INTERNATIONAL GR UP W RESEARCH REACt' 1£5 NOVEMBER 4- 5- 6 1996 AIX- EN- PRO VENL'E / FRANCE III h. ~tec nlcatome INTERNATIONAL GROUP ON RESEARCH REACTORS 5th MEETING NOVEMBER 4-5-6;1 1996 AIX-EN-PROVENCE / FRANCE GRAND HOTEL ROI RENE PROCEEDINGS PREFACE Overview The fifth meeting of the International Group on Research Reactors (lOORR-V) was held in Aix- en-Provence, France, on November 4-6, 1996. Attendance was excellent (between 70 and 80 participants from 21 countries). Thirty-one papers were presented in four sessions over the two- day meeting, and written versions of the papers or hard copies of the viewgraphs used are published in these Proceedings. A book of abstracts was published and handed out at the beginning of the meeting, which proved to be very useful to the participants. The meeting was a huge success, and we look forward with much anticipation to lOORR-VI. I should like to congratulate Klaus Boning on his election to Chairman of lOORR. I am certain that he and his colleagues will do an excellent job. It has been my pleasure to be associated with - the IGORR membership, and I hope to meet you all again some day. Kathy F. Rosenbalm OPENING SESSION Colin D. West - ORNL Yannick Le Corre - Technicatome Rene Ginier - CEA Welcome to the fifth meeting of the International Group on Research Reactors (lGORR). I should like first to thank our French colleagues, and the Commissariat A l'Energie Atomique (CEA) for their splendid efforts on organizing this meeting, which has a very strong international participation. I should also like to thank our Technical Program Coordinator, Kathy Rosenbalm, whom most of you know but who, unfortunately, could not be present. IGORR has become stronger, over the years since it was founded. Clearly, our organization has a useful role to play within our community of people working on new and improved research reactor facilities. Its' vitality is evidenced by the increased number of participants in this meeting and by the continued evolution and changes in our agenda. For example, in making the preparations for this meeting we found so much interest in and so many contributions concerning cold neutron sources that we have included for the first time a whole session devoted to papers on that topic. In addition, this year there are reports on two separate surveys that were initiated by the last IGORR meeting: one, 1;>y Albert Lee, on containment design criteria and one, by Doug Selby, on cold neutron cross sections. Another innovation this year, thanks to the CEA organizers, is the preparation, before the meeting, of a book of abstracts of the talks: I think we will all find this helpful in the next couple of days. Of course, the full Proceedings will be produced and published as soon after the meeting as practicable: please submit your contributions to the organizers on time. I should particularly like to welcome, for the first time, representatives of some of the institutes in former Eastern European countries. I am sure their experiences and ideas will be interesting and helpful to us all. Mter the technical sessions, there is to be a visit to Cadarache, which I know will be a new experience for many of us. In addition, I should like to thank very much the mayor of Abc, who has kindly and graciously invited us to a reception on Monday evening: a very nice welcome and gesture from our host city. The agenda is a full one, and full of interesting papers. The session chairmen have been instructed to keep speakers strictly to the time limit, so that their colleagues later in the session will not be deprived of their fair share of timel Let us begin IGORR-V. .&;1it ech n.l catome 1 IGORR 5 MEETING NOVEMBER 4, 1996 f?! WELCOME ADDRESS ~ M. YANNICKLE CORRE (TECHNICATOME Chairman) Dear Colin WEST, Dear Bertrand BARRE, Dear audience, I am really happy to welcome you for the 5th meeting of the International Group on Research Reactors : I am really happy because it's the second time since the creation of IGORR in Nineteen Ninety (1990), that TECHNICATOME has the pleasure to organise in a close collaboration with the Commissariat a I 'Energie Atomique the IGORR meeting in France. In Paris in June Nineteen Ninety Two (1992) IGORR Two (2) gathered Forty Five (45) participants. Today, in Aix-en-Provence, we are more than Eighty (80), including for the first time new representatives of some Eastern Europe countries. This shows the interest of the Research Reactor Community for lGORR activities. Thanks to Colin WEST for having created this group. I am also happy because many of you, coming from foreign countries will have the opportunity to discover the fascinating city of Aix-en-Provence, the «City of Arts ». Ilitechnicatome 2 The Nuclear Research Reactors have also an important past. They have played a key role in the history of the Nuclear Energy and my opinion is that they will still play an important part. More than Three Hundred and Seventy (370) Research Reactors and critical mock- ups were erected during the past Fifty (50) years in the world. More than Two Hundred and Eighty (280) are still in operation, and among them Eighteen (18) in France. Let me remind you some aspects of their contribution to the Nuclear Energy development: • with the first critical mock-ups, the feasibility of the fission nuclear chain reaqtion announced in August Nineteen Thirty Nine (1939) by Albert EINSTEIN to Franklin ROOSEVELT, on the basis of the work performed by Frederic JOLIOT, Enrico FERMI and Leo SZILLARD was demonstrated: the first reactor CPl went on criticality in Chicago on December Nineteen Forty Two (1942) . In France, ZOE started on December Nineteen Forty Eight (1948) ; It was the beginning of French Nuclear Reactor history which led to celebrate in Nineteen Ninety Five (1995) the Fifty (50) years of existence of the «Commissariat it l'Energie Atomique» with the background of one of the world biggest nuclear programs and a major contribution to the independence of our Nation. • The first generation of Research Reactors enabled the determination of numerous physical parameters and gave the confirmation of the behaviour of fission Nuclear Reactors, therefore giving the basis for the development of large nuclear power plants. • Afterwards many irradiation programs had to be performed in Research Reactors to qualify fuel and material technology for the future reactors. For this purpose sophisticated in core loops were also introduced in Research Reactors allowing fuel testing in representative thermodynamic conditions. • Up to now in the world, Research Reactors have demonstrated their extended possibilities : o in the field of fundamental and applied research, o in the field of material, component and fuel testing, o in the field of radioisotope production, o in the field of teaching and training. Ilitechnicatome 3 • Very early safety aspects have also arisen, leading to specific programs on dedicated facilities, like CABRI and PHEBUS in France. CABRI allowed the study of the cinetics of cores in case of sudden reactivity introductions. On PHEBUS, the consequences of severe degradations of fuel are studied under an international collaboration. I know that you will visit these facilities in Cadarache on November Six (6). • For the scientific community, Research Reactors appeared to be also powerful tools for fundamental research and medicine applications. In the Nuclear developed countries, the high level of demands led to the specialization of the Research Reactors. In France, it's the reason why after the first generation of multipurpose reactors like MELUSINE, SILOE and then OSIRIS, specialized high flux reactors were built, in Nmeteen Seventy Two (1972) the RHF in Grenoble, and in Nineteen Eighty (1980) the ORPHEE reactor in Saclay. In a complementary