c et , i) drippag liquif eo d fro fuee mdi l elemen transfee di n to r from reactor top, etc. ii) handlin f rupturego d fuel element transfen so f fueo r l elements between fuereactoe e di di l d storagran e pool iii) wastes whic arisy hema from fuel cladding rupture, e,g. gaseous fission products 88Rb (t^=18 min), daughte 88f 2.8hrinitia= rKo e | t rdi ( t a l) stage 88ane Sddi r followin decae gdi f 88 yafso Kd oran 137 137ßCd san a 8 2 2727 iv) wastes fro reactoe mdi r coolant, e.g Al,. ) AAL(n,rt)24N1OM a 55pe, , etc. 182 In future the spent fuel originated from the research reactor will be safely stored in the spent fuel storage rack adjacent to the reactor core. So far cladding failure of any fuee oth fl element t occurno d .di s Therefore leakago , n ther s f fissioewa eo n product reactoe th n si r system. Options for rîdioactive waste management The main options in the field of waste management particularly in case of low level and short-lived wastes presently adopted in the country include releas atmosphere th o et e under controlled conditions discharge to the sewers on evaluation of safety aspects disposal to the normal municipal land fills conditioning and treatment safe storage. Management of radioactive wastes from nuclear medicine centres and other organisations : No significant wastes arise from the therapeutic usesI .o fTh131 e excertions are controlled at the place of origins. The radioisotopes presently used for clinical and other purposes include I3IJ, 125i799Mo-99nrrc,55Fe,57cOand5lCr. Solid waste temporarile sar y store ultimateld dan y disposemunicipalite th t a f do y land dependin radioactivite th n go y contents, after sufficient delayin decayingd gan , only after attainment of international exempt levels as per IAEA safety series 89, IAEA, Vienna (1988). The liquids are mostly short-lived and are temporarily stored, diluted and released to die sewers under controlled conditions. The gaseous wastes, presently 4 1 Ar only from the reactor and 1 31 1 from the radioisotope production facilities are discharged to the atmosphere through the stack after proper monitorin controld gan . The sources from outside organisations commonly include 60Q>, thorium nitrate, 226Ra-needles, etc. A data base inventory of all spent or used sources including ~6Ra-needles is under preparation. After collection of the spent sources depending on the half lives and radioactivity contents particularly, the ^Ra- ncedlcs wil conditionee lb .centralld dan y stored. Technical Assistance from IAEA : a) Title : Radioactive waste management and related environmental problems. T.C. : Projec . BGD/9/005No t . Period covere : d 1984-87. 183 Status : Completed. b) : Title Safe management of radioactive wastes. T.C. : Projec . No t BGD/9/007. Status : On-going. c) Research Contract : Title of Project : Development of improved liquid radioactive effluents treatment technolog precipitatioy yb ion-exchangd nan e anrelatee dth d analytical control system. Project No. R/C No. 6794/RB. Period covered 15/11/9 14/11/91o t 15/03/92& 14/03/943o t . Under the scope of the RC project ferrocyanide ppt. method has been found to be suitable for the treatment of low level liquid radioactive wastes containing isotope of Cs.The bes valuH tp removo et fros radioactivee C mth e solution. 10 founs si e b o dt Tae contamination factors for different pH values (Cs-137 activity = 3.7 kßq) is show figurtabld n i an . e1 2 e Regulatory aspec f radioactivo t e waste managemen: t In exercis powere th f eo s conferre sectioy db nNucleae th 3(a f o ) r Safetd yan Radiation Control Act (Act No. 21 of 1^93) Bangladesh Atomic Energy Commission(B AEC) is the competent authority in the country to formulate necessary TABLE I. INVENTORY OF RADIOACTIVE WASTE AT AERE, S AVAR (Jan. 1987- July 1994) Typ wastf eo e Max.container surface Amount/volume Origin dose rate (mSv/hr) levew Lo l solid 0.26 0.53m3 RR levew Lo l liquid 1.00 litre2 25 s RIPL IFRB Spent ion exchange 0.02 16. 7kg RR resins RR ===> Research Reactor RIPL =====> Radioisotope Production Laboratory IFRB====> Institut Foof eo Radiatiod dan n Biology 184 TABL . DECONTAMINATIOEII N FACTOR DIFFERENR SFO VALUEH Tp S (I37Cs activity kBq7 3. : ) Amountf so Amountf so Amounts of Expt. No . pH K4[Fe(CN)6] NiSO4 CsCI DF 0.75M 0.001 M 1. 2 1ml 1ml 10ml 220 2. 4 1ml 1ml 10ml 98 3. 6 1ml 1ml 10ml 347 4. 8 1 ml 1ml 10ml 497 5. 9 1ml 1ml 10ml 618 6. 10 1ml 1ml 10 ml 654 7. 11 1ml 1ml 10ml 598 8. 12 1ml 1ml 10ml 93 regulations or policies or issue orders or instruction for the management of radioactive wastes and to take appropriate steps to implement them. managemend sectior A an regulat y spe e radioactivus ma e ne th C 3(f)ef th to AE ,B e wastes. As per section 4(1 )-b, any person without holding the pertinent licence(s) shall not brin makr go e entranc vehicly an f eo e into Bangladesh operate nucleay db r power ro carrying radioactive materials or radiation producing equipment or radioactive wastes. 700 FIG.pHvs1. DF. 185 Regardin radioactive gth e waste disposal recene th , t IAEA Safety Series, IAEA Technical Reports Serie IAEd san A Technical Documents shal consultede lb . The Commission shall provide guidelines and procedures to be followed by the licensees for waste management and disposal. In addition, Commission shall adop IAEe tth A Safety Serie Regulatio: s6 e th r nfo Safe Transport of Radioactive Material, Vienna (1990) for the guidance of transport of radioactive materials. Future directions : 1. To formulate frame, prepare and finalise regulations, codes of practice, guidelines, etc. pertinen safe ih e o management disposad tan radioactivf lo e wastes arising from the uses in medicine, industry, agriculture and research. 2. To plan, design and construct a central waste storage cum treatment facility at AERE, Savar. studo T y geologica . 3 hydrologicad lan l aspect selectioe di r s fo shallo f no w land repository site. For the materialisation of the above, IAEA technical assistance is badly needed in near future. 186 WASTE TREATMENT OPTIONS AND PRACTICES CHEMICAF O E US L PRECIPITATION PROCESSER SFO LIQUID RADIOACTIVE WASTE TREATMENT V. ZABRODSKY, N.B. PROKSfflN, A.S. GLUSHKO Institute of Radioecological Problems of Academy of Sciences, Minsk, Belarus Abstract The process of removal of 137Cs and 90Sr from simulated and real liquid radioactive wastes by chemical precipitation followed by centrifugal ion was investigated depending on sequence of different factors. These factors are: chemical compositio solutiof o n n (salt content, pH, surfactants, nature and concentration of chemicals used for precipitation), ageing time, typ centrifugf eo e roto othersd ran . 1. INTRODUCTION Large amount liquif o s d radioactive waste generatee sar d during decontaminatiof no industrial facilitie d equipmensan t contaminated afte Chernobye th r l accident liquie Th . d radwastes contain Cs [(10- -10- )Ci/l], Sr [(10- -10- )Ci/l], large amounts of surfactants, l37 9 8 90 9 8 poly-phosphates, heavy metals and others components. Specific feature of these radwastes is the high concentratio f salt suspendeo d n san do thi t solids se reasoDu . n chemical coprecipitation method seem preferable b o st r theiefo r treatment constructioe Th . piloa f no t scale installation consisting of chemical reactor with device for overflowing of clarified solution, sand filters and tank for purified water is now at the final stage The cation exchanger OH BAH [1], clynoptiloiite and quartz sand are supposed to be used as filtering materials. The work on developing the methods of increasing of sorbtion ability of clynoptilolyte and cation exchanger OMBAH towards radionuclides 137Cs and 90Sr were fulfilled within the framework of the State Programme of Mitigation of the Chemobyl Accident Consequences. It is possibl increaso et sorptioe eth n abilit thesf yo e sorbent severay sb l hundread time usiny b s g a method of grafting of complexing groups to the surface of sorbent Besides using of gravity other method solid/liquif so d separatio beine nar g develope r Institutou n i d e .Those ar e ultrafiltratio centrifugationd an ] [2 n . 2. EXPERIMENTAL The specific sorbentr sfo Cd Ss an rnicke- l ferrocyanid hardld ean y soluble saltf so J37 90 calciu mwer- e forme solution i d mixiny nb g K.4[Fe(CN)6] (0.025 mol/1) with NiCl2 (0.05 mol/l) or Na3?O4 (Na2C2O4) with CaCl2 by following a stoichiometric course. All solutions as a rule were prepared using a tap (running) water. Ge-Li semiconductor detector was used for determination of ^7Cs concentration in solutions. The following operations preceded the determinatio 90f no usinSy rb g ß-counter daughtee Th . r radionuclid s coprecipitatewa Y ^ e9 d with iron (III) hydroxid thed an n 7.0 H 5 separatep 7. -t ea d fro solutioe mth filtrationy b n . Microfiltration through polyethylene terephtalate nuclear membranes with different sizf o e pores (developed at JINR, Dubna).and sedimentometric analysis were used for the determination of dispersivity of sorbents being formed in the solution through chemical precipitation. Testing of solid/liquid separation was carried out on flow separator with a plate or two-chamber rotors . The separator with plate rotor has following technic characteristics: 189 • Throughput - (2-50) litres/hour • Rotational speed - 10 000 rpm • Plate diameter (on the generatrix). • maximal - 77 mm • minimam m 2 3 - l • Volume of sludge space - 0,008 I. carry-ovee Th solie th df ro phas e together wit centrifijgee hth d effluen determines wa t d by its filtration through a nuclear membrane (pore diameter 0. Î Spm) folllow weighting of the dryed membrane. Accordin experimentae th o t g l data obtaine increase dcentrifugth e th f eo e throughput (decreas residencf eo e tim solif eo d particle fiele f th actio do n si centrifugaf no l forces) leads to the rise of the carry-over of solid phase with the centrifugate. It should be noted that at the same throughput the carry-over is less for slightly soluble salts of calcium the nicker nfo l ferrocyanide largee Th . r densit calciuf yo m salt obviousls si maie yth n reasof no that. Apart from the plate rotor a two-chambers rotor has been tested. A larger volume of sludg econsideree spacb y ema it's da s advantage carry-ovee th t Bu . nickef ro l ferrocyanidn ei the case of the two-chambers rotor was larger then for thin-layer one. The centrifuge throughput correspondin minimao gt l carry-ove solif o r d substance with centrifugat bees eha n choosen for subsequent experiments. 3 RESULT DISCUSSIOD SAN N dependence Th décontaminatioa f o e n factor (DF varioun )o s factor bees sha n investigated These factors are. histor f solutionyo s (simulate r reado l liquid radwastes) f solutionso H p , , content of salt, content of surfactants, ageing of suspension, nature and concentration of sorbents formed in solution during the chemical precipitation. 1 Chemica 3 l compositio treatee th f o nd solutions S.l.I.pH. Table 1 contains the 137Cs decontamination factors derived for suspensions formed at different pH. According to the data the efficiency of the treatment process practically does not J.WV • 1 I . ! * Fig Retentio. 1 . l^Cf no s during 90 - ^ microfiltratio f nickeo n l ÖsUo . ferrocyanide suspension (2.4- 1(T3 mol/1). Pores diameter: 70 - B 60- 1- 1.35 urn: 2,3- 0.06 urn. *1 Ageing:1.2-freshry prepared; 3- 0 5 % R, H2 ———————————— ———————* — 1 24 hours. 40 -—— AA3« —————————————————————————————— 30 - 2U - 10 -1 * . ft . . , 10 12 pll 190 Fig.2. Efficienc removaf o y f lo fro solutios e mC th ? ' n versus salt concentralion. {Thin-layer separation. Ageing time j hour. Distüled water.) 0 15 0 10 50 200 [NaCIJ^l precipitatioe th f o depen H p suspensionf n ndo o . Microfiltratio f freshlno y forme d agean n di d 1 day suspensions of nickel ferrocyanides has been carried out by the use of the nuclear membranes with various pore sizes (Fig . Accordin) .1 resulte th o gt s obtained thero n s ei substantial chang dimensione th en i f nickeso l ferrocyanide particle intervaH p n si l froo t m5 10.5. The most of these particles have dimensions greater than 1,35 urn immediately after the formation of suspension (reducing of l37Cs retention at pH>10,5 is caused by the increase of nickel ferrocyanide solubility at this condition). According to the data obtained, pH values in the 90 interval 9-10,5 may be suggested for one-stage precipitationC osf 137 anSrd. The data obtained made it possible to assume high speed of ion exchange reaction of 137Cs with nickel ferrocyanide. The formation of ferrocyanide immediately in the solution (in situ) and absence of a diffusion stage in sorbent phase confirms the high speed of sorption. 3.1.2. Salt content. The experimental data obtained allow to make a suggestion that following factors have an influenc efficience th n eo treatmenf yo sale th t f concentratioi t solutioe th n i increaseds ni : • increase of carry-over of precipitate during centrifugation by encreasing of difference between density of solid substance and density of dispersing medium; • enhancement of nickel ferrocyanide coagulation; • aggravatio 137Cthe s ionof n exchange sorptio nickenby l ferrocyanide. A resulsa fore tcurvth f m versuo F e"D s salt content" (Fig.2 rathes i ) r complicated. 3.1.3. Surfactants Usually surface decontamination compositions contain detergents which are the mixture of surfactants and complexing agents. For example, detergent "CO-2V is commonly used for cleanin surfacee th f o gindustriaf so l equipment contaminated afte Chernobye th r l accident. That detergent has the following composition: 50% of sodium polyphosphates, 25% dodecylbenzenesulphonate and 18% of sodium sulphate. According to the legislation of the Republic of Belarus the discharge of effluents containing dodecylbenzenesulphonate into sewage collection system is prohibited. Two methods of the treatment of liquid radwaste for separation of the surfactant are being developed: • precipitative method by using of calcium salts, • destruction of the surfactant by ozonization. 191 Table 1. Removal of Cs from the solution by chemical precipitation and thin-layer centrifugation. Throughpu 13/ {5.6-8.6)l/s ti h Solution composition PH [K4(CN)6]. [NiCl2]: additional component Ageing DF moi/1 moll timen ,mi 10.0 5.0-10-4 -IQ"10 3 60 5±3 30.0 1.0-10-3 2.0 -IG"3 60 10±6 10.0 3.0-10-3 6.0 -10-3 60 20 ±~ 6.0 2.5-10-3 5.0-10-3 40-60 26±13 8.0 2.5-10-3 5.0-10-3 40-60 31X8 1 n n _ ^ '-.- «1 ;. 3 c r> -•r.~~- -."'-<-, " . î y. i '•••. 5 • I G" " 5 .(: -II'.' " ,C<4;-2y. t-^ui ~ ~_ . - 30.0 2.5-10-3 5.0 -lu'3 ] 5-20 Î ± 11 : 10.0 2.5-1 0-3 5.0-10-3 [FeCK|=l-K)- moll 60 ]]0r5(- 1 0.0 2.5-10-3 5.0-HÎ-3 60 ]:(.':. 71; 10 ') 2.5-10-3 7.5 -ÎO-3 r~. : •- -_«,(,< 1,1 i,i 5.75-10-3 5 (. .in- ? i , " :.'-^• ' 1 Ju.ü 1.U-1Ü-3 2.U -Kr-" Real radw aste 4U-6U 2u±5 Table 2. Removal of °°Sr from the solution by chemical precinitation and thin-!a>ei centrifugation. pH=10. Precipitate, moll Dispersing i Calcium Calcium medium Ageing time DF oxalate phosphate 5.0-10-2 simulated 24 hours 3.3 solution 2.5-10-- real radwaste 24 hours 4.1 2.5-10'2 simulated Freshly prepared 73 solution 1.0-10-2 simulated Freshly prepared 43 solution 1.0-10-2 real radwaste Freshly prepared 48 lattee Th r metho mors di e efficien generatd an t e less amoun wastef o tsolidified e b o st t A . the same tim precipitative eth e method doe requirt sno additionay ean l equipment. The quantitative data obtained in this work show the reduction of the solution decontamination versus the increase of the surfactant concentration (Table 1). 3.1.4. Rati chemicalf oo s use r precipitatiodfo n It is known that excess of FeCCN^,4' in solutions iasilitate the formation of stable ferrocyanide sols. There are some data concerning chemical precipitation of nickel ferrocyanide 1 4 4 suspension t variousa s Ni"2 / Fe(CN)c,- ratios (Tabl . Increasine1) f Fe(CN)(3go - concentration i solution decreasin e leadth o st f decontaminatioo g n factorscontrarye th n increasinO e . th , f go above-mentioned ratio lead greateo st r efficienc decontaminationf yo . Thi compliancsn i date ar a e with the results of studies on the kinetics of nickel ferrocyanide sedimentation 192 «u • Fig.3. Efficienc removaf yo f lo A i3"Cs from the solution by W chemical precipitatio thind nan - A » . layer separation versus ageing time of nickel ferrocyanide 40 suspension. DF 30 • ^W / ' A 10 ^ /A 0 20 O IS 0 10 O S 0 Time,nMnute The data in Table 2 indicate that higher strontium decontamination factors are achieved in the case of the use of calcium phosphate as sorbent, than in the case of calcium oxalate. The conclusion may be made that higher decontamination reached during centrifugation in the case of calcium phosphate is due to it's larger density (3 14 g/cm ) in comparison with that of oxalate 3 (2.29 g/cm3). 3.2. Suspension ageing According to the experimental data the increase of the ageing time of suspension leads to increase th decontaminatiof eo n efficiency (Fig.3) suggestioe Th .mad e b y ee thanma th n i t cours timf o e e fine particles stick togethe produco rt e larger particles. Sedimentometric analysis of freshly precipitated and aged for one day nickel ferrocyanide has been carried . The results obtained show that agein suspensiof go n caus increase th e particlf eo e dimensions froo t m4 18 urn. That data are confirmed by the results obtained by using optical microscopy. So ageing of the precipitate facilitates the phase separation. But it should be noted that dimensions and mass of freshly precipitated particles are also big enough to perform quantitative isolation of radionuclides (Table 1,2) 3.3. History of the solutions The coprecipitation study has been performed with both simulated and real liquid radwastes The last ones were collected during the decontamination of different industrial equipment. This Gomee worth n i kl provinc Belaruf eo beins s i e Statth g carrieey b specialize t dou d enterprise "Polesje". Characteristics of surfaces to be cleaned require the strong base {(5- 10)%} decontamination composition necessars wa usedt e i b o o mako yt st S . e it's partial neutralizatio pH<1o nt 0 before proceeding wit chemicaf ho l precipitation e formatioTh . f no stable sol f so polynuclea r hydroliti existinl A solution e d th can n g i form n zinkates Z s a f so d an s aluminates took place durin neutralizatione gth . Accordin resulte th a go s t f obtainedo e us e ,th high speed centrifug a perspectiv s ei e metho r clarifyindfo f thesgo e suspensions. About (90- 95)% of suspension material can be precipitated by use of separator with rotational speed 10 000 rpm. So the conclusion may be made that sufficiently high ceasium and strontium decontamination reachee factorb y rear sdma fo l liquid radwastes. 193 . 4 CONCLUSIONS Two kinds of centrifuges are used now for liquid radwaste treatment. At first stage decanter is use separatior dfo relativelf o n y heavy particles fro solutione mth clarifiee Th . d solutiod nan dewatered sludg e produceear thit a ds wast e stagth f eo treatment. More fine purificatiof no solution separatorf takeo e sus concentratee e placth .Th y eb d suspensio f finno e particlen o s the outlet ofthat device is directed to the decanter for further dewatering. It seems expedien introduco t procedure eth chemicaf eo l precipitation into above mentioned flow sheet between decanter and separator. In that case the process of chemical precipitation and sedimentation would be taking place in solution clarified from the bulk coarse-dispersion settle suspended dan d particles. REFERENCES [1] SOLDATOV,V.S., SHUNKEVICH,A.A., SERGEEV,G.I Synthesis, structure and properties of new fibrous ion exchangers. Reactive Polymers, Ion Exchangers, Sorbents 7 (1988) 159. [2] ZABRODSKY, V.N., DAVIDOV, Y.P., TOROPOV, I.G., GLUSHKO, A.S., EFREMENKOV V.M. "Treatmen liquif o t d radioactive waste using combinatio f chemicano l processes with ultrafiltration", Nuclear Waste Management and Environmental Remediation (Proc.1993 Int. Conf. Prague), Vol.3 (BASCHWITZ,R., Ed.), ASME (1993) 719-722. 194 ENVIRONMENTAL IMPACT ASSESSMENF TO OPERATIONAL PRACTICE PROCESSINR SFO G LOW LEVEL LIQUID WASTES IN THAILAND P. YAMKATE, F. SINAKHOM, P. SUPAOKIT Waste Management Division, Office of Atomic Energy for Peace, Bangkok, Thailand Abstract effluente Th f averago s e activit f 0.0yo 1 Bq/L gross alph 0.2d aan 2 Bq/L gross beta, about 300-900 m3 per year have been discharged from the centralized low-level liquid waste treatment plant int canale oth s connectin Chae th oo gt Phya River since 1965 impace Th . n o t the environment is routinely monitored since then. The current results revealed that the radioactivity in the vicinity of the OAEP is kept under control and acceptably low. This indicates that there is no potential hazard to members of the public. 1. INTRODUCTION Sinc e 'atomieth c energ r peaceyfo ' programm bees eha n introduce Thailano dt n di 1962, the application of nuclear technology has gradually increased. At present, nuclear technique quite sar e well-know mosy nb f Thao t i scientists, academicians, medical doctors, industrial people, etc.The principal nuclear technologies are those used in the medical treatment, therapy and nuclear medicine, research work, agriculture and industry.The developmen f nucleao t r technology, however, create certaisa n negative aspecte , th tha s i t increasing generation of radioactive wastes. worldwida s i t I e practice that radioactive wastes kephave b o tet under contro thad an tl their potential impact on man and his environment has to be acceptably low.lt is, therefore, police e Thath th f i yo Atomi c Energy Commission thaOffice th t f Atomio e c Energr yfo Peace (OAEP) has to render the service on the management of the radioactive wastes arising fro l radionuclidmal e user Thailand.ln si e Wast dutth e f th yo es i tManagemen t Division (WMD) to fulfill this mission properly. As the main radioisotope users are those in the medical sector, it is found out that the major waste producers are the hospitals in the central region of Thailand. However, regarding to the volume of waste produced, 80% of the total volume can be assigned to the OAEP, wher Thae eth i Nuclear Research Reacto othed an r r relevant facilitie locatede sar . Since 1965, the well equipped chemical co-precipitation plant for waste water treatment s bee t intha npu o operatio e OAEth t na P premises waste Th . e water collected frol mal institutions using unsealed radioactive substances in Thailand including OAEP internal laboratories s beeha , n delivere treatinr dfo thin gi s plant.The liquid wast predominantls ei y aqueous solution, with low content of salt, and small amount of organic liquid, the quantity of untreate dr year.Th wastpe 3 m abous e i e 0 mai80 t n radionuclides containe waste th n edi are, 14H C3 : ,32 P, 35S, 36C1, 45Ca, 51Cr, 55Fe, 57Co, 59Fe, 60Co, 65Zn, 90Sr, "Tc, 1251,13II, 137Cs, 232 238radioactivite d TUhan Th . y liqui w containera de wasteth n i d betwees si n 3.7-37 Bq/L. Low level waste is treated by using the alum-coagulation process at the treatment plant. The concentrated sludge residue is kept in the form of solid radioactive waste. The effluents 195 • - Sompling ilotion - Thowo Sunthon T»mpl« X.2 - Lord Yoo Priion L3- Bong Kh«n Conol Wot«r Got« L4-Tong Luang T»mpl« L5-Kino Mongku» » irutltvjt« of îl«hflol«gy 1 L6- PoX Nom Ttmplt L7~8ong Ktiun Ti»n Conot L8-North boundar f Kotttwro y t Univmily L9- Bong Bua Morku LlO-Tung Song Hong L U-Klong Pro Po FIG. Environmental1. sampling point OAEPat site. have been discharged int OAEe oth P pon eventualld dan y disperse canala dvi s int Chae oth o Phya River. The release of the treated effluents by OAEP is permitted only after the radioactivity conten waste th s bee f eha o t n analyze compliancn i found e b d an o dt e with authorized limits. The average activity in the effluents at the discharge station is 0.01 Bq/L gross-alpha activity, 0.22 Bq/L gross-beta activity mBq/3 mBq/8. ,2 3. 137f Lo d 90C f LSo san r respectively. Since the radioactive effluents are discharged into the aquatic environment, those are subjected to consumption of the people in the downstream of Chao Phya River, where Bangkok is situated. The assessment of health detriment is then very imperative in two respect ; sfirstl t providei y quantitativa s e measur radiologicae th f eo l impac f efflueno t t discharge secondld san sourca essentiae s th i f t yi eo l informatio r improvinnfo liquie gth d waste treatment systems.The system of dose limitation recommended by ICRP [1] is used as basie controllinr th sfo g exposure considerationo .Tw emphasizee sar d: 1. All exposures shall be kept as low as reasonably achievable, economic and social factors being taken into account (optimization of protection). . 2 Committed effective dose recommende individualr dfo ICRy sb Pexceedee b shal t no l d (dose limitation). radiation I n protectio memberr nfo publice th f so , ICRP recommend seleco st groua t p populationf o calleo s , d "critical group" groue assumeo th , pwh receivo st highese eth t dose consequencs a wastf eo e discharged committee Th . d effective critica e dosth o et l group will be calculated moso Tw .t important radionuclides discharge Bangkheo dt n Cana 137e d Clar san 90Sr. Both radionuclides are then used for dose assessment. 2. METHODOLOGY The methodology for evaluating the impact of routine release is based on guidelines of ICRP exposuree Th . radionuclideo st s that originat effluene th n ei t were converte estimatdo t e of doses to critical group using data from the environmental measurements. The approach used estimato t s i dose eth e ove whollife e reth th individuan f timea o f eo years0 l 7 (take e b , o nt from intak particulaa n ei r year). The exposure pathways associated with radionuclides discharged into freshwatee th e ar r consumption of water, aquatic populations, irrigation leading to contamination of foodstuffs and external radiation from sediments thin I .s assessment pathwaye th , whicy sb h members of the public are most likely to be exposed are through drinking water and ingestion of water plants.They are then used in the calculation of dose to the critical group [2] resided along the canal. Routine monitoring is conducted to measure the dispersion of radionuclides in the aquatic environment. Eleven sampling station locatedownstreae d sar an p du m fro release mth e poin. 1 shows g a t Fi n ni 3. RESULTS AND DISCUSSION Tabl eshowI e annuasth l average concentratio surface I37f 90th no d SC n i rsan e water and water plants (swamp cabbage, epomoeareptans) from 11 locations. Maximum concentrations of 137Cs in water and swamp cabbage are 9.9 mBq/L and 0.063 mBq/kg 197 respectively. Maximum concentration waten i swam d S ran 90 f spo cabbag mBq/5 3. d e Lear an 0.051 mBq/kg respectively. Table II shows committed dose to critical group via ingestion of surface water and swamp cabbage. Maximum organ dose to the critical group which maximally consume water and swamp cabbage are shown in Table III. TABLE I. ANNUAL AVERAGE OF RADIONUCLIDES IN CANAL WATER AND IN WATER PLANT (SWAMP CABBAGE) LOCATION Canal Water Water Plant 1"Cs(mBq.L') "SKmBq.L1) 137Cs(mBq.Kg-1) 90Sr(mBq.kg-1) Downstream L01 8.3+1.9 3.2+2.0 0.063+0.018 0.042+0.011 L02 2.8+2.0 1.3+2.0 0.064+0.004 0.027+0.010 L03 6.2+2.1 3.5+2.2 0.002+0.018 0.024+0.011 L04 8.2+2.0 2.6+2.2 No Sample No Sample L05 6.5+2.0 0.03+2.2 No Sample No Sample L06 6.0+2.0 3.0+2.1 No Sample No Sample L07 6.8+2.2 0.7+2.2 No Sample No Sample Upstream L08 7.6+2.0 1.6+2.0 0.044+0018 0.051+0.012 L09 9.9+2.5 1.9+1.9 0.045+0.019 0.037+0.012 L10 7.3±2.2 2.3+1.8 0.009±0.019 0.031+0.011 Control L11 2.6+2.1 1.8+2.3 No Sample No Sample TABLE II. COMMITTED EFFECTIVE DOSE FROM OAEP DISCHARGE VIA INGESTION PATHWAY Committed effective dose , x 10-" Sv Pathway Radionuclide 1y 10 y 15 y Adult Drinking Water s 13C 7 2.72 6.93 9.01 9.01 90 Sr 11.5 10.8 7.35 6.86 Water plant 137 Cs 0.69 2.46 3.60 3.28 90 Sr 6.7 8.76 6.72 5.71 Oassessmene nth criticaf dosee o t th o st l group, maximum concentratio 137f no d C san ^S in water and swamp cabbage are used. It is assumed that the critical group consumes all vegetable grown in the canal and drinks the water from that canal as well. At the same time, the food consumption rate for members of the Thai population at different age group (i.e. ly, adultd an takee y )ar 5 n1 1fro, 0Fooy e mNutritiod th dan n Division Ministre th , Publif yo c Health achievo t , assessmente eth dosee unir Th . spe t intak differenf eo t radionuclides ("dose coefficients") are taken from the publication of the National Radiological Protection Board Unitee th f o d Kingdo whic] m [3 mos e base e th h ar tn do recen t recommendation ICRf so . P60 198 TABLE III. MAXIMUM ORGAN DOSE FROM OAEP DISCHARGES VIA INGESTION PATHWAY Maximum orgav nS dose0 1 x , iy 10 y 15y adult Drinking Water s 13C 7 3.26(stomach) 7.62(uterus) 9.70(uterus) 10.40(uterus) 90 Sr(Bone surf.) 124.8 127.4 98.0 95.6 Water plant 137 Cs 0.83(stomach) 2.71 (uterus) 3.88(uterus) 3.78(uterus) 90 Sr(Bone surf.) 72.8 103.5 89.6 79.6 Bangkhee statue th th f o so t e n Du Cana l usew whic s sewagda no s hi e drainagr efo domestic and industrial discharges the quality of canal water is not suitable as water supply for drinking purposes, recreation or even for fishing. Swamp cabbage is found hi the canal all the year. It is the most dominant exposure pathway for radionuclides released into the canal.The calculated total committed effective dose from ingestion swamp cabbage to represent member fougroupf e so ag r s (ly,10y,15 adultd 0.2ye an , )0.2ar 2 d 0.2, 9 an 7 , respectively0.2Sv 5 totad an , l max. orga bone nth doseo t surfac2.3 s ev yeai n S 1d te rol o et child. The result of dose assessment revealed that the normal operation of the low level treatment plant at OAEP, insignificantly contributed to the annual dose limit for the public around its vicinity comparing to the dose limit recommended by ICRP (1990) as shown below. Impac criticao t l group Dose limit mSv/a mSv/a Total committed effectivl e dose 2.9x10"4 Total organ dose, bone surface 2.3xlO"3 life time organ dose = 3.5 Sv REFERENCES ] INTERNATINA[1 L COMMISSIO RADIATION NO N PROTECTION, Recommendations of the International Commission on Radiological Protection, ICRP Publicatio (1991)3 , Ann1- n60 .s ICR.No , P21 ] INTERNATIONA[2 L ATOMIC ENERGY AGENCY, Effec f Ionizino t g Radiation no Plants and Animals at Levels Implied by Current Radiation Protection Standards, Technical Reports Serie . 332sNo , IAEA, Vienna (1992). ] NATIONA[3 L RADIATION PROTECTION BOARD, Committed Equivalent Organ Dose Committed san d Effective Doses from Intake f Radionuclidesso , Chilton, NRPB-R245 (1991). 199 TREATMENT PROCESS AND FACILITIES FOR URBAN RADIOACTIVE WASTES Y. ZHANG, Z. CHEN, J. DAI Municipal Radioactive Waste Disposal Experimental Centre, Shanghai, China Abstract Luojin Site of Shanghai Municipal Radioactive Wastes Disposal Experimentation Center is involve e treatmenth n di f low-leveo t l radioactive wastes coming from Shanghai city. e treatmenTh t proces facilitied san describee ar s thin di s paper. Key words: City low-level radwaste, Compaction and sealed packing, Microwave drying, Ultrasonic decontamination, Incineration, Solidification. I . Introduction Since the 1950s, radioisotopes and nuclear techniques have been extensively used in hospitals, research laboratories, industrial and agricultural premises. The increasing and diversified use f radioactivso e materials resule productioth n i t f largno e quantities of radioactive wastes and their disposal requires careful and coordinated management. Luojin Site of the Shanghai Municipal Radioactive Wastes Disposal Experimentation Center has been set up ( by the Shanghai Municipal Bureau of Environmental Protectio o t provid n) a commoe n approac e centralizeth o t h d managemenf o t intermediate-level and low-level radioactive wastes in Shanghai area, and it is under the administrative control of national agency of environmental protection. This paper draws mainly upo facilitiee nth treatmend an s t process use Luojin di n Site. . Facilitie I Treatmend an s t Process The waste treatment site, facing Changjiang river and occupying about 10 acres, is situated at Luojin of Baoshan District, 45 km away from the center of the city. It has facilities such as solid waste sorting and compacting shop, animal carcass drying unit, ultrasonic decontamination unit, solidification shop, solid waste incineration shop, liquid waste treatment shop, wastes storage repository, etc. e volumTh f wasteo e s minimizei s ensuro t d e bestth e physical stabilite safth e d yan storaging conditions e schemTh . f treatmeneo t proces s representesi Fign di . 1 . 1. Packaging, collection transportatiod an , n Radwaste e collectear s d accordin e relateth o gt d national regulation e differenth d an s t categorie f wasteo s kepd containersn an i t s marked with special symbols e standarTh . d plastic bags lite0 6 , r fiberglass litedrums0 2 rd stainlesan , s steel drum e madar s n ei 201 Decontamination — Recovery, be discarded Noncombustable Nonmetal Compalion Weighingr Presorting I-- Noncombustable Animal corpse Microwave drying C organiH , c 3H liquid waste Spent resin, sludge Monitoring, Drain off water Liquid waste treament Liquid waste treatment Pumping house Fig. 1 The treatment process of Luojin Site the Site. The plastic bags are categorized into four uses, i. e. for combustible materials, noncombustible materials, metal, and liner of 60 liter drum. The containers, in which different kinds of wastes are stored separately and packed, are collecte y sitb de workers regularly .o specia Thertw e lar e e truckSitee th on n ,i s (loaded-weight 1. 