way, OSIRIS and SILOE activities were focused on irradiation experiments and radioisotope production; both still operate with heavy programs. What's future for the Research Reactors? Today we have to acknowledge than most of existing Research Reactors are of pool type. They proved their versatility and their easy adaptability to new irradiation programs. High flux reactors obtain higher fluxes with the help of an additional reflector made of Beryllium or heavy water. Pool type Research Reactors appear to be a good investment on the condition that enough attention is payed to their design, taking into account the experience acquired by the years of operation of the previous Reactors. In the countries who developed for many years a nuclear program, a high level of demand exists with a resulting specialization of the research facilities and with even more high neutron flux requirements. There are needs for: • on one hand, irradiation reactors for qualification of fuel and materials including fast reactors and fusion programs, and radioisotope for medicine and industry, • on the other hand, reactor for neutron beam delivery with a growing demand for cold neutrons. , In a complementary way, safety assessment studies will certainly need the continuation of ongoing programs on dedicated facilities. III h. ~tec mcatome 4 For the countries who recently decided to develop a Nuclear Research or Energy program, a multipurpose pool type Research Reactor is certainly a good choice if it is done in association with a coherent development policy for the national scientific community. This kind of Research Reactor will allow access to most of the applications I mentioned before. The France who successfully managed a huge electronuclear program has developed the highest level of expertise in Research Reactor design, as well in CEA as in the associated industrial partners. Numerous collaborations with nuclear developing countries are on the way now and will benefit from this high skill level. TECHNICATOME, with CEA, is willing to contribute to this effort. Mister Chairman, I do believe that Nuclear Energy will remain a key technology for the future. Research Reactors can be in the next decades useful and versatile tools, contributing to this future in the frame of qualification programs for fission and fusion energy. The Reducing Enrichment in Research and Test Reactors program, I mean RERTR program, aiming at demonstrating that most of existing Research Reactors can be converted to low enriched uranium silicide fuel, with a comparable level of performances, is of a great interest, opening the field for wide peaceful applications. New projects will of course benefit from this new fuel technology. In France, we are preparing the necessary renewal of the existing irradiation Research Reactors. Mister Chairman, dear audience, there are still a lot of things to do ; doubt and humility are necessary in any scientific approach. After years of success the major error would be to believe that all technical aspects have been solved. Chemobyl and TMI are here to remind us this reality. Fighting against high economical constraints, our countries need to maintain for the future a scientific and engineering community with the highest skill level in the nuclear field. I am sure that lGORR, facilitating exchanges between countries, is working in that direction for the world scientific development. I wish you three interesting working days of technical debates and visits. -/. Mister Cbainnan, Ladies and gentlemen "- " I am very glad to have the opportunity to say some words for introducing this looRR meeting_ But before that, I have to apologize Mr Barre for not being here as it was anticipated. To day he has to be in Paris and as Deputy Dr I will do my best to represent him here. Our Direction , the Nuclear Reactor Division is in deed strongly interested in the looRR works, firstly because we are operating a certain number of research reactors, OSIRIS and ORPHEE in Saclay, SILOE in Grenoble, PHENIX in Marcoule, as well as critical mock-up like EOLE and MASURCA, in Cadarache, but also because we are great users of results obtained with all .these Research Reactors. ''Nuclear energy will continue to playa central role in french energy policy". This strong declaration is not from me, ladies and gentlemen, although I quite agree with it, but was made recently by the french industry minister. To support this declaration there is a lot of good reasons but there is to my feeling two basic conditions : public acceptance and competitivity Public acceptance, great efforts are made to-day in direction of the public, in conferences, through media. but one of the best way to get the public acceptance is to show that power reactors can be operated without any significant accident. To prevent such event a lot of works has been done and is still under way on experimental loops or experimental reactors like Phebus or Cabri that some of you will visit after- to-morrow. There, in the frame of international programmes, are performed tests to determine the limits of service of fuel elements under accidental transients. The second condition, competitivity will be, more than ever, the key factor. Due to the progress made by other sources of energy, nuclear energy has to improve its own competitivity. All the constructors are giving a high priority to these aspects: extending the life of the reactors to 40, 50 even 60 years, getting better availability through longer fuel cycles and reducing fuel costs with higher bum up rates. All possibilities are explored and, nuclear designer will demand more and more to the materials: fuel. cladding, internals material. We must better understand their comportement under irradiation, better predict their evolution under transients and make the better choice for these materials. New need appear to-day related to the necessary stabilization of plutonium stock-piles, transmutation of minor actinides. To achieve these goals important programmes are under way in experimental reactors like OSIRIS and SILOE and will be pursued in the next years. However the neccesity of economy applies also to research reactors and our Division will concentrate all the programmes in OSISRIS in Saclay, the Siloe reactor being shut down at the end of the next year. The importance of nuclear energy in France justifies that CEA continues to have to its disposal research reactors specially for material testing. It's the reason why we are soon thinking to the following reactor after OSIRIS which will be in operation still about a decade_ You will have to-morrow a first presentation of this future reactor ; it has soon a name : Jules Horowitz Reactor, from the name of a prominent physicist of CEA who died last year_ J. You see that we are confident in the future of nuclear energy : nothing is definitly acquirrecl, we' have to compete but we will give us the b~st tools in order to maintain the economic advantages of nuclear energy In that perspective, research reactors have an important role to play and I am sure that lGORR will contribute to draw a better profit of our tools, I wish to lGORR a fruitful meeting Thank for your attention, AGENDAS OF THE MEETING OVERALL AGENDA 1GORR 5 SUNDAY NOV 3 I8HOO I 20HOO J Registration MONDAY NOV 4 08H30 09H30 Registration 09h30 IOh15 Opening session • Opening IGORR 5 C.WEST ORNL • Welcome speech Y.LECORRE TA R. GINIER speaking for B. BARRE CEA/DRN Practical information IOhI5 10h30 Coffee break 10h30 12h30 Session lA Operating research reactors (6 papers) I2h30 14hOO Lunch 14hOO 16hOO Session 1B Operating research reactors (5 papers) I6hOO 16h15 Coffee Break 16hl5 IThOO Session IC Operating research reactors (2 papers) IThOO 17h45 Session2A Research Reactors in design. or (2 