2 ton) is equipped with lead shield, and the other (loaded-wight 1 ton), with a liquid wastes SS tank. 2- Sorting and Compacting Shop A YB90-100F compactor, i.e .a horizontal hydraulic triple action presss i , recommende r compactindfo g solid wastes e wholTh . e system consist f compactino s g box, compactor, rolling sealer, program-controlled electric control, compacted products rolls, etc. The solid wastes such as contaminated rnetals less than 3 mm in thickness and nonmetals, such as cotton, paper, wood, rubber, plastics, etc. can be easily compacted into a 40 liter iron-made drum and then ejected into a 60 liter fiberglass drum before storage. 3- Incineration Shop Incineratio r reducinfo e y bese th nvolumwa th gt e seemd b weigh an eo t f s o t combustible wastes. After incineration, the reduction is by 98. 9% for the volume and by 94. 5% for the weight. The incinerator consists of combustion, purifying, control, and ventilation systems. (1) Combustion system This system consists of feeding, burning chamber, and ash removal. There are two combustion chambers in the incinerator to ensure the complete combustion of the wastes e combustioTh . e firssecond th an r t nd 100fo dtemperaturean C " 0C " 0 85 e ar s chamber respectively during operation e capacite incineratoTh th . f o y s i abour t 50 kg/h. The ashes are usually solidified in the second stage prior to disposal. (2) Purifying system This system includes rotary dust trap, electrostatic precipitator, heat exchanger, electric heater, high-efficiency filtration, etc. The total efficiency is 93. 9%. (3) Control system The main items in the control system are electric lighter, static high-voltage adjuster, blower, feeding time, burner's temperature, negative pressur chambere th n ei , etc. (4) Ventilation system n ordeI o ensurt r e safetth e f workero y d avoi e an cross-contaminatios th d e th f o n systems ther e threear e set f ventilatioo s n systeme centra th e sho r th pfo n l i s control room, operatio removalh n as hall d an , . 203 4- Radioactive liquid waste treatment shop This shop consist medium-levea f so l liquid waste treatment system low-levea , l liquid waste treatment system and a central control room. The schematic diagram of treatment proces representes i s Fign i d- 2 . Medium— level Low— level liquid waste liquid waste Tank Residue - * e tob Tubular mixer coagulants solidified I Sloping tubular Sludge b o t e précipitr ato solidified Monitoring Wood flour Spent wood flour adsorption column to be incinerated Monitoring ,, Back blush water Filter to collection tank Monitoring _ High—level to medium- Electro osmosis level liquid waste tank Monitoring Spent resio t n Ion exchanger be solidified Discard tank Discard tank Fig 2 Proces. s schem f mediumo e d low—an — level liquid waste treatment e facilitieTh f liquio s d waste treatmen e unde ar te contro th r f systemo l s equippea n di general control room e item Th f contro. o s l includ e leve th eliqui n i l d tank, operation and discharge, sampling, pressure, flow, acidity, voltage, electricity, etc. e tesTh t show e radionulcideth s n liquii s de effectivelb wast n ca e y removed after passing throug whole hth e totae systeth ld decontaminatioman n efficienc s 10i y 2~104. The purifying curve of simulation water for I4?Pm is shown in Fig. 3. 204 concentration 800 700 600 500 400 300 200 100 0 origin slopping wood flour filtration n electrio o water tubular adsorption osmosis exchange precipitation Fig. 3 Purifying curve of simulation water for 147Pm 5- Synthesis treatment shop This sho composes pi f plastido c solidification, microwave dryin f animago l carcassd an , ultrasonic decontamination. ) Plasti(1 c solidification unit Plastic solidificatio applicabls ni asheso et , sludge, spent resin, concentrates, etce .Th treatment proces f plastio s c solidificatio s showi n Fign ni 4 . (2) Ultrasonic decontamination unit Ultrasonic decontamination ultrasonatorW involveK 5 a s , acidic trough, alkaline trough, and washing trough. The contaminated wastes are to be decontaminated effectively as a result of the use of immersion and ultrasonator. e wasteTh s which have been decontaminate ultrasoniy b d c procese reuseb n r do ca s discarde ordinarn a s da y wastlevee th f contaminatio f o eli identifies ni e 1/5b 0o dt less than the limit specified by the national regulations. ) Microwav(3 e dryin f radioactivgo e animal carcass unit The main equipment of the unit is a 2450 MHz microwave heater. 205 ash conveyer tank electro-magnetic plastic trough electro-magnetic jshake feeder electron-magnetic feeder storage drum Fig. 4 The treatment process of plastic solidification scheme The microwave heater is designed to achieve: Power : 1-10 KW adjustable Frequency: z MH 0 2455 ± 0 Chamber size: $ 600X800 mm Dewater rate: t loweno r thaKg/K8 . n0 h W • Energy leakage: 6- Radioactive wastes storage This building wit n effectivha e storage capacity 117s specialli 3 4m y designee th r dfo storag f useeo d l sourcekindal f solid o s an sd wastes which have been treatede Th . storage buildin constructes gi reinforcemeny db t concrete e disposaTh . l pits consistf o crowds of frame and trough underground, the drums shall be stored separately according to their surface dose rates. The storage building has electricity room, dose room, exhaust filter room, and the entrance for shoveler. The 3 ton bridge crane, 2 ton shoveler, cable TV monitor system, and dose alarm system are also available in the building. Twelve drume filleb o dt s with different waste e assemblear s a fram d loaden di an e d into a pit for disposal. 206 A FILTER STUDY FOR RADIACITVE LIQUID WASTE TREATMENT YE YUCAI TIANBAOU N W , XI O GU , Institute of Nuclear Energy Technology, Tsin Universitya gHu , Beijing, China Abstract The paper introduces the bench-scale research on the feasibility of a Chinese-made pleated filter cartridge to be used for radioactive liquid waste filtering in a nuclear power station (NFS) A standard loess dust made in China was used as simulative suspensions. The gravimetric method using 5 m nuclear - track porous membrane as a filtering medium was used to determine the filtering efficiency of the filter cartridge in removing suspensions wit hparticla e -size more tham n5 The filtering tests were recycla earne n i t edou loop resulte .Th s showed that PP-cartridges which were developed dan made by Hangzhou Water Treatment Development Center, SO A, could meet the needs of the radioactive liquid waste filtering sinteree th , d micro porou tubeE sP s coul usee db makdo t reusable eth e filte treao rt t radioactive liquid wastes with high-conten suspensionf to filtee th rd woulsan d hav elona g service life whil residue eth e strippind gan regeneration process were used . INTRODUCTIOI N The reactor cooling-water must be filtrated to keep the Pressurized Water Reactor (PWR) m normal operation. There are other water streams in PWR power station, for instance, mid - level radioactive liquid wastes, the ground washing waters and the laundry drains They contain suspensions that adsorb and carry lots of radioactive substances and also need filtrateration to reduce their specific activity prior to further treatments such as evaporation, ion exchange etc S oseriea filterf so radioactive th s r musfo equippee S tb eNP liquia n di d waste filtering. The liqui waste- d NPSa dra sm , which filtrated neee dividee b b o dt y dma , into three kinds accordine th go t concentration suspendef so d particle sourcee th d wastesf so an : (1) Primary loop cooling water in cooling circuits of NPS with suspension concentration 1 mg/L, (2) Mid - level radioactive liquid waste and ground washing water with suspension concentration range of 100- 150mg/L; (3) Laundry drains with suspension concentration 150 mg/L. A compacresule sa th designf S filterto e usuallth NP te l sar , al y installed togethe individuaa n ri l roo ordemn i o rt facilitat change filtee eth th f re o cores onlfilted e an y,equippes on i r speciaa r dfo l waste (sometimes wit sparha e one) discardee Th . d filter core normalle sar y directly solidifie cementatioy db n abroad whicn i , filtee hon rt corpu s ei int onormativa e drum capacit L wit0 thehd 20 y nan temporar e kepth n ti permanenr yo t radioactive waste repository, so the size of the filter core must be less than 2.LABORATORY FELTERATION STUDY Based on the design requirements for the radioactive liquid waste filters of NPS, the objectives of this study are as follows' efficience (1)Th removinf yo g suspended particles with filtelargee th highes i ry rb tham r ntha5 % n98 (2) The cartridge size is less than O500X 500(mm). (3) The flow rate is 20 - 27m /h for the liquid waste with 1 mg/L suspension, 10m /h for waste with 100 - 150 mg/L suspension and 5m /h for the laundry drain with 150 mg/L suspensions. filtee Th r) cartridg(4 e mus ionizine tb g radiation-resistan filtee th rd characteristictan s mus changt no t e after absorbing Y - dose upto 1043y and every filter element must work for as long as more than half a year 207 3.SELECTIO SIMULATIVE TH F NO E SUSPENSION As limited by conditions, it was impossible to utilize a real radioactive liquid waste in the test, the simulative suspensio uses substituta ns dwa reaa e th lr one efo reportes i t .I d tha compositioe tth f liquino d radioactive waste form NFS is complicated. The shape and size of particles as well as the physical and chemical characteristics of the suspensions are variegated. In China, diatomite dust is usually used as a simulative suspension in the research for normal water filtration. Diatomite with medium diameter of &Mn was selected as simulative suspension at the beginnin thif foungs o s wa study dt i that bu ,t diatomit gooo to dd filtratioeha n characteristi experimentae th d can l results coul shot dactuano e wth l filtration behaviou liquif ro d wastes from NFS investigationsf o . Aftet lo a r e ,w realized tha tsuspension e mosth f o t liquie th n sdi waste produced werfroS falloue meNF th t dust excepr s ai fro e mtth from some corrosion products and a few crystalline salts. So according to the situation in China, the standard loess dusselectes simulative wa t th s da e suspension compositios It . particle th d enan size distributio similae A nar US o rt standar dusf dAc t (See Tablaccordancn i , e1) liquiA e dwitUS e radwasthth e filter test standard. Tabl . eCompariso1 n Betwee Standar e dusf nth Ac t A d. S Losse . U e s th Dus d an t Particle Density Bulk Median Accumulative fraction under tbe Distribution Category Color density diameter limited particle Size Dx.(%) Shape (g/cai:) (g/cms) Cum) 0 5 10 20 40 >40 iossesdust Brown Irregulä . 6~22 r .8 0.62 3—1 1. 53—00 4 2 .33± 3 49—2 75±3 0 9110 — 3 Ac0 f 10 i 8< Gre 9 6 y Irregula1 5 6 . S~2.7 3 2 3 r 7.2—9.S . 0 —059 S 1 . 7 0 5 . DETERMINATIO4 FILTEE TH F RNO EFFICIENCY filtee Th r efficienc defines yi removine th s da g suspende e ratith f oo d particle certaif so nsimulative sizth n ei e waste, particl e ratii.e.e th th ,f o o e concentration before filtratio particle th o nt e concentration after filtration analytie Th . c method determinatioe th r sfo n of suspension concentration involv choicee turbidimetrye th th f so particle th , e size distrimetr gravimetrie th d yan c method etcturbidimetre .Th onls yi y suite limitea r dfo d rang suspensiof eo n concentratio generalls i d nan use t yno d for accurate determination of the suspension concentration less than Img/L due to its low sensitivity. So it was given particle up.Th e size distrimetry, theoretically, coul usee db determindo t particle eth e numbe variouf ro s e sizeth n si liquid, but finally we adopted the gravimatric method owing to the limitation of conditions. In order to determine the removing ratio of suspended particles larger the 5«m in diameter, the specimens taken from the inle outled filtean te tth r were filtrated separately wit hSua mporoun i s diameter standard nuclear track porous membrane and the membranes were dried and weighed, and the filtering efficiency was calculated from the weight of the residue. Despite tha methoe tth tims dewa consuming sampline ,th analysid gan s procedures coul made db sitn ei e and the analytical errors were reduced 5. THE RESULTS OF THE FILTERING TESTS pleateresulte 5.e 1th Th r sdfo PP-cartridge By screenin variouf go s filter elements mad China n efouni s wa t di , tha PP-cartridge tth e whic developes hwa d dan mad Hangzhoy eb u Water Treatment Development Center AO , S coul, d mee requiremente tth filtratioe th e r th s fo f no radioactive liquid waste. The elements were consisted of polypropylene superfine fiber with resistance to the radiation largtemperature d th an e Xid 0 flo pressur8 , an w o wt lo p rated u ean drop were achieve foldiny db g structuro et enhanc filtratinreusins - eit n selectes no ge gwa cartridgesth t areai s da o S . . Tablell show result e testse sth th f ,s o becaus elemente eth s wer producet eno sam e th en di batc pore th he sizs ewa not uniform tese th t conditiond an , samee s wer th result e t ,th e no s showed some differences. Durin tese gth ta suspension with a special concentration was pumped into the filter from time to time, and the changes of the pressure drop with time were recorded, and the maximum pressure drop of the filter was controlled at 0.25 MPa according to the engineering requirements typicae .Th l experimental dat pressurshowe e aar th Tabld n i an e I edroI ptim- e curves are shown in Figure 1. 208 0 2 4 6 8 10 12 14 16 r:g. l pressure drop - time curves of oie PP - 25 cartridges in test for the laundry drainage In tests No. 2 , 3 , the cartridge were irradiated by Co rf-ray for absorbing dose 1 10 Gy before test, to simulate the irradiation effect in the primary loop of the reactor. According to experience, the pore size of the elements made of high polymer fiber would enhance a little by ionizing radiation, but in fact the irradiation process didn't effect obviousl performance th n yo f filteeo r element theid san r filter efficienc thi n engineerinyi e sth test d ,an g requirement could be satisfied. Tablell shows that PP-5 cartridg trean eca t liquid waste more than 200 0witm h suspension concentratio mg/n1 L or treat about 100 m3; waste with 100 mg/L under 0.25 MPa pressure drop under the test conditions (the simulative matter was added at times to keep the required concentration). It is noteworthy that such amounts of waste correspond to half a year's waste discharged from a small to medium NFS, The life of the elements would be even longer if the concentration suspensioe th f so lowee nar r than that mentioned above. 5.2 Study on the reusable filter cartridges reusable Th e filte thawhoss e i r ton e filtrating effect coul recoveree d b lif e eth timd dan e coul prolongee db y db residue dischargin regenerationd gan abroat A . generalls i t di y mad sinterinf eo g metal, ceramics ,wir r fabrio eg cba precoates i d netan d with filter-ai improvo dt e filter precisio facilitato t d nan e discharging residueprecoatine th t Bu . g a. :=. 3. pressure Th e2 F. droig p tim- porou- E eP curvese tubth f seo suspensioe inth tesr fo t n witmg/0 h10 L 209 Table2. Test Results for the PP—5 Cartridges Filter Suspension Content Row Maximal pressure Continuous operating No. efficiency Note (mg/L) Crn'/h) drop(MPa) time (U) (JO 1 1 26 0.75 79.5 99.5 absorbin -dosgy e 2 1 2S 0. 04 31 99. I y G ' 10 o t p u absorbing v — dcse 3 1 26 0.03 96.5 99.1 up to 1 0' Gy 4 1 2S 0.25 58 99.3 5 1 25 0.25 32.0 99. 5 6 100 10 0.07 12.0 7 100 10 0.25 11 93.4 i 8 1ÜÜ 10 0.25 12 9 150 1C 0.25 3.5 10 ISOfLaundry drain) 5 0.25 14 98. cartridg5 32 — PP e e us U toOCLaundrv drain) o 0.25 17 98.2 —cartridg5 P 2 P e us e 12 150 (Laundry drain) 5 . 0.24 M 98.3 use PP —cartridg5 2 e 1^ tu 150 (Laundry Jrain) 3 0.25 15 use PP— 25 cartridge would bring about disadvantages suc comples ha x operation, high cosproductiod tan radioactivf no e solid wasteso S . the sintering PE-porous filtrating tube without precoating was selected to make a backwashing-regenerating test. The tubjointls ewa y develope madd d an Shangha y eb i Medicine Industry Institut Dond eWatean u gO r Treatment Equipment Factory, located in the city Wenzhou of Zhejiang Province, and the test results are satisfactory. The characteristics of the filter cartridge are as following: (1) Sintered micro-porous PE - tubes are made of extra - high polyethylene resistant to strong acid and strong alkali and can be used with pH ranging from 1 to 14. The polymer is odorless, tasteless and nothing will be dissolved and stripped from it. In order to raise the operation temperature certain materials are added into it and the operation temperature will be up to 110— 120 "C • tube cellulas Th e i ) (2 structurtube n rfluii e th eth n dmovei d ean 3-dimensionan si same th ler flowthicknesfo o S . s filtee ofth r cak resistance eth e increas tub- E less ei P sn e i tha nporoue thath n ti s membrane wit similae hth r retention restraineefficience size th f th e o d yan d particle smaller fe s si r tha actuae nth l por etube sizth ef ebecauso e residue ofth ebridg- e hig e effectth ho filterinS . g accuracy wil establishee lb d onc residue eth e layer e formth n so porou- surfac E P f seo tube . (3) The filter cake of PE-porous tube is easy to discharge from the smooth tube surface. As the elasticity of the long and thin tube, the backwashing and flowing of compressed air is adopted as the regeneration means industrially to prolon tube gth e life. Under certain condition residu producey e b dr s n eca d witporou- E h P volum e s th tub d f eo an waste is greatly reduced. (4) PE-porous tube can resist ionizing radiation and the physical-chemical characteristics won't change obviously wit habsorbea d dosage upt. Öol *Gy (5) PE-porous tube can be manufactured in various forms and sizes according to the requirements of the user. Owing to the low cost and the simple production process. (6) Being high strength, PE - porous tube can sustain 0.3 Mpa pressure and even O.oMPa with internal line. Its crash - esistant characteristics is far higher than sintered ceramic or metal tube because of it's good flexibility. PE-porous tube was tested and the results are shown in table III. The flow rate data in Table 3 are in engineering scale, which are converted from the results of the bench test. The engineering scale poroufilte7 7 mads f i ro sp tubeeu s with size 500(mm)th031X 5 X2 e total filtering are abous ai t 3m». PE-porous tube isn't suitable laundr e eitheth r liquie rfo y th drainagr d liabls blockei e fo wast b t r i o o e t s , a ed up with low suspension contents as the pressure drop of the filter will rapidly rise because the suspended particles fail to build trappe e bridgsurfac e ar tube t th th en f bu de o eo int tube oth ewaste poresth r e witFo . h suspension concentratio mg/0 hig10 e Lf th n o h filtering precision wil providee lb filtee th rs d a cak formes ei d rapidlye th s A . simulative suspensio standars nwa d losses dust with mid-diameter 8-10Aresistance th filted e man th rf e o cak s ewa high, the pressure drop rises rapidly and the continuous working time is very short. But in the filtrating process, 210 TableS. Test ResultPP—e th r 5sfo Cartridges rater Maximal pressure Continuous operating Suspension Content Flow efficiency Remark N °" (ig/L) (mVb)drop(MPa) ) CS ttortO y G ' absorbin10 X 9 f go 54 21 1 26 0.50 Y 'dose 22 1 26 0.50 53 22 1 26 0.25 30.5 24 100 10 0.25 99. 3 21. 5 paus timee4 operation sI n 2S 100 10 0.25 4 . 99 27.5 pause S tunes in operation 26 100 10 0.25 23 paus antese2 operation i n 27 400 10 0.25 15. 5 pause 1 tunes in operation 28 ISO to 0.258 1 5 . 99 pause 5 cimes in operation when the pumping sloped it would lead to backward flow and pressure in the bube in negative pressure so the filter cake would fall off easily and even more effective falling of the cake could be got if backfwashing by compressed air radioactive adopteds i th n i t Bu . e sites other method residue th r sfo e discharging adoptehave difficulb s i o et t i s do a t t collec cleansd tan fros of e backwashine fega th m th g process. Therefo tese discharginf rth to g residu stoppagy eb f eo pumping was conducted. The results are shown in Table 3 in which the stoppage of pumping shown in " Remark" is for a study on the situation of residue stripping through change of the operational pattern. In the test No. inlee th ts valva 2 turne, 7s ewa d off befor stoppine e th filtration e o risinpressurth t e f d gth o f le g o t i , e e droth n pi following filtrating operation. This means that the residue on the tube has not fallen off. Fro teste mth s described above sinteree ,th d PE-porous tube foune onle sar b yo dt suitabl mediue th r efo m leve- l radioactive liquid without colloidal matter in NPS and the sintered PE - porous tube filter could be adopted as the reusable filter for the waste with suspension 100-150mg/L, for the residue could be stripped by the negative pressure equipmene th filtee inlet d a th f ran tcollectino r tfo g residu installes ei prevendo t watee tth r flow from disturbine gth discharging of residue so as to make the filter work for a long time. When the residues stripped and captured on the tube are cumulated to a certain volume and the pressure drop rises to a limited value, the feeding is stopped and the filte openeds ri filtee ,th r cartridge residud san ecollectin- g barretreatmentr liftefo e t ar l dou . RESULT. 6 DISCUSSIOD SAN N determino T ) (1 filtee eth r performanc radioactivr efo e liquid waster treatment selectioe th , standare th f no d loess dust mad Chinsimulativn e i th s aa e suspension, whic foun s representative b hwa o dt corresponded ean e th do t U.S. A.standard specimen is reasonable. (2) The PP-cartridge developed and made by Hangzhou Water Treatment Development Center is able to meet the need radioactivf so e liquid waste filtering proces NPSn si characteristics It . hige s ar resistanc- hflo- w wlo ratd eean and the efficiency could be up to 98% for the suspended particles with larger than 5*Mn in diameter. Besides, the ground washing waste wate laundrd ran y drainag alsn filtratee ca o b d wit filtere hth . sinteree Th ) d(3 PE-porou usee sb tub manufacturdo n t eca e reusable filte radioactivr rfo e liquid wastesuitabls i d ,an e for the waste with high suspension content, and its service life can be further extended by the practice of residue-. stripping and regeneration process under negative pressure. REFERENCES 1.LIU Aifang,e Stude Th , y al tResult N90f so 7 Road Dus BF-d tan 2 Simulate Dustr dAi , Labour Protection Science Technology, NO.3, 1989. 2.A.H.Kibbey,H.W.Godbee Filtratiof o e Us Treae o nt ,Th t Radioactive Liquid Light-Water-Coolen si d Nuclear Reactor Power Plants, NAREG/CR -41, ORNL/NAREG-41,1970. 211 VOLUME REDUCTION OF SYNTHETIC RADIOACTIVE WASTE BY THE THERMOPRESS HEYDENR DE N . DEBIEVP ,VA . P E Belgonucleaire S.A., Brussels, Belgium Abstract The managemen f radioactivo t e majoth rf o econcern e l wasteon al s f i o s organizations involved in the nuclear field and the volume reduction of such waste thus i smajof so r interest, plane mainlth to yt operators . The Thermopress is a compactor designed especially to treat the synthetic wastes. In order to avoid the disadvantageous springback effect of synthetic wastes after compression e Thermopresth , s combines compactiod an n slight external fusioe synthetith f no c pellets hydraulis It . c power system ensure A regulate a scompactio . kg o 160t 0d p electricau n l heating system allow o controt s e thicknesth l e f externafusioo sth f o ne l th par f o t synthetic pellets after compression automatin A . r extractioai c n system maintains the compaction chamber at a negative pressure and assures the coolinsynthetie th f o g c pellets afte heatine rth g cycle. This paper describes the historical development of the thermopress, from the non nuclear standard equipment designed by the Dutch Company ICORDE, to the actual equipment improved by the Belgian Engineering Company BELGONUCLEAIRE for nuclear application subsidiars it d ,an y TECNUBEL. The main technical characteristics and advantages of the thermopress will be reviewed in this paper as well as the economical aspects in favour of this type of equipment. I. HISTORICAL DEVELOPMEN THERMOPRESE TH F TO S The Company ICORDE has developed a new press to reduce the volume of synthetic wast e Dutc- Winneth e f ho r environmental Award 198e th 9 - thermopress combines compactio d slighnan t external meltincompactee th f o g d wastes (pellets). The pellets are cooled before releasing the compaction, , keeso p d theian r shape, avoidin e disadvantageouth g s springback effect well-known with other type pressf o s . e importanth o t e t Du amoun f synthetio t c e nucleawasteth n i sr field, BELGONUCLEAIRE has approached ICORDE to develop together a thermopress for nuclear applications. The "nuclear" thermopress development has been achieved in two steps. desige e prototypFirston th f y no ,b e (rectangula n nucleano r r shaprfo s a e aplications) in order to : *• adapt the equipment for nuclear applications by : maintainin• depressurga e insid compactioe th e n chamber durin entire gth e treatment process; • improvemen confinemene th f o t d increas ventilatioe an t th f o e n capacity r coolinfo ; 0 m0 -»3/h25 (4 g•) • replacement of the mechanical press by a hydraulic press, with the hydraulic grou contaminatee pth locatef o t ou d d compaction chamber; 213 • improvement of the process control command, and the general assurance quality level; • provid separatea e d control comman thermoprese th df o pane t ou ls itself; *• improve the performance of the equipment and the quality of the pellets issued from the process by : • increase the compaction force and the cooling capacity; • uniformisation of the heating by electrical tracing and better control by several PT 100; • improvemen controe th f o tl comman d possibilitan d chango t y e easile th y parameter f proceso s n ordei s o fine t rbes th d t parameter o increast s e the volume reduction and the quality of the pellets, taking into account the composition of treated wastes. Many tests have also been performed with this prototype in order to check the compaction performance with different types of synthetic wastes, and mixin synthetif go c wastes generally treatenucleae th n i dr field. e prototypTh s alswa eo a demonstratiouse s a d n unir manfo ty potential customers suc s nucleaa h r power plant d wastan s e treatment facilities, nuclear waste management agencies,.. II. TEST CAMPAIGNS WIT PROTOTYPE HTH QUALIFICATIOD EAN PROCESE TH F NO S Many tests have been performed with the prototype (see Photo below) in order to : *• check the influence of the different process parameters; *• measure the volume reduction of typical specific wastes, or wastes mixing; *• chec e possibilitth k f incineratioo y e pelletth f o ns issued froe th m thermopress (qualification of the process). Prototype 214 II.I. Influenc procese th f eo s parameters Tests campaig s beenha n achieve n ordei d o chec t re influenc th k procesf o e s parameters on the performance of the thermopress : heating temperature, heating time, force of compaction, duration of compaction cycle, and cooling temperature. teste Alth l s have been performed witf same o hth . e 1 typ0 wastef 15 o e ± : s polyethylene sheets (thickneshavin) ,u 0 totaga 10 : sl weigh abou. f to kg 5 t e resultTh s have been compared fropoine th m: vie f to f wo »• pellets quality ; + pellets density (volume reduction); »> duratio e completth f no e cycle (treatment capacity). Main conclusion followine th e sar : g P- The heating temperature influences the quality of the pellet (optimum temperatur r polyethylenfo e 0 °C)17 .s i e Tests with other materials have showed that the heating temperature is to be adjusted according to the treated material. *• The extension of the heating time increases the duration of the complete cycle, without significant improvemen e pelleth f to quality. f The compaction force influences directly the volume reduction, without any influenc duratioe th complete n th eo f no e cycle r thiFo s. reason wile ,w l use for all the tests the maximum compaction force. *• It appears that the cooling temperature is the most important parameter for the pellets density (see Graph below). THERMOPRESS N P l LIST OF SYNTHETIC WASTES ALREADY TREATED • Polyethylene or similar • PVC • Overshoes • Mixed Wastes • Cellulose • Rubber -Latex - Polyethylene -PVC Influence of cooling temperature 90 Temperature C 215 As "low" is the cooling temperature, as "high" is the density (volume reduction) of the pellet. As a matter of fact, a low cooling temperature increases the hardness of the external melted part of the pellets, avoiding so the springback effect. On the other hand, a low cooling temperature extends the duration of the complete cycle. An optimum must be found between density of the pellets and the capacity of treatment (number of pellets/hour). II.2. Volume reduction of typical specific waste Tests campaigns have been performed to check the possibility of typical synthetic waste compaction (or mixing) similar to wastes coming from nuclear power plants as : polyethylene• , polypropylene, polystyrene,...; « overshoes; rubbe• r (mask, gloves,; .) . • paper; • clothes; • heat insulation... THERMOCOMPA.CTION TESTS RESULTS PRODUCTS WEIGHT FINAL VOLUME R V FACTOR Ï 0 10 T PE 5,6 20 1 7.5 Overshoes 100 X 5,4 kg 18 1 8.3 Plastic laboratory 2 kg 11 1 13.6 bottles (fine) 100 X X 0 5 T PE 3 kg 34 1 4.4 Rubber (Mask) 50 X PET 33 X 6 kg 27 1 5.5 Rubber 33 X CellulosX 3 3 e X 0 4 T PE 6 kg 29 1 5,2 RubbeX 5 2 r Cellulose 25 X Heat insulatioX 0 1 n PET 25 X 6 kg 46 1 3.2 RubbeX 0 2 r Cellulose 20 X Heat insulation 10 X PapeX 5 2 r . t 5 5 T PE 6 kg 31 5 1 4 8 CellulosZ 6 e Heat insulatioX 6 n Paper 33 X r 2 3 T PE 6 kg 38 1 3 9 Cellulose 20 Z PapeX 0 2 r Heat insulation 8 X PET 50 X 6 kg 23 1 6,4 Rubber masks 25 X Rubber gloveX 5 s2 Remark Initia• s l volum1 e 0 frobage 15 nth s • Tests vere made with a double PET bag • P F. T is polyethylene or similar \ factoR • Reductio- r n Volume factor - Initial Volume/Final Volume product. ' weighn i • e tar s 216 The table gives the volume reduction obtained with various mixed synthetic wastes. The main conclusions to obtain a good quality of pellets and a volume reduction of about 5 are : > wastes must be closed in a simple (or preferably double) polyethylene bag; *• percen soff to t plastic or/and soft rubbe % (weight) 0 r5 mus> e ;tb *• percen clothef to s and/or pape% (weight 0 r4 mus< )e t b (se e Photo below); »• percent of heat insulation must be < 10 % (weight). II.3. Qualification tests In order to qualify in Belgium the process of the thermopress, a campaign of incineratio pelletf no s prepared witprototype hth s beeeha n organized with the assistance of the Belgian Agency for wastes management NIRAS/ONDRAF. 700 kg of synthetic wastes, similar to those issued from the Belgian nuclear power plants, have been compacted by the thermopress in polyethylene bags of compositioe Th . 1 0 thosf no 15 weightn i e % waste( ): s swa % pape• 0 2 : r • clothes : 20 % • PE, PP, PS : 50 % • Latex, rubbe% 5 : r % 5 : C PV • 217 120 pellets have been realized and burnt with success in the CILVA incinerator (capacit kg/h0 10 f )yo locate BELGOPROCESSt a d . III. TECHNICAL SPECIFICATIONS OF THE NUCLEAR THERMOPRESS After many tests performed with the prototype, BELGONUCLEAIRE has decided to construct a standard equipment for the nuclear market taking into account the experience of those tests. Nuclear application and cost reduction for the construction of a standard thermopress for the nuclear market were the main objectives for the design e commercializeoth f d product. For these reasons, following modifications have been decided : *• Circular therraopress instea f rectangulao d n ordei r o stort r e pelletth e s n drumi s; *• Compaction chamber in stainless steel instead of carbon steel; *• possibility of remotely waste loading and pellet unloading; *• possibilit t thipu so t ycompactio n chambe n glovi rx (fobo e r alpha applications); > simplification of the control command taking into account the main parameters havin n influenca g performancee th n o e . e maiTh n technical specification e nucleath f o sr thermopres e describear s d hereafter : This standard equipment coul e reviseb d d accordin o specifit g c requirements of the customer for particular applications : > Wastes compaction chamber : • standard shap cylindrica: e r rectangulao ) lmm (diameten 0 (i 58 r ± : r option) • nominal capacity : ± 200 1. *• Compaction force : up to 1600 kg, assumed by an hydraulic group (maximum pressur bar0 8 : e) > Electrical heating : 16 Kw. The heating temperature is adjustable between . °C 0 19 d an C ° 0 13 »• Depressure insid e compactioth e n chamber assumen exhausa y n b dfa t connected to the site ventilation via a prefilter and a HEPA absolute filter. *• This exhaust fan is also equipped with high speed (250 m-^/h) in order to ensure the air cooling of the pellets after the compaction and thermal cycle. e coolinTh *• g temperatur adjustabls i e. °C 0 4 ed betweean C ° 5 n8 >• The full cycle time (compaction - thermal - cooling cycle) is about 15 minutes. > Material : • internal chambe stainles: r s steel • heating part (removable) covered by special coating • other part : carbon steel painted with decontaminable coating • wastes inlet and outlet designed for remotely loading and unloading (as an option) • quick connectio a separat o t n e control panel equipped with safety interlock alarmsd san . e procesth l s automatedi sAl A .selectio8 differen o t p u nt process programmes coul suppliee b d n optiona s a d . • internal chamber equipped with a smoke detector (as an option) 218 • overall dimensions : - length 700 mm m m -0 widt80 h - height 220m 0m • shipping weight : ± 200 kg. IV. ADVANTAGES OF THE THERMOPRESS The storage on the site of producers of nuclear synthetic wastes, as well as the transportation to a centralized treatment facility or to an interim/final disposal results in high costs. The thermopress reduces the volume of such wastes (average volume reduction pelletn i ) 6 betwee sd acceptablan n3 disposar fo e later o l r treatment. maie Th n advantag thermoprese th f eo s coul summarizee db d hereafte: r >• Volume reductio variouf no s synthetic waste n pelletsi s eas handleo yt ; *• Reduce the volume of interim storage ; > Increase transport efficiency; * Possibilit f incineratioyo synthetie th f no c pellet; > Possibilit o integratt y e compactioth e n chambe a smal n i rl confined cell; > Possibility to adapt easily the dimensions of the pellets according to the customers requirements; »• Small compact machine; >• Easy for operations and maintenance ; > Safety controlled and automatic process; > Low investment (less than 100,000 US dollars); * Safe environmen moneyd tan . 219 WASTE CONDITIONING RADIOACTIVE WASTE-MORTAR MIXTURE FORM CHARACTERIZATIO PHYSICO-CHEMICAS IT O T E NDU D LAN MECHANICAL PROPERTIES OBTAINE ACCELERATEN A DN I D NON-ACCELERATED AN D LEACHING PROCESS A. PERIC, I. PLECAS, R. PAVLOVIC, S. PAVLOVIC Radiation and Environmental Protection Department, Institute of Nuclear Sciences "Vinca", Belgrade, Yugoslavia Abstract Mortar as a matrix for the radioactive waste materials of the low and intermediate level f bondeo d activit s investigatei y e "Vincath n i d " Institut n accordanci e e with IAEA standardized procedure recommendationsd san propertiee th f o e s On .tha t could characterize mortar as an appropriate matrix for radioactive waste materials accept of the most commonly used radionuclide leach rate is its mechanical strength. In the performed experiments two group orthocylindee th f so r shaped mortar matrix samples doped with 137CsCl solution, after curing in the atmosphere of defined parameters for 28 days were treated on leaching in distilled water using accelerated and non-accelerated processes. The first group of samples was treated on leaching for period of 292 days in aim to obtain cumulative 137Cs leached fraction. 137Cs leach rat f suceo h treated samples basa s establishinn ei wa , g cycle r acceleratefo s d ageing processes, that understands samples immersion, heating, immersion and freezing in each cycle. The second group of samples was treated on leaching using cycles in accelerated ageing of mortar matrices. By introducing the mortar-radioactive waste mixture form into the environment with temperature extremes, we could obtained the radionuclide 137Cs leach-rate, tha adequats ti nearle th yea e r yeon fo r non-accelerated leaching condition almose th n si t eight cycle accelerate e stepth f so d ageing process. Mechanical strength of such treated samples was investigated using hydraulic press. Decreasing of the empirically obtained values for mechanical strength of samples treated by accelerate standardized dan d Hespe leaching method were noticed, being consequence th f eo corrosion effects of water on mortar and synergistic influence of mortar matrix exposure to the elevated temperatures before each immersion step. These experiments were performed in aim to predict mortar matrix behaviour due to leaching and mechanical characteristics in the prolonged period disposalf so , assumin mose gth t undesirable environmental conditione th n so disposal site. 1. INTRODUCTION e cementatioa resulth s f A o e t radioactivth f no e d wastan ew materiallo e th f o s intermediate level of bonded activity immobilization process, solidified radioactive waste- mortar mixture form is obtained. This solidified form is a part of the engineer trench system, consist of: matrix-radioactive waste mixture monolith, inside the concrete container, which itself is posed in the concrete made trench [1]. Mortar matrix, accept its role as primary barrie immobilizee th o rt d radionuclide migration fro radioactive mth e waste monolite th o ht environment satisfo t othes w yfe ha , r tasks that arise fro safete mth y handling, transpord an t radiation protection poin viewf to , suc goo: has d mechanical characteristics, stabilit matrif yo x structur prolongee th n ei d perio storinf do disposad gan l time, fire resistance, thermal stability, resistance on the influence of the corrosive environment, resistance on the irradiation, etc. In this paper, mechanical strengt radioactive-mortae th f ho r mixture forms measura s a , theif eo r 223 stability durin e manipulatiogth disposald nprolongean e th n i , d tim disposan o e l sites i , discussed. These investigation hav studiee e e effectcorrosio basea th di th f n n i eo o s n influence on the mortar matrix structure, caused by the water environment of the matrix. One grouorthocylindee th f po r shaped samples, wher sample eth e surface completely exposeo dt the environment, was treated on leaching in the distilled water, as a leachant [2]. The second group of the samples, prepared in the same batch as the first group of samples, was treated on leaching in the accelerated ageing process. Accelerated ageing of the sample understands repeating of the cycles, in which mortar-radioactive waste mixture sample is introduced to the step : immersionof s , heating, immersio freezingd nan eacn i , h cycle establishinn I . e gth numbe cyclese th f ro , tha appropriate tar purpose th o et thesf eo e investigations, leach-ratf eo the radionuclide 137Cs from the mortar-radioactive waste mixture form, in the one year leaching test experiment ordinare , th carrie n i t y dou manner takes basa wa ,s nea parameter [3]. Both groups of the two way treated samples were, after prescribed experimental period, investigated on the mechanical strength, in aim to compare two sample groups characteristics, due to a named property and in attempt to define possibility of the predicting the matrix material properties in the prolonged periods of time, spent in the undesirable environmental condition disposae th n so matrie l th sit o et x material structure. . EXPERIMENTA2 L Fopurpose rth describef eo d mortar-radioactive waste mixture form mechanical strength investigations, two groups of the otrhocylindrically shaped samples, H=D=4.5 cm, were prepared. Formulation of the mortar-radioactive waste mixture form is given in the Table I. Table I - Materials used in the radioactive waste-mortar mixture samples preparation. MATERIAL FORMULATION (g) CEMENT PC-45 (MPa) 1320 SAN (mm2 D0- ) 335 DISTILLED WATER 450 137CsCl solution, pH=1.2 25 MIXING ADDITIVES 10 Water-to-cement ratio, W/C, for the chosen formulation is 0.36. Prepared samples have average apparent density p =2.145 g/cm3 and a porosity s =0.22. Radioactive waste-mortar mixtur s prepareewa usiny db g planetary mixing device. Mixture material, preparee th n di batch botr fo , h group samplesf so poures ,wa d int plastie oth c mold hardeno st . Afte daye on r , samplee th s were take t fro molde nou m th cureatmospherd n a s an n di percen5 6 f eo t relative humidit temperaturd yan e T=20°C perioa days8 2 r f fo ,.do Hardened sample l haval s e approximately the same level of bond activity, A,, »10 kBq. After the curing time period, the first group of samples was placed inside clear plastic beakers, where the completely exposed surfac samplee th f eo s were brought into contact with leachant, distilled water. Experiments were carried out at a room temperature of T=20±l°C.The leachants were renewed periodically. Afte e preaccepteth r d immersion time, grou f threpo e samples were investigatee th n o d mechanical strength resistance postulatinn I . acceleratee gth d ageing process, base parameter was one-year leach-rate value for the 137Cs, measured in the non-accelerated process. The 224 same leve f activito l y adequat leach-rata o et e measure non-acceleraten di d processs wa , obtained hi an eight successive cycles of sample's: immersion, heating, immersion and freezing. Each immersion steperformes pwa distillen di d rooe wateth mn o r temperaturf eo Tj=20°C; heating was performed in the air of temperature Th=70°C; freezing steps were performetemperaturr ai e th n do Tf=-20°Cf eo . Afte acceleratee th r d agein numbee th n gi f ro cycles appropriat certaie th eo t n non-accelerated leaching experimental time equivalent, sample were investigated on their mechanical strength characteristics. 3. RESULTS AND DISCUSSION Resulte performeth f o s d experiment e showar s n numericall d e Tablan th I n I ei y graphicall Figure th n y. Obtainei e 1 d valuemechanicae th f so l strengt orthocylindee th r hfo r shaped samples, H=D=4.5 cm, have to be multiply by factor 8, when describing the properties of the standardized cube sample, length a=10 cm, that is in accordance to Yugoslav standards. According to a performed experiments and investigations, it could be deduced that there is a certain decreasing trend of the mechanical strength characteristics, when investigating the radioactive waste-mortar mixture samples, treated solely in the distilled water. The highest value for mechanical strength is obtained after t,=100 days, when hydration process of mortar is finished. Decreasing valuemeasuree th f so d characteristi dedicates ci corrosioe th o dt n effect of the water on the mortar composition, and it is in correspondence with Ca2+ depletion, caused by leaching from the matrix composition. Certain trend to the saturation values of the mechanical strength is noticed for the samples treated in distilled water for 365 days and longer. Obtained values for mechanical strength of an accelerated aged samples have shown that, when the mortar matrices are introduced to the influence of the temperature extremes and immersion, the end of the hydration process is postponed. TABLE II. MECHANICAL STRENGTH OF THE RADIOACTIVE WASTE-MORTAR MIXTURE SAMPLES TREATE NON-ACCELERATEE TH DN I ACCELERATED DAN D LEACHING PROCESSES. EQUIVALENT MECHANICAL STRENGTH RESISTANCE (MPa) LEACHING TIME (day) Ordinary aged Accelerate aged 28 4.90 4.90 45.5 5.85 5.10 100 8.84 6.50 292 8.56 7.65 365 8.35 7.56 547 8.10 6.85 730 8.05 6.32 Experimental conditions postulate n acceleratei d d ageing leaching process caused widening of the matrix interim pores, better contact of the leachant with matrix, faster leaching of the Ca2+ and consequently, destruction of the matrix structure. 225 9— ...... / """""••o--, ""••o— .... ""— — O-. — ..... _ 8 ——— o / x-*--^ 7 / /' *\ ' / / '"S-^ ë"co" / er"" ----.. 7 O- / - ^"^ 6 ^ / / • MechanicaB — Q-— l strengh accelerate th f o t e aged sampl 5 4 C 8 0 60 0 40 C 0 20 0 t (day) FIG. 1. Comparative curves of the mechanical strength resistance of the two-way treated sample groups. Successive treatment of matrix, in the conditions that are destructive its structure, is resulting in faster decreasing of the mechanical strength of the accelerate aged samples. Using linear fittinvaluese th f go , that represent mechanical strengt groupo tw r treatef hso fo d samplese th , lowest values of matrix resistance on the influence of external forces are obtained after ten years of continual described attacks for the accelerate aged samples and twenty years for ordinary treated samples, as it is shown at the Figure 2. — 1-Lineai—— V fitted cuv f ordinareo y aged samples o--——o Mechanical strengh ordinare th f o t y aged samples ———— n-Uneary fitted accelerate curvth f eo e ageds B——a Mechanical strenght of the accelerate aged sampl y = -000128X1 +891 var000722 max dev00937 -000251Xy= »82 0039r 6va ; 5m ix dev0219 1 2000 4000 6000 t (day) FIG. 2. Linearly fitted curves presenting mechanical strength of the ordinary and accelerate aged treated radioactive -waste-mortar mixture sample. 226 4. CONCLUSION Performed experiments were carried out in the frame of the quality testing experiments of the matrix materials for the radioactive waste materials of the low and intermediate level bonded activities thesn I . e tests triee w ,confir do t m performance mortaf so r matrix materials due their mechanical characteristics and radionuclides retention capabilities, even in the worst environmental conditions, that migh t site occuth e n planne o rfinae th r l ddisposafo f suco l h waste materials. Results obtained in the accelerated ageing processes of the mortar-radioactive waste form appliese serveth indicatio n f a o ds e s a mortar-matrith f no x formulation quality and possible mortar matrix behaviour in the extended disposal time. REFERENCES [1] PLECAS I.,PERIC A., "Quality testing methods used in radioactive waste solidification process in "Boris Kidric" Institute of Nuclear Sciences-"Vinca", Invited paper at the RCM "Use of Inorganic Sorbents for Radioactive Liquid Waste Treatmen Backfild an t Undergrounr fo l d Repositories", IAEA, Belgrade(1990). ] PERI[2 C A.D.,PLEC PAVLOVId an A. I , "EffectCR. f relativo s e surface ared aan leachant composition on the Cs leach-rate from cement waste forms", Scientific Basis for Nuclear Waste Managemen 137 t (Proc.Inter.Symp. Boston) Vol.296. (INTERANTE, PABALAN, Ed.)PublisherS ,MR s (1992) 241-246. ] HESP[3 E E.D., Atomic Energy Review, 9,1971, p.195. 227 USING BITUMEN SOLIDIFICATIO ILLR NWLLLFO & W ZHANG WEIZHENG, LI TINGJUN Beijing Institute of Nuclear Engineering, Beijing, China Abstract The paper describes the results of the research and development work done on the bituminizatio intermediatd an w lo f no e level radioactive wastes generated fro nucleae mth r fuel reprocessing plant. The purpose of the work was to select a suitable bitumen for various radioactive wastes determino t , e operation condition propertiee th d e th san f so bituminized waste forms. Wite constructioth h e developmenth d an n f atomio t c energy industr n Chinai y , large amoun f radioactivo t e waste gradually accumulates and awaits treatment, especially radioactive liquid waste produced from nuclear fuel reprocessing plant. In the early stage, according to its radioactive concentration and chemical properties this kind of liquid waste was stored separately in o severamort en te thale thousan on n d cubic meters underground tanks. In consideration of the safety of long-term storage and environmental pollution caused by damage of the tank, research personne d engineeran l n Chini s a havt researcse e n transforo h m the liquid waste into solid waste since the mid stage of the 60s. Bitumen solidification is one of the researches that can be used to transfor above-mentionee th m d ILL LLL& W W into solidification waste form. Bitumen solidificatio e advantageth w s rat f lo ha volumno e f o s e reduction, low leach rate. Much attention has been paid by many s beecountriesha nt i develope d an , d used an dn Chin I . a since 1969, basic researche s beeha e sn bitumeth madn o ne solidification and small-scale solidification unit test has been given. The purpose is to select suitable bitumen for various radioactive liquid waste, to determine operation conditions and the properties of bitumen solidification waste form. At the 229 beginnin f 70so g ,a serie f middle-scalo s e simulate tess wa t made. Based on performance experiment, scrape thin-film evaporator was optimized as main engineering device of bitumen solidification. Many researche testd san s have been completen o d thermal stability of bitumen solidification waste form. Soon afterwards industry-scale bitumen solidification workshos wa p designed and constructed which now is put into operation and has treated neathousane on r d cubic meters radioactive liquid waste. This paper chiefly provides information about the general engineering situation of this workshop and problems that should paie b d attentio bitumen i o nt n solidification. 1. Representative radioactive liquid waste for bitumen solidification. Chemical composition: total saltness 400g/l + NH 4 0.5g/l alkalinity(OH~) 1—1.5N Main nuclide: 137Cs 90Sr average radio activity 1.85*10 Bq/1 Max radio activity 3.7*l08Bq/l 2. Description of the process e e nexfigurth Se tn o epage . Bitume s filtratei n y filter(2b d ) and then is pumped into bitumen supply tank(4) by bitumen gear pump(3) , where additive is put in by fixed quantity and the bitumen wil e requiree preheateb th l o t dp u dtemperature . After being analyzed, radioactive liquid waste is transported into liquid waste tank(6), then pumped into regulation montejus tank(7) whicn ,i hadjuso o t 10—1t HNO n H i P t 3n ,t 2i pu wil e b l accordance wite chemicath h l compositio e liquith f o dn waste,and some e tankadditiveth .n i Thee alse liquit ar sth npu o d waste wil pumpee b l d into supply tank(10 user )fo . 230 condensote soltriificolton solididcoiton to reoosilofy temporary storage . 2_ 20 2* 23 iiii» bilumin «01*17 cod look supply tank prtheof«' jcropir fhifi-/ilm I/op lank monitonng boi 6 U 15 M_ 16 18 bilurmn gear Ihirmal (niulaticri liquid »osli lonk regulolion sert* m( It r mg mlel drum 601 tforag« bjnir If on j If r box cooler trap urdtnurloc* cenlrfu^ol r purng« p pump monlt|us lank pump pump Fig. When solidification system begin o operatet s , Thermal insulation gear pump(5), screw metering pump(ll) will star n orderi t , fixed quantity of hot bitumen and liquid waste preheated by preheater(12) will be delivered in proportion into scrape thin- film evaporator(13) whicn i , h bitume d liquian n d wast o througg e h e uppe th d lowean r r distribution plat succession i e d wilan ne b l e e insidth throwth f n e o eevaporatoo n th y wal f wa o l- by r centrifugal force produce distributioy b d n plate rotationo t e .Du the scraper's rotary stir, bitumen and liquid waste gradually mix in for f filmo m , flow down spirally wa wild e heatean e yb l th n o d by steam in the jacket of the evaporator. The water in the mixed material will evaporate by degrees, the salt and radionuclide in the liquid waste will be covered by bitumen until well- distributed liquid mixture of bitumen and salt can be obtained at the end of the evaporator. Solidification process is completed by four boxes and a drum transportation line. The whole process is controlled by computer. Batch loadin adopteds i g , each batc s fouha hr drum d eacan sh drum wil e fille b ltimes4 n e controi dTh . l ste: fouis pr empty drum of each batc o lineg h , pass through protection sealing dooo t r inlet drum box(15) ,o feet the do ng box(16) through sealing door, in which four drums make a round trip and will be filled at the fixed loading position with liquid bitumen and salt mixture flew from storage tank(14 ). After four cycles, full drum wile b l sealed at fixed position and goes to decontamination and monitor box(17) through sealing door. After surface decontaminatiod an n monitoring, the drum goes to transfer box(18). Finally drum filled with bitumen-salt mixture will be transported to the solidification temporary storage roo y digitab m l control crane and piled up in good order for cooling. After solidified 232 completely, the drum as the bitumen solidification product will be transported to the repository by special purpose truck. 3. Process parameter of scrape thin-film evaporator 1 Paramete3. equipmenf ro t heat conduct area 2.5m 2 diamete f hearo t conduct area 99—99.5KPa temperature control: waste liquid feed 90—95°C bitumen feed 130 *C steam <2.5MPa product outlet 165—170 *C Productio. 4 n capacit producd an y t target 4.1 Production capacity Abou drum0 3 t s (2001)of solidification waste forr day.pe m 4.2 Product target total saltness »40%(w) moisture content 233 soften point >65*C initial exothermic temperature -£240*0 spontaneous combustion temperature f solidificatioo n waste form -£30C 0° 5. Engineering problems should be paid attention to on bitumen solidification 5.1 The main equipment of this workshop is scrape thin-film evaporator technologicae .Th l difference between scrape thin-film evaporator and other evaporators is that liquid waste heated till boiling and separation of secondary steam almost happen at one heating areas in form of film, the film-like liquid mixture of waste and bitumen should not be overheated during the evaporation o avoit d explosive boiling y insiddr , e wall, cutoff flod an w V decomposition selectee .Th d control parameter flowrate ,th e enter the equipment and temperature should be suitable to obtaining qualified product. During test operation, dry inside wall once occurre d cause an de sal th dt scarred seriousl e evaporatth n o y e surface t alsI . o happened that high moisture conten n produci t t caused outlet pipe blocked up. 5.2 Special attention should be paid to fire prevention when using bitumen solidification process, because bitumen itself is inflammable, especially whee liquith n d waste contains oxidant such as Manganese ion, and salt such as nitrate which can produce oxygen when being heated. For example, it is found that when using solidification sample to test its thermal stability and if the solidification sample contains 40% NaNO3, the nitrate begins to melt when the temperature reaches 280 °C. Due to different specific gravity, the liquid bitumen and nitrate separate and the bitume ne e surfacexperimentath th stay n f o o se l instrument. Becausw healo t f o econductivit w healo t d convectioan y e th f o n bitumen t emi,no t hea n froe ca bitumet th m n surface. Whee th n 234 sample keeps on being heated, the nitrate will decompose and give out oxygen so that it cause the sample burn severely or explode when using airtight container. Therefore preventive measures that are taken during workshop constructio s followsa e nar : a. Overheating of bitumen and solidification waste form shoul e strictlb d y prohibited during transportation, production and storage. Automatically temperature controlled system is installed. b. Before solidification each batch of radioactive liguid waste shoul e analysedb d . Solidification samples shoul e madb d e o somd in o eordet representativ d o analyst ran A DT e e experiment under constant temperature resulte usee .Th b adjuso t dn e sca tth process control condition of the workshop. c. The drum can not be transported from workshop to repository until the waste solidified completely and it's temperature drop normao st l atmospheric one. d. Warning devic d automaticallan e y fire fighting facilities shoul e installeb d e importanth t a d e tworkshop th pos f o td an , emergency ventilation system should also be installed at the same time. 235 RADIOACTIVE WASTE FORMS: A REVIE COMPARISOD WAN N R. C. EWING Department of Earth and Planetary Sciences, University of New Mexico, USA Abstract Borosilicate t presenta glas , is e swast th , e for f choico m r mosfo e t countries and for most waste compositions. The selection of borosilicate glass is based n mainlanticipatea n o y d eas f processino e g (glas e wast th e mixed s ar d efri an t , melte t a relativeld temperaturesw lo y d pourean , d into canisters) e facth t , that e technologth s i wely l demonstrate r actuafo d l (radioactive) waste d finallan e , th y assumption that the glass as an aperiodic solid will easily accommodate wide variation n i wasts e stream compositions whic e extremelar h y comple d variedan x . There are, however, alternative waste form e ssingl b whic y r polyphaso e ma h e crystalline ceramics. Principal ceramic nuclear waste forms include: Synroc, tailored ceramic = supercalcine) ( s , TiOz-matrix ceramics, glass ceramics, monazite, synthetic "basalt", cementitious materials, and FUETAP concrete. In addition, there are a number of "novel" ceramic waste forms which have been developed to only the most preliminary stages (e.g., crichtonite and cesium- titanates) d there an severa,ar e l multi-barrier strategies which encapsulate on e ceramic waste form in another. Finally, in recent years, spent fuel has become n importana t waste form. This paper will briefly describ e importancth e d an e types of ceramic waste forms that have been developed and review their advantages and disadvantages. 2. INTRODUCTION During the period from 1977 to 1982, there was a tremendous diversity in the types of nuclear waste forms under development. In the United States, much f thio s work ended wite decisioe borosilicatth hus o t n ee wastth glas s ea s forr fo m defense wast t Savannaa e e constructioh th Rive d e Defensan th r f o n e Waste Processing Facility. Synroc, a ceramic waste form, was selected as the alternative waste form t furthebu , r developmen e Uniteth n i dt State s e absencendeth n i f do e funded programs. Major research and development programs for the developmen f o Synrot c continue n i Australid a culminatin e constructioth n i g f o n a "cold" Synroc pilot-scale processing plant. Basic researce propertieth n o h f o s Synroc have been continue e Australiath t a d n National Universit e latth e y b y Professor Ted Ringwood and his colleagues and at the Australian Nuclear Science and Technology Organization with collaborative work at the Japan Atomic Energy Research Institut d AERan e E Harwele Uniteth n i dl Kingdom. Synroc remains perhap e mosth s t thoroughly studied ceramic alternativ o t borosilicale e glass. Investigations into the properties and performance of other ceramic waste forms have revived during the past ten years for application to special waste stream compositions t LawrencA . e Livermore National Laboratory a Mixe, d Waste Management Facility is being developed in order to demonstrate an alternative to incineration e wastTh . e a derivativfor s i m a Synro f o e c composition originally develope e immobilizatioth r fo d f o reprocessen d residue t a Savannas h Riverd an , typical phases include zirconolite, perovskite, spinel, nephelin d rutilean e . Although the radioactivity is low in this waste form, this does illustrate the ubiquity of a rather limited set of phases, some of which will be discussed in this presentation e Idahth t oA Nationa. l Engineering Laboratory n iron-enrichea , d basalt waste form has been under development, and the addition of ZrO2 and TiOi has produced zirconolite crystals as an actinide host in a silicate ceramic, i.e., a 237 basalt. Most recently, spent fuel - a metal clad, ceramic oxide - has received important consideration as a waste form because its use eliminates the need for reprocessing s i highl t i yd stablan , e under reducing conditions. 3. IMPORTANC WASTE TH F EEO FORM Despit e greath e t challeng f o handline g chemically complex wastes, which are highly radioactive and of great volume, the greatest challenge still lies in the development and evaluation of the long-term durability of waste forms. Materials scientists will have to design materials to performance standards that t measureno e ar n decadesi de ^ years Th 10 t rathe o bu .t ,e measure ar ^ r 10 n i d issue of long-term durability is unusual in materials science and requires interdisciplinary research programs with rather unusual combinationf o s subdiscipline processing/synthesi— s s technologies, materials properties, mineralogy and geochemistry. Thus, even after issues of technological feasibility and cost are considered and settled, the most difficult scientific issue remains: What is the long-term durability of the waste form? and What is the effect of improved durability performancethe on assessment calculatedthe and doseto humans ? Why this interese wastth n ei t form, whe o mucs n h effor s devotei e t th o t d final disposa f nucleao l r wasta geologi n i e c repository n recenI ? t yearse th , preponderance of effort and attention has been on the geologic repository as the long-term barrier. Performance assessment a well-designe f o s d geologic repository have focused on the development of models that attempt to describe the complex interaction of geologic, hydrologie, geochemical and geophysical barriers over long periods. Much less attention has been paid to the long-term behavior of the waste form; however, the waste form can be the first and final barrier to the release of long-lived nuclides, such as the actinides (Pu, Np and Ir 129o ." Tc r Am)o , Thus, research program n radioactivo s e waste forms will require: ) carefui. l consideration f o synthesis d processinan s g technologies; ii.) a detailed characterization of the wastes and the waste form after immobilization; iii.)n extensiva e data bas f o corrosion/alteratioe n experiments ovea r wide range of conditions, and in some cases, for extended periods (both repository-relevant and special experiments designed to elucidate the corrosion mechanism); iv.) kinetic e modelcorrosion/alteratioth f o s n procesd an s thermodynamic models that predic e formatioth t d stabilitan n f o phasey s which will control solution compositions; j v studie f relevano s t natural phase r systemo s o confirt s m experimental and extrapolated results. 4. PRINCIPLES OF NUCLIDE ISOLATION IN CRYSTALLINE CERAMICS n contrasI o glast t s waste form n whici se radionuclideth h n principli e ar s e homogeneously distributed throughout the waste solid, ceramic waste forms incorporate radionuclide o waystw n :i s ) (1 Radionuclide y occupma s y specific atomic position e periodith n i s c structures of constituent crystalline phases, that is as a dilute solid solution. The coordination polyhedra in each phase impose specific size, charge, and bonding constraint e nuclideth n e o incorporatesb s n thaca t d inte structureth o . This means that ideal waste form phases usually have relatively complex structure types with a number of different coordination polyhedra of various sizes and 238 shape d witan sh multiple subsiitutional scheme o t allo sr chargfo w e balance with radionuclide substitutions. Extensive nuclide substitution can result in cation and anion vacancies, interstitial defects d finallan , y change n structuri s e e typeOn . expects, and finds, the formation of polytypes and twinning on a fine scale. The point defect n themselveca s s becom e radionuclidese th site r fo s . Excepn i t unusual situations (e.g., monazite, CePU4) e complexitth , e wastth f eo y composition results in the formation of a polyphase assemblage (e.g., Synroc consists of phases suc s zirconolitea h , CaZrTiaO?; perovskite, CaTi0 d "hollandite"an 3; , BaAlaTieOie), with unequal partitioning of radionuclides between the phases. In Synroc, actinides partition preferentially into the zirconolite phase. The polyphase assemblages are sensitive to waste stream compositions, and minor phases form, including glass, segregated along grain boundaries. Ideally, all waste stream elements, radioactiv d non-radioactivean e e importanar , t componente th n i s phases formed n somI . e rare cases a singl, e phase (e.g., monazit r sodiuo e m zirconium phosphate, NZP n incorporatca ) e radionuclideth e f nearlo l al y s inta o single structure. ) (2 Radioactive phases, perhaps resulting from simply dryin e wastth g e sludges, can be encapsulated in non-radioactive phases. The most common approach has been to encapsulate individual grains of radioactive phases in Ti02 r AhOso , mainly becaus f theio e r extremel solubilitiesw lo y . This usually requires major modifications to the waste stream composition and special processing considerations to maintain temperatures that are low enough to avoid volatilization of radionuclides. A similar approach may be taken with low temperature assemblages (e.g., mixing with concrete) n thii t s bu cas, ee therth s i e possibility of reaction between the encapsulating phase and the radioactive phases. Both of these types of waste forms are specifically fabricated for the incorporation or encapsulation of radionuclides. Spent fuel - a metal-clad, U02 ceramic - is designed without consideration of its waste form properties. The properties of spent fuel as a waste form are determined primarily by the e reactorth irradiatio n i . 2 U0 Radionucliden e historth f o y e distributear s d throug e fueth h l matri s a interstitiax l defects s exsolved/precipitatea , d phases, along grain boundaries n void i e d fue crackr th an o ls , f o shees t gap. . 5 ADVANTAGE DISADVANTAGED AN S F CERAMIO S C WASTE FORMS e maiTh n advantag f ceramio e c waste form e facs th tlie n i thas t they hold the potential for engineering a phase assemblage which provides unique structural hosts for specific radionuclides. Ideally these hosts should be thermodynamically stable, but for most repository environments this is unlikely (a notable exception is sphene, CaTiSiOS, in the ground waters of the Canadian n alreadca e shield)yon demonstratt bu ; e greater stabilit e cerami r th somfo yf o e c phases than for metastable borosilicate glass. The development of ceramic waste forms which are stable at high temperatures has even more important implications: 1) higher thermal stability provides the possibility of higher waste loadings d thu an a ,reductios e amounth n f i nmateriao te handledb o t l ) highe2 ; r thermal stability means that disposal can occur in rock units at greater depth or in canister arrays with closer spacings. The instability of borosilicate glass at higher temperature s i wels ln facti known d , an disposa, l conceptr fo s borosilicate glas e limitatione shapeth ar s y b d s imposes thermait y b d l instability. Becaus e thermath e e decale resulth eventh f fissiof o ys o i tt n products (137d Can s 90Sr) whic e shorar h t lived (half live f 30.o s d 28.2an 1 years, respectively), temporary, ventilated storag s beeha e n propose r vitrifiefo d d waste prio o t finar l disposal. Even without the higher thermal stability, higher waste amounts are incorporate n ceramici d s becaus f o theie r higher density. A final consideration is that many of these ceramic phases occur in nature (e.g. zirconolite, pyrochlore, perovskite, zircon, monazite, uraninite, etc.). This provide e possibilitth s f evaluatino y e long-terth g m durabilit f theso y e phasen i s 239 the presence of aqueous solutions and with a-decay doses that reach values comparable to those which the waste form will experience in the first 1,000 to 10,000 years after disposal e abilitTh o .validatt y e projected long-term behavios i r a critica e performancl th par f o t e assessmen e nucleath f o tr waste containment strategies. The disadvantages of ceramic waste forms are an inherent part of their complex microstructures. First, the extrapolated corrosion history of a waste form s i criticas evaluationit o t r lpolyphas Fo . e ceramics e corrosioth , n process i s inherently more complicated than modelling the corrosion process of an essentially single phase glass. Despit a considerable e amoun f experimentao t l data, o confirmethern s i e de long-terth mode r fo l m corrosio f Synroco n . More importantly, thers neveha e r bee a nperformanc e assessmen e lonth gf o tter m behavior of a ceramic waste form. Second, the atomic periodicity of crystalline material s disruptei s y a-decab d y damage. Thus ,e possibilittherth s i ea f o y radiation-induced transformation e froperiodic-to-aperiodith m c state (amorphizatio r metamictization)o n . a welThi s i ls known proces n minerali s s which contain uranium and thorium and is observed in actinide-doped phases (e.g. Pu-doped zircon) e procesTh .s i mitigates y naturab d l annealing (which increases with higher temperatures) and some phases anneal at low enough temperatures that they are only found in the crystalline state (e.g. uraninite and monazite). Still, this transformation can result in decreased chemical durability and the volume expansion associated with the transformation can cause microfracturing wit n increase a hsurfacth n i ee area expose o corrodint d g fluids. <>. CONCLUSIONS Ther s everi e y reaso o expect n t that waste form performanc e mucb n h ca e improved over what is now accepted for borosilicate glass. Ceramic waste forms, such as Synroc, have already demonstrated this improved performance under certain conditions (e.g. hydrothermal, up to 300°C). Prudence requires that research and development of these second generation waste forms continue: y strategAn f ) isolatioo i. y n should emphasiz e near-fielth e d containmene th f o t radionuclides. This is primarily a function of waste form or "waste package" performance. Strategies that -»rely solely on long travel times, dispersal or dilution, implicitly presume releas d movemenan e f radionuclideso t . ii.)e Th long-term performance assessment of the success of radionuclide containment require e developmenth s f deterministio t c e modelfuturth f eo s physicad an l chemical behavior of each part of the barrier system. Although difficult, it is almost certainly easier to model the chemistry and physics of corrosion and alteration of waste forms, with the subsequent release or retention of radionuclides over some rang f conditionso e o develot , s i tha t i pn coupled hydrologie, geochemical and geophysical models of the movement of radionuclides throug e far-fiel th a hgeologi f o d c repository e extrapolatioTh . f o n corrosion behavior over long periods a firmerest n o sr scientific foundation than e extrapolateth dn examplea behavio s a , of , r hydrologie systems e thasitar te specific and highly dependent on idealized boundary conditions (e.g., climate and recharge). Hi.) Finally, natural phases (minerals and glasses) provide an approach to "confirming" the hypothesized long-term behavior of waste form phases in specific geochemical environments. Indeed, "natural analogue" studies have become an important component of performance assessment. The very specific use of natural analogue phases (i.e., naturally occurring phases that are structurally and chemically similar to waste form phases) in determining the corrosio r o alteration n behavio f o wastr e form phases provides fundamental data e thasignificanar t r performancfo t e assessment. 240 ACKNOWLEDGEMENTS This paper is a summary of a collaborative effort with Professor Werner Lutze that spans some fifteen years at the Hahn-Meitner-Institut in Berlin, the Kernforschungszentrum Karlsruhe, and now, finally, at the University of New Mexico. The product of this collaboration is Radioactive Waste Forms for the Future (North-Holland, Amsterdam, 1988). BIBLIOGRAPHY CHAPMAN, N., "Natural analogues: The state of play in 1992", Proceedings of the Third International Conferenc n Higo e h Level Radioactive Waste Management (TULENKO, J.S., Ed.), American Nuclear Society (1992) 1695-1700. EWING, R.C., "The rol f naturao e l analogue n performanci s e assessment: Applications and limitations", Proceedings of the Third International Conference on High Level Radioactive Waste Management (TULENKO, J.S., Ed.), American Nuclear Society (1992) 1429-1436. EWING, R.C. (Ed), Thematic issu n Nucleao e r Waste. Journa f Nucleao l r Materials 190 (1992) 347 pages. A recent review of nuclear waste form research, particularly of spent nuclear fuel. EWING, R.C., LUTZE, W., High-level nuclear waste immobilization with ceramics, Ceramics Internationa 7 (19911 l ) 287. HENCH, L.L., CLARK, D.C., High-level waste immobilization forms. Nuclear & Chemical Waste Management 5 (1984) 149. A summary of waste form properties comparing borosilicate glass alternativeto waste forms. LUTZE , EWINGW. , , R.C. (Eds), Radioactive Waste e FutureFormth r fo s, North- Holland, Amsterdam (1988). A thorough summary of waste form types and state-of-knowledgepropertiesthe to up in 1987. HATCH, L.P., "Ultimate disposa f radioactivo l e wastes". American Scientis1 4 t (1953firste th f )o proposals 410e On . f alternativeo waste forms. . ROY"SciencR , e Underlying Radioactive Waste Management: Statu d Needsan s " Scientific Basis for Nuclear Waste Management (Proceedings of the Materials Research Society, Boston, 1978), vol .l (McCARTHY, G.J., Ed.), Plenum PressY N , (1979) 1-20 n unusuallyA . prescient summary f researcho requirementsr fo waste form development. RINGWOOD, A.E., KESSON, S.E., REEVE, D.D., LEVINS, D.M., RAMM, E.J., "SYNROC", Radioactive Waste e FuturFormth r efo s (LUZE , EWINGW. , , R.C., ed.) North-Holland, Amsterdam (1988) 233-334. A comprehensive survey of the principles of nuclide containment in a titanate-based ceramic. 241 DEVELOPMEN NUCLEAA F TO R WASTE DRUM OF CONCRETE WEN YING HUI Beijing Institute of Nuclear Engineering, Beijing, China Abstract The paper describes the selection of raw materials for a nuclear waste concrete drum and the propertie e materialsth f o s e formulatioth , propertied an nconcretee e th th f o sd an , specificatio technicad nan l quality requirement drum.The th r sfo e manufacture essentiald san technology experimente th , checkd san effective wels th s a s a l e quality contro qualitd an l y assurance carried out in the course of production are described. The nuclear waste drum has a simple structure, and is made from easily available raw material containd san rationasa l formulatio concretef no compressive Th . e strengte th f ho dru s mgreatei r tha MPa0 7 ne tensil th , e strengt s greatei h r tha MPa5 n e nitrogeth , n permeability is 3.6-2.16xlO'18 m2. The error of the drum in dimensions is ± 2 mm. The external surface of the drum is smooth. The drum meets the China standards regarding sandy surface, void and crack. The appraisal results show that quality of the drum is as good as of the same foreign product. Our research shows China has the ability to develop and produce nuclear waste concrete containers and lays the foundation for standardization and series of the nuclear waste containe wastr rfo e packin transportind gan Chinan gi . 1. INTRODUCTION A large amount of radioactive waste is generated from nuclear facilities. According to the properties of radioactive waste and the requirements for interim storage and disposal, the concrete container is widely used for LLW and ILW in the world. Now Qin Shan and Daya Bay nuclear power plants have already been operated concrete th , e drum develope alss dha o been manufactured and utilized there. The development of the concrete drum was successful, the qualitie drue th mf so camadvanceo t p eu d standard same th f eso foreign products. 2. CONCRETE DRUM CONSTITUENTS AND PROPERTIES A reinforced concrete drum, intended for loading with low and intermediate level radioactive waste, shall mee followine th t g requirements: Confinement during the various phases of handling, transport and storage; Biological shielding against radiation emitted. TABLE I. DRUM SPECIFICATIONS thickness (cm) type D (m) H(m) base side wall Cl 1.4 1. 3 15 15 C2 1. 4 1.3 30 30 C3 1.4 1. 3 40 40 C4 1.4 1 3 15 15 243 The four types of a concrete drum are shown in Table I. Cl, C2 drums are used for storage of solidified cemented low level spent resins, concentrates or sludges. C3 drum is use storagr dfo solidifief eo d cemented high level resins. Figur showe1 concrete sth e drum. 1. concrete drum 2. s tee 1 ins ide drum 3. concrete drum cover FIG. 1. Diagram of the nuclear -waste concrete drum. 3. TECHNICAL REQUIREMENTS 3.1. Raw materials The technical standards of sand, gravel, cement, mixing water and admixture required must comply with Chinese standards. 3.2. Propertie concretf so e requiree Th d propertie concretf so showe ear Tabln i . Froe2 above mth e requirements, it is clear that the concrete must have high tensile strength and density. TABLE II. CONCRETE PROPORTION AND CURING PARAMETERS i tern scope of selection content of cement 400-550kg/ma rati f sano o d 25-33% . admixture superplast icizer (FE) 0.5-1%, silica fume 4-10% curing system natural curing steam curing (40-70'C) 3.3. Properties of drums The acceptance criteria of the drum are: (1) Reinforcement bar: Visible inspection of the external surface and the bottom of the container mus carriee checo b tt t kdou that therreinforcemeno n s ei r appearinba t n go surfacee th . 244 ) Finish(2 externae Th : l conditio containee th f no r mus suce tb h that whe t comeni f o t sou the mould no further manual or mechanical action is needed. Any trace of smoothing, polishing or stopping, either in the bottom or on the external surface, is the cause for refusal. (3) Sandy surface: All visible surfaces with a shallow depth of some 1/10 mm maximum are considered as sandy surfaces. Surfaces are not cumulative. Included in this fault e chipar s cause y remova b de dru th mf o l fro e mouldy impactmth b r o , s during handling faule Th t. must hav esurfaca r otheo ee rareon greateo n an i rcm tha0 n10 2 followino tw e th 0 rectanglgf 2 a o r x shapes o ; m c cm e5 : 0 eithe1 x squarra m c 0 e1 cm. (4) Voids: Voids are all those faults concerning a small surface but of considerable depth. Example thif so s typ faulf elargo e ar t e bubble r lacs o f fillin ko g aroun r belodo e wth U-bar. To avoid rejection, the fault must lie within the allowed surface area, viz, 25 differeno other o tw f e o r cmon t n shape2i followss sa r o ; : eithecm 5 x squarra m c e5 a rectangle 10 cm x 2.5 cm. The depth of all must be checked. When penetration is such as to cause doubt as to the bearing capacities of this reference surface, the containe rejecteds i r . ) Crack(5 : Crac bottoe t surfac th lengts ou k it d e mwidtles, s i ean th h sn mm hthao 1 n0. is less than 20 cm. On the outside surface, crack width is more than 0.1 mm but is less lengts bottoe it th , n less hi mo mm s; tha cractha3 cm n0. nk2 widt mors hi e tha1 n0. mm but is less than 0.3 mm, its length is less than 5 cm. Crack width greater than 0.3 t acceptablemno ms i . ) Insid(6 e quality: Cutting drums vertically intpartso otw , visual inspection f insidso e quality, on all cutting surfaces, no cracks and no voids deeper than 1 cm were inspected. The aggregates are well-distributed. The reinforcements are in good location and distribution, and the quality of concrete is good. (7) Dimensional checks: The following tolerances shall be checked and complied with: m m 2 Externa ± l diamete heighd an r t Internal diameter (entrance, lower, upper) ± 5 mm Concentricitm m 4 f inne yo outed ± an r r axis: (8) Seepage: After the concrete has dried, there must be no seepage from containers filled with water. 4. TEST RESULTS The entire process of the development was divided in two stages.The first stage included design of the drum, selection of raw materials, concrete ratio, manufacture of a test drum and cutting check.The second stage included developmen f machino t e tools, commercial manufactur acceptances it drue d th mf an eo . 4.1. Selectio materialw ra f no s followine Th g materials have been selecte drume th r :dfo Sand: The Yongding River sand Gravel: Crushed limestone (size fro) 5-2mmm 0 Mentougou rock plant Cement: 525* Portland cement produced by Jidong Cement Works of China Mixing water: Potable water Admixture: The superplasticizer and silica ash. 245 4.2. Concrete ratio Accordin properte th go t y requirement manufacturd san e conditions basie ,th c proportion and curing system were determined with testing over and over again. These basic proportion and curing syste showe mar testin e Tabln ni Th . geII resultconcrete th r sfo e propertiee sar show Tabln ni resulte eth IIIsatisfactoryl e sar .Al l itemAl . s mee requiremente th t s because the superplasticizer and other admixture were used in the concrete, the strength and density increase and other performances of the concrete are improved. TABLE III. CONCRETE PROPERTIES test designation Acceptance test ing creter ia results slump test m 3c ± 5 2m . 6c 6-4. weight loss (28 days) <30kg/cm8 14. 1-17. lkg/cm3 shrinkage (28 days) <600 Hm/m 242-30m m/ 3M nominal compressive strength (28 days) >55MPa 77. 9-82. 2MPa nominal tensile strengt 8 days(2 h ) > 5. OMPa 5. 5-6. iMPa [resistance to compression (or traction or flexion) 25 cycles] Q flff > ÖU» >96% resistance to compression (or) on specime 20'Ct a n C ±1' resistanc o freezint e g defreezing cycles <500 M m/m 0-6 v- m/m swell ing at 25 cycles permeabilit o nitroget y n afte 8 day2 r s 5X10' 18m2 3.6-2. 16XlO-18m2 4.3. Concrete drums From July 1989, whe ncontraca signes wa t d through July 1992, more thadrum0 n70 s have been produced, each drum inspectes (10wa ) 0% d carefully rate qualifief th ,eo d products was more than 90%, quality came up to the technical requirements as specified in Section 3.3. Through the overall acceptance checks, properties of the nuclear waste concrete drum developed by us came up to the advanced level of same foreign products. The first batch products, 596 concrete drums, have been transported to Da Ya Bay nuclear power plant and have been used to the satisfaction of Chinese and foreign experts. Figure 2 shows the nuclear waste concrete drum develope. us y db . PRACTIC5 E We adopted the way that the institute and the factory jointly develop the drum by stages. The institute is responsible for: examination and approval of the design establishment of technical requirements establishmen f acceptanco t e programm A criteriQ d aan e selectio materialw ra f nconcreto d san e proportion overall quality management from desig manufacturo nt surveillancd ean assuro et e eth drum quality. 246 FIG. 2. The concrete drum. The factory is responsible for manufacturing. The following works were performed in course th f manufactureeo : (1) The raw materials (sand, gravel, cement, admixture etc.) were controlled strictly accordin technicae th o gt l requirements. Sand and gravel followine Th g check tests shal performee b l d regularly: Ever drums0 y10 : granulometry, cleanliness, méthylène blue test, chlorin contenn eio t Ever drums0 y20 : flatness index, colorimetry Every 500 drums or every year whichever is the more frequent: Los Angeles coefficient, wear resistance. If needed, all aggregates older than 6 months shall be removed from the storage area and not allowed to be used. Cement followine Th g test performede sar : For each delivery day: setting test, hot expansion test, specific area surface and density, false setting test. Each month: mechanical properties at 28 days, heat of hydration.. Cement shall neve storee b r morr dfo e tha monthsn3 . (2) The operating parameters (concrete ratio, curing, unmoulding, etc.) were controlled in the course of manufacture. 247 (3) The concrete properties were tested systematically. The acceptance tests are to be carried out with nominal-formula concret consis d followinge ean th f o t : Slump test: 28 days compressive and tensile strength test, loss of weight and shrinkage test, permeability to nitrogen test, freezing-defrosting cycle test. The concrete production unit shall be able to maintain the following tolerance in actual mix compositions with respec theoreticao tt l composition weight)n s(i : sand ±3%, aggregates a wholsa e ±3%, cemen wate% ±2 tr ±2%, admixture ±5%. drume Th s ) wer(4 e strictly controlle acceptancer dfo . Durin acceptancee gth drue r ,on m pe type will be checked as follows: Surve inspecd yan t installatio weldef no d wire mesh, casting, vibrating, curing. Afte day7 r removaf so l from mould, chec drue kth m dimension, seepage tese ton by one. The drum's exterior quality is a direct window for user assesses product quality, visuae th made hola b l s o dchecea t points kha , each drum mus checkee b t d strictly, each drum itself has working sheet on which strength, overall dimension, surface qualit seepagd yan e condition shal recordede b l workine th , g shee alss i t o qualita y evidenc qualitd ean y certificate. (5) The quality management was performed seriously. From design to manufacture, overall quality management has been performed, effective QA surveillance measures have been undertaken drum'e .Th s development adopted PDCA circulating procedure, tha Plans ti - Do-Check-Action management method e organizatioTh . n system assured thae th t development of drum was always under control. 6. CONCLUSION (1) The nuclear waste concrete drum developed by BINE has the following advantages and properties: easy to obtain raw materials, rational formula in concrete, the compressive strength of the drum is more than 70 MPa, the tensile strength is more than 5 MPa, the permeability of nitrogen is 3.6-2.16xlO~18 m2. The error in dimensions of the drum is ± 2 mm. The external surface of the drum is smooth. The drum complies with China standards regarding sandy surface, void and crack. The appraisal results show that qualit drue goos a th msams s f di a yo e foreign product. (2) While developin concrete gth e drum, high precision moulds, high-frequency vibrating table, special grab, cutting machine and nitrogen-seepage meter have been developed. The drum construction is rational, special grab can be remotely-operated, it opens a new treato t y , packwa , transpor intermediatd an stor d w an t elo e waste from every nuclear power plant in China. (3) Overall quality management, effective QA, QC surveillance is the assurance of the success in the drum's development. REFERENCES [1] Draft contract for supply of concrete drums for the delivery years 1991 to 1995 inclusive [2]. Design criteria for solid waste treatment unit metallic and concrete drums Issued by SOFINEL, Internal identification number GD/RE-87, 0534/NEAW. 248 A STUDY ON THE WET CHEMICAL OXIDATION AND SOLIDIFICATION OF RADIOACTIVE SPENT ION EXCHANGE RESINS TIANBA , GUICHUOWU N YUN, JIAQUA , YUCANWU E IYI Institute of Nuclear Technology, Tsinghua University, Beijing, China Abstract This paper describe researce sth h decompositioworke th n so Ion-Exchangf no e Resins(IERs H2Û2n )i - Fe2+/Cu2+ catalysis systems for volume reduction and improvement of immobilization in cement. The resins used in the study were polystyrene strong acidic and basic resins containing about 45% of water. 13 90 134 5l The. radioactive spent resins loadinCo'Csg60 Cs ,Sr, anCdr with a radioactive activity level of 4GBq/m3 were obtained fro mreactoa r installation bees ha nt I .foun batcn di h scale experiment that many factor influencs sha decompositioe th n eo mose DERsf th no t d importanan , t one H2Oe sar 2 dosage, H2Û2 dose rate, temperatur value H bese p d t.Th etemperaturan e rang 97-99°Cs ei pH-value .Th resif eo n slurry chosen in this study is 2.0-3.0. The appropriate dosage of H2O2(30%vol.) is 200ml/25g wet mixed resins decompositioe Th . n rati 100os i mor d catior % an aniofo d e nthaan % n n90 ffiR s respectively, while it is 85% for mixed resins(as TOC-value). The analytical results indicates that the radioactive nuclides loaded in the spent resins are concentrated in decom-position solution and solid residues. No radioactivity enters int off-gase oth , whil condensate eth e fro reactioe mth n syste radioactiva s mha e activity of 1.65Bo/l. Foaming is a problem associated with resin dissolution. Addition of a little amount of anti-foam agent can solve this problem very well. Three cementation materials have been chosen for encapsulatio decompositiof no nthree residueth f eo kinl solidificatioAl f .d o n material producn sca e qualified cemented products with excellent propertie lonr sfo g term storage adoptee .Th d volume reduction(VR) proces significantln sca y reduce waste volum solidifief eo d product decrease% 40 y sb compared with that of original spent resin. 1. Introduction Radioactive spent resin originated from nuclear power station and other nuclear installations is usually directly encapsulated in cement. Cementing spent resin in status quo ante inevitably increases the ultimate disposal volum wastf e o thu d disposae ean s th l expens correspondingls ei y increased. Therefore, it is suggested that spent resin be treated firstly and disposed in a volume-reduced state. Spent resin can be mineralized by means of incineration, pyroiysis, acid degradation and low-temperature wet oxidation as well. It is well known that the catalytic decomposition of hydrogen peroxide with the catalysis of either ferrous or ferric salt chaia s si n radical reaction which yields hydroxyl radical hig,a h reactive radical. This- proces bees sha n popularly applie destructioe th dn i refractorf no y organic substances containe wastn di e wate recenn I r. t yearsbees ha n t ,i widel y investigate prospectiva s da e treatmenoptioe th r nfo spenf to t resin temperaturw Lo . oxidatiot ewe n process take advantagee sth higf so h reactive radicals suc HO-s ha , O-, etc. generated fro decompositioe mth oxidantf no s suc catalyzeh HsQjs ha C , ,Oj one-electroy db n trans.-metal ions, alkal yellod ian w phosphorus etc Catalyze. d low-temperatur oxidatiot ewe s nha evident superiorit rad-wastn yi e treatmen moderats it r tfo e operation conditio sufficiend nan t volume- reduction effect^. objectivee Th thif so : s studto s ywa -decompose the spent resin before solidifying with cement; -reduce the volume of the waste product immobilized in cement; -improv qualitiee e th wast e th f eso producd tan -provid preparatioe datth r afo resif no n decomposition procedure H^O-y sb j oxidation process. 249 ion-exchange Th e resin nuclea e useth n di r installation Chinn si sulphonates awa quaternaryaminater do d 2 t 2 cross-linked polystyrene resin. H2O2-Fe " "/Cu * syste selectes decompositiome wa th r dfo spenf no t resin. determino T volume-reductioe eth n processfactoe th f o rdecompositioe ,th n residu cementes ewa d with various propertiecemene th producd e th tan f so t were tested. A Fentoo st n reactio n, Habe Weisd postulated ran sha followine dth g chain reactions'2': Fe^+HjO, —— > Fe* +OH- +HO- (1) ) (2 2 — X° HO° —++ HOO^2 »- - HOO- + \O2 —— > Ö2 +Hp +HO- (3) HO- + Fe2* —— > Fe3* -i-OH- (4) 34 24 ) (5 Fe " +H » — 2FeO— 2 " -i-HOO * +H - 3 24 ) (6 Fe * +HOO - * —+H — >2 FeO + " High reactive hydroxyl radical reacts with organi eithey cb r abstractio hydrogef no additior no o nt an unsaturated site to yield organic radicals which are readily oxidized by O or oxidant ions such as Fe34" , Cu^.etc.P). 2 ) (7 +HOH - R +R - —O —Hj > R-C=C-+HO- ——> • R-C-C-OH (T) J T l > RCOO- (8) ) (9 RCOO- ——CO >R+ ' R- +CU24-2 — »CU++R+ . (io) . +FCR * ——>Fe 2f+ +R (11) Fe+ 34+ - —Cu — > Fe24" +Cv» (12) 2 Cheves Walling pointe that dou t Cu*2 - would induce synergetic catalytic effect with Fe* owingo t reaction (12)W. The resins studied were sulphonated or aminated cross-linked polystyrene. The ideal results of the resin destruction by H^Gh might be presented as follows: CjH+H2SOO Hj gHSO0 3 2O4+2 +2 3 2 2 —CO — 8 > (13) (CH)n +25 n HjO, —— > n CO2 +3n HjO (14) C12H19NO + 3 1 Hj'Oj —— > 12 HjO +38 Hp +NH4OH (15) 3. Laboratory experiments (1) Material and equipment resine Th s use thidn i s study were fresh cross-linked polystyrene strong acidi basid can c resins (in granular form )which marked 732 and 717 respectively, and the radioactive spent resin was obtained from a nuclear installation which loaded nuclides of «Co, «7Cs, 134Cs,90Sr, "Crwith radioactivity leveGBq/ra4 f lo reagente 3.Th s use thidn i s study were hydrogen peroxide; FeSO^THjO, CuCNO^; citric acid; sulfete-resistant cement(SRC); acrylate co-polymer(ACP), epoxide plastic-polyamide-styrene (EPPAS); The analytic instruments were DID AC 800 Multichannel Analyzer( Intertechnique, France) fo -emittinry g nuclide analysis; BH121 backgrounw 6lo ß d an da analyzer(mad Chinan e i emittin - ß r )fo g nuclide detection. TOC Analyzer(ShimadzC O-1 BTO u Company, Japan) for analysis of TOC (Total Organic Carbon) of dissolution residue; SP-2305 Gas- Chromatography(mad Chinan ei conten2 ofmeasuremenr O e f)fo th d gas n an ti 2 . CO f to (2) Process A lot of batch-scale studies were performed to investigate the optimum decomposition process. 25 g resin slurry with 0.9g critic acid 30%(vol, ) hydrogen peroxid 0.0d FeSOean 1 M ^ CuO^O^ solution were adderate e ml/min1 th f t sd o 0.2a d .an 5 ml/min. respectively. Off-gas fro vene mth t lins ewa cooled through a water condenser and simultaneously analyzed with a gas -Chromatograph to observe the release patterns of O2 and CO2 in off-gas. Mechanical stirrer was used to provide homogenous mixing. Whe reactioe nth n mixture reached 85°C dissolutio, initiates nwa proceeded dan d rapidly. The decomposition temperature was kept under 99°C to prevent from foam over. At the end of the reactio decompositiovalue C th f nTO eo n residu analyzes ewa determindo t degradatioe eth n percentage of the tested resins. Radioactive spent resi thes nnwa tested unde same rth e condition mentiones sa d above. Specific radioactive activity of the decomposition product of resins including off-gas, condensate, 250 decomposition solutio solid nan d residue were detecte examindo t distributioe eth radioactivf no e phases nuclidega e ,th liquin si d phas solid ean d phases the d decompositioe ,an n th n residus ewa immobilized in cement after evaporation to compare the volume reduction effect of this process with direct encapsulation in cement. . Result4 Discussiod san n 4. 1 . Radioactive spent resm can be partially mineralized through hydrogen peroxide oxidation catalyzed by "" The operation conditions and the results of the decomposition process are listed in Table I Tabl eI Operatio n conditio result d resie nan th nf s o decompositio n hem Cation resm(732) Anion resm(7J7) Mixed resin* H2Qj/resm**(kg/kg) 3.7 5.0 3 8 Catalyzer/resin reaction time A**» 1.67x10^ l ICxlO-A: 1 A- 1.52x10-* (kg/kg) ,mm B:«"** 149x10-* B 1 88x10-* Citnc acid/resin (kg/kg) -0.06 Anti-foam agent/wet resin 0 001 001 (kg/kg) Temperature ("O 97-99-C Initia valuH p l e 3.0 2.0 30 Reactio resin(hrst we g n tim25 )r efo 2.5 3.5 25-30 TOC value of liquid residue(ppm) . Fro2 economi4 e mth c poin f viewo t unnecessar s i t i , mineralizo yt spene eth t resins completelye Th . decomposition reaction ends up with the high ratio of CO7 to O2 content in the off-gas. A typical gas releasing histor procese th f yo gives si Figurn ni e1 0 20 40 60 80 100 120 140 160 180 duration(min ) F:g. I CO, and O, releasing history 251 4.3. Bench-scale experiments demonstrates that many factors can influence the decomposition of ion exchange resins. Within these factor mose sth t important one HjOe sar dosage, H^O^ addition rate, temperatur reactiovalue H th p f ed o nean system. Batch -scale tests reveals that comparativel7 y high temperature can prompt the decomposition reaction. Reactions conducted under different temperature shows that when the temperature raises, the reaction rate increases rapidly and simultaneously the decomposition of resins occurs more completely The upper limit of the temperature to be controlled is belo boiline wth g poin f waterto 100°C. e i , preveno t , t fro foam-ovee mth r pH e reactiovaluth f eo n mediua ver n i y s mimportani t positio r bot e catalytinfo th h c effecf o t ", Cu overale 2+th /Cu d lan + utilizatio f hydrogeno n peroxid e experimentaTh e l results indicates tha effece tth aciditf reactioo e t th f yo n mediu mmucs i h more marked becaus f hig eo valu H hp e hydrogen peroxide no longer reacts with ferrous salts as the uncharged molecule , but rather as the ion HO2- In a neutral solution all the ferrous ion remains m solution while all the feme compounds has been precipitated High pH-value causes also a self-decomposition of hydrogen peroxide(H,O2- H^O +1/2 (X) on one hand. On the other hand it is necessary to consider the requirement for alkalinity in the subsequent cementation process beforehand Therefore , the pH value chosen in this study was as high as possible i e . pH=0 3 20- As oxidant, H,O decompositioe 2 rolplayth y eke m sa n reactio stude Th ny indicates that insufficient hydrogen peroxide leads to the incomplete destruction of the cross-linkages and thus results m only partial decomposition of resin. The more H,O, is added, the more complete the resin is decomposed and the highe treatmene th s ri tviee cost th economw f n o I .proces e th f practicabilityd o san immobilizatiof yo f no decomposition residu cementn ei experimentae ,th l results exhibit tha spene tth mineralizete resib n nca d incompletely. Ther optimun a s ei m H,,O2 dosage that satisfy both econom practicabilitd yan e th f yo process Practicability can be defined here as the tolerable limit of maximum salt content for encapsulatio concentratee th f no d decomposition residu cemen n ehighesi e th r o t t permissible radioactive solidificatioe leveth f o l n product fro e viemth w poin radiatiof to n protection Figur ereflect2 e sth relationship between H^Oj dosage and the TOC value of decomposition solution in the batch-scale experiments catiowitg 5 exchanghn 2 nio e resi mixer Fo n d spent resm(cation.anion resin=2y b "1 weight), the appropriate dosage of hydrogen peroxide (30% vol.) is 200 ml/25 g resins s alswa Iot found thae organith t c substance containe e decompositioth n i d n solution degraded continually after the reaction was ended and the solution was laid up for a few hours This post- decomposition might be caused by residual trace amount of H^ and organic peroxides produced m the cours decompositionf eo . Foaming is a problem associated with resin dissolution and particularly with the anion exchange resin Foaming wilt onlno ly contaminat e condensateth t alsebu o lea o dsecondart y contaminatio off-gaf no s line by radioactive nuclides at higher temperature This problem was practically eliminated by increasing the stirring ratmaintainind ean dissolutioe gth n temperature below 99° ordeCn i minimizo t r e foaminn gi the reaction system mosn I . t case ssulphonatea d type organic anti-foam agent marked XP- uses 1o wa dt prevent foamin thin gi s study XlOOOppm 0 20 40 60 80 100 120 140 160 180 200 HfeOi dosage ig - Relationship between H,(X dosage and the TOC value of decomposition residue 252 radioactivite Th 4 4. y analysis results obtained during spent resin decomposition process indicates that the radioactive nuclides loaded m the spent resins kept concentrated in the decomposition solution and solid residues. No radioactivity existed in the gas and liquid phase of the decomposition. 4.5 To prepare these decomposition residues for cementation it is necessary to neutralize these residue valuH p f 8—1NaOeo o y st b 0 H solutio thed nan n evaporat t 99°ea C unti salte th l s contenf to evaporated residulimitatioe 40%(wto th t o p t e u e saltf no )du s conten cementationr tfo . Three cementitous matrice solidification si n system were chose immobilizatior nfo decompositiof no n residue. which were sulfate-résistant cement (SRC), AEP-SRC, EPPAS-ARC. The formulations of encapsulation processes are listed in Table II. Table It Solidification parameters and product properties Cement Sulfate-Résistant Cement Polymer no ACP EPPAS Polymer content 0 4% 8% Water/Cement (in weight) 05 05 05 Salt/Cemen weightn (i t ) 035 035 035 Bleed no no no Set time (hrs) Initial 20 16 19 Final 22 2 1 23 Compresstve strength Maintaine day8 2 r sdfo 400 308 41 3 (MPa)* Irradiated" 350 294 387 High temperature stability*** Fine Fine Fine 2 2 Accumulative leaching ratio of total ß for 998x10- 745xlO- 7 17r]0-: 42 days (cm)*»** Dens«y( g.cm°) 206 203 1 92 Volume reduction £actor(VRF)***** 039 037 034 "Tested according to National Standard GB-177-62 "Physical test for cement" **Total irradiate ddosager . 2.8x 105Gy "Tested accordin ASTgo t M D63-74 *** Tested according to National Standard 7023-86 "***VRF=(Onginal volum dissolvef eo d resin—volum finaf eo l cemented productXonginal volume of dissolved resin 5. Conclusion basiO e seriea n th f s o resif so n decomposition conclude s testi t i s, d (1) Radioactive spent ion exchange resin can be successfully destructed by K^Oj in a Fe^/Cu2*"—citric acid system . The resins transform from a solid phase consisting of organic matrix into a liquid phase containing a little amount of organic components and the decomposition ratio is approximately 100% for catio exchangern nio , mor anior fo e mixenr tha% exchangefo n90 % d resi85 d n. ran (2) The radioactive nuclides loaded in the spent resin in the period of decomposition are concentrated completely in the decomposition solution and solid residue, no radioactive contamination associates with ventee b thn ed ca off-ga t directli atmosphero e s th , s yo t e withou furthey tan r treatment; concentratee Th ) (3 d decomposition residue successfullb n eca y immobilize n cemendi t with proper formulatio cementee th d nan d product termn si qualitf so y meet regulatory requirement long-tera r sfo m storage of IL W such factors as compressive strength ( lOMPa), high temperature stability, freeze/ thaw cycles, gamma radiation stability and leaching ratio. (4) The volume reduction percentage of the HjO2 oxidation process is up to 30—40% compared with the volum directlf eo y cemente exchangn dio e resin whic voluma s hha e incremen 80%f o t . 253 , 1"We.al BAXTERt e t , oxidatioA. . R , f organio n c ion-exchange resins with hydrogen peroxide— radwaste process development", radioactive waste management 2. BNES, London, 1989. 2. HABER, F., and WEISS, J., 'The catalytic decomposition of hydrogen peroxide by iron salts", Proceedings of the Royal Society of London, serie , LondosA n (1935). 3. NONHEBEL, D. C. , WALTON, J. C. , "Free-radical chemistry", at the university press, Cambridge(1974). 4. WALLING, C. , and KATO, S. , "The oxidation of alcohols by Fenton's reagent, the effect of copper ion", Journal of the American Chemical Society, Aug., 1971. 254 RADIOBIOLOGICAL WASTES TREATMENT: ASHING TREATMENT AND ASH IMMOBILIZATION WITH CEMENT S. FENG . WANGB , . GONGL , . WANGL A , SH . L , China Institut Radiatior efo n Protection, Taiyuan, Shanxi, China Abstract The possibilit f biologicayo l wastes treatmen discusseB I t usiny b d a RAF-g 3 type rapid ashing apparatu e immobilizatioth d an h wits as hf o cemenn s studiedi t . This apparatus, developed by China Institute for Radiation Protection (CIRP), is used for pretreatment of samples before chemical analysis and physical measurement. The results show that it can ash 3 kg of animal corpora by a batch, ashing time is 6-7 h and the ash content, < 4%(wt). The ashing temperature not exceeding 460"C »e used withou risky an f thig o s h e losseimmobilizeb f radionuclidesn o sca h as de witTh . h cement usin a gin-drur a mixing procedure e optimuTh . m formulatio f cementeo n d waste for e 36±6%(wtmi e Datonth f o ) g Portland^ cement,29±2%(wt f wateo ) d 36±6%(wtan r ) of ash.The cemented waste form e homogeneouar s densed an s s densitIt . s ~1.78g/cm°i y , compressive strength is >7.7MPa. At the 42nd day the leaching rate of ^""Cs and Sr 8e s i 2.5xlO~* cm/ d 9.0xlO~dan " cm/d respectively e coefficienTh . f volumo t e reduction is about 1.6 for ash immobilization with cement. 100 kg of biological wastes becomes a less than f cementeo 11. g 2k d waste form after ashin d solidifyinggan . 1. Introduction Radiobiological waste e mainlar s y produced fro e mradioisotopeth s applicationn i s the radiobiological tests and the assay of radiation medicine. It is not possible to store a lonthe r gmfo period because the e putrescibleyar e otheth n r O .hand , with radiation harmfulness, they must be treated to become harmless waste forms. The RAF-3 type rapid ashing apparatus, developed by the CIRP, is used for pretreatmen f biologicao t l sample before chemical analysi physicad an s l measurement£T1. a emai r lFo amoune biologicath f o t l wastes e apparatuth ,e use b treao t dn ca st them. This report give e feasibilitth s y stud f biologicao y l wastes treatmen usiny b te th g apparatu described an s e resultn investigatioa th s f o se immobilizatio th f o n h as f o n with cement. 2. Ashing treatment 2.1 Apparatus The rapid ashing apparatus CTI consists of a rapid ashing furnace, an oxidation/ reduction gases supply systea temperature-programmin d man g cabinet. Figur showe1 s its exterior view. 2.2 Ashing 2.2.1 Animal aample.3 The animals—the rats and the rabbits are used in the cold tests. The animale radioactiv th e use b n i do t s e ratse th test e . Beforar s e ashingf o l 1m , radioactive solution containing tracer radionuclide *a*Cs, "Sr, °°Co, ""Zn r "o 0l>, Ps ui injected into the each rat, then the rats are all killed with ether. 255 exterioe Fig.Th 1 r vie f rapiwo d ashing apparatus 2.2.2 Ashing Processta] Biological wastes will undergo the charring phase and the ashing phase in the ashing furnace e charrinTh . g proces s conductei s d unde3 atmospher N e iner s th r ga t e in order tha a safrapid t an e d charrin a highe s carriet i a g t rou d temperaturee Th . ashing is conducted under the circumstances of oxidizing atmosphere (0» and NO*) in order to accelerate the oxidation/reduction process. The whole process usually takes 6 — 7 h, varied with waste types and there is no contaminated black in the container wall. The ashing temperature is 360—450'C, and the ash content is < 4 wt.%0 2.2.3 Whereabout f radionuclideo s s after ashing Three safe temperatures have been chose t thina s test: 360'C ,400'C .»öd 460'Ce Th . test results show tha e recoverth t f radionuclideyo s ( *Cs, Sr, °Co, °"Zd n»Puan ) TS ae e BSI can arriv e losse t 100th a e f radionuclided o s% an e undetectedar s , i.e e radionuclideth . s contained in biological wastes remain in the ash (see Table 1). Tab.l Recovery of radionuclides after rats ashing (7 hr of ashing time, at three ashing temperatures) 350 'C 400 'C 450 'C nuclides injection amount measure- recovery measure- recovery measure- recovery ment value ment value ment value o/ o/ of cpm cpm /o cpm /o cpm /o 134Cs 2151 2184 101 ±6 2138 99.4+1 2148 99.9±2 1 8eSr 5765 5739 99.5±1 5611 97.3+ 3 57954 + 0 10 60Co 1416 1406 99.3±1 1408 99.5 ±1 1421 101 ± 0 8DZn 1186 1187 100+1 1182 99.7 ±1 1172 98.9+1 23Spu 350-561 103 ±2 96.3+ 1 98.6+4 256 3. Ash immobilization with cement h characterizatioas e Th 1 3. n There are a lot of brittle bone-black in the ash generated from rapid ashing furnace, Be particle-sizOth e distributio d volume changeb nan n ey ca b d densit h as f o y stirring. a stronge s ha rThh absorptioas e n capacit r waterfo y . Afte hour4 2 r f o swate r immersion, the volume of ash was not changed, but the water content in ash could attai 48.2%(wto t n filteriny b o )41.4%(wtt d an g pumpingy b ) . The adsorption distribution ratio (Kd) and desorption coefficient (Rd), for as*Cs and SBe 0.8Sar r 9 ml/g, 82.6 128ml/gd %an , 21.8% respectively. 3.2 Immobilization with cement 3.2.1 Specimens preparation the Datong cement, which is one type of ordianry Portland cement, was chosen as e immobilizatioth n r matriexperimentou n i x . Cement, water,an n certaii h as dn amount were combine o providt d e propeth e r weight percentage compositions. After stirring for 3 minutes at a speed of approximately 137 rpm, the mixture was poured into the 4-6 cm diameter cylindrical glass container (beaker). Two samples were made for each formulation. After weighting, the specimen containers were covered to mimimize evaporation loss of water and then set aside to cure for 28 days .at ambient temperature.Daily checks were made for the presence of observable free standing water. The well cured specimens were taken out from the glass container (break the beaker), d bottoan p mto surfacee th f specimeno s s were abraded witcoate* 0* h d abrasivo t e mak s height-to-diameteit e r 1:1—l:l.le ratib o ot iel. 3.2.2 Formulation development 3.3.2.1 Requirement r procesfo s d productan s s To successfully desig a nformulatio e immobilizatioth r fo n h waetae n f cementi eo n , a numbe f proceso r producd an s t requirements e satisfiedhavb o t e . Process requirements e produc e mixinTh th ) n i (1 t:g stage mus e fluib t d enougo t h give a homogeneous product. According to teste, the consistencies of cement-ash mortar e rangth shoulf 10-40mra n o ei freo e N e b d) wate(2 . r (bleed) remaine surfacth n o s e aftehours4 2 ) Settinr (3 . g tim f leso e s tha day5 e ndesirable1. ar s . Product requirements: According to the National Standard, the characteristic requirements for cemented waste form are listed below (see Table 2). 3.3.2.2 Formulation design Water-to-cement ratio is used as an important parameter in formulation designcei m. When the ash is solidified, a portion of water is absorbed by the ash and thus is not directly available for hydration of cement. The ratio of weight percentage (W) of total water included in formulation to that (C) of cement can be expressed by W/C-(WAb../C)4(Whyd./C) (1) where W»*,«. — weight percentag f wateo e r absorbed withih as n Wftytf. —weight percentag f wateo e r use r hydratiofo d f cemento n . 257 Tab.2 Characteristics requirement for cemented waste form No. item characteristics requirement free standing water no free standing water compressive strength >5MPa leaching rate at the r 13 The weight of water absorbed within ash is related to the weight of ash by ) (2 a W.X n b ..= wher —weigha e t percentagh as f o e n —ratie weighth f o t percentag f wateeo r absorbed withithao t f asho h t nas . As for Datong OP cement, the ratio of water-to-cement is 0.26—0.34 for hydration of cement a matte s A .f convenienceo r e Wth b3, ,d./ Cs i considere s 0.3a d e media,th n e havw value ed An . a + C -f W - 100 (%). (3) The following relation can be derived from (1),(2) and (3) W-[100Xn-(n-0.3)xC]/(Hn)) (4 . According to equations (3) and (4), a variety of formulations can be designed. 3.3.2.3 Ternary diagra f cement-ash-wateo m r systed man acceptable formulations of cemented waste forms On the basis of the formulation designed above, the test of formulation comparison has been conducted. According to the formulation test results, we can obtain a ternary compositional phase diagram of cemented waste form.Figure 2 is the compositional phase diagram illustrating the region of formulation acceptable for ash immobilization with cement. Formulations which contai e minimuth n m water necessar o t forya m homogeneous mixable mortar fall on the line labeled "mixability limit" . None , of these specimens exhibited free standing water after a 24 h cure time, but the formulations e regioith n f acceptablno e formulation t containo d di ns thos r whicfo e e observablth h e free standing water afte hour4 s absorbe2 rwa s r o dcombine d e intwastth o e form weekso tw r withio . e on n 3.3.2.4 Optimum process formulation Formulations which fall at the boundaries of acceptable limits may not provide good reproducibility s i necessar t I . o yt kee p away from these boundaries a r fa s a s possible reasonablA . e formulation should hav a gooe d workability,o t whic s ha h consider the fluidity of cement-ash mortar, setting time, and the principle of as more h contenas n wasti t e for s possiblema . Thus e optimuth , m process formulation fal n rangi l f acceptableo e formulations.The median formulation is 36%(wt) of cement, 29%(wt) of water, and 36%(wt) of ash. Variation of process parameter is ±5%(wt) of cement, ±2%(wt) of water, and ±6%(wt) h respectivelyoas f . 258 100 7o go Water, %(wt) D acceptable formulation area Elîl optimum process formulation aéra Pig.2 Ternary compositional phase diagram for immobilization of ash with cement 3.3.3 Characterizatio f optimuno m formulation specimen i The specimens were prepared with the median of optimum formulation, i.e. '36%(wt) of cement, 29%(wt) of water, and 36%(wt) of ash waste. The results of specimen property measurements are listed in Table 3. Tab.3 Characterizatio f optimuno m formulation specimens No item result 1 cement-ash mortar consistencym m , 25 2 free standing water, % no 3 setting time, hr 36 4 maximum centre temperatureC ' , 65.5 5 volume density, g/cm3 1.78 6 compressive strength, MPa 7.7 7 compressive strength (afte a weer k immersion)a MP , 8.4 8 compressive strength (after freeze/thaw test), MPa 8.3 9 compressive strength (aftera flamMP e test), 7.9 10 integrality after fall-down test cracking 11 total porosity, % 47 12 leachability at a34Cs LR, cm/d 2.5 xlO-4 CLF, cm 2. 10-. 1y 1 the 42nd day 86Sr LR, cm/d 9.0 xlO-4 CLF, cm 4.4 xlCT2 13 appearence: Homogeneous, dense and free standing solid .— — , — __—————————————————————————————————————— — ~—— — 259 4. Conclusion (1) The rapid ashing apparatus is used for biological waste treatment. Its features e rapiar d charring, rapi dashinw ashinlo d g gan temperature e apparatuh Th as . n ca s 3kg animal corpora by a batch.The ashing temperature can be chosen within the range of 360—460'C- The ashing time is 6—7 h. The ash content is < 4%(wt). If the ashing temperature e t risexcee th f higno ko t o hd no sd losses 460'C ha f radionuclidee* o s* , . a typ f s lighi o e h t as wast e (2Th )e material whic s liablhi o disperseet . According e radioactivtoth e waste disposal requirements t i mus, immobilizee b t d wit a hmatri o t x form a free standing solid. The in-drum mixing procedure is an effective and reliable treatment way for ash immobilization with cement using the formulation proposed in this work. (3) The optimum formulation of cemented waste form is 36±6%(wt) of cement, 29± 2%(wt f watero ) d 36±6%(wtan , e f performanceasho ) Th . f wasto s e form wern i e compliance with the technical equirements except for the impact resistance. By the aif calculationo d s concludei t i , d tha e coefficienth t f o volumt e reductio s i nabou t 1.6 for ash immobilization with cement, 100 kg of biological wastes will produce a less than 11.2 kg of cemented waste form after ashing and solidifying. (4) Cement and ash waste both are porous materials. Cement-ash waste form is also a porous solid, and its total porosity is as high as 47%, which has an effect on e propertieth f o wasts e form, suc s a impach t resistance, compressive strength, leachability. etc. (6) further work is required for ash immobilization with cement in furture, such as choice of additives, development of formulation centered on the improvement of impact resistance, compreesive strengt d leachabilityhan tha o e s wast,th t e n forca m meet all the requirements for disposal. REFERENCES MeisunN [1JI ] t al. e ,e Studa Larg th ,f o ye Rapid Ashing Apparatus, Radiation Protection, Vol.6, No.6, pp366-369 (1986). [2] JIN Meisun, et al., Study of a Rapid Ashing Method for Biological Samples, Radiation Protection, Vol.6, No.6, pp3G7—364 (1986). ] IAE[3 A Technical Report No.118, Vienna (1970). [4] Corlsson , LosseG. , f Radionuclideo s s Relate o t dHig h Temperature Ashing, INlS-mf-10046 (1986). [6] CHEN Xianjun ProposaA , r Determininfo l g Leach Facto f Fixatioo r n Producf o t Radioactive Waste, Radiation Protection, 2(1), 16 (1982). [61 G.Aronld., Waste Form Development Annual Progress Report, October 1981-September 1982, BNL 51614, UN-70 (1982). [7] C.G.Honard, et al., Immobilization of Ion-Exchange Resins in Cement, Final Report 1326R (1991)N EU 2E , . [8] CHEN Baisong, et al., Study on Cement Monolith Solidification for Immobilizing Intermediate Level Waste from Reprocessing Plants, IAEA-TECDOC-668, p.83-106 (1990). 260 DEVELOPMENT OF THERMOPLASTIC SOLIDIFICATION PROCESS FOR URBAN SOLD) RADIOACTIVE WASTES JING WEIGUAN Municipal Radioactive Waste Disposal Experimental Centre, Shanghai, China Abstract Urban radioactive solid wastes come mainly from laboratories and hospitals using nuclear technolog radioisotopesd yan . Mos f theo t mcombustibls i the d treatee ean yar y db incineration intwhich oas dispersivs h i easil y ma yd contaminatean environmente eth thir s.Fo reason the immobilization of the ash is required. Spent ion exchange resins are also dispersiv thed ean y convertee neeb o dt d into stable waste forms. This paper describee sth technological process and operation conditions for polymerization of the incineration ash and n spenio exchangt e resins e witth thermoplastih c solidification unit. 1. Introduction As compared with cement, bitumen, glass, or complex material solidification process, the thermoplastic solidification process is characterized by simplicity in technology, less equipment, safe operation, and low investment. There are some other advantages in thermoplastic solidification process such as easy operation polymerizationo n , curino n d gan , treatmen f solidifieo t d products. This paper also gives the results of physico — chemical characteristics and radiation resistanc f solidifieeo d wastes. 2- The process and equipment According to the principle in thermoplastic solidification, the thermoplastic material (the solidifying agent) becomes soft and plastic so that other materials (contents) can be incorporated wit s heatedh a t return i solie d th an d,o t sstat s cooledea . e procesIth n e solidifyinth s g agen d contenan t t were pretreate y screeningb d , dewatering, and mixing steps. Then the mixed material was fed to screw extruder and remixing, softening, compacting, and plasticating steps were carried out at certain operating temperatures. Finally solidified products were produced through pelleter and packaged into storage drums. Thermoplastic solidification unit is shown in Fig. 1. e followinTh g parameters were controlle processe th n di : The particle-size of the solidifying agent and content were controlled below 40 mesh ; Dewatering temperature was 85 *C in thermostat oven; Accordin e softeninth o gt meltind an g g pointe solidifyinth f so g agenmeasurede b o t t , the operating parameters (content ratio, screw temperature, screw rotational speed) were determined. The process parameters of thermoplastic solidification are shown in Table 1. 261 - pellete0 - dru1 - 3 moto- scre 2 0 4 0 m0 w r extruder 05,0 8- hoppe - mixe 6 0 r 07,10,1 2- vibratio n - vibratiodru h conveye1 as 1 m - 9 n0 rscree n Fig. 1 Thermoplastic Solidification Unit Table 1 Parameters of thermoplastic solidification process Temperature °( sC ) Screw Solidified speed waste Feeding Softening Plastication Extruding zone zone zone zone (rpm) PVC+ash, resin 140 0 18 160 170~180 10 PS+ash, resin 120 0 17 140 160—170 10 PE+ash, resin 95 0 11 102 110 10 Optimum Solidifying agent/content (ash , content ratio (weight) resin) = 1/0. 3 3- Physico-Chemical Characteristic Solidifiee th f o s d Wastes 3-1 Appearanc densitd ean y e experimenTh t shows that r resino regardles h ,as providef o s d thermoplastic solidificatio s i carrien t witou d h above process parameters e solidifieth , d waste appearanc f compacteo , solid surfacd an , e crackfre s obtainedwa e . 262 The density of the solidifying agent, content, and solidified wastes is shown in Table 2. densite Th Tabl f solidifyin yo - 2 e g agent, content solidified an , d wastes Pure material Solidified wastes density (g/cm3) density (g/cm3) + 30% resin h as % 30 + PE 0.91 1.02 0.87 PS 1.04 0-98 1.02 PVC 1.44 1.31 1;38 Catio1 3 . 1 n resin Ash 0.86 3. 2 Compressive strength and tensile strength 120 specimens fro same mth e experimene batcth f ho t were sen Shanghao t i Institutf eo Architecture to measure compressive strength and tensile strength. The results are show TablTabln d ni an . e3 e4 compressive Th Tabl . 3 e e strengt f plasticho solidified san d wastes Solidifying Solidified Solidified waste agent waste y G afte 6 10 r accumulated (Mpa) (Mpa) (Mpa) PE 14.5 h as 0 3 P E+ 17.4 19.6 PE + 30 resin 13.4 17.8 PS 80.0 PS + 30% ash 59.0 47. 1 PS + 30% resin 11.8 PVC 66.5 PVC + 30% ash 75-3 76.2 PVC + 30% resin 35.9 24-7 Table 4. The tensile strength of solidified wastes Solidifying 30% ash 30 % resin agent solidified waste solidified waste (Mpa) (Mpa) PE 8-6 5-3 PS 23.4 12.2 PVC 31-5 14. 1 seee b nn froca t mI tha4 e compressivTable d th t an 3 s d tensilan e e strengte th f ho incorporating ash is higher than the incorporating resin, mainly because the cation resin and solidifying agent posses a poos r compatibility betteA . r mechanical strengtf o h solidified wastes is favorable for final storage. Exposed by a higher dose rate of 7 radiation, solidified wastes would sho obviouo wn s change compressivn si e strengtht I . means that thermoplastic solidified wastes have suitable radiation stability. PVC 263 solidified waste has a best compressive and tensile strength. Moreover, the mechanical solidifiee strengtth l al f dho waste meen e storagca s th t e requirements. 33 . Shock strength Whe e specimennth e solidifieth f o s d waster e Tabl . e5 Oxygen inde flamind xan g phenomen f solidifieao d wastes OI Flaming phenomena Room Specimen (0, temperature vol. %) Melting Dropping Crimping Charring Smoking Smell CO PE+ash 21-0 Y Y N N Y Y 20±5 resi+ nPE 21-0 Y Y N N Y Y 20±5 PS+ash 21-0 Y N N N Y Y 20±5 PS + resin 21-0 Y N N N Y Y 20±5 PVC+ash 50.2 N N N N Y Y 20±5 PV Cresi+ n 46.0 N N N N Y Y 20±5 Oxygen index means minimum O2 concentration needed to keep buring in a balanced state when plastic specime measurine th n i s ni g apparatu floe th w d rats an f mixeeo 2 dO with N2 is 4 + 1 cm/sec. Table 5 shows that the PVC solidified waste contains higher OI than others while PC and PS solidified wastes melt when burned. It demonstrates that PVC solidified waste ha bettea s r flame resistanc solidifieS P d e an tha dE nP wastes . 35 - Penetration Specime r cros fo s soake t e penetrans nwa dayth 0 ou sectio thed t n d12 i an s ncu r nfo t and longitudinal section e deptTh .f penetratio ho s measurednwa . •v e incorporatin e Th solidifieth f o E P h do penetration n as d gwast an d ha eC PV . incorporating resin had depth of 0. 2 mm. PS incorporating resin had depth of 0. 5 mm. e resultTh s sho wsolidifiee th tha l al t d waste gooa d dsha penetration resistance. 6 Leachin. 3 g resistance methoAccordinO IS leachine e dth th o gt g resistanc usind s determinean ewa gC " 0 7 t da deionized wate r leachinfo r weeks4 g1 . 264 Incineration ash contained isotopes of 147Pm, ]37Cs, 13T, 12Ï,32 P, etc. In this test ''"Pm was used as tracers. Specific activity of U7Pm was measured by liquid scintillation counter e leachinTh . g result e showsar Tabln ni - e6 Tabl - e6 Leaching rate percentagd san f accumulateeo d leaching Solidified Leaching rate Accumulated leaching waste cm/d % PE + resin . 14X10-3 8 8.77 PE + ash 4.28X10-8 7.27 PS + resin 1.42X10-8 2.39 h as P S+ 0. 82X10'8 1.32 PVC + resin 7. 10X10'8 5-64 PVC + ash . 10X10-7 8 3-80 37 - Radiation resistance Specimen forn i s f 25X10X1mo barm 0m s were sen Radiatioo t t n Cente f Shanghao r i Institut f Nuclearo e cobalt-6A . 0 gamm an externa s a usecel s wa a ld l irradiation source. The radiation dose received by specimen was 106 Gy (dose rate was 5. 9 KGy/ h). After irradiation e dimensionath , l change r o structuras l degradation froe th m macroscopic poine obviou th f vie o td wan s change colourn i s , compressive tensild an , e strengt he specimens th wer t l founal eno n i d . 38 - Weathering resistance Weathering resistanc s evaluatewa e3 differen y b d t tests e freeze-thath : w test, long-time soakin deionizen gi d water weatherind an , g test. The freeze-thaw test Specimen sizn si f Long-time soaking Specimens were soake n deionizei d d wate t rooa r m temperatur days0 e 12 Th r . fo e results of the test showed that all the sollidified wastes had no changes in volume and weight except expansio weighd nan t increasin incorporatinS P n gi g resin. Weathering test Specimens were kept outsid e flath et rood sufferean f n overala d l n lighactiosu f t no ultravialet ray, temperature, wind, rain, oxygen, ozone, etc. for 375 days (126 days of sunny day2 f 15 cloudy,o s day6 7 ,f rain o s day1 2 ,f acio s d foggyday8 f d o s an , 265 rainstorm). The results of the test showed that the stripping or cracking were not found in the solidified wastes except the PS incorporating resin. There were no changes in volume, weight, and colour except a slight colourfade in the PE incorporating resin. 4. Conclusion 41 . Thermoplastic Solidification Proces s beesha n develope incorporato dt e city solid radwastes, including incineration ash, spent radioactive resin d contaminatean , d plastics. It is simple, and feasible and requires low investment. e characterizatioTh 42 . solidifiee th f no d wastes shows tha incorporatinS P t s i h gas characterize y goob d d radiation resistance (10 , ) goo y 6G d mechanical strength (compressive strength 59 MPa),good leaching resistance (8- 2 X 10~9cm/d at 70"C), d volum an n optimua s ei reductioS mP solidifyin• ) 6 n7 . rati2 ( go agenr fo t incorporating incineration ash and PVC and PE have the similar advantages when incorporating resin. 43 . 11,000 piece f radioimmunassao s y kitr yea consumee pe s ar r Shanghain di . Total 3 solivolumm d7 f wasteo e s deliverei s e thermoplastia yearth n i dn I . c solidification process, the spent PS tubes can be used as solidifying agent in incorporating reduco incineratiot s a eo s storag h nas e volum f radwasteeo storagd san e cost. 266 WASTE DISPOSAL AND SAFETY ASSESSMENT ROCK CHARACTERIZATIO SITN I E SELECTION A.E. OSMANLIOGLU Çekmece Nuclear Research and Training Centre, Turkish Atomic Energy Authority, Istanbul, Turkey Abstract General information about the waste management activities hi Turkey is presented. Recent site selection studiefuture th r es fo necessitie Turkef so mentionede yar . Preliminary studies and programmes of rock characterization in site selection process are described. Candidate host rock formations and sampling points are shown in figures.Initially, several geomechanical test e appliear s granitn o d e sample laboratoryn i s . Then test resulte ar s evaluated. 1. INTRODUCTION In Turkey, only low level radioactive wastes are produced by industry, hospitals and research laboratories thir sFo . reason, detailed disposal site selection studies hav t beeeno n started yet. High level radioactive wastes should be permanently isolated from the environment and remained saf verr efo y long periods mineA . d geologic metho disposaf do bees lha n considered to be a preferred solution. Ground control is one of the major problems in deep repositories due to the nature of bearing strata. Geomechanical properties of host rocks play an important role in an underground repository stability [1]. This study will comprise preliminary investigations of the site selection studies in Turkey. The aim of this study is to collect informatio propertien o n f candidato s e host rocks before detailed investigation sitn o se selectio carriee futuren ni ar t thin dI ou . s study, candidate host rock formations whice har available in our country are generally investigated with a particular emphasis on granite rock. 2. WASTE MANAGEMENT IN TURKEY Recent activitie wastf so e managemen Turkei briefle h t b n yca describe followss da : (a) Sealed sources: conditione cemen n dmiddli e th stee a n f i te o l drum wasteA (bRI ) s produce y hospitalb d d laboratoriesan s . This typ f wasteo e s include especially plastic tube injectorsd san . Compacted withi drume compactona th y sb r rfo reduction of their volumes. (c) Liquid wastes produced by nuclear research centres. Chemical precipitation is applied to these liquid wastes in a waste treatment plant. Precipitated sludge of the liquid waste is mixed with cement in the drum. The upper part of the drum is covered with pure cement composite. All these drum takee sar n intr storagoou e building nearb waste yth e treatment plann i t Çekmece Nuclear Research and Training Centre, 269 . PRELIMINAR3 Y SITE SELECTION STUDIES Selection of a suitable repository site is one of the most important aspects of radioactive waste disposal. In this study, the candidate host rock formations which are available in our country were generally investigated. Basic host rock formations are shown in Fig. 1. General tectonic conditions of our country can be described according to the basic fault zones and previous earthquakes. Our country can be separated into five earthquake levels. Especially North Anatolia fault is more effective in this separation. These earthquake level zone showe sar Fign i . 2 . MEDITERRANEAN 6RANITE FORMATIONS ED TUFFS FIG. Candidate1. host rock formations FIG. 2 Earthquake level zones 270 . ROC4 K CHARACTERIZATION It should be noted that the repository is presently planned to have an operational lifetime of twenty-six to thirty-five years, and to provide an option for retrievability for fifty years after waste emplacement. Therefore, a variable which has a major impact on host rock performance is time. A second variable anticipated to have a significant impact on nuclear waste emplacement is the effect of high temperatures on rock (100-400°C). A third variable which the repository will experience, t encounterewhicno s i h conventionan i d l underground excavations radiatioe th s i , n [2]. r thisFo reason, numerous rock characterization test plannee beginninsar e th n di thif go s study. Some of these tests have been completed. But some of them have not completed yet. Planned tests are shown in Table I. TABL . EROCI K CHARACTERIZATION TESTS Index Tests Georaechanical Tests Special Tests Specific gravity Triaxial compressive Thermodynamic properties Water contents Uniaxtal compressive Radionuclide migration Porosity Tensile Strg. Permeability Heat transfer Rock characterization test e initiallar s y applie granitesn do . First, numerous block sample takee sar n from north-east regio Turkeyf no . These sample preparee sar inder dfo d xan geomechanical tests. Core specimens are taken from these block samples. Several geomechanical tests are applied on these core specimens. Results of these tests are shown in Table II. 271 TABLE H. ROCK CHARACTERIZATION TESTS Planned Completed Sample Test Test Initial PROPERTIES No Number Number Results Density , (g/cm3 ) 20 200 2OO 2.364-2.652 Water Content) ,(% 20 200 2OO O. 173-O.678 Porosity) ,(% 2O 2OO 200 O. 385-1. 868 Uni .Comp. Strg. (MPa) 1O 1OO 30 120-324 Tensile Strg. (MPa) 10 1OO 35 25-110 Poisson Coefficient 1O 30 1O O. 26-08 .3 Elasticity Coefficient 1O 3O 1O 265000-44OOOO Internal Friction Angle (*) 6 12 6 34-52 Cohesion, (Kg/m2 ) 6 12 6 65OO-98OO Strg. Under ther loadm. s 1O 30 Thermal Conductivity 6 6 Radionucl ide Migration 1O 10 . CONCLUSIO5 N Geomechanical properties of the north-east granite rocks are seemed to be convenient for detailed investigations accordin roce th ko g t test s which have been datee donth o .et Aftee th r completion of the planned tests, the north-east granites will be classified in accordance with the existing classification system. REFERENCES [1] OSMANLIOÖLU, A.E., "Stability of Rock Pillars in Underground Waste Repositories", SPECTRU 4 InternationaM'9 l Topical Meetin Nuclean o g Hazardoud an r s Waste Management, Atlanta (1994)A US , . ] BIENTAWSKI[2 , StratT. . aZ , Contro n Minerai l l Engineering . BalkemaA . A , , Netherlands, (1987) 183. 272 COMPETITIVE ADSORPTION OF ^Sr ON SOIL SEDIMENTS, PURE CLAY PHASE FELDSPAD SAN R MINERALS S. H. SAKUMA, S. AHMAD Malaysian Institute for Nuclear Technology Research, Bangi, Malaysia Abstract Laboratory batch experiments were conducted to determine the adsorption of ^Sr by a soil sediment, mineralogically pure clay phases (vermiculites, smectites and illites) and feldspar minerals (adesine, albite, microclin d oligoclasean e a functio s a ) f ionio n c composition. The clay minerals were present at different proportion in the soil sediment. The important adsorbing adsorptio e phaseth d san n mechanism(s determinee b n )ca d fro stude mth - ies. Twent stoco ytw k solutions were prepared with concentration majoe th f sro cation, sCa werd an 0.0031 eo t a varieN 0 M 0. d 0.0016 o 2 t d g0.0031an o t fro0 an 0 , 0. m, 50. M 2M , respectivelyM . S5Sr trace uses spiko drwa t higs solutione it eth h o t specifie sdu c activitd yan short-life experimente Th . s yielded adsorption coefficient value thaj sK t coul describee db d by equations using samples from the sediment, pure clay minerals and feldspar minerals. Theoretical slope purr valufo e1 e- ion - exchange mechanis strontiumf o m adsorption ont- oCa saturated cladescribeds ywa slopee Th . s obtaine experimente th n di s represente averagn da e of adsorptio severan no l different mineral surfaces having different relative affinitier fo s strontium, calciu magnesiumd man . Experiment results showed that strontiu adsorbes mwa d to ion-exchange site thad san t calciu magnesiud man m cations were effective competitorr sfo these sites. Vermicultes, smectite illited san s clay minerals yielded adsorption coefficients that could be described by equations slopes-1.0 similar to the theoretical value. The feldspar minerals yielded slope ranges from -0.72 to -1.13, and the sediments slope value of -0.81. These suggest that ion-exchange was the dominant adsorption mechanism for strontium. Slopes make othef so r tha sugges0 n1. t that other mechanis operativee b y mma . Distribution coefficien valued K t s calculated fro experimente mth s would mak t possiblei accuratelo et y predict future concentrations of ^Sr in groundwater from sediments. This can be done by estimating the distribution of ^Sr in the area sediments, current ^Sr concentrations and major concentrationn io s data fro areae mth . 1. INTRODUCTION e nee r propeTh fo d r disposa f radioactivo l e wastes containin s causeha r d^S g considerable interest in its adsorption behaviour on minerals of the type found in and around the various types of disposal sites. Adsorption studies are needed to estimate the rate of evene th f groundwaten o ttranspori r ^S f o t r penetration intthrougd oan hdisposaa l sites which may contaminate drinking water. resulte mean y adsorptio e b batcstudTh a a th f f r sn o o h y o ^S techniqu f no a r efo number of clay minerals in solutions of sodium salts can be approximated by ideal ion- exchange equations [1]. It was found that distribution coefficient values at very high salt concentration were very low. This caused migratio nr relativ rate^S f so wateo et r flow, through geologic formations whose adsorption behaviour was dominated by clay minerals, likele higb t higha o t y h salt concentrations thin I s. system e mass-actioth , n equilibria equations can adequately describe the Sr-90 adsorption reactions [1-2]. 273 Previous work [3-4] on the measurements of Kd for mineralogically pure segregates indicated that vermiculites was an effective adsorbent phaseSr . Relationships between K and dSr exchangeable calciu sedimentmn o competind san g cation concentration, indicates electrostatic forces primarily contro sorptioe th l f strontiumno . Besides clay minerals, hydrous metal oxides, primary aluminosilicate organid san c matter effective ar , e absorbents s wa [5-7] r ^S . strongly , suggestincorrelateMn d an l gd A witspecifi , hFe c adsorptio thesy nb e metal oxides [8-9]. In addition to adsorption mechanism, strontium may be retarded during groundwater transpor precipitatioy tb n processes, eithe SrCO s coprecipitatioa ry b r o 3 n with CaCO3 [10]. purpose Th thif eo s stud gaio t s nyi better understandin adsorptioe th f go n propertief so strontium onto soil sediments, mineralogically pure clafeldspad yan r mineralw ho d an s change groundwaten si r chemistr affecn yca t these properties maie Th n. concerns were eth competitive effects of calcium and magnesium cations on strontium adsorption. The results and Kd values calculated with equations determined will make it possible to predict future concentration groundwaten i r ^S f o st disposaa r l site. This accomplishewa s ) (a y db identifying the mineralogical composition and properties of the soil sediments, (b) determining the distribution coefficient of ^Sr between solid and aqueous phases at different ionic composition, (c) statistical analyses on the experimental data, (d) forming distribution coefficien equatioj K t eacr nfo h mineral soid slan sediments ) identifyin(e d an , pure gth e clay minerals and feldspars minerals adsorbents associated which effectively adsorbed the ^Sr in the soil sediments. . THEOR2 Y Mechanism of Strontium Adsorption by Ion-Exchange Interpretation of the K^ results in terms of an ion-exchange mechanism can be explained as follows. The thermodynamically rigorous mass-action equilibrium expression for a binary cation- exchange reaction, such as strontium adsorbed onto a Ca-saturated clay is aSrb+ + b(CaX) - a(SrX) + bCaa+ (1) where a is the valence of calcium ion, trace th es i componen r S strontiumf o t , valence th s i strontiuf eo b m ion, calciue th s i m componena C binarn i t y system, and X is the solid adsorbent of soil sediments. equilibriue Th m expressee constantb n ca : , dK as , K = ([SrX]a[Caa+]b)/([CaX]b[Srb+]a) (2) where brackets , indicat] [ , e thermodynamic activities assumee on f I .s thaexchange th t e capacitye solith f do , adsorbenC , constans i t t (equivalen r unipe tt weight s i thad r )an S t w tracpresenlo e t a concentrationt , the e concentrationth trace th ef n o constituents adsorbed, (SrX), is much smaller than C, and the concentration of calcium ions adsorbed on 274 exchange sites, (CaX), is approximately equal to C/a in terms of moles per unit of weight, because C = (SrX)b + (CaX)a. The distribution coefficient can be represented by: b Kd = (SrX)/(Sr +) (3) where (Sr""solutioe th "s )i n concentratio trace th ef n o constituen equilibriut ta m wit solide hth . 1 By substituting the relationship: [A] = y{A}.(A) (4) where [A] is the actvity (moles), {A} is the activity coefficient, concentratioe th s i an) d(A n (moles). Equation (2) can be written as: a a+ b b K = [(Kd) (M ) /(C/a) j. ö (5) activit e ratie wherth th f oo s i y e 3 coefficients : {SrX}[ ) = (6 3 a {Caa+)b/[ (CaX}b {Srb+}a] For ideal ion exchange of the Sr constituent in which the exchange capacity C is b constant, the ratio of activity coefficients for the adsorbed ions, {SrX}/[a {CaX}, is constant. For low ionic strength solutions, the ratio {Ca } / (Sr " "} is also constant. The 3 becomes a+ b 13 1 3 constant. Using these condition assumptiond san logarithmia d san c transfor equatiof mo , n5 the dependence of distribution Kd of the Sr constituent on the calcium ion concentration, reduces to -b/a, the ratio of Sr ion charge to the calcium ion charge. Therefore, for ion- exchange of Sr2* for Ca2+, -b/a is -1.0. 3. MATERIALS AND METHODS A serie f batco s h adsorption experiment conductes swa determino dt strontiue eth m adsorption properties of soil sediments and the individual minerals composing these sediments as a function of the equilibrating solution composition [11-14]. Samples from the soil sediments were analyze X-ray db y diffraction (XRD) metho identifo dt e individuath y l minerals which were know effective b o nt e adsorbent phases. Experiments were conducted usin soie gth l sediments, mineralogically pure clays suc smectitess ha , illite vermiculited san s that compose majoe dth r portion sedimentse th f so severad an , l feldspar minerals. Prioo rt these experiments, supercentrifuge equipment was used to segregate clay minerals from the soil sediments by particle density effects. The experiments were not able to segregate completel e clath y l mineralsyal s decide finalld e mineralogicallwa an us , t i yo dt y pure minerals in the experiments. Twenty two stock solutions of 500 ml were prepared with the concentrations of the major cations Ca, Mg and Na varied from 0.0 to 3.12 x lu* M, 0.0 to 1.65 x 10'3 M, and 0.0 to 3.1 10'x , 2respectively 3M compositione Th . stoce th f ko s solution listes i s Tabln di . eI 275 Experiments 15 through 17 all have the same initial calcium and magnesium concentrations. The solutions were spiked with 85Sr tracer and the pH of each stock solution adjusted to 8.2 ±0.1 with sodium hydroxid hydrochlorir eo c acid. TABLE I. EXPERIMENT NUMBER AND SOLUTION COMPOSITION (MOLAR CONCENTRATION) Experim. Ca(NO3)2 4H2O Mg(NO3)2 6H2O NaCl Final pH 1 0.0 0.0 0.0 8.230 2 0.000624 0.0 0.0 8.200 3 0.0 0.000165 0.0 8.207 4 0.000624 0.000165 0.0 8.210 5 0.0 0.0 0.000312 8.170 6 0.000624 0.0 0.000312 8.215 7 0.0 0.000165 0.000312 8.229 8 0.000624 0.000165 0.000312 8.226 9 0.0 0.0000823 0.000156 8.155 10 0.000624 0.0000823 0.000156 8.208 11 0.003120 0.0 0.001560 8.194 12 0.003120 0.001650 0.001560 8.186 13 0.003120 0.000823 0.0 8.180 14 0.003120 0.000823 0.003120 8.234 15 0.003120 0.000823 0.001560 8.195 16 0.003120 0.000823 0.001560 8.200 17 0.003120 0.000823 0.001560 8.197 18 0.001560 0.0 0.0 8.210 19 0.001560 0.000823 0.001560 8.208 20 0.001560 0.001650 0.003120 8.160 21 0.001560 0.001650 0.0 8.200 22 0.001560 0.0 0.003120 8.239 Each batch experiment was conducted in a 40 ml polycarbonate centrifuge tube. One gram sediment feldspad san r minerals gra1 0. md claan , y minerals were precisely weighed and added into their respective tubes. Twent spikef o l ym d solutio thes nnwa added into each tube. All the tubes were equilibrated for 14 days shaken at room temperature. After the equilibration period, 1.0 ml solution sample was collected with a 0.45 //m disposable filter, acidified by adding approximately 100 /A of concentrated hydrochloric acid or nitric acid (6 M or 12 M) and then analyzed foSr85 gamma. Solid sediments were removed from each tube by filtering throug h0.4a . filtethed 5^m an nr allowe dryo solie dt Th .d sediments were transferred to counting tubes for gamma analysis. 276 All the solutions and the sediments were collected at the bottom of counting tubes of the same geometry. Countings were conducte minimiz o samt e y th d edecada e th e o yerrot e rdu Sr of the strontium radionuclide. The distribution coefficient Kd was calculated by : Ksr=dpm/g (?) dpm/ml where dpm/g and dpm/ml were the activities expressed in disintegrations per minutes per gram of sediments and per milliliter of solution, respectively. Ionic strength which measures the total concentration of charge in a solution was calculated by: ) (8 [mjzS 5 0. 2 = I where nij is the molality or concentration (m) of the ith species of Zj charge [15]. The ionic strength (I) parameter was used to calculate activity coefficient of the solution. At higher concentrations < 0.5M, Davies equation was used to calculate activity coefficient Y which represented better experimental data than other equations in the literature [16-17]. Activity coefficient y was calculated by: ) (9 [(Vl)/(2 Zi 7 -1 l I n+V/I1 - Y 0.21- i)= ] Expression for activity was calculated by: Activity (moles concentratio= ) n (moles) x activity coefficient Y (10) Detail statistical regression and variance analyses were done for each batch experiment to yield values for the adsorption coefficient Kd. . RESULT4 DISCUSSIOD SAN N TwentvalueK/ o sytw were obtaine eacr dfo h batch experiment adsorption so Sr(IIf no ) r on the soil sediments, the minerallogically pure clay (vermiculites, smectites and illites) and feldspar minerals (adesine, oligoclase, albite and microcline) at different solution composition. For experiments 15 through 17 which have the same initial calcium and magnesium concentrations resultd K e sth , were ver batc8 ye hsimilath l experimentsal n i r t I . can be deduced that the experimental results produced were accurate and can be used with Sr confidenc produco et Ke deth equation eacr sfo h minerals. K/' values for pure clay minerals showed the highest values compared to soil sediments and the feldspar minerals. Average Kd values for smectites and microcline mineral showed to have the highest and smallest values, respectively. The average soil sediments Kd values were relatively higthid han s indicated that some individual minerals componene th n i t sediment selectively adsorbed strontium differene e abilitth Th f .o y t mineral adsoro t s b strontium varied considerably, but the most reactive phases were smectites, vermiculites and illites hige Th h. proportio clae th y f nmineralo sedimente th n si s further gav hige risth ho et adsorptio strontiumf no abilite Th .feldspaf yo r mineral sedimente th n si adsoro t s b strontium were much less compare clae th y o dmineralst . 277 Figur Figuro t 1 e , shoe8 wequilibriue th Kf o m value msu functioa e s sa th f no dSr calciu magnesiud man m concentration solutio n si soie th l r sedimentnfo mineralsd san . Table II lists the adsorption coefficient Kj for soil sediments and minerals described by the equations. Sediments Adsorption Smectites Adsorption _ 2 T 4 . .- oc • • g 3.5 i '+5 O Q. J o S o 2.5 m en 0 1 T3 2 • 1 '-, A 1.5 X 5 0.5 S 1 U> en 0.5 ° 0 ^ 0 -4.5 -4 3 - -3.-2.5 5 -4.5 -4 -3.5 -3 -2.5 log (Ca +Mg) log (Ca +Mg) 85 85 Figure l. SrKd values againsequlibriuf o m tsu m Figur. e2 SrK d values against sum of equilibrium CaandMg (moles/L) for sediments CaandMg (moles/L) for smectites Vermiculites Adsorption Illites Adsorption — 4 T — 4 T g 3.5 . g 3.5 . •f 3 . * • Ë. 3. • - 5 2. o •h o 2.5 . — M n •0 2 * ^^ •Mo o2 *• •A 1.5 - . 5 1. A 4^ 5 1 S 1 en 0.5 - en 0.5 . ^ 0 0 0 -4.5 -4 3 - -3.-2.5 5 -4.5 -4 -3.5 -3 -2.5 lo a g+Mg(C ) a +Mg(C g )lo 85 85 Figure 3. SrKd values against sum of equlibrium Figur. e4 SrK d values againsequilibriuf o m tsu m Ca and Mg (moles/L) for vermiculites CaandMg (moles/L illiter )fo s Albite Adsorption Oligoclase Adsorption T 2 . 1 _ T 2 . —1 .2 0-{ • -.| 0.8 . • Q. • S 0-4 1 0.4- ^3 •o •«• A 0 1* . 0 4S. S -0.4 •i m 5 -0.4. • • ^f^ CO en —^n^^ »k 2 -0.8- 9^ -2 -0.8 4 - -4.5 -3.5 -3 -2.5 -4.5 4 - -3.2 - 5 3 - 5 log (Ca +Mg) a +Mglog(C ) 85 85 Figure 5. Sr Kd values against sum of equlibrium Figure 6. Sr Kd values against sum of equilibrium g (moles/LM d Caan r albit)fo e (moles/Lg M d Ca an oligoclasr )fo e 278 Microcline Adsorption Andesine Adsorption 0 T o 1.2- .1 -0.2 \ 1 ° fa. -0.4 .. I. 0.8 • 8 -0.6 a o.e. S -0." 8- "S 0.4 . Ig v. -o 0.2 . • S » £ -1.4 5 -0.2 . -4.5 -4 -3.5 -3 -2.5 4 - -4.-3.5 3 - 5 -2.5 a +Mglog(C ) a ~+Mglog(C ) 85 Figure?. ^Sr Rvalues against sum of equlibrium Figure 8. Sr Kd values against sum of equilibrium g (moles/LM d Caan r microclin)fo e Ca and Mg (moles/L) for andesine TABL . ADSORPTIOEII N COEFFICIENT K EQUATION EACR SFO H MINERALS DSr AND SOIL SEDIMENTS Minerals Adsorption Coefficient Equatiod K g Lo n R sediments 8 1 1 .35562 1 1 - . -0.8074 ) 0 7± Mg + 50.0396a ± (C g 9lo 0.96 smectites 0.6774- ) -1.0095Mg 00.0960+ ± a (C 4 3±0.0g lo 1 313 0.98 vermiculites 0.6677- -0.9770) Mg 0.09902+ ± 0.03245a ± (C 1g 3lo 0.98 illites -0.99509 ± 0.02521 log (Ca + Mg) - 1 .04392 ± 0.07695 0.99 albite 3.3846- -1.0470) Mg 0.06966+ ± 30.0228a ± (C 0g 0lo 0.99 oligoclase 3.5085- -1 ) . 1343Mg 30.0894+ ± 00.0292a ± (C 3g 9lo 0.99 microcline 3.0220- -0.7389) Mg 0.16897+ ± 0.05536a ± (C 9g 5lo 0.91 adesine -0.72563 ± 0.02989 log (Ca + Mg) - 1.96493 ± 0.09125 0.97 r purFo e ion-exchang singla n eo e typ sitef theoreticaee o th , l slope woul -1.00e db . Figure Figuro t 1 s shoe8 w that very good linear correlation exists. Slopes obtainer dfo smectites, vermiculites and illites ranged from -0.98 to -1.01 suggesting that ion-exchange was the dominant adsorption mechanis strontiumr mfo . Slope oligoclasr sfo albitd ean e minerals were -1.13 and -1.04, respectively, which also indicated that ion-exchange was the dominant adsorption mechanism. However, slope for microcline and adesine -0.74 and -0.73, respectively, indicated that ion-exchange was the dominant adsorption but another mechanism may also be operative. The soil sediments slope -0.81 was close to the theoretical value -1.00, which suggested that ion-exchang dominane th s ewa t mechanis r strontiummfo , however mineral component sedimente th n i s s which reacte anotheo dt r operative mechanisy mma exist. The clay mineral phases having higher cation exchange capacities and dominant of ion- exchange mechanism for strontium contributed to the high Kd in the sediments. For feldspar minerals, eventhough ion-exchange mechanism was dominant, another mechanism may be operative. K values indicated that strontium adsorbed was small compared to the clay minerals. Desorption d ^Sr experiments [8] indicated a very strong correlation exists majoritA . 80%~ Mn ( y d betweeextractabld f adsorbes )an o an , wa r Fe r n^S , d ^S eAl exchangeably adsorbed ,e remainde mosth f o ts apparentl wa r y specifically adsorbey b d hydrous meta nonexchangeablels oxidewa d san . From Ref. [6] catioe ,th n exchange reactions 279 f feldsparo strond ha s g specific effects catioe Th .n exchange capacitie f feldsparso s were highly variable, dependin nature displacine th th f n ego o released gan d cations. Certain cations have a very strong tendency to be fixed. In microcline which contains both Na and K, Na was preferentially liberated by water and acid over K. In albite and oligoclase, Ca was preferentially exchange dstatee b ove . Thusn drNa ca that i , t strontiu adsorbes mwa catioy db n exchange with Ca or Na on the feldspars surfaces. Experimental studies on cation exchange reactions of feldspar surfaces were relatively few compared to similar studies on clay minerals. 5. CONCLUSION r A metho s beedha n developed which calculate KvaluS e fro equationn ma s i t I . possibl accuratelo et y predict future concentration groundwaten i r ^S f so sitea t a r. d Thin sca be done by estimating the distribution of ^Sr in the area soil sediments with the equation, curren r concentration^S t majod datn an s io ra fro e varioumth s area wells. Once th e distribution of ^Sr in the sediment is calculated, future ^Sr concentrations in a nearby aquifer coul e calculateb d y combininb d e equatioth g n with future majo n concentrationsio r . However mineraloge th , aquifee th f yo r material shoul t variedno d significantly froe mth sedimentary material that was used in the experiments or otherwise the Kd values will varied significantly. This could result in large errors in future ^Sr concentrations within the site wher sedimente eth s were sampled. From this stud betteya r understandin adsorptioe th f go n propertie strontiuf so m onte oth soil sediments, mineralogically pure clay mineral feldspad san r mineralchangew ho d n si san groundwater chemistry can affect these properties. This was shown by the competitive effects of calcium and magnesium cations on strontium adsorption. The results also showed that mass action equilibrium laws adequately predicte behavioue dth strontiuf ro solutiomn i contacn i t wit mineralse hth . Adsorption behaviou f clao r y mineral frequentls swa y relate higho dt , relatively constant ion-exchange capacity competine Th . g cation concentration calciuf so d man Sr magnesiu msolutionn i s correlate linearly wit measuree hth d Kd values. Acknowledgement authoe Th . CantrelsgratefuJ - s i r . K o t l from Battelle, U.S.Ar fo . his comments and discussion, the IAEA for its financial support and the management of MINT, Bangi, Malaysia. REFERENCES [1] RAFFERTY, P., SHIAO, S.Y., BINZI, C.M., MEYERS, R.E. ADSORPTION OF SR(n CLAN >O Y MINERALS: EFFECT SALF SO T CONCENTRATION, LOADING AND PH, Inorganic Nuclear Chemistry 43 (1981). [2] SHIAO, S.Y., EGOZY, Y., MEYER, R.E., ADSORPTION OF CS(I), SR(II), EU(ÏÏI), CO(II) AND CD(II) BY ALA, Inorganic Nuclear Chemistry 43 (1981) 3309. [3] PATTERSON, R.J., SPOEL , LaboratorT. , y measurement strontiue th f so m distribution coefficient kd for sediments from a shallow sand aquifer, Water Resources Research 17 (1981) 513. [4] HIGGO, J.J.W., Review of Sorption Data Applicable to the Geological : SafetEnvironmentUK ye th n i W Interesf sDeeo e LL th d r p an fo tDisposa W IL f o l Studies, Nirex Radioactive Waste Disposal, NSS/R162, British Geological Survey, Nottingham (1988). 280 [5] BRADY, N.C., The Nature and Properties of Soils, Macmillan Publishing Company, New York (1990). ] [6 Mineral Soin si l Environments, Soil Science Societ Americaf yo , Madison, Wisconsin (1977). ] JUO[7 , A.S.R., BARBER, S.A. retentioe Th , strontiuf no soilmy b influence s sa y db pH, organic matte saturatiod ran n cations, Soil Scienc (19709 e10 ) 143. [8] JACKSON ,INCH R.ED AN . , K.J., Partitionin strontium-9f go 0 among aqueoud san mineral species in a contaminated aquifer, Environmental Science Technology 17 (1983)231. ] JACKSON[9 , R.E t al.e . , "Adsorptio radionuclidef no fluviaa n si l sand aquifer: measuremen distributioe th f o t n coefficien (strontiumd K t (cesiumd K d d )an )an identification of mineral adsorbents", Contaminants and Sediments, Vol. 1, Ann Arbor Science Publisher, Ann Arbor, MI (1980) 311. [10] HALEVY , TZURE. , , SoiY. ,l Science (1964. )98 [11] Batch-Type Procedures for Estimating Soil Adsorption of Chemicals, Technical Resource Document EPA/530/SW-87/006-F, United States Environmental Protection Agency, Washington D.C. (1992). [12] STRICKERT, R., FRIEDMAN, A.M., FRIED, S., The sorption of technetium and iodine radioisotope variouy sb s minerals, Nuclear Technolog (19809 y4 ) 253. [13] MAHONEY, J.J., LANGMUIR, D., Adsorption of strontium on kaolinite, illite and montmorillonite at high ionic strength, Radiochimica Acta 554 (1991) 139. [14] TAMURA, T., STRUXNESS, E.G., Reactions affecting strontium removal from radioactive wastes, Health Physic (1963s9 ) 318. [15] STUMN , MORGANW. , , J.J., Aquatic Chemistry Introduction A , n Emphasizing Chemical Equilibria in Natural Waters, John Wiley and Sons, New York (1980). [16] NORDSTRÖM, O.K., MUNOZ, J.L., Geochemical Thermodynamicse Th , Benjamin/Cummings Publishing Company, California (1985). [17] KRAUSKOPF, K.B., Introductio Geochemistryo nt , McGraw-Hill International Series in the Earth and Planetary Sciences, McGraw- Hill Book Company, New York (1979). 281 ENVIRONMENTAL IMPACD AN TW STUDLO R YFO INTERMEDIATE LEVEL RADIOACTIVE WASTE DISPOSAL WANG ZfflMING China Institute for Radiation Protection, Taiyuan, Shanxi, China Abstract The work on disposal of low and intermedjate level radioactive wastes (LLW and ILW) has been already started-and pre- generatestagW IL ed wordan disposaf froko W mLL nucleaf o l r power plant is now proceeding in China. Corresponding assessments have been conducted and safety assessment methodology for disposal of s beeha n systematicallW IL d an W yLL studied aimin t ga evaluatin g their impact environmentn o s . Some aspects which need solvinn i g safety assessment are put forward in this paper. To make the assessment results meee actuath t l situation e followinth , g four problems should be solved: 1) make the inventory of radionuciides to be disposed clear and put stress on key radionuciides; 2) accurately determine the release rate of radionuciides from a repository, particularl relatioe yth n between release arrangemenratsizd d ean an t of waste forms as well as water content in the repository; 3) find out radionuclide migration behavio e geologicath n i r l medium, especially relation between retardation coefficient and velocity of unsaturated wate rstandardizatio) 4 flow d an ; f modelsno . essentiae Th l purpos f disposaeo f radioactivo l e waste isolato t s si e them from human environmen mako t e sure thasubsequeny an t t retur huma e f thenth o mo t n environment wilresult lno undun ti e radiation exposur necessare mano et Th . y isolation degree depends on the actual radionuclide content and properties of wastes to be disposed under consideration. And the practical isolation capacity depends on performance of the whole disposal system. Similar to the other nuclear activities, safety assessment is necessary in disposal of radioactive wastes objectivs It . analyzo t s ei e expected performanc disposae th f eo l system relate o safetydt , especiall e possibilityth f returo y f radionuciideo n s released froma repositor humao yt n environmen comprao t d an résulte t eth e analyzed with acceptable judgo t criteri e acceptabilits a eth o s ae disposa th f yo l syste expected man d activities. Usually, safety assessment can be divided into two types, i.e., generic assessment and site- specific assessment. Generic assessmen usefue b makinr n fo ltca g programmatic decisions regardin choice g th disposa a f eo l concepappropriate th d availablf o tan e eus e resources sa well as in gaining recognition of the feasibility of a disposal concept. Site-specific assessment is necessary for decisions affecting siting, design, and licensing for construction, operation, shutdow sealind nan repositorya f go nowassessmento e t th p , U . s for disposal of LLW and ILW are mostly concentrated on the generic. 283 contexn I f safeto t y assessment, model computed san r codes have been developed dan employe n mani d y countries accordin o theit g r respective conditions. PRESTd Oan PRESTO-II codes wer . EnvironmentaS e . developeU e th y b dl Protection Agency (EPA)[l'2 calcubto t j e moisture movemen wels a treleases a l , transpor exposurd an t o et f radionuclideso n ma developes A . generir dfo c assessment, they have been actually used for such commercia levew llo l radioactive waste disposal site Barwellt sa , Beatt Wesd yan t Valley[2]. GEN n code was developed by the U. S. Nuclear Regulation Committee (NRC) to review and approve the license application and this code together with the other codes can be used to perform an independent assessment calculation for disposal of LLW PL BIOS code is used for assessment in Englandf4!. Japan Atomic Energy Research Institute (JAERI) used SAMSON-I-STA to make generic assessment^] and Central Research Institut f Electrieo c Power Industry (CRIEPI) use combinatioda f FORADOno , CORF, FEGM and FERM codes to make site-specific assessment of Rokkasho LLW Repository in Japan["3. Even though the codes have been used under various situations, such aspects need further improving especially as no consideration of the effect of unsaturated conditio radionuclidn no e migration when calculating their release froma repository o consideration ; e applicabilitth f o n f distributioo y n coefficient , whekj , n calculating radionuclide migratio geologicae th n i l medium consideratioo ;n effece th f nto of chemical speciation of and competition between radionuclides on release and migration necessars i t I . inquiro yon t o s f them o ed deeplan , e aspecte b effecte yth y th s a ma ss considerable. The work on disposal of LLW and EL W has been already started and pre-stage work of radioactive waste disposa r nucleafo l r powe rproceedinw planno s i t Chinan gi e Th . radioactive waste disposal activitie e conducteb o t s d have been evaluated accordino gt Chinese laws and regulations in order to evaluate the effects of radioactive waste disposal on environment. Recently a computer code PRESDSA used for safety assessment of whicn i disposaW h somLL f o le -improvements were trie s beedha n develope Chiny db a Institute for Radiation Protection (CIRP)[7>83. The following four problems should be solved to make the results of assessments meet the actual situation. 1. Make the Inventory Clear and Put Stress on Key Radionuclides e differencTh propertien i e f wasteo s contentd an s f radionuclideo s affecy e ma sth t choice and the decision of disposal options. To make the inventory clear is very important determinatioe inth f disposano lanalysi e optio effecth e th d environmentn f o ntan s o s i t I . the foundation not only for selection of disposal site and design of disposal option but also for safety assessment Becaus f differenceo sourcen ei f wasteso distinctiod san f collectionno , conditiond san contro r wastesfo l e radionuclidth , e compositio e d contenb e waste an nth y n i ma st considerably different!^ ^,9-15] Therefore, the properties, especially chemical character of waste s wela ss compositioa l contend nan f radionuclideso t f wasteo e dispose,b o t s d shoul e madb d e clear before implementin a ggive n disposal practicee licensth n I e. application of siting for disposal of wastes, analogical method can be used for providing rough dat f inventorao casn yi f lac eo f applicabl ko licensn ei t dataBu e .applicatio f no buildin operatind gan disposae gth l facility measuree th , d data shal givee b l conduco nt t design and assessment with a definite aim. The models and computer codes for shallow land disposal of wastes have been develope mann di y countrie estimato t s e their radiological consequencesU'^lMSd ]an use r somdfo e actual disposal sitess founi t I . d from many results that even thouge hth 284 contents of some radionuclides such as Co, Sr, -Cs and Cs are higher, the radiological effec publin o tothe e lowers cth i rd radionuclidean , s suc 14s hCa , ^Tc, 129j 22 and ^Ra ^ih lower content may be the main contributors to dose to man[5>MM61. The principal reason is that the radionuclides need transport in the environmental medium, including aerated zon saturated ean d zone, betwee .releasd n an exposur n e ma fro o met repository e periodTh .n general i , f transporo ,e mediuth n i s tmlongeri e Th . radionuclides, therefore, with shorter half-lif highed ean r sorption capacit t reacno r hn o yca can reach in a small quantity human environment. Only those with longer half-life and lower sorption capacit reacn yca h human environment questionA . , thereforee b y ma , raised: in conducting safety assessment research, which radionuclides should be selected? In fact, they should be selected according to the ratio between transport time from release to exposure to man and half-life of radionuclides. It is not necessary to study them in detail if the ratio is greater than ten. So the notable attention should be paid to migration behavior of such radionuclides with longer half-life and lower sorption capacity as ^4C, 99Tc, I29!, 226Ra, etc. rather than 60Co, 90Sr, 134Cs and 137Cs in safety assessment. 2. Accurately Determine the Release Rate Release rat radionuclidef eo s mose froth mrepositorf ta o importan e on s yi t parameters in evaluating environment impact. It is a complicated function which may be affected not onl y wastb y e forms themselve t alsy integritbu b so d retentioan y n capacite th f o y engineered barriers. Moreover, it also is affected by the external factors such as wind, flood and so on. Up to now, useful data used for assessment are obtained only in some througr o linkt i f so h some tests with smaller-scal shorter-timed ean release Th . e ratr efo an actual disposal is obtained only by calculation mathematically. The release pathway is principally through groundwater in case of ground disposal of wastes t presentA . e maith , n mechanism considere n simulatioi d f releasno e through groundwate leachins i r g whic comprehensiva s hi e effec f suco t h processe s diffusiona s , dissolution and surface rinse. Diffusion release, in general, is an essential process. The models developed and used for describing diffusion release related to specific surface of waste forms and diffusion coefficients in the waste forms and in the other engineered baniers[17'22]. They, however, did not consider the effect not only of waste form size but als watef oo r conten leachinn o tliteraturese th n gi t leadI .overestimat o st e obviously the release rates and hence release effects. In order to estimate reasonably the release rate, through the tests, it is found that the two factors mentioned above affect obviously release rate and that the difference is about 1-3 orders of magnitudel''15,23] ^ reasonable correction should be made when the results obtained from laboratory are used for mathematic models calculatin effece g th actua f o t l disposal because actual waste forme ar s larger tha solidifiee nth d waste forms use leachinr dfo g test wated san r content under real disposal is lower than those under leaching tests. For this reason, it is necessary to study deeply the effects of both solidified waste form size and water content on leaching to get leachine th g rate which correspond reae th l disposao st l situation. . Fin3 t Radionucliddou e Migration Behavio Geologicae th n ri l Medium Usually, equilibrium adsorption models in which distribution coefficient, k 4. Standardization of Models Six types of models are involved in a comprehensive assessment of disposal of LLW and ILW: source term release model, movement mode f wateo l r flow, transport modef o l radionuclide groundwatee th n si surfacd an r e water, transport mode f radionuclideo l n i s the atmosphere, transfer model of radionuclides in the food-chain and dose model, in which some submodel includede sar comprehensivA . e assessmen tforme e modeb n dca l by combination of models. At present, ther mane ear y assessment units engage assessmenn di t wor submitted kan d numbea f environmentao r l impact reports d (EIRsscope an th f disposaW n eo i ) LL f o l DLW in China. Some differences exist in choice of models and parameters in the EIRs. From long-tera m poin viewf o tunifie e th , d models shoul developede db , liksituatioe eth n U.S.A.n i competene th C y b ,NR d t an authoritieA byEP s through organizing experts with various specialitie standardizo t s a o ss e safety assessment helpfus comparisoi e t th I o .t l n of the results and review of the EIR. Moreover, parameters which can not be measured within a shorter-term should be collected and reviewed to provide the best parameter value variouo t s s type f assessmentso s . Some parameters f courseo , , e b whic n ca h obtained withi nshorter-tera m shoul determinee db measurementsy db . REFERENCES [1] LITTLE C. A., FIELD D. E., EMERSON C. J., and HIROMOTO G., Rep. ORNL/IM-7943, Oak Ridge Natl Lab., TN (1981). [2] FIELDS D. E., EMERSON C. J., CHESTER R. O., LITTLE C. A., and HIROMOT , RepOG. . ORNL-5970 Ridgk Oa , e Natl Lab. (1986)N T , . [3] KOZAK M. W., CHU M. S. Y., HARLAN C. P., MATTINGLY P. A., NUREG/CR-5453, SAND 89-2509, Vol.4, Sandia Natl Lab.(1989)M N , . [4] LAWASON G. and SMITH G. M., NRPB-R169, National Radiological Protection Board, Oxon (1984). ] MATSUZUR[5 al.t e , , ProceedingUH. 198e th 9 f so Join t International Waste Management Conference. Vol. L pp. 515-520 (1989). ] KAWANISff[6 , IGARASHM. l , MAHART. I , KOMADAY. d an , AH. MAK , WastY. I e Management'87, Vol.3 . 175-18pp , 0 (1988). ] CIR [7 JAERId Pan , Safety Assessment Methodolog Shallor yfo w Land Disposaf lo Low Level Radioactive Wastes (Final Report). Vol (1993).4 . WAN d , PRESDSAan M. , . J. GZ . Z Compute] U ZHO :A , WAN [8 G G. , . S. U H . G J r Code Used for Shallow Land Disposal of Low Level Radioactive Wastes, CIRP, TY (1993). 286 [9] PINNER A. V., HEMMING C. R., and HELL M. D., NRPB-R161, National Radiological Protection Board, Oxon (1984). , NRPB-R138 HIL] d D. PINNE . an [10 L, M V. . RA , National Radiological Protection Board, Oxon (1982). [11] SMITH G. M., FEARN H. S., SMITH K. R., DAVIS J. P., and KLOS R., NRPB-M148, National Radiological Protection Board, Oxon (1988). . AGENCS . U TOXI R ] YFO [12 C SUBSTANC DISEASD EAN E REGISTRY, PB90-141714 (1989). IAEA-CN-43/470, aL ] t TOSTe [13 , P. . EA , IAEA, Vienna (1984). [14] MEYER G. L., IAEA-207/64, IAEA, Vienna (1976). [15] WANG Z. M., and LI S. S., Guideline of Safety Assessment for Shallow Land Disposal of Low Level Radioactive Wastes, Atomic Energy Press, Beijing (1993). [16] MARTVOET J., and ZEEVAERT T., BLG 629, EUR 13042 EN, CEN/SCK, Mol (1990). [17] PESCATORE C., Improved Expressions for Modeling Diffusive Fractional Cumulative Release from Finite Size Waste Forms (1990). ] SULLIVA [18 SUEd , BNL-NUREG-43926an J. . , NC M. . NT , Brookhavan Natl (1990Y N , ) Lab. [19] MATSUZURU H., and SUZUKI A., Waste Management, Vol.9, pp.45-56 (1989). IS.H , SU ided man , quoJ. dM Ref.5KI , H. , . C O CH , , CHOS. L. . . C K I M KI ] [20 pp. 383-388 (1989). ] KEMP [21 , WastR. . FC e Management Vol.1, 1 88 . 549-56pp , 0 (1988). ] SUARE al.t [22 e , ideA. . mZA quod Ref.5 . 503-50pp , 8 (1989). [23] WANG Z. M., YANG Y. E., and KAMTYAMA H., idem quod Ref.7, Vol.5, No.28 (1993). [24] OGAWA H., et al, idem quod Ref.7, Vol.5, No.3 (1993). [25] MUKAI M., WANG Z. M., LI Z. T., and LI S. F., idem quod Ref.7, Vol.5, No.5 (1993). [26] WANG Z. M., et al., Analysis of Radionuclide Migration Behavior in Loess Medium, to be published in the MRS'94 (Kyoto). 287 A STUDY ON PROTECTIVE COVERS FOR LOW AND INTERMEDIATE LEVEL RADIOACTIVE WASTE DISPOSAL IN NEAR-SURFACE FAGOTEE S- CHINA' S EXPERIENCE . ZfflWENF U G . C , China Institute for Radiation Protection, Taiyùàn, Shanxi, China Abstract Coveimportann a s ri t protective barrie low-and-intermediatr fo r e level radioactive waste disposa near-surfacn i l e facility e performancTh . f coveo e s i r dependen sitn to e meteorological conditions, site characteristics, soil properties, cover structur othed ean r factors. Foreign Experiences show that performanc covef eo n rca not be effectively assessed and predicted only through laboratory tests and small scale field tests. China is planning to construct five regional disposal sites for low-and- intermediate level radioactive waste wound 2000. China started cover study in 1988, concept design of cover testing, systematic literature survey on cover testing, cover testing plan were sequentially carried out. and some apparatus were purchase by the end of 1991. However, due to funding shortage, the testing work has not been gone further any more since then. We think funding shortage is the issue that commonly exists in developing countries. The paper looks back our wort in the are» of cover study for radioactive waste disposal in near-surface facility, summarizes our experiences and lessons, which are expected to be instructive and helpful to other countries, especially to developing countries who are performing or will start cover study for radioactive waste disposal in near surface-facility. Introducing, digesting and transplanting the developed countries' technologie modeld principae san th s si l approac recommene hw developino dt g countr aree f th covea o n yi r stud low-and-ii.-termcdiatr yfo e level radioactive waste disposa near-surfacn li e facility, whic foune hw d eventually make vere sth y limited funding workabl productived ean . Introduction A protective cover is an important barrier for radioactive waste disposal in near-surface facilities, plays a critical role to long-term safety of radioactive waste disposal'1-">. In general specific functions include'4': (1) Minimizing infiltration through covee th r from precipitatio surfacd nan e runoff Minimizin) (2 ; contace infiltratege th th f to d water with waste through drainage layelow-permeabilitd an r y barrier layer ) Minimizin(3 , g differential settlemen subsidenced an t ) Minimizin(4 ; g surface erosion; (5) Providing resistance to biological intrusion; (6) Providing resistance to freeze-thaw attack; and (7) Maintaining long-term stability withou e nee th tf activ do e maintenance t shoulI . notee db d thae long-terth t m stabilit f thesyo e cover functionindependenn a t no s i s t function requiremen covereflectioa a f t o t rbu time-dependenf no t variation othee th f rs o cover functions. A number of countries have carried out cover development in varying degrees"'5- '• '• *• '•I0- "'. Through comprehensive analysi theif so r research programs three trend availablee sar firse tTh . engineering-scaltrenn a s di e cover testing. With further development of the research it has been recognized that laboratory-scale or small-scale cover tests could not be effectively used for long-term predictio f coveno r performance. Therefore, engineering scale tests were carried out, typical cases including permanent protective barrier development program by the Westinghouse Hanford Company and Pacific Northwest Laboratories in the United States'10'; waste cover tests by S.Melchior, Hamburg University in Germany*1"; bio-intrusion barrier testing by Los Alamos National laboratories in the united States(7). The second trend is modelling approach in cover research. Since radioactive waste disposal involves large space scal lond ean g time scal f hundredeo s years (for low-and-intermediate level radioactive waste disposal). Although engineering-scale cover tests can provide solution to the issue of large space-scale of a waste cover impossibls i t i , conduco et t cover testtima r esfo perio hundredo t p du s years. Therefore, modelling approacs hi being increasingly used and enhanced in the field of cover research to predict long-term performance of covers. A lot of codes have been develope r covefo d r analysi d assessmenan s t like CREAM, HELP02', UNSAT1D" e Uniteth y db 2' States, MARTHE France03y 1b TOUCH*d an , Spainy 6b thir'e Th d. tren eacs di h countr coven yow rcarrie s researcit t testssd ou han . Since the performance of waste covers is determined by such factors as site meteorological conditions, site characteristics, soil properties and cover structures and so on. Therefore, although there has been much experience in cover research and tests'3- M), including generally accepted methodology, specific test is still necessary for specific design. Currently, researc hroughle b wor n kca y divided into three catelogue termn i s f researco s h approaches e firsTh t. approac s tharesearchi e (h t conductes hi d mainl testingy yb , like China' Germany'"'d an - thin I . s case large amounf o t ) 5 funding and time are required, and long-term performanc5 e assessment is still quantitative without efforts in modelling. The secon testing/modellins di g approach i.e. Hanford Permanent Protective Barrier Program' 10s followini ' approache th g . This approach also requires large amoun fundinf to firs e timd th t s gan oneea . Certainly, this approac significanf o s hi t valu boto et h engineering design and long-term performance assessment of waste covers because testing is in combination with and supported by modelling mutually. The third is modelling approach which is used in Spain'6'. By this approach the cover research is done through computer simulation and with essential parameters obtained in laboratory experiments. Through different viewpoints exist single th , e modelling approac wastr fo h e cover researc regardes hi mose th ts da effectivf o y wa e cover researc casn hi difficultie f eo fundingn i s . China's cover researc s beehha n develope regulated dan d wit s developmenhit radioactivn i t e waste disposale Th . progress and development of cover research in the field will be described in the following sections. And finally, recommendations to future cover research program will be given with consideration of current cover research trends and cover research practic Chinan ei . 289 Cover Researc Low-and-Intermediate-Lever hfo ! Radioactive Waste Disposa Chinn i l a 1. Background China will build up three regional sites to dispose of low-and-intcrmediate level radioactive(L/ILW) by the en f thio d s centur a solution s a y o radioactivt s e wastes arising from nuclear fuel cycle, nuclear power plants, and nuclear technology applications in China. Figure 1 shows distributioe th L/ILf no W disposal sites. To meet the demand of radioactive waste disposal the cover research was intiated in China Institute of Radiation Protection in 1988. Up to now the work can be divided into four stages in the cover research i.e. conceptual design, testing design, testing preparation and modelling research. 2. Cover Research Activities 2.1 Stag Conceptuaf eo l Design With the development of Qinshan Nuclear Power Station, work concerning the South China L/ILW disposal Site was put forward and site pre-selection was conducted in Zhejiang Province, in 1988 and cover research was Fig. 1 .Distribution ofL/ILWdisposal sites followed s thoughwa t I .t that0*: "fo ra disposa l celf o l radioactive waste disposal system it is necessary to include: (1) Bio-intrusion barrier, (2) Cover to minimize water infiltration, (3) Drainage and collecting system of infiltrated water, (4) Base liner to prevent groundwater contamination, (5) Backfill materials to retard radionuclide release, and (6) Corrosion-resistant waste container." Based upo e abovnth e knowledg d locaean l meteorological condition f annuao s l mean precipitatio f 1290mmno , annual mean temperatur f 15.6°eo potentiad Can l evaporation rat 800mmf eo followine ,th g tests were designed. (1) Test Design of Bio-intrusion Barrier Biological intrusion generally includes animal intrusion, plantation penetration, and inadvertent human intrusion. Here only plantation intrusio consideres nwa teste th . n dAccordini locae th lgo t geologica meteorologicad lan l conditions, three 10*8m blocks were planned to be selected in the Gaoyu area, Anji county, Zhejiang Province, to place different biological barrier treatment in the three blocks and observe continuously for 5 growing seasons to measure root depth and degree of disturbance to the bio-intrusion barriers. The schematic design of the test is shown in Figure 2. C> mT \ /~T^ I— —-^9 «— —Lii_.X . 19n _i , 10m x AH ^ X' N c £ 1*0.03 Ca-Bentonlte OravsK-aon) OobKLe Oravel/Ca-Bentonite (5-20cm) Fig. 2. Conceptual design of bio- intrusion barrier test(3> 290 (2) Test Design of Infiltration Barrier A multi-layer laye p covemixtur a s to designe s i r e tesrwa e f Th locath teo r dfo l bentomt clayd ean , subsequently beneath is gravel layer The principle objective of the test was to determine the anti-infiltration capability of the mixture layer of bentomte and clay According to the plan, some experiments should first be conducted in laboratories to determine the maximu densityy mdr , permeability, strengtd •an bentomte/claf ho y mixtur varioun ei s ratio Then cove,a r woul constructee db d in the field at bentomte/clay ratio of maximum strength and minimum permeability Figure 3 shows the schematic design of the watee tesTh tr infiltratio d distributioan n colume th n i nn woul detectee b d d unde e conditionth r f locaso l annual mean precipitation, maximum precipitation density, and maximum precipitation duration Fi gConceptua3 l designof infiltration barrier test(3) However, because of slowdown of the East China L/ILW Disposal Site and in funding the above two tests were not be abl como et realito et y Only basic propertie locaf so l bentomte were teste resulte dTh liste e sar Tabln di e1 Tabl Basie1 c propertie locae th f ls o bentorute o N Items No 2 Particle density 271 269 Fluid limit(%• ) 9930 5442 Plastic lunit(%) 7020 4168 Plastic index 29 10 1274 Maximu density mdr y 097 1 38 Optimum water content(%) 620 31 1 Total porosity(%) 642 487 Swell/shrinkage ratio(%) 7608 5682 Saturated permeabihty(cni/s) 063*10"* 1 09*105 Air-dned water content(%) 683 934 2.2 Stage of Testing Design With the development of the Northwest China L/ILW Disposal Site, relevant work was set out in 1991 The site is in very and area with annual mean precipitation of 62mm, annual mean temperature of 7 9°C, and potential evaporation rate of 3577mm/yr Therefore, evapotranspiration through covers were considere testinn di g design besides water infiltration Water movement contro mose thoughs th te wa l b critica o t t l functio L/ILa f no W disposal cover Considerin situatiow ne d e gnan th previous cover research experienc ethree-yeaa r LILW disposal cover researc t uphse <5 plas ) nwa 291 1990 Literature surve cover yfo r research Cover testing desig" n Conceptual desig experimentaf no l apparatus 1991 Sketch desig experimentae th f no l apparatus Manufacture of the experimental apparatus Purchase of related experimental equipment Apparatus Installatio tnad nan l test Material collection and preparation 1992 (!) Infiltration Test — simulated cover test for Southeast China L/ILW Disposal Site Evaporation/transpiratio) (2 n Tes simulatet— d cover tesr Northwestfo t China L/ILW Disposal site e InfiltratioTh n Tescomposes twa followine th f do g activities (i) Field distnbution measurement of temperature, depth and moisture in the near-surface, (11) Study of water movement through covers (a) Water infiltration tests with soil depth of 0 5m and 1 Om and under compact and loose conditions, Low-permeabilit) (b y material tests - bentomte/backfill mixture, at ratio of 1 9, 2 8, and 5 5(wt), - cement/backfill mixture, at ratio of 1 9, and 3 7(wt), - bitumen/backfill mixture, at ratio of 0 05 0 95(wt), and - compacted clay, Wic) (c k effect test, Multi-laye) (d r cover test Figur eshow4 profile sth e desigmulti-layee th f no r cover (e) Modification test of top soil Add 2% bentonite, 0 75% NaCl and 0 75% Na2CO3 into the top soil and compact them to thickness of 0 5m, and (f) Plantation test Compare infiltration difference through covers wit withour ho t surface plantation Top soil layer 0 5m evaporation/transpiratioe Th n test were composes a f do Compacted clay 0 3m follow- Gravel layer 0 5m (i) Field distnbution measurement of temperature, depth, and moistur near-surfacee th n ei , Compacted bentonitm 3 e0 (u) Study of water movement through covers ) evaporatio(a n through covers with , thicknes5m 0 f o s Compacted backfilm 5 0 l , Om 3 d an , Om 2 , 5m 1 , 1Om (b) evaporation through backfill layer without further engineering actions, with thickness of 0 5m, 1 Om and 1 5m Fig 4 Schematic diagram of respectively, multi-layer cover rfo (c) evaporation through backfill layer compacted to the infiltration test maximu densityy mdr , , 5m wit 0 hd thicknesan m 3 0 f so evaporatio) (d n through 0-15cm gravel surface treated covers, (e) evaporation/transpiration with plantation treatment, and (f) evaporation/transpiration through multi-layered covers The profiles of the covers are shown m figure 5 Top m Soi5 0 l Topm Soi5 0 l Compacted Soil 0 5m Compacted soil 0 5m Cobble Layer 0 5m Cobble Layer 0 5m Compacted Clam 5 y0 Compacted Clam 5 y0 a Planted surface b Bare surface Fig 5 Schematic of multi-layer covers for evapotranspiration tests 292 An apparatus was designed specially for the cover tests'5' Figure 6 shows the apparatus schematically As show Figurn apparatue ni th , e6 s consistf so the following subsystems (1) Container It is 1*1 *3m carbon-steel with thickness of 5mm It is composed of six section and connected each other with flanges, (2) Filling system, (3) Spnnkler system Spray water with a H»«t Supplier spnnkler, (4) Evaporation heating system The heat required woul e supplieb d d with infrared light accordin o locagt l solar radiation duration and density to assure the surface temperature of the filling not lower than that of the field surface, (5) Drainage system, (6) Temperature/moisture measure system, and (7) Supporting systems, including pump, electrical control, support equipment, cranc eet As planned test woul conductee db Chine th n dai Porous Pipe« Institut r Radiatiofo e n Protection, Taiyuan, Shanxi r «Itoo r Bed Province 2Stag3 Testinf eo g Preparation Du insufficieno t e t funding e wholth , e s teswa t postponed A lysuneter system was purchased in 1991 and Fig 6 Schematic diagram of apparatus a laboratory was reconstructed specially for the tests in of the cover test 1992 Because it was realized that the application limitation of cover testing results obtained in the laboratory of China Institut r Radiatiofo e n Protectio n desigi n d an n performance assessmen coverf o t eithen si Northwese th r t Southease Sitth r eo t Site e gian,th t cover research program was suspended temporanly 2 4 Stage of Modelling Under an IAEA technical assistance project entitled "Low-and-Intermediate Level Radioactive Waste Disposal (CPR/9/014)", a fellowship training on L/ILW disposal cover design was held at Battelle Pacific Northwest Laboratones, the United States in 1993 It provided us opportunity to overview current work of waste cover research and development around the world obtaio t , n systematic understandin paste th ,o g t presen t situatio futurd nan eHanfore plath f no d Permanent Protective Bamer Program, and to acquire perceptual knowledge of cover research The first of two main activities is a laboratory-scale evaporation experiments of different surface treatments tine sands and pea-gravel The experiments show that gravel layer on the surface can greatly reduce evaporation and as a result increase infiltration The work is to simulate cover performance with HELP code whic EnvironmentadevelopeS s U hwa e th r dfo l Protection Agenc revieo yt assesd wan s waste cover designe Th s work demonstrated tha HELe th t P effectivt codno s ei Hanforn ei d becaus underestimatt ei e evaporation rate Anyway this work gav systematis eu c knowledg modellinf eo g approac coven hi r researc assessmentd han , whiceconomin a s hi effectivd can e » approac coven hi r research Cover research by modelling approach was initiated in China Institute for Radiation Protection in 1994 The modelling approac covef ho r researc consideres hi d suitabl presenr efo t actual condition Chinn s i wor e plannes Th ake wa b o dt earned out by two stage First, input parameters required by code shall be prepared through laboratory experiments and literature survey Then secon e code th ,th en dAfte ru woul o t r e comprehensivdb e sensitivity analysis preferre,a d cover design suitabl Chinr efo a Southwest conditions woul givee db referencs na factuaf eo l engineering adapte e desig b futur e o t th r n dei o field cover test when favorable conditions are available Conclusion Recommendationd san s 1 Recalling China's history in L/ILW disposal cover research shows that China's cover research may be divided into two phase researcy sb h approaches, testing phas modellind ean g firse phasth t n phaseI e achievee w , belor dfa w expectew dNo we are just in the beginning in the modelling phase however we are confident that we will achieve much more than that in the testing phase 2 Performanc L/ILa f eo W disposal cove closels ri y connecte sito dt e charactens very much site-specific Thereforet i , is recommended that site-specific cover tests be conducted as long as conditions permit However, in case of insufficient funding, modelling is truely an effective approach to cover research, which is more applicable to the developing countries 293 3 Cover desig assessmend nan t remai quantitativn i e stag Chinan ei ) Mor0514 e important, L/ILW site widele sar y distributed in China and conditions of these sites are very different Therefore, we recommend China's competent authorities to increase funding to LILW disposal cover research so as to improve results of cover research by modelling approach and provide funding to field scale cover tests as long as conditions permit 4 We recommend that the developing countries start their cover research with modelling approach, e g performing cover testing mainl modelliny yb g analysi assessmend san t with necessary input parameters obtaine laboratorien di o t s a o ss make very limited funding more effective and productive We recommend that the developing countries to make full use of technological results and models accumulated by the developed countries as basis of our LDLW disposal cover research And when situations permit large scale cover tests wil performee lb steo dt p into cover researc testing/modellinf ho g REFERENCES IdahoG c d 1986In ,an , G E "Safet[Iy] Assessmen Alternativef o t Shalloo st w Land Buna f Low-Leveo l l Radioactive Waste Volume 1 Failure Analysis of Engineered Barriers", NUREG/CR4701, August, 1986 [2] nimois State Geological Survey, "A Study of Trench Covers to Minimize Infiltration at Waste Disposal Sites Task 1 Report Revie Presenwf o t Practice d san Annotated Bibliography", NUREG/CR-2478-V1 Huang d "Concep, an W L Y , C t, DesigGu Simulater n[3fo ] d Test Low-Levef so l Radioactive Waste disposal", December 1988 unpublished [4] Bennett , "Recommendation D Soir R , fo l C CoveNR re Systemth o st s Over Uranium Mill Tailing Low-Leved san l Radioactive Wastes", NUREG/CR-5432 Vol 1, February 1991 , "Statu L C Covef , so Gu r Researc[5] Testind han g Design", Journa ENERGf lo ENVIRONMENTd Yan , 1992 2o N , ] [6 Gravalos Santiagod an "Analysi, M L ,J ,J Infiltratiof so n Throug hMultilayea r Cover Syste Wasta mt a e Disposal Cell", IAEA Research Coordination Meeting, October 21-28, 1991, United Kingdom [7] Hakonson etc, ,E "Preliminar,T y Assessmen Geologif to c Material Minimizo st e Biological Intrusio Lowf no - Level Waste Trench Cover Future e Pland th sr an sfo , ORNL/NFW-81/34,1981 [8] Femmore , "Evaluatio W ,J Bunaf no l Ground Soil Covers", DPST-76^27, November 1976 [9] Nyhan etc, , W "CorrectivJ , e Measures Technolog shallor yfo w Land Site Bunad sAn t Fiella d Studie Bion so - intrusion Barrier Erosiod san n Control", LA-10573-MS, March 1986 [10] Wing, N R and Gee.G W , "the Development of Permanent Isolation Surface Barners, Hanford Site, Richland, Washington, USA", 357-363, Proceedings of the International Symposium on "Geology and Confinement of Toxic Waste", 8-11 June 1993, Montpellier, France (II] Melchior, S etc "Water Balance and Efficiency of Different Landfill Cover Systems", 325-330, Proceedings of Internationae th l Symposiu "Geologmn o Confinemend yan TOXJf o t C Waste", 8-11 June 1993, Montpellier, France [ 12] Electnc Power Research Institute, "Companson of Two Groundwater flow Models - UNSAT1D and HELP", EPRI/CS- 3695, October, 1984 [13] Thiery, D , "Tn-dunensional and Multi-layer Modelling of Transfer in Unsaturated Porous Media" , Proceedings Internationae ofth l Symposiu "Geologmn o Confinemend yan Toxif o t c Waste", 8-11 June 1993, Montpellier, France [14] US Environmental Protection Agency, "Technical Guidance Document Final Covers on Hazardous Waste Landfills and Surface Impoundments", PB89-233480, July 1989 [15] Gu, C L and Fan, Z W , "Recommended Modification to Shallow Land Disposal Option of Large-Volume Grouting Matnx", 1992 unpublished etc, , F "FeasibilitD , Wu y ] Analysi[16 s Repor Day r Low-and-Intermediaty fo t aBa e Level Radioactive Waste Disposal Site(Draft)" 199y 4 Ma , 294 SCREENIN SORPTIOF GO N MATERIAL RADIOIODINR SFO E AND TECHNE1TUM J. ZENG, D. XIA, X. SU, X. FAN China Institut f Atomieo c Energy, Beijing, China Abstract By using batch sorption experiment s beeha n t i scomplete o determint d e sorption TcOd an ratio~ jI anionsr sfo . Several kind f industriao s l product minerald an s s from China were tested. Tests were performe t 25°a d c constant temperatur tracen i e d pre- equilibrated wate r deionizeo r d water e tesTh t. results show thasorptioe th t n ratiof so tiemannitr I~fo activated ean d carbo f apricot-pi nordeo e th f ro 10e ar 3tml-g~ 1,while eth sorption ratio of TcO^ for jamesonite is of the order 10*ml-g~.Thes1 e materials can be considered as selected backfill materials to improve capability of depository for preventing iodine and technetium anions from migrating. Comparing results of sorption ratios of I~ TcO^d an T from bot pre-equilibraten hi d wate d deionizean r d water,in most cases values e nearlar samee yth . Althoug difference hth f sorptioo e n ratio betwee solutiono ntw r sfo materialw fe times,is a hig5 a s s ha si stils i t l feasibl screeninr efo g sorption materialy sb using traced deionized water system. 1. INTRODUCTION In nuclear waste geological disposal^he long-lived fission products "Tc and ml, nor- mally existing as anions, could be hardly sorbed by bedrock surrounding repository,such s granite,tufa basalt,ad an f s wel s backfila l l bentonite (1-3)e purposTh . f thieo s study is to search for materials with good sorption properties for TcO^ and I~,in view of their mixing with backfill materials. s beeha n t I reported that naturally occurring minerals bournonit d tetrahedritean e strongly sorb TcO^l 2000),ouo t p s valueu )(R e ar sr research results show that TcOj is strongly sorbe y stibnitb d e (4),the sorption rati s abouoi t 2xl03ml-g~ e preTh 1-. sumptio thas ni t some other Sb-containing minerals should have good sorption capacities . Frofo Tc rvarioue mth s mineral f thio s s kind availabl Chinajamesoniten ei , antimony ocher,kermesit d antimonitean e whose sources differ fro e sourcmth f stibniteo e usen di previous work have been chosen in this work.Batch techniques have been used for sorption and desorption experiments. 2. EXPERIMENTAL Material1 2. s e materialTh s were crushe d sieved dan fractioe Th . n between 60-120 mes s colhwa - lected for experiments. Pre-equilibrate2 2. d water Pre-equilibrated water was prepared by contacting deionized water for at least two week t 25°sa c with ground mineral t beesno nthad sievedha t ratie th ,solutiof oo n volume 295 solie th do t weigh 20(v/w)s wa t phasee Th . s were separate centrifuginy db t 1800ga m 0rp for one hour. 2.3 Tracer 9 The tracer of "Tc used for experimints was NHf Tc04 in O.lM NH4OH solution,with activitn a 2xl0f yo 7Bq-ml~1 obtained fro Radio-chemicae mth l Centre Amersham,United Kingdom. The solution of Na I without reduced agent and with the specific activity of 5.8xl0 125 7 Bq-ml"1 (from Isotop Departmen f CIAEo t uses thin wa )di s work. Sorptio4 2. n procedure 0.25g of materials (60-120 mesh) was weighed and added to a stopped polyethylene centrifuge tube thad beeha tn weighed,5m spikee th f o dl preequilibrated wate s thewa rn adde it,meanwhilo dt blana e k test uses tha determino wa dt t e initiaeth l activits ywa prepared by only placing 5ml of spiked preequilibrated water in a tube in which there is materialy noan s selected. tubee Alth l s were place a water-bat n di h shaker wher e wateth e r temperaturs ewa set at 25°c. The shaking frequency was 120 oscillations per minute. Contact time varied e shakinth f o gd perioden e aqueoue daysth 5 th 3 ,t froo A t . ms7 phas s separateewa d fro solie mth d phas centrifuginy eb houre on r . t fo 1800ga m 0rp An aliquot of the supernatant was removed and placed in a scintillation counting vial in scintillatiof whico l h7m n solutio beed nha n added.The sampl thes ewa n measuree th r dfo remaining activity.The rest of supernatant that was taken out as much as possible was valuH p provide e e th determination.Th r dfo e volum e finath f l o eliqui d remained with solid,which was determined by weighing,was used later in the caculation of desorption ratio. 2.5 Desorpton procedure Desorpton experiments were carried out from the samples previously used for the sorp- tion experiments.5ml of non-spiked preequilibrated water corresponding to the materials s use eacr wa dfo h experiment.The same procedur sorptioe th n ei thas n experimentwa t s was used for separation of the phases and for the radioactivity measurements. The des- orption contact time was 23 days. . CALCULATIO3 N The sorption uses ratii expreso s dt o R abilite th sI f materialyo 125 r o soro c st b"T and calculated by the following formula V_ _ D-AfAt - RS~ At 'W where spikee Af125th r o :n Ii d initia c wate"T f l o ractivit l m r ype supernatane th n i : activitl At m r t ype solutio n afte contace th r t time v: volum f liquieo d phase 296 : weighW solif o t d material used : dilutioD n correctness factor thin I s work,sorptior fo e equas Di on n o experiments,whilt l desorptior efo n experiments Vd ~ ^Vd + Vr '' where Vd: volume of solution added for desorption Vr: volume of final residual liquid with solid phase after sorption experiment 125r o I : fractiosorbec Fs "T f do n on/in solid phase 4. Results and Discussions 4.1 Sorption and desorption of "Tc on minerals c sorptio"T r fo varioun valuee o ns R Th f o s Sb-containing mineral day5 3 f t o sa s contact time are present in Table 1. Sorption ratio data for "Tc on other materials Table 1. Sorption and Desorption of "Tc on materials Materials Contact time Sorption ratio PH Desorption ratio pH days R,, ml-g-1 sorption R«*, ml-g-1 desorption stibnite 1 143 5.87 99.9%(42)* 5.22 7 380 6.22 99.8%(36) 5.32 14 520 4.53 99.8%(29) 4.78 21 1032 4.64 99.5%(22) 5.15 28 1980 3.84 99.2%(15) 4.71 42 1980 4.67 99.7%(1) 5.99 molybdenite 1 1.3 7 0.6 14 2.6 21 1.3 35 2.7 rare-earth 1 5.1 hematite 7 1.2 14 1.0 21 2.8 kermesite 35 44 7.51 antimonite 35 2660 5.21 2220(23) antimonite 35 25800 3.75 34500(23) ocher jamesonite 35 61200 3.75 37700(23) iron powder 35 1910 7.86 reduced apricot-pit 34 18000 8.67 activated carbon * The figures in parentheses are contact time for desorption,and the percentages are the desorption fraction remaine mineralsn do . 297 which were considered as potential backfill candidates for retarding "Tc (5),taken from the reference (6),were also listed as compared with those above mentioned. It is seen that Sb-containing minerals used for testing,except for kermesite,strongly sorbed "Tc. The Rs value antimonitr ordee sfo compatibls i th magnitudf f d o ro an s 3 ewa 10 e f ewito date hth a provided by previous study (4). This has again proved good Tc-sorption capacity of anti- monite. Tc-sorption capacities of both antimony ocher and jamesonite are much greater than that of antimonite,the Rs values are 2.58xl04 and 6.12xl04ml-g~1,respectively. Dependenc contacn o s R tf eo tim e indicated thasorptioe th t n ratio r antimonitesfo , antimony oche jamesonitd an r e increase with increasing contact time sorptioe Th . n equi- librium wasn't reache day5 3 r s dfo contact kemesitr .fo Variatios R e e th wit f no h contact time differed from above minerals, the highest value of Rs occurred at 7 days, in the following days dropped gradually case,thesy an n I . e results correspon conclusioe th o dt n mad R.Strickery eb l (l),thaa t e t equilibriun a t m distributio reaches ni d slowly (>100h) for bournonite and,presumably,for similar minerals. Tabl . Sorptioe2 desorpiod nan 125f no materialn Io s Materials Contact time Sorption ratio pH DesorptioH np ratio 1 days Rs.ml-g- sorption Rd,ml-g-l desorption chalcopyrite 1 41.5 7.73 99.0%(22) 8.15 3 47.8 8.03 99.8%(20) 7.71 8 49.0 7.64 99.4%(15) 7.84 15 66.6 8.11 99.0%(8) 8.10 21 96.9 8.24 98.0%(1) 7.95 galena 1 118 4.27 98.0%(22 6.71 3 83 4.83 98.0%(20 6.62 8 63 3.53 99.0%(15 6.64 15 62 4.64 99.1%(8) 6.34 22 137 6.14 99.6%(1) 6.36 rare-earth 1 1.2 7.91 hematite 3 1.6 8.24 8 1.2 8.34 22 1.3 8.31 molybdenite 1 0.4 7.70 3 1.1 7.62 7 3.0 7.64 21 0.9 7.89 stibnite 1 2.7 5.42 7 2.6 5.66 15 1.5 4.72 21 2.6 4.54 pyrite 1 2.7 7.05 3 2.5 7.83 7 6.0 8.03 15 3.6 8.10 21 5.9 8.10 diatomite 34 250 7.64 480(23) 7.66 cinnabar 35 500 8.20 tiemannite 35 2300 5.74 800(23) 6.70 apricot-pit 35 2800 8.54 4600(23) 8.58 activated carbon *The figures in parentheses are contact time for desorption,and the percentages are des- orption fraction remained on minerals. 298 Desorption experiments were carrie t witdou h stibnite, antimonite, antimony oched ran jamesonite. Desorptio s use nr expressin wa dfo rati d oR g experiment results e samth , e formula used for calculating Rs was used for caculating Rd. The results are presented in Table 1. By comparing Rd with Rs listed in Table l,it can be seen that Rd values are smaller than Rs value for both antimonite and jamesonite, while Rd value is larger than s valueR antimonr sfo y ocher. these results indicate that slight desorption occur botn so h antimonte and jamesonite, while no desorption occurs on antimony ocher during 23 days of contact time. 4.2 Sorption and desorption of 125I on materials Sorptio kind0 1 125f nf mineralo n o sIo r inorganiso c material investigateds swa e Th . results obtained are shown in Table 2. The sorption ratios of 125I on diatomite, cinnabar, tiemannite and apritcot-pit activated carbon were much higher than on other materi- als,suc stibnits ha e ,molybdenite, rare-earth hemahite,chalcopyrite,galen pyritd aan d ean s valueR . son arouno s d 102-103ml-g~1 were obtained. Desorption experiments were carried out with chalcopyrite,galena diatomite and apri- cot -pit activated carbo showed nan d that sorbee morth f eo d tha 125% n99 I remainen do these materials after 20 days of desorption. Therefore, sorption of 125I on the materials mentioned above seems also to be irreversible. . CONCLUSIO5 N Some Sb-containing minerals,such as antimonite ocher and jamesonite,have very high sorption capacities for "Tc,their sorption ratios are competitive with that for both iron powder reduce activated dan d carbon. The results obtained also indicated that some inorganic materials,such as diatomite, cinnabar,tiemannite and apricot-pit activated carbon have good sorption properties for 125I.From desorptio seee b n n n testthaca t 125r si t o somn sorptioI o c e "T materialf no s investigated seems to be irreversible.lt. appears that these materials can be taken into consideration as backfill candidate for retarding migration of "Tc and 125I. REFERENCES [1] Strickert,R.,et al, Nuclear Technology,Vol. 49,No.49,253( 1980). ] Wolfsberg,k.,Sorption-desorptio[2 n Studie Nevadf so a Test Site. LA-7216-Ms,(1978) [3] Erdal,B.,et al., CONF-781121-6,(1978). [4] Zhung Huie,Zeng Jishu,Zhu Lanying, Radiochimica Acta 44/45,143(1988). [5] Westsik,J.H.,et al.,Scientific Basi Nuclear sfo r Waste Management Vol. 6(ed.Topp,S.V.). North-Holl and,P326(1982). [6] Zeng Jishu,Xia Deying,Annual Report,Institute of Atomic Energy,Beijing, China(1988) 299 A SYSTEMS APPROACH FOR QUALITY ASSURANCE IN WASTE CONDITIONING, STORAG DISPOSAD EAN L E. R. MERZ Jülich Research Centre, Jülich, Germany Abstract An integrated managemen e systeth r mfo radioactivf o t e wastes refers to the complete spectrum of background policy, safety, environmental protection, and actual practices which define the classification, control, movement, conditioning, quality assurance, storag disposad ean f lo wastes. Emphasi s placesi d upon demonstrating tha radioactivl tal e wastes casafele nb y isolated frobiosphere requiree mth th r fo ed time. Four domains dictate the requirements concerning properties and quality of the wastes to be disposed of: - handling and transport conditionin- g interi- m storage - final disposal. Thus, waste produc canisted tan r quality assurance measures muse tb oriented towards criteria derived from their overall safety assessments. The most stringent requirements originate from long-term safety aspects geologicae ofth l repository. In evaluatin engineerinn a g g projec o differenttw t way achievinf so g the specified goal are customary: either an inductive or deductive analysis approach. It is proposed here that the deductive method, which first analyses the total system as a whole and then draws inferences for each single step, is the more advantageous way. The criteria to be set up for any kind of radioactive waste disposal must always be put in perspective: (1) what are the waste characteristics? (2) what time period for safe isolation is of interest? (3) which geo- logical disposal alternatives exist? Different approache usee b dy sma in the short- and long-term perspective. In either case, a general pro- cedur recommendes ei d which involves concentrating, containind gan isolating the source of radiotoxicity as far as practicable. Governmental agencies determin e requirementth w eho efficienn a r sfo t quality assurance metsystee b n . Compliancca m e with these authentic standards will be assured by independent product and quality control measures. 1. Introduction The objective r ensurinsfo g quality assurance waste (QAth en ) i manage- ment prograo providt e ar me confidence tha e integratetth d radioactive waste management system will prevent disposed waste from returning to the biosphere and will operate safely in accordance with legislature and re- gulatory requirements. The program needs to provide assurance that the waste management system will perfor s programmatiit m c functions reliably and efficiently. The quality assurance program must cover all the elements of the waste management system. thesf Almoso l e al telement uniquee sar ; there ear some similarities with nuclear power plant quality assurance measures /!/, but with notable differences certaia n .I n way, nuclear power plant quality assurance programs may serve as the model for the waste management program, 301 but it has to be developed so as to be appropriate for the unique operations of the conditioning, transportation, storage, and final disposal elements of the whole system. An effective program of quality assurance is essential for demonstrating that the technical performance of the waste management system and its ele- ments meets regulatory standards. The licenses needed for the operation of the various waste management steps require the implementation of a quality assurance program that satifies all relevant governmental stan- dards, orders and directives. As an example the German Report "Produkt- kontrolle radioaktiver Abfäll Schachtanlag- e e Konrad citee b y d "ma /2/ . The principal regulatory requirements that apply to the waste management quality assurance progra containee ar m thin i d s report. An essential elemen e developmenth f to d implementatioan t sucf no a h quality assurance program is the instruction and training of all personnel participatin specifie th n i g c national program overale .Th l government- dependent quality assurance plan envisages developing training modules and conducting training sessions to ensure that all personnel partici- e prograpatinth n i gm fully understan e managemendth t system theid san r responsibilities for quality. Consistent application of quality assurance requirements not only en- sures the accomplishment of work by all participating organizations to the same required quality alst ,bu o facilitates systematic verificatiof no quality achievement. Quality assurance practices are a component of good management and are essential to the achievement and demonstration of high quality in products and operation. Organizational arrangement r sounsfo d quality assurance practice e requisit sar partiel al r sefo concerne provido dt cleaa e - rde finition of the component groups'responsibilities and channels of communicatio d coordinatioan n n between them. The objective qualitf so y assurance measure theid an s r integration into the overall waste management system are illustrated by the following outline e Figur,se . 1 e 2. The systems approach In evaluatin engineerinn a g g project competeny ,an t professiona- len gineer or technical manager would prefer to use a systems approach to interconnec e subtasktth optimun a o st m functional entirety simpln .I e terms, this means that the engineer would like to quantify, on a comparable basis, different ways of achieving the specified goal by considering all the aspects and effects of each option. In principle, eithen ra - inductive analysis approach or a deductiv- e analysis approach caappliede b n . The inductive route draws conclusion analyziny b s g each individual step separately subsequentld ,an y putting them togethe o forrtotat e mth l system, whereas the deductive method in a reverse mode first analyzes the total system as a whole and then draws inferences relating to each single step. 302 Radioactive Waste Management Waste Packaging Interim Final Wasle Conditioning Transportation Storage Disposal . ... ^ . . 1 1 __ 1 Hazard Potential of Radioactivity Execution Surveillance Technical Measures, l Qual i Y Assurance [ Method d Procedurean s s Systems Approach Collectio f Wastno e legislatio d Responsibilitnan y Treatmen y Varioub t s of the Operating Organisation Techniques Evaporation Safely Culture Solidification Quality Assurance Compacting Implementation Packaging Transport, Storage Quality Control of Waste Disposal Product d Canisteran s s Management Audit and Application of proven engineering fecfiniqucs Inspection Program Safety Assessmen Verificatiod an t n Safety Analysis Report Risk Analysis Fig. 1: Outline of radioactive waste management structural elements Clearly, the deductive approach is the more appropriate one, although in most cases the inductive route has been applied up till now in waste management. This is due to the fact that mostly experts try to solve their proble themselvey b m thein si r special field. Many mistake pase th t n si originate from this. Many nuclear waste management and related safety critics, when making presentation non-technicao st non-scientifir lo c audiences, will pontifi- cate on the amazing new concept known as the holistic approach in so far as it is any different from normal quantifiable value judgements of a social, moral, or ethical nature. If this is done, it has the inestimable advantage to the critic of making meaningless comparisons between diffe- rent concepts, sinc critie e th concludn cca e tha e holistitth c approach prove particulae sth momenscheme t superioe b rth pe y f o an o et o rt alternative, without having to go to the bother of quantifying the case for it. 303 Inductive Approach Deductive Approach Packaging Interim Conditioning Wait* Predetermined Safety Features Transportation Storage of Waste Disposal of the Overall Sy»l«m Packaging Interim Waste Waste Transportation Storage Conditioning Disposal Best Individual Performance i> i 1 i 1 Deduced Requirements i r < i ' r 1 r Derived Safety Features Optimized Optimized Optimized Optimized Packaging interim Waste Waste From Putting Togethe Overaln a o t r l System Transportation Storage Conditioning Disposal Fig : System2 . s approac r wasthfo e management Advisory Commissions K SS RS . K BMFT BMU Objective Project Responsibility with geoscientific assistance Execution Nuclear f Projecto s DBE Research Establishments d Researcan h BMFT: Federal Minister for Research and Technology (Bundesminister für Forschung und Technologie) BMU: Federal Minister for the Environment, Nature Conservation and Reactor Safety (Bundesministe Umweltr rfü , Naturschut Reaktorsicherheitd zun ) BfS: Federal Agenc Radiatior fo y n Protection (Bundesamt für Strahlenschutz) BGR: Federal Agency for Geosciences and Raw Materials (Bundesamt für Geowissenschaften und Rohstoffe) DBE: German Company for Construction and Operation of Waste Repositories (Deutsche Gesellschaft zum Bau und Betrieb von Endlagern für Abfallstoffe) Fig: Responsibilit3 . r disposafo y radioactivf lo e wast n Germanei y 304 3. Legislation and responsibility of the operating organization Nuclear power stations in the Federal Republic of Germany are privately owned and operated. Consequently, the electricity utilities are obliged to take care of the back-end of the nuclear fuel cycle. German government and federal authorities have laid down in a special order that nuclear power stations will only be licensed if sufficient evidence of secure ultimate waste disposal practice discloseds si . The ultimate disposa wastf lo e remain responsibilite sth Fedee th -f yo Governmentl ra Germae .Th n Atomi cappoint/ Energ/3 t e FederaAc y sth l Radiation Protection Agency (Bundesamt für Strahlenschutz, BfS) in Salz- gitter to undertake waste disposal on behalf of the Federal Government. Thus, the BfS has to organize, construct, and operate the repositories for radioactive wastes. An organization chart is depicted in Figure 3. In the future there may be a certain change. The commitment to govern- ment responsibility may be changed by an amendment to the Atomic Energy Act /3/ within the aim of transferring the responsibility for installing and operatin repositore th g o industryyt . Ultimate responsibilit e variousafete th th f r o ys yfo waste management installations rests with the operating organization. The operating organi- zation establishe police sth adherencr fo y safeto et y requirements, establishes procedures for safe control of the installations under all conditions, including maintenance and surveillance, and retains a competent, fit and fully trained staff. In Germany, responsibilit protectinr fo y g public healt assurind an h g safety of the radioactive waste has been vested in the Federal Ministry foEnvironmente rth , Nature Conservatio Reactod nan r Safety (BMUn i ) accordance wits enablinit h g legislatio subsequend nan t law regud san - lations. In Germany radioactive waste is controlled by the Atomic Energy Act /3/ and the Radiation Safety Ordinance /4/. Construction and operation radioactivf o e waste management installations onln accordinca y / /3 o gt be grantee designeth f i d d operatio e installatioth f no precaul al d -nan tions taken against damage represen e artstate th tth .f eo Accordin o t g /4/ the technical design and operation must fulfill the safety objectives laid down in the stipulations described therein. Similarly e constructio,th operatioe th d nucleaa nan f no r repository requires the official approval of a plan according to section 9 b of the Atomic Energy Act. According to section 24 paragraph 2, the highest authorit e Federath n i y l Stat whicn i erepositore hth estabe b o t - s yi lished is responsible for this. For example, in Lower Saxony, where the licensing procedure for the Konrad mine is currently in progress, the Lower Saxony Environmental Ministr authorite th s i y y responsiblr efo plan approval. Preliminary waste acceptance requirement Konrae th r d sfo min e have been formulated by the Federal Agency for Radiation Protection, BfS /5/. Theformulatee ar y tha y n suci wa d t a hthe y first describ generae eth l aspects which mus fulfillee tb waste th ey db package thed san n list more specific requirements for the waste forms, the packagings, and the radio- nuclide inventories. Safe operation and permanently effective enclosure of radioactive materia e guaranteeb n ca l meany b d f constructiono s , control- ad d ,an 305 ministrative measures. Every future licensing procedur nucleaa r efo r waste treatmen storagd an t e plant must confor safeto mt y criteria stipu- lated by the government. In nuclear technology the multibarrier principle is applie t onl no n nucleadi y r power plants alst bu wast,n oi e treatment and storage facilities for any radioactive waste. An ingenious technical safety system allow malfunctionl sal s that havreasonably an e e probabilit occurinf o y e controlleb o gt n sucdi a h way that the population and operation personnel do not suffer any serious damag healtho et . Operational disorder malfunctiond san s that could occur e planinth t mus e examinetb malfunctioa n i d n analysis. After being grante licensea d , every nuclear plan subjecs ti o t observatio supervisiod nan staty nb e authorities durin s entirit g e life- time. They must make sure tha operatine tth g company adhere o lawst s sa otheo welt s la r rule regulationsd san followd ,an s licensing conditions. Representative e supervisorth f so y bod expertr o y s appointe they n b d ca m ente timey plane an rth t ,ta inspec demand examind an tan , d it einfor - mation frolicense mth e holder. 4. Safety culture Radiation protection standards have been established setting limits to exposure which are not to be exceeded. Site selection and facility design ensure that any radioactive discharges in normal operation do not lead to exposure limit reasonabls sa beinw lo gs a y exceede e achievablar d an d e (ALARA principle) . The applicatio e ALARth f Ano principl e reliecorrece th n so t attitude of the staff. The correct attitude can be encouraged by the process of "Education for Safety Culture" and there is now some experience of this in the field of reactor operating training. The "Code of Practice", set up by a EURATOM Directive, spells out the need for training of "Radiation Protection Advisers". Safety is the quality of being unlikely to cause or occasion an injury. Safety culture is regarded as an important feature of operational safety at any installation and it should be an important part of radiation pro- tection practices. Safety culture seem o determinst limite eth safetf so y performance that can be achieved. An important component of nuclear lessonf o e us s e safetlearneth s i y d from mistakes viee restt .I th w n so that mistakes wi\Ll be made but the consequence will not be a total failure if some other person can be told how to avoid that mistake. Some for protectiof o m requires i n thosr fo d e reporting their mistakes. If this attitude is to take root, avoidance in reporting near misses in the interest of maintaining a good record must be seen as a serious fail- e safetth urn i ey culture e concep.Th f ALARnaturaa to s i A l componenf o t the safety culture. Safety culture is present in all aspects of radiation protection which attitude depen e th worker n th o d f educationae eo .Th l process must be sufficiently flexible to permit the behavior of the trainee observee b possibles o sa t modified r an d fa this n .a I d s respect, experience of realistic situations, the gaining of job experience and maturit e importantar y . 306 5. Quality assurance implementation High qualit n equipmeni y human i d n an tperformanc f utmoso s ei t impor- tance in nuclear plant safety. The process in which high quality is achie- ved are subject to control and verification by quality assurance practices. The primary rol qualitf eo y assuranc l activitieal n designinge ei th f so , procuremen constructiod an t n o phasassurt s i e e tha e involvetth d activi- tieproperle sar y organized, define thed an dn implemented firse .Th t acti- vity is to have in place defined QA requirements provided in official pub- lished technical inspection orders (e.g. KTA rules, DIN norms, VDI general directions e implementorth r )fo adhero st . Numeroueto s code standardd san s have been adopte r nucleafo d r use, after formulatio professionae th y nb l engineering communit d approva an ye appropriat th y lb e agencies. Approved codes have simultaneoueth s objective reliabilitf so safetyd an y . Thee ar y based on principles proven by research, past application and testing. The quality assurance entity must assure that therpropea s i e r program to verify tha thesl tal e regulatory requirement identifiede sar . Quality control and product control will then verify that they are being complied with. Quality control will be present during the construction and fabri- cation of the equipment and products. This includes the assurance that the materials have been properly selected and not damaged, as well as verification tha e itemtth s have been constructe fabricated an d accorn i d - dance with specified requirements. All safety related components, structure systemd san classifiee sar d basie o nth theif so r function significancd san e with regar safetyo t d . Quality assurance practices thus cover: - validation of designs - supply and use of materials manufacturing- , inspectio testind nan g methods - operational and other procedures to assure that specification mete s.ar The associated document subjece sar strico t t procedure verificationr sfo , issue, amendment and withdrawal. Formal arrangements for the handling of variation deviationd san importann a e sar t aspece processth f to . An essential component of quality assurance is the documentary veri- fication that tasks have been performed as required, deviations have been identified and corrected, and action has been taken to prevent the recurrenc errorsf o e . A ticklis he reviee implementatio th tasth s f i kwo produce th f no t quality assurance progra independenn a y b m t organizatio r commissionno , so-callee e.gth . d Product Contro le BfS Grouth , f o paccommodate KFAt a d- Jülich, or the Technical Control Board (TÜV = Technischer Überwachungs- verein). In the audits released by them, their staff will be examining selected technical products and procedures, such as a design or a site characterization test program with a team of technical and quality assurance specialists. The qualitwore assesseds th ki f o y , with particular emphasis giveo nt any QA program breakdowns that allowed problems to occur, in the case of observation audits, the inspection staff determines whether the audit team belonging to the BfS is adequately assessing technical products in addition to the product controls. It must be borne in mind that to adequately assess the quality of work in audits, the technical audit 307 team members must have a thorough understanding of the work themselves and they must give an account of their independence. Rules should be established which in performing chemical and physical analyses must guarantee that a service delivered by the analytical lab- oratory will be of consistently good quality. As examples the rules known as QA-EN 2900 r QA-IS0o e standardize quoteOe th b 900d y an d1ma d pro- cedures described therein applied individuae .Th l laboratoro t s ha y acquir qualitea y assurance (QA) certificate. Compliance wit presee hth t standard verifies si accrediten a y db d independent institutioy b r no the custome performiny rb g audit o chec st e accordancth k e applieth f eo d procedures with the standards and the efficiency with their procedures are followed. The advantages of such a quality assurance system for the customer are obvious. But also the executing laboratory may profit from it. The gai transparencn i n workinf o y g condition consequenca s sa cleaf eo r definition competencesf so , responsibilities workind ,an g rules leads increasn a o t n motivatioei efficiencyd nan . 6. Quality control of waste products and canisters Radioactive wastes properly conditioned and packed have to meet accept- ance criteria specific to a particular repository under the expected environmental conditions operatoe .Th nucleaa f ro r installation will make all efforts to satisfy these requirements since he bears responsi- bilit r compliancefo y . Therefore plane ,th t operato capacits hi n ri s a y the waste conditioner has to install an adequate control system of owns hi . Irrespective of such self-regulation, the BfS is charged with the duty of installing an independent redundant checking institution. The BfS has thirf o made dus e partie fulfilo st s dutielit s specifie offician a n i d l order /6/. It has a contract with the Research Center Jülich, KFA, for the performance of the quality control for radioactive wastes. Contractual assistance is also provided by the Federal Agency for Materials Research and Testing (BAM = Bundesamt für Materialforschung und -prüfung) and the Technical Control Board (TÜV = Technischer Überwachungsverein). The product quality control group supports the observance of the waste acceptance requirement followine th y sb g measures /?/: qualificatio- conditioninf no g processes - control of the conditioning processes and inspections rando- m alreade testth t sa y conditioned waste packages - checking of the documentation. The performance of quality control measures requires contracts between the waste producer whicn i conditionerd S hsan Bf detail e th f d o s san executio determinede ar n . The favored procedur employee b o et n futurmethoe i d th f process o dei s qualificatio wela f lo n instrumented conditioning process with subsequent process inspections. Process data from production operations will be used to determine the actual waste form and some of the characteristics of the canister. Operatio facilite th f no y mus carriee compliancn b ti t ou d e with the officially approved operation handbook. In the case of high-level waste glass product e compositiosth n wil determinee lb d primarily b y analyzing the liquid waste before melting and then establishing that it has been vitrified within known process operating parameters. 308 Analytical calculations will be used to acquire information such as the quantity of certain isotopes, canister temperature, and criticality prevention. Administrative controls wil usee r observanclb fo d specificationf o e s absence th suc s a restrictehf eo d materials, e.g. quantitie gasesf so , free liquids, explosives, pyrophorics, chemical toxicants, corrosives and fermentable reactants. Control e repositorth t sa y entranc dictatee ar e healty b d h physics sa well as legal accident prevention regulations. They comprise: visua- l inspectio waste th ef no packag e contro- f dimensionlo weighd san t - measurement of the dose rate and - control of surface contamination. The corresponding reports are included in the waste package information to fulfil waste lth e acceptance requirements importanl .Al t data arising at the waste producers, the control points and the repository are docu- mented in a central data library. 7. Management audit and inspection program The approach to ensuring that the waste treatment and final disposal facilities are designed and constructed in a manner that minimizes poten- tial contaminant releases depend e interactioth n so followine th f no g programs: - regulatory controls imposed during the period of site selection, facility design constructiod ,an n incorporated into licensing procedures and regulatory requirements engineeref o e -us d control minimizo st e effluent releases - monitoring program facilite th t sa y designe o providdt n a e early warnin unplannef go d release environmene th o st t - radiation protection programs, consisting of administrative and operating controls designe minimizo dt e worke publid ran c exposur o contaminanet t sources - procedure o mitigatst effecte eth accidentf so naturad san l catastrophes. The plant management should require that all principal work assignments are conducted in accordance with standard written procedures that include consideratio relevanf no t safety practices. Periodic management audits of procedural and operational efforts should be performed to maintain emission exposured san s obeyin ALARe th g A principle. Tabl liste1 e sth items that shoul reviewee b d audin radioactiva a f n to i d e waste management facility. 309 Table 1: Material to be reviewed in an "Audit" of a radioactive waste management facility 1. Safety meeting reports, e.g. fire contro chemicad lan l hazard protection systems, reduction of radiological source concentration 2. Control of liquid and airborne releases of radioactivity, recordin emissione th f o g s 3. Employee exposure records showing trendr sfo categorie workerf so s 4. Radiological monitoring program 5. Radiological survey and sampling data, contamination survey 6. Inspection log entries 7. Report n overexposurso workerf eo s 8. Operating procedures reviewed during the time period Documente. 9 d training program activities 10. Records of control equipment use, maintenance and inspection, emergency procedures REFERENCES /!/ IAEA, "Basic Safety Principles for Nuclear Power Plants, A Report by the International Nuclear Safety Advisory Group". Safety Series No.75-INSA6-3, International Atomic Energy Agency, Vienna (1988) /2/ Martens, B.-8. (Ed.), "Produktkontrolle radioaktiver Abfälle- Schachtanlage Konrad, Stand Januar 1994". Bundesamt für Strahlenschutz, Salzgitter, Report BfS-ET 19/94 (1994) /3/ Federal Ministr Interiore th f o y , "Gesetz übefriedliche di r e Verwendung der Kernenergi Schutd eun z gegen ihre Gefahren" (German Atomic Energy Act Julf , 1985o ) 15 y , BGBl.I,p.l565, last amendment Feb.18,1986, BGBl.I.p.265 / Federa/4 l Ministe Environmente th r rfo , Nature Conservatio Reactod nan r Safety, 'Verordnung über den Schutz durch ionisierende Strahlen" (Radiation Protection Ordinance) Oct.13,1976, p.2905, last amendment June 30,1989, BGB1.I, pp.1321-1375 310 /5/ Brennecke, P. and Warnecke, E. ."Requirements on Radioactive Wastes for Disposal (Preliminary Waste Acceptance Requirements, April 1990) Konrad Repository". Bundesam Strahlenschutzr fü t , Salzgitter (1990), BfS-Report ET-4/90 /6/ Federal Ministry for the Environment, Nature Conservation and Reactor Safety, BMU, "Richtlinie zur Kontrolle radioaktiver Abfälle mit ver- nachlässigbarer Wärmeentwicklung niche ,di einn ta e Landessammeistelle abgeliefert werden". Bundesanzeige (1989)Nr.631 r_4 a 111 Odoj, R., Warnecke, E., Martens, BrR., "Quality Control prior to Dis- Konrae posath n dli Repository". Kerntechni 1 (1987)104-105 k 7 311 LIS PARTICIPANTF TO S ALBANIA Mr. Feng, Shengtao Mr. K. Dollarn China Institut Radiatior efo n Protection Institut Nucleaf eo r Physics Taiyuan, Shanxi 030006 Tirana . TingMr , Weiguan ARGENTINA Shanghai Municipal Radioactive Waste . Pahissa-CampaMr . J , Disposal Experimentation Center Comision Nacional de Energia Atomica Roadu 209T e ,4 Xi Shangha i 200032 Av. Libertador 8250 Buenos Aires 1429 Luo. Mr , Shanggeng China Institute of Atomic Energy BANGLADESH PO Box 275(87) MD. Munsun Rahman Beijing 102413 Bangladesh Atomic Energy Coommission- 5 15 x Bo . O . P Mr. Ye, Yucai Ramna, Dhaka-1000 Tsinghua University Institute of Nuclear Energy Technology BELARUS 1021x Bo , BeijinO P g Zabrodsldi. Mr , V.N. Institute of Radioecological Problems Mr. Yun Guichun Academy of Sciences Tsinghua University Sosny, Minsk 220109 Institut Nucleaf eo r Energy Technology PO Box 1021, Beijing BELGIUM Mr. P. Debieve . ZenMr g Jishu Belgonucleaire China Institut f Atomieo c Energy Avenue Ariane 4 275(93x POBo ) Brussels B-1200 Beijing 102413 BRAZIL Zhang. Mr , Weizheng Mr. Miaw, S. T. W. Beijing Institute of Nuclear Engineering Centro de Desenvolvimento da Tecnologia Nuclear PO Box 840 Supervisao de Rejeitos Beijing 100840 Caixa Posstal 941, Belo Horizonte 30161-970 Mr. Zhang, Yinsheng CHINA Shanghai Municipal Radioactive . FanMr , Xianhua Waste Disposal Experimentation Center. China Institute of Atomic Energy 2094 Xie Tu Road, Shanghai 200032 PO Box 275(93) Beijing 102413 . WanMr g Zhiming China Institut Radiatior efo n Protection Mr. Fan Zhiwen 0 12 x Bo O P China Institut Radiatior efo n Protection Taiyuan, Shanxi 030006 Taiyuan Shanxi 030006 . WenMr , Yinghui Biejing Institut Nucleaf eo r Engineering PO Box 840 Beijing 100840 313 CROATIA Mr. Moncouyoux. P . J , Mr . Subasi.D c CEA Hazardous Waste Management Agency-APO Centr Valléa l e eRhond u ed e SavskaCesta41/IV 1 17 P B Zagreb 41000 Bagnols sur Ceze Cedex F-30207 CUBA GERMANY Mr. L. A. Jova Sed Mr. Merz, E. R. Center for Hygiene and Radiation Protection Forschungszenerun Juelich Calle 18A y 43 Miramar PO Box 1913 Apartado Postal 6094, Playa Juelich D-52425 Habana GHANA CZECH REPUBLIC Mr. E. O. Darko Mr. Holub, J. Radiation Protection Board Nycom Ghana Atomic Energy Commission Radiova1 PO Box 80, Legon 7 Prah2 2 a10 GUATEMALA Mr. Janu M. Mr. Rodriguez Jimenez, S. R. Nycom Direccion Genera Energie ld a Nuclear Radiova1 Avenida Petapa, 24 Calle 21 Praha 102 27 Guatemala 01812 EGYPT HUNGARY . Marei. Mr A . S , Mr. Czoch, I. Hot Lab. and Waste Management Centre Hungarian Atomic Energy Commission Atomic Energy Authority PO Box 565 101,El-kasr-El-Eini Street Budapes4 137 t Cairo INDONESIA Mr. Emara M. Mr. Yatim, S. Aatomic Energy Authority Radioactive Waste Manag. Techn. Center 101,El-kasr-El-Eini Street Tangeran0 g1531 Cairo IRAN. ISLAMIC REPUBLIF CO FRANCE Dehghani Tafti, A. Mr. Brosser. H . ,R Atomic Energy Organization of Iran NPPD Ministr Industrf yo y Nuclear Installations Safety Tandi7 . . AfiighNo sst . aAve Directorate Tehran BP6 Fontenay-aux-Roses F-92265 ISRAEL . BrenneMr . rS . FerniquJ-C . eMr Ministry of the Environmental Institute CEA/ANDRA for Environmental Research 31-33 rue de la Federation c/o Permanent Mission of Israel 75752 Paris cede5 x1 Anton Frank Gass0 e2 Vienna 1180 Austria 314 KENYA THAILAND Mr. D. Otwoma Mrs. Supaokit, P. National Radiation Protection Laboratory' Offic f Atomieo c Energ Peacr yfo e Radiation Protection Board Vibhavadi Rangsi. tRd PO Box 19841 Chatuchak, Bangkok 10900 Nairobi TURKEY KOREA. REPUBLIF CO Mr. Osmanlioglu, A.E. . H . J m Ki . Mr Cekmece Nuclear Research Korea Atomic Energy Research Institute Trainin& g Center Taejon Korea 305-353 PKIHavaalani Istanbul MALAYSIA . SakumaMr , Syed Hakimi USA Nuclear Energy Unit Mr. R. Ewing Ministry of Sei., Tech. & Environment Dept f Earto . Planetarh& y Sciences Komplek Pusspati, Bangi, Kajang 43000 Universit Mexicw Ne of yo Albuquerque Mexicw ,Ne o 87131-1116 ROMANIA . TurcanuMr , C.E.E. VDZTM NA , Institut f Atomieo c Physics Nguye. Mr n Thi, Nang Radioactive Waste Treatment Dept. Nuclear Research Institute PO Box MG-6, Bucharest R-76900 Dalat RUSSIAN FEDERATION YUGOSLAVIA Mr. Latypov. ,E Peric, A. Federal Nuclear & Radiation Safety Authority Institute of Nuclear Sciences RF Gosatomnadzor "Vinca" Taganskaya st.34, Moscow 109147 2 52 x Bo O P Belgrade SLOVAKIA Mr. L. Konecny ZAMBIA Nuclear Regulatory Authority Mr.MwaleK. Nuclear Installation Inspectorat Radiation Protection Board OkruznaS Ministry of Health , Trnav64 8 a91 Lusak, 3 x aBo O P SWEDEN IAEA Gustafsson. B , Saire. ,D Nuclear Fuel & Waste Management Co., 8KB NENF Waste Management PO Box 5864 0 10 x Bo O P Stockholm S-108 4 2 Vienna A-1400 Austria SYRIAN ARAB REPUBLIC Takrit. S . Mr i Tsyplenkov. ,V Atomic Energy Commission NENF Waste Management PO Box 6091 0 10 x Bo O P Damascus Vienna A-1400 Austria 315 QUESTIONNAIRE ON IAEA-TECDOCs It would greatly assist the International Atomic Energy Agency in its analysis of the effective- ness Technicalofits Document programme could ifyou kindly answer followingthe questions return address and formthe the to shown below. Your co-operation greatlyis appreciated. 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