papers) construction 19h30 Reception at the City Hall TUESDAY NOV 5 08h30 IOh30 Session2B Research Reactors in design. or (5 papers) construction IOh30 IOh45 Coffee Break IOh45 12h30 Session3A Cold Neutron Sources (5 papers) 12h30 14hOO Lunch I4hOO 15hOO Session3B Cold Neutron Sources (4 papers) 15hOO 16hOO Session 4 Workshop on containment survey (2 papers + discussion) 16hOO 16h15 Coffee break 16h15 17h15 Session 5 Workshop on R&D needs ITh15 18hOO Business meeting and Closing session C:\I996\igonvlbasdonlagenda CEA CADARACHE TECHNICAL VISIT PHEBUS - CABRI - CASCAD - TORE-SUPRA FACILITIES 1 Wednesda.y November 67 19961 8.45 a.m. • Meeting point at the hotel 9.00 a.m. • Departure to Cadarache 9.45 a.m. • Arrival at Cadarache 9.45-10.00 a.m. • Formality requirements 10.00-10.30 a.m. • Welcome speech 10.30-10.40 a.m. • Bus trip to Phebus Group 1 Group 2 10.40.10.50 a.m. • Bus trip from Phebus to Cascad 10.40-11.30 a.m. • Visit of Phebus 10.50-11.25 a.m. • visit of Cascad 11.25-11.35 a.m. • Bus trip from Cascad to Tore Supra 11.35-12.20 p.m. • Visit of Tore Supra 11.45-12.30 p.m. • Visit of Cabri 12.20-12.30 p.m. • Bus trip from Tore Supra to Cabri 12.35-12.45 p.m. • Bus trip to lunch 13.00-14.30 p.m. • Lunch 14.30-14.40 p.m. • Bus trip from lunch to Phebus 14.40-14.50 p.m. • Bus trip from Phebus to Cascad 14.40-15.30 p.m. • Visit of Phebus 14.50-15.25 p.m. • Visit of Cascad 15.25-15.35 p.m. • Bus trip from Cascad to Tore Supra 15.35-16.20 p.m. • Visit of Tore Supra 15.45-16.30 p.m. • Visit of Cabri 16.20-16.30 p.m. • Bus trip from Tore Supra to Cabri 17.00 p.m. • Departure (Arrival in Aix en Provence at 17.45 p.m.) IGOW DETAILED AGENDA Page: 1 1 OPERATING RESEARCH REACTORS 4h45 SESSION lA - Chairman Bernard Farnoux IS HFIR UPGRADE PROPOSALS CD WEST-MD. FARRAR OAK RIDGE NATIONAL ORNL USA LABORATORY 6 OPERATION EXPERIENCE AND CURRENT STATUS OF HANARO KlL LEE, HR. KIM - BJ. ruN - KOREAN ATOMIC KAERl KOREA JB.LI!E ENERGY RESEARCH INS1TIUI1! 13 CURRENT STATUS OF RESEARCH REACTORS RA AND RB AT 1HE M. KOPECNI - M MATAUSEK- VINCA INS1TIUI1! OF VINCA YUGOSLAVIA VINCA INSTITUTE OFNUCLEAR SCIENCES D. STEFANOVIC NUCLEAR SCIENCES 9 STUDSVIK.'S R2 REACTOR - REVIEW OF RECENT ACTlVlTIES M.GROUNES- STUDSVIK NUCLEAR STUDSVIK SWEDEN C. GRAsLUND- M CARLSSON- AB T. UNGER-A.LASSING 28 MAIN EXPERIENCES IN RENOVATION OF 1HE DALAT NUCLEAR TRANH.A. - PHAM VoL- NGUYEN INSTITUTDE IRN VIE1NAM RESEARCH REACTOR N.D.-NGOP.K. RECHERCHE NUCLEAIRE VIE1NAM 2 UPDATEON TIlEBR2 REFURBISHMENT E.KOONEN STUDm CENTRUM SCKlCEN BIlLGlUM VOOR Kl!RNENERGJIl I SESSION IB - Chairman Christian Desandre I 27 RESEARCH AND SERVICES OFLVR-lS REACTOR IN REZ J. KY.ML4 - O. ERBEN - NUCLEAR RESEARCH NRlR CZl!CH V. KNOBLOCH -1. BURIAN -V. BROZ INS1TIUI1! REZ PLE REPUBUC 1 H1FAR MAIORSHUTDOWN REPORT SoIIlM AUSTRAUAN ANSTO AUSTRALIA INSPECTION OF REACTOR ALUMINIUM TANK AND REPLACEMENT OF NUCLEAR SCIENCE SECONDARY COOUNG WATER CIRCUIT PIPEWORK AND TECHNOLOGY ORGANIZATION 26 MODIFICATION OF JRR-4 To NAKAJIMA - M BANBA - 1AERl 1AERl JAPAN Y. FUNAYAMA - Y. HORIGUTI- MlSSH1KI 23 OVERVIEW ON TIlE MAIN ENGINEEllING WORKS PERFORMED ON P. ROUSSELLE - G. DE SAINT OURS TECHNlCATOMFJCEA TAiCEA FRANCE FRENCHRESEARCH REACTORS 1HESE LAST YEARS J. GUIDEZ - C. JOLY - M MAZIllRE H.GUYON 19 UTILIZATION OF TIlE BUDAPEST RESEARCH REACTOR L I'lDOI'SZKY KFKl ATOMIC ENERGY KFKl HUNGARY RESEARCH INS1TIUI1! SESSION Ie - Chairman Kir Konoplev 7 DESION MODIFICATION OF HANARO REFLECTORCOOUNG SYS1EM J.So WU-S.Y.HWANG- Y.K.KIM KOREAN ATOMIC KAERl KOREA ENERGY RESEARCH INS1TIUI1! 35 !ABA ACTlVlTIES ON RESEARCH REACTOR SAFETY F.ALC4L4RUlZ INTERNATIONAL !ABA VIENNA ATOMIC ENERGY AGENCY Bold & italic name = speaker .:II996\igomo\badonltimingdoc ~0Bt0 DETAILED AGENDA Page: 2 1 RESEARCH REACTORS IN DESIGN OR CONSTRUCTION lh45 SESSION 2A - Chairman Albert Lee 21 STATUS OF 1HE TRR II PROJECT LF. LIN - WM. ClDA- INSTIT1JTE OF lNER TAIWAN CY. YANG - US. SHEU - NUCLEAR ENERGY ROC CC. WANG-DISHIEH- RESEARCH 10 THERMOHYDRAULIC AND MECHANICAL ANALYSIS OF A J. ADAMEK - H. UNGER RUHR UNIVERSITAT RUB GERMANY SCALE FRM-ll CORE DUMMY BOCHUM SESSION 2B - Chairman Jean-LuI: Minguet 4 A STATUS REPORT ON 1HE PROPOSED CANADIAN AG. LEE - W.E. BISHOP - ATOMIC ENERGY AECL CANADA IRRADIATION RESEARCH FACILITY G.E. GILLESPffi OF CANADA, Ltd 20 PRELIMINARY STUDY OF CORE CHARACTERISTICS FOR JT. YANG-LS.KAO, INSTIT1JTE OF lNER TAIWAN ROC TRR-ll HM.USffiH-DY. YANG NUCLEAR ENERGY JA JlNG - SK CHEN RESEARCH 12 STATUS OF 1HE FRM-ll PROJECT HJ. DIDIER - G. WffiRHEIM SffiMENS AG KWU SffiMENS GERMANY 14 CONSTRUCTION OF 1HE HITR AND ITS IRRADIATION M. ISHIHARA - T. KIKUCIll, JAPAN ATOMIC JAERI JAPAN PROORAM J. AIHARA - H. MOOI -T. ARAI, ENERGY RESEARCH T.TANAKA INSTIT1JTE 18 1HE JULES HOROWITZ REACTOR (R.1.H.) S. FRACHET - P. MARTEL, COMMISSARIAT A CEA FRANCE 1HE CEA FUTURE TOOL FOR TECHNOLOGICAL B. MAUGARD - P. RAYMOND- L'ENERGffi IRRADIATIONS F.MERC1IIE ATOMIQUE CADARACIffi __ - --- -- ---- -- ---- --_ .. ------- -- Bold & italic name = speaker c:\I~\tImIns.doc - ----- - ------ ---~--- ~ow DETAILED AGENDA Page: 3 3 COLD NEUTRON SOURCES 02h45 SESSION 3 - Chairman Hans Joachim Roegler 3 EXPERIMENTS WITIl COLD SOURCES FOR NEUTRON PHYSICS K A. KONOPLEV - VA PETERSBURG PNPI RUSSIA ANALYSIS KUDRJASROV - G.D. PORSEV - NUCLEAR PHYSICS I.A. POTAPOV - VA TRUNOV- INSTITIJTE GJA VASILEV - A.S. ZAKHAROV 8 THE COLD NEUTRON SOURCE AND OTHER IN-PILE K.. GOBRECHT- TECHNICAL GERMANY EXPERIMENTAL FACILITIES OF THE NEW RESEARCH REACTOR E. STEICHELE UNIVERSITY OF MON~ FRM-n IN GARCHING MUNICH 11 POST IRRADIATION EXAMINATION OF Z6 NCT 2S STAINLESS M.MAZIERE COMMISSARIAT A CEA FRANCE paper presented by STEEL FROM SF2 COLD SOURCE CELL OP THE REACTOR ORPHEE L'ENERGm B.FARNOUX ATOMIQUE SACLAY 33 SUMMARY OF HFIR COLD SOURCE PROJEci"s D.SELBY OAKRJDGE ORNL USA NATIONAL LABORATORY 37 PELLImZED MElHANE COLD SOURCE CONCEPT A. T. LUCAS - PERFORMED BY D. OAKRJDGE ORNL USA SELBY NATIONAL LABORATORY 29 PERFORMANCE OF THE NIST UQUID HYDROGEN COLD SOURCE 1M. ROWE - P. KOPETKA- NATIONAL NIST USA RE. WILLIAMS INSTJ.TUrE FOR STANDARDS AND TECHNOLOGY 32 DESIGN REVIEW IN CONCEPT PHASE OF CNS AT HANARO CO. CHOl- KN. PARK, JM. SOHN, KOREAN ATOMIC KAERI KOREA SH. PARK, MS. CRO ENERGY RESEARCH INSTJ.TUrE . 34 COLD NEUTRON CROSS SECTIONS - HISTORICAL REVIEW D.SELBY OAKRIDGE ORNL USA NATIONAL LABORATORY 36 GKSS - CNS : THE POSSIBIUTY OF NATURAL CONVECTION OF THE Iv. KNOP - W. KRULL GKSS GKSS GERMANY GASEOUS HYDROGEN MODERATOR 4 SESSION 4 -WORKSHOP ON CONTAINMENT SURVEY - Chairman Robert Williams OlhOO 5 RESULTS OF A SURVEY ON THE DESIGN BASIS FOR RESEARCH A.G.LEE ATOMIC ENERGY OF AECL CANADA REACTOR CONTAINMENT/CONFINEMENT BUILDINGS CANADALdt 22 FRENCH RESEARCH REACTORS - DESIGN OF REACTOR BUILDING P. ROUSSELLE - JL MINGUET- TECHNlCATOME TA FRANCE IN ACCORDANCE WITIl SAFETY APPROACH AND IAEA F.ARNOULD RECOMMENDATIONS i 5 WORKSHOP ON R&D NEEDS AND RESULTS- Chairman Klaus Boning OlhOO ~ 6 . II BUSINESS MEETING & CLOSING SESSION II OlhOO ~ Bold & italic name = speaker o:I1996\lptrv\basdoltimlns-dnc SESSION 1 OPERA TING RESEARCH REACTORS PAPERS ~- ~-~ HFIR UPGRADE PROPOSALS c. D. West and M. B. Farrar Oak Ridge National Laboratory Oak Ridge, Tennessee, USA presented to the 5th meeting of the International Group on Research Reactors November 4, 1996 ORNL-OWG 96Z-696B Why Upgrade The HFIR? • There are important scientific and isotope production needs that can only be met with a high-powered research reactor . - This has been recognized by all major reviews of the national needs for neutron sources • The HFIR has the highest neutron flux (the single most important measure of a neutron source's capabilities) of any research reactor in the western world • The HFIR is a multipurpose research reactor, and its upgrade will benefit many areas of science - Neutron scattering - Isotope production - Neutron activation analysis - Materials irradiation testing - And others • With minor improvements, the HFIR can operate for many more years (at least until 2030) ------ ORNL-DWG 96Z-6969 Besides Possible Improvements To Existina CaDabilities. What M · Q. Q P pi '( , I !.fI~t I . ..L.JI'-': 0 i o;;ut;:::'~I .. ncc ~ "b" • Cold neutron beams - Cold neutrons (low energy, long wavelength neutrons) have become increasingly important as neutron scattering science has tackled more and more subtle and complex problems - HFIR is the only major research reactor in the world without a cold neutron source (which would boost the flux of cold neutrons by an order of magnitude) • Instrument space - All new and recently modernized research reactor facilities have large halls or beam rooms to accommodate many big scattering instruments and their shielding - HFIR was originally designed and optimized for isotope production and has only' very limited space, inside the reactor building, for instruments ORNL-DWG 96Z-6970 Other Desires In A HFIR Upgrade Program • Schedule the upgrades to minimize risk, cost and down time • Coordinate with other neutron facility improvements to provide continuous service to user communities • Capitalize on our ANS experience (reactor and instrumentation design, cold source development, thermal neutron guides, etc.) ORNL-DWG 96Z-6971 A HFIR Futures Group • In the spring of 1995, a HFIR Futures Group was formed to consider enhancements to the reactor's performance and capabilities • The group, representing the interests of all parties at the laboratory, met over a period of several months • The group, with much helpful input from many colleagues, prepared the basis of an upgrade proposal package that ORNL later presented, in a slightly modified form, to DOE's Basic Energy Sciences Advisory Committee - The committee endorsed these proposals and recommended their implementation ORNL·DWG 96Z·6972A Summary of Proposed HFIR Upgrades • Return to 100 MW operation • Life extension improvements: Be replacement/upgrade, etc. • Cold source: New beams & new beam hall: cold neutron intensities > ILL • Thermal neutrons: 5 new thermal beam guides & new guide hall. Thermal neutron flux at samples > 5xlLL (using new ideas to match the area and shape of the source to the acceptance angles of the guides) • Isotope production: Improved spectrum and access • Materials irradiation and testing: Improved access • Neutron activation analysis: Improved access and handling, more irradiation space • Remote handling facility for processing, maintenance and packaging -a c .ta. .. •• ••••• •~c. •.•: '=· , .0U..0u. . taG) """ - DG:)1- .... Ie :e .--a:.:-l en= ra-t' .2.1 BRIEF DESCRIPTIONS OF UPGRADE, COLD SOURCE, AND BERYLLIUM CHANGEOUTnsIPROPOSALS Create Detailed Monte Carlo Model A detailed Monte Carlo (MCNP) model of the High Flux Isotope Reactor (HFIR) is essential for the evaluation of any proposed modifications that impact the reactor assembly. The model has been completed and will be modified and updated as appropriate in the future. Prepare Safety Analysis for 100 MW Operation By selection of fuel elements with favorable channel dimensions from the present inventory. and later by adopting improved fuel plate inspection/selection procedures that were developed during the ANS Project., it will be possible to operate HFIR at 100 MW again with the present coolant inlet pressure and temperature while maintaining safe operating conditions for the embrittled reflector vessel. This will increase the neutron flux by 18%. benefiting all the missions of HFIR. Cold Neutron Source HFIR is the only high-power research reactor in the world whose capabilities do not include a source of cold (Le., very low energy, long wavelength) neutron beams: such beams are now used for much of the most important neutron scattering research and basic nuclear physics experiments at the reactors. The liquid hydrogen cold source to be installed in HB-4 will provide gain factors of 5 to 30. depending on the wavelength, in the low energy neutron flux available at HFIR. The intensity of the new beams will be as high as or higher than those available anywhere in the world. although geometrical and space limitations will only allow three or four beams. at most, to be extracted, compared with at least eight at ILL. New Beam Tubes The present beam tubes at HFIR are about 4 in. in diameter, subtending only a small solid angle at the core, which greatly restricts the number of neutrons that make it all the way down the tube without being lost in the walls. Most scattering experiments would benefit if the tubes were flared between the reactor pressure vessel and the concrete shield around the pool to give a larger area and therefore pass more neutrons. Enlarging the beam tubes inside the reactor pressure vessel. to the extent allowed by the vessel openings and the core support structures, will also increase the number of neutrons available for scattering experiments. Harden Spectrum in Flux Trap The central hole in the HFIR core, a region of about 5 in. in diameter and 20 in. deep, is called the flux trap or target region and has the highest steady-state neutron flux in the world. Hydrogen in the light water coolant effectively slows down (moderates) neutrons released from fission reactions in the core, so that the neutron energy distribution in the trap is that of any material at a temperature of around 160"P. However, it is now known that a slightly higher energy (faster, "harder." or more epithermal) spectrum would be more effective in producing some of the transuranic elements. Displacing some of the water in the trap with solid aluminum rods - aluminum is a less effective moderator than hydrogen - would give a harder spectrum and, hence. increase the transplutonium production rate. This is a very simple, low-cost enhancement., and experiments are already underway to measure the benefit. Another approach under consideration is to fabricate the transplutonium production capsules from stainless steel, which would absorb thermal neutrons. leaving a harder spectrum within the capsule, instead of aluminum as at present Multiple Hydraulic Rabbit Tubes The optimal production of some specialized isotopes requires that the target material be irradiated for only a fraction of a normal operating cycle. The existing hydraulic rabbit facility provides space for up to nine rabbits (2-in. long capsules) in the highest flux region of the reactor; its primary purpose is the production of medical R&D isotopes and some production medical isotopes. More such facilities are needed as demand increases, and a new magazine-fed multiple-tube hydraulic rabbit facility design will allow the irradiation of rabbits in six positions. This will increase the production capability from 9 to 54 rabbit capsules. Rabbit Holders in TarGet ReGion Special "peripheral target assemblies" to be inserted in one of the highest thermal flux regions of the reactor, are to be designed and fabricated to hold up to eight "hydraulic rabbit capsules." These targets will allow the irradiation of isotopes that are needed in some small quantities without the fabrication of a full-size target: the method will be most useful in the production of medical isotopes, such as Sn-117, that require irradiation for more than one operating cycle. The shipping and processing costs of the rabbits are lower than for full size targets. The rabbits are also much easier to handle and ship than larger targets, resulting in a quicker response to demand, which will result in a more cost-effective and responsive medical isotopes distribution program. Neutron Activation Analysis Pneumatic Rabbit Tubes There are currently two NAA rabbit tubes located in the HFIR core. The current facility is limited in the volume of the sample that can be irradiated. A new design will 'provide larger flight tubes and loading stations to handle larger rabbits. This will allow the irradiation of larger samples or even multiple samples in each rabbit Additionally, provisions for delayed neutron counting will be added. Neutron Radiographyffi>mography Facility In tomography, images or radiographs of successive planes in an object are combined to give a complete, three-dimensional view (CAT scanning is a familiar example from the medical world). The laboratory directed research and development (R&D) funding has been granted to demonstrate neutron tomography, which will add to the existing X-ray tomography technique the penetrating power of neutrons and their ability to "see" light materials such as lubricants or inclusions. For example, it will be possible to observe, by electronic gating of the signals, the changes taking place inside an operating engine at different times during the cycle. Assuming the demonstration is successful, a world-beating operational facility will be installed as part of the upgrade program. New Gamma Irradiation Facility The spent fuel elements from HFIR. are highly radioactive, producing copious gamma rays, and even though the activity declines fairly quickly, they have been used as a source for testing the effects of gamma irradiation on small quantities of various substances. The control cylinders, movable neutron absorbers between the core and the beryllium reflector to control or shut down the reactor, are somewhat less radioactive at the end of their useful life than a fuel element, but their radioactivity decays more slowly. A new facility, to be inserted into a spent control cylinder, will provide a high flux (106 R/hr) irradiation capability for tests on larger samples. New Neutron Scattering Instruments Existing scattering instruments will be improved and new instruments (e.g., a small angle neutron scattering machine and a residual stress analyzer) constructed to take full advantage of existing and new beams from HFIR. Monochromator Shields Monochromators are used (like a prism or diffraction grating for light) to select, and send in a desired direction, only neutrons with the energies or wavelengths desired. Other wavelengths are deflected in other directions, and so the monochromator location must be heavily shielded. We have been awaiting replacement of the present shields, which are 20 years old and have a very restricted aperture, for ten years. In addition to the beam size restrictions, the mechanical components, such as bearings, are worn. The monochromator shields need to be replaced regardless of any other upgrades. HB-2 Guide HaIl This project will put five thermal neutron guides into the existing HB-2 beam port to bring neutrons from close to the core out to a new guide ball that will have space for up to 15 instruments. HFIR already has the world's most intense beams of thermal neutrons, and this upgmde will make them brighter still and with lower background because of the inverse square weakening of the gamma and fast neutron contamination down the long guides. Remote Handline Facility At present. there is no remote handling (hot cell) at HFIR, and irradiated materials, (which include isotope targets, materials irradiation capsules, and assorted scrap materials) must be loaded directly into shipping casks positioned underwater and suspended from a crane. With a remote handling facility at HFIR, irradiated materials could go directly into a hot cell from the pool, and specimens or target materials could be retrieved and prepared for their final destination, thus bypassing the shipment to a disassembly cell. Excess hardware and scrap materials could be shipped directly to tbe ORNL burial ground facilities. If possible within the cost limits, the remote handling facility would also be designed to accommodate examination of control plates and canning of spent fuel prior to dry stomge. The addition of this facility would improve safety and reduce person-rem exposures attributable to bandling radioactive materials at HFIR and allow faster and more cost-effective handling of materials irradiation experiments and isotopes. Beryllium Redesign for Plutonium Production DOE has a need for an isotope of plutonium, Pu-238, to be used as a heat source for the production of electricity in remote locations and in space. Funding has already been received to redesign the vertical holes in the beryllium reflector to accommodate production of this, or other, isotopes. When the beryllium is next replaced (in FY 1999), the new reflector will be built to this modified design. In-Senrice Ipspection and Ber:yllium Chan~Qut This task must be carried out. regardless of upgrades, to continue operation beyond FY 1999. DOE Orders require an in-service inspection (IS1), and radiation embrittlement will necessitate another routine changeout of the beryllium reflector. A delay in the lSI may be requested so that it can coincide with the reflector replacement, minimizing cost and reactor downtime. Neutron Sciences SUllll0rt Building During every changeout of the "permanent" beryllium reflector, research equipment in the HFIR building must be physically removed to a safe climate-controlled storage place. In the past, trailers and other temporary storage have been used, but with present concerns and regulations concerning contamination, that is no longer an attmctive option. We propose to erect a preengineered building to serve this purpose for the 1998-1999 changeout and for later ones. Ip between these operations, the building would be used as space for additional beam scattering instruments on the HB-4 (cold neutron) beam line. Replace Iransfonners The existing transformers are oil filled and were identified at least as long ago as a 1990 DOE Oak Ridge Operations appraisal as in need of replacement as an environmental and fire hazard. Funding requests for this action have been made but were preempted by other, higher priority needs. Replacing these transformers with others, placed at another side of the HFIR electrical building, will also free up valuable space that can be added to the equipment support facility/experimental area. Other Reactor Improvements Given that the HFIR is to be upgraded and to give a further scientifically useful life of 35 years or so, aging electrical, mechanical, and control systems will be replaced or refurbished to keep the facility operating in a safe and reliable manner. General Notes All of the above improvements are independent of each other, and each would be beneficial on its own (although there are synergies - the whole upgrade is naturally worth more than the sum of the parts). Likewise, the two buildings (the Neutron Sciences Support Building, available for experimental space between beryllium changeout operations, and the HB-2 guide hall) and the remote handling facility are valuable improvements on their own but are also synergistic with other enhancements. Operation Experience and Current Status of HANARO Kye Hong Lee, Hark Rho Kim, Byung Jin Jun and Ji Bok Lee HANARO Center, KAERI P.O. Box 105, Yusung, Taejon, Korea, 305-600 ABSTRACT It has been almost two years since the first fuel loading at HANARO. Presented in this paper is the operation experience through the reactor commissioning and power operation during that period. The current status of utilization and installation of utilization facilities is summarized and some of research projects are introduced. INTRODUCTION To be a safe and useful research reactor, a lot of activities are going on In HANARO. The physics parameter measurement continues following the nuclear commissioning for safe operation and understanding reactor characteristics in parallel with the calculations for comparison. Maintaining operator resources would be a basis to have operators with high quality. In a sense that operator training program is valuable for safe operation, groupwise self-education and procedure review program not only broaden operator's knowledge but also revise operation procedures from experience. Due to the decommissioning of TRIGA n & m, the radioisotope production in HANARO, only remaining research reactor, for domestic demand is absolutely required. Neutron beam facility, irradiation facility for NTD (Neutron Transmutation Doping), and FTL (Fuel Test Loop) facility ins~lation are in progress although those works sometimes require reactor shutdown. The first trial of reactor material test using non-instrumented capsule has been completed and is under PIE (Post-Irradiation Test) at IMEF (Irradiated Material Examination Facility). The design of instrumented capsule will be completed by July 1997. As well as the safe operation of HANARO, the reactor operation team involves in various projects for better environment of HANARO users and supports the utilization and development team. BNCT (Boron Neutron Capture Therapy), DUPIC fuel irradiation test analysis, IPS (In-Pile Section) design for fuel pin irradiation test, TRIGA decommissioning, etc. are the present project items. The number of visitors to HANARO is over five thousands in 1996, which shows that HANARO contributes to nuclear public acceptance. CHRONOLOGY After the successful approach to the initial criticality with 17 fuel bundles on Feb. 8, 1995, the first operation core was configured by loading seven additional fuel bundles on May 10, 1995. Everytime a fresh fuel bundle was loaded, the control absorber rods (CARs), shut-off rods (SORs), and fuel bundle worth as well as the critical CAR position were measured and compared with the predictions. With this first cycle core, zero power tests, power ascension tests, and long-term operation tests were followed consecutively. Zero power tests include noise analysis, and the measurement of void coefficient, fuel bundle worth, thermal and fast flux distribution, assemblywise power distribution, gamma dose rate distribution, photoneutron effect, and isothermal temperature coefficient. Starting with natural circulation test, thermal power, transfer function, loss of electric power test, decay heat, power defect and xenon effect measurement were carried out in power ascension tests. Loading of four more fuel bundles led to the second cycle core on June 25, 1996. In order to have an appropriate excess reactivity and CAR position for the measurement of CAR worth by swap method, four fuel bundles were loaded in two times two at a time. In September, two more fuel bundles were loaded and the third cycle started with the operating power of 24 MW. The fourth cycle is scheduled to start in January of 1997. At each BOC and EOC, several important physics parameters were measured. The tests include the measurements of the control rod worth, isothermal temperature coefficient, xenon load, and power defect. Whenever the fresh fuels are loaded, the critical CAR position is predicted by diffusion code with bumup calculation. The operation power in first and second cycle was 15 MW and raised to 22 MW in the middle of 2-2 cycle. The core average bumup at the end of cycle 2 was 39.11 GWDIMTU (24.16 %U-235). The water conductivities of primary cooling system, spent fuel pool, and heavy water system are measured everyday. The conductivity is maintained much lower than the technical specification limit value. The water quality is analysed weekly and the pool water radiation is measured weekly by HpGe detector. The radiation level around beam tubes was below 3.0 f.J. Svihr and 2.5 f.J. Svihr in the control room at 22 MW. The radiation area is categorized in four level; clean, normal operation, temporary operation, and contaminated area. The control room is in clean area and the radiation limit is 6.25 f.J. Sv/hr and the access time is unlimited. Around beam tube belongs to normal operation area and the radiation limit is 12.5 f.J. Svihr and the access time is limited to 40 hr/wk. The first overhaul was carried out from March 4 to March 30, 1996. The Seismic Monitoring System surveillance was performed separately for 2 week from Feb. 27, 1996. The Neutron Flux Monitoring System linearity was measured and the position of each detector was calibrated accordingly. Since January of 1996, radioisotope production has been tested for isotopes such as 1-131, 1r-192, Mo-99, Tc-99m, Ho-166. Test of air balance and system performance test on Radioisotope Production Facility have been completed and will be licensed for isotope production facility in this November when the normal production is started. From that time, the domestic demand on 1-131, 1r-192, Ho-166, P-32, etc. will be fulfilled. The feasibility study of Fission-Moly production is in progress. Seventeen lead hot cells and four concrete hot cells are currently in operation and eleven more lead hot cells will be ready by 1997. The radiation emergency plan has been established to ensure the adequate response to the emergency which would cause a significant risk to the KAERI staff and the inhabitant neighbouring the HANARO site. The emergency organization consists of spot support team, administrative support team, and technical support team. The emergency state is divided into four levels. Whenever the emergency occurs, the emergency headquarter notifies KINS (Korea Institute of Nuclear Safety) and City Administration and reports MOST (Ministry of Science and Technology). The second emergency drill was performed on May 3, 1996 following the first one last year. The Safety Action against war has been also prepared for the security of HANARO fuel. lAEA inspection is being carried out on in-core, fresh, and spent fuels and monitoring camera film is replaced annually for the transparency of HANARO operation. EXPERIMENTS AND FACILITIES Sixteen hexagonal fuel bundles irradiated during fast flux distribution measurement were scanned by in-pool gamma scanner installed in the reactor service pool for the assemblywise power distribution measurement. The photo peak of long half-life radioisotope, La-140 was selected and measured by NaI(Tl) detector due to more than 3 months of cooling time. For more accurate measurement, the gamma scanning system was designed, which can rotate and move up and down the fuel bundle with very high precision. The same sixteen fuel bundles were scanned by this system on Zr-95 and Nb-95 photo peaks with NaI(Tl) detector for axial power distribution measurement at IMEF. Pin power distribution analysis is in progress. Non-instrumented capsule was inserted into the central trap for the lifetime evaluation of existing nuclear material and its comparison with the newly developed material. The test materials were stainless steel (core structural material), Zr-2.5%Nb (CANDU pressure tube material), and low alloy steel (reactor pressure vessel material). Six eutectic alloys as a temperature monitor and Fe wire in Cd tube as a fluence monitor were also included in the capsule. The evaluation of neutron irradiation effects include tensile, impact, fracture toughness, hardness, thermal annealing recovery tests, magnetic property measurements, micro structure analysis by optical microscope and TEM (transmission electron microscope), component analysis, etc. Instrumented capsules will be loaded in CT, IR, OR, or IP next year. To verify the HANARO fuel at high linear power, two types of test fuel assemblies were loaded at IR. or CT. Each test fuel assembly has six fuel elements at the peripheral positions between the corners and 27 AI dummy and 3 hollow elements. Two type-A fuel assemblies were loaded at IR.l & 2 and the discharge burnup will be 40 and 65 alo. A type-B fuel assembly with three instrument tubes and thermocouples were loaded at CT and the discharge burnup will be 90 alo. In type-B fuel bundle, Rhodium SPND (Self-Powered Neutron Detector) and Platinum SPGD (Self-Powered Gamma Detector) were installed for power history and gamma flux measurement. In the pre- and post-irradiation test, dimension measurement, fine structure investigation, mechanical/thermal characteristics, and bending test of fuel element are performed. Gamma scanning, blistering test, and chemical burnup analysis of fuel element are carried out additionally in the post-irradiation test. Neutron beam facilities are installed in three phases. At the end of the first phase (1993-1997), four units of neutron spectrometers (High Resolution Powder Diffractometer, Four Circle Diffractometer, Polarized Neutron Spectrometer, Small Angle Neutron Spectrometer) and a neutron radiography facility will be available. In the following phases, the neutron beam research will further extend its scope and depth such as in advanced technological materials, polymer science and biological substances by installing a cold neutron source and guide hall neutron beam facilities. Fuel test loop is on manufacturing and IPS will be installed at LH in mid 1998. It is a steady state fuel test loop for irradiation test of a fuel assembly such as CANFLEX, DUPIC, and advanced PWR fuel bundles. The safety analysis report IS on review by KlNS for installation license. After the completion of this FTL, it IS planned to install a transient fuel test loop for fuel pin test in power ramping. OPERATOR PROGRAM Three shift six group system has been established by operators. Each group has four personnel; one senior reactor operator, one reactor operator, and two system operators. Out of six group, one group is on education, and another supports the reactor management. There are 211 reactor operation procedures and those are assigned to six groups for review and revision. HANARO simulator is being used for the personnel to learn basic operation and reactor kinetics and will be further developed to include full reactor control algorithm. Operator training program is open annually in which six classes are with instructors who designed HANARO systems or who is an expert in his own field. The program of visiting to power reactors, foreign research reactor is prepared for operators annually. Intensive health checkup is carried out annually and radiation exposure is monitored quarterly. PROJECTS The design project of an IPS for the steady state irradiation of fuel pins has been initiated in collaboration with AEA recently. This IPS is for steady state irradiation of up to 3 fuel pins and will be coupled to the FTL out of pile system by means of pipework and flexible hoses within th HANARO pool. It is not intended that the main FTL IPS and this IPS will be operated at the same time. The preliminary design will be completed in about five months. It is more than ten years since KAERI and its associate, Korea Cancer Research Hospital were interested in BNCT. With the operation start of HANARO, however, the interest in BNCT was vitalized and the feasibility study is in progress since last year. Boron compound survey, HANARO neutron beam feasibility study, gamma/fast neutron filter investigation, medical treatment technique, etc. are the items in BNCT project. For the DUPIC fuel irradiation test in HANARO, the proper fuel assembly configuration with the instrumentation for monitoring was designed and the fuel enrichment to satisfy a required linear power condition was evaluated. The burnup calculation will be followed in next project year. The low flow CHF test will be carried out using a flow loop in KAIST (Korea Advanced Institute of Science and Technology) at the end of this year. This test is required to validate the thermal margin of HANARO fuel at natural circulation condition and validate the accident analysis results. Thirty finned and 10 unfinned fuel element simulators are reserved for this test. Currently, the test section and the heater are being manufactured. CURRENT STATUS OF RESEARCH REACTORS RA AND RB AT THE VINCA INSTITUTE OF NUCLEAR SCIENCES M. Kopeeni, M. V. Matau§ek, D. Stefanovi~ VINCA Institute of Nuclear Sciences, Belgrade. Yugoslavia ABSTRACT Both research reactors at the VINCA Institute of Nuclear Sciences. the heavy water moderated and cooled 6.5 MW research reactor RA and the heavy water moderated zero power research reactor RB. commissioned in 1959 and 1958 respectively, belong to the generation of research reactors which have played an important role in developing nuclear power technology. as well as different scientific, medical and industrial applications of nuclear energy and radioactivity. This paper presents basic facts about the RA and the RB research reactors and describes the present status of their systems and components. Different options for the future status of the RA reactor are specified. Problems related to the RA reactor spent fuel storage and the long term safe disposal of irradiated fuel, which are identified as the most urgent ones. are discussed. Some experiments. as well as practical applications. performed at the RB reactor are displayed. Conclusions concerning future status of the research reactors, as well as some general conclusions about building and maintaining research facilities for the future of nuclear energy. are drawn. 1. INTRODUCTION Both research reactors at the VINCA Institute of Nuclear Sciences, the heavy water moderated and cooled 6.5 MW research reactor RA, designed and constructed by USSR, and the heavy water moderated zero power research reactor RB, designed and built by Yugoslav scientists, commissioned in 1959 and 1958 respectively, belong to the generation of research reactors which have played an important role in developing nuclear power technology, as well as different scientific, medical and industrial applications of nuclear energy and radioactivity. The research reactor RA was and still is the largest research facility in Yugoslavia and the most valuable installation in the VINCA Institute. It was first operated with 2 % enriched uranium metal fuel of Soviet origin. In 197680% enriched uranium oxide fuel dispersed in aluminium was purchased from USSR. The decision to reconstruct and innovate the RA research reactor was made during the early eighties and a number of improvements, including complete exchange of electronic equipment, was initiated. Unfortunately, for both technical and political reasons, these actions have never been completed. Having in mind the long shut-down period of the RA reactor, together with the overall technological, political and economical changes, the status of the RA research reactor is now a subject of serious reconsideration. In the RB reactor, besides its original natural uranium metal fuel in the form of solid rods, the RA reactor fuel can also be used. Since criticality is achieved by regulating the moderator level, a large variety of core configurations with different lattice pitches and fuel arrangements have been studied over the years, making possible experimental verification of different theoretical models and calculational schemes in neutron physics and neutron transport theory. Most recently, a coupled fast-thermal reactor system was realized, having a central zone where the neutron flux energy distribution is similar to the one in the fast research reactors. In the meantime, complete exchange of control, safety and radiation protection equipment of the RB reactor has been performed. The following paragraphs present basic facts about the RA and the RB research reactors and describe the present status of their systems and components. Different options for the future status of the RA reactor are specified. Problems related to the RA reactor spent fuel storage and the long term safe disposal of irradiated fuel, which are identified as the most urgent ones, are discussed. Some experiments, as well as practical applications, performed at the RB reactor are displayed. Conclusions concerning future status of the research reactors, as well as some general conclusions about building and maintaining research facilities for the future of nuclear energy, are drawn. 2. RESEARCH REACTOR RA 2.1. Basic Facts about the Research Reactor RA Operation The USSR designed and built 6.5/10 MW thermal heavy water moderated and cooled research reactor RA [1,2] started operation in 1959. Until 1976 the reactor was operated with 2% enriched uranium metal fuel, when 80% enriched uranium oxide fuel dispersed in aluminium was purchased from USSR. The new fuel, having the same geometry and the same amount of 23SU per a fuel element as the old one, was supposed to enable increase of the neutron flux due to diminished parasitic absorption in 238U. From 1976 to 1979, the reactor operated with mixed 2% and 80 % enriched fuel core. In 1979, during regular refuelling, deposits were noticed on the new fuel cladding. The reactor was shut down and started operation again in 1981 with completely fresh 80% enriched fuel core. After a year of running the reactor at decreased power of 2 MW, deposits on the fuel cladding were noticed again, in spite of regular heavy water purification. After 1982 the reactor core was once again completely exchanged, the deposits were again detected, and since then the reactor has never been operated at full power and in the nominal working regime. In the meantime, serious problems were also encountered with some other reactor systems. The reactor was finally shut down in August 1984 in order to reconstruct and improve practically all vital reactor systems. Examinations performed at that time showed that the reactor building, mechanical parts of the reactor and the reactor vessel, were in reasonably good condition. Institute VINCA decided to build new systems for emergency core cooling and emergency ventilation, a new irradiated fuel handling system, a system for purifying water in the spent fuel storage pool and a new power supply system. A complete exchange of electronic equipment and instrumentation had to be performed through the technical assistance program of IAEA. New equipment was purchased from Soviet suppliers. However, for a number of political, administrative and technical reasons, this reconstruction has not been completed. In the meantime, the old instrumentation of the RA reactor had been dismantled, awaiting for arrival of the new one, and the reactor RA was left in an intolerable situation without control, safety and radiation protection systems. In the course of preparations for restarting the RA reactor, a significant amount of new fuel was purchased from the former USSR, sufficient for the life time of continuous operation of the reactor. The Institute VINCA built the new emergency core cooling system and the new emergency ventilation system. The new irradiated fuel handling machine was completed, but software for its remote operation is not installed. The new system for purifying water in the spent fuel storage pool is finished but not tested yet. Although the reactor facility is financially supported and basic expenses and salaries of the stuff are covered, during the last several years the reactor RA crew suffered serious losses in skilled and trained personnel. With all this in mind, during the year 1995, the Institute VINCA, together with the Yugoslav largest consulting and engineering organization ENERGOPROJEKT, produced a document [3] describing scenarios of the possible options for the future status of the RA reactor: present state, 2 restarting the reactor, conservation or decommissioning. The following main conclusions were reached: the present state of the research reactor RA can not be tolerated any longer; all other options would require international technical and/or financial support; the spent fuel storage pool is in a very bad condition and should therefore be the cause of considerable concern; the problem is primarily neither scientific nor technical, but it should be considered from the nuclear safety and radiation protection point of view. For this reason, Institute VINCA addressed both the national authorities and the IAEA, emphasizing the necessity for immediate action concerning safety, security and radiation protection problems at the RA reactor. Main results of this effort and actions taken so far are summarized in the following paragraphs. 2.2. Reactor RA Main Systems and Features Research reactor RA is a thermal heavy water moderated and cooled reactor. Its main parts are: the central body, Figure 1, including the active zone situated in the inner aluminium reactor vessel; the graphite reflector surrounding the inner vessel and situated in the outer stainless steel reactor vessel; the biological shield of water and heavy concrete; the heavy water primary cooling system; the ordinary water cooling systems, i.e. the secondary technical water cooling system and the graphite cooling system; the He-system with He-atmosphere at the upper part of the reactor vessel; the control systems for technological parameters; the reactor power control system; the dosimetry system for measuring direct radiation and the water, He and air activity; the fuel transport system; the auxiliary systems like power supply system or special ventilation system. The fuel element (slug) is an aluminium cladded hollow cylinder, 11 cm long, with an outer diameter of 3.7 em, containing uranium in tubular form, Figure 2. Fuel elements are inserted in alumini1!m tubes (11 slugs/tube), forming a fuel channel. A maximum of 82 channels form a square reactor lattice having 13 em pitch. The D20 coolant, which is circulated by pumps into the reactor vessel from the bottom, flows upwards through the fuel channels on both sides of the fuel and flows down through the core moderator space into the outlet tube at the bottom of the reactor vessel. Water from the Danube river is used as a secondary coolant. Two safety rods containing cadmium absorber are operated by two independent and identical safety mechanisms for automatic and immediate shut down (scram) of the reactor. Another two cadmium rods are used for automatic control of reactor power, while seven cadmium rods are used for compensating long term reactivity changes. Control and safety rods are inserted into the reactor core through aluminium channels, isolated from the primary circuit and filled with helium when the reactor is in operation. Since the existing control rods are partially burned-out, particularly their lower active parts, new stainless steel control rods, with gadolinium as a burnable absorber, were purchased together with the new electronic equipment. The main characteristics of the present status of the RA reactor are the following: the core contains 480 fuel slugs, initially 80% enriched and with relatively low average burnup; safety rod, rods for automatic reactor power regulation and rods for compensating long term reactivity changes are all in the core in the lowest possible position; heavy water is drained from the primary circuit and placed in the heavy water tank; helium is drained from the inner reactor vessel and the gas system, so that the reactor vessel is filled with air under normal pressure; the automatic power control system, the system for measuring the reactor parameters and the relevant technological parameters, as well as the system for radiation control and protection do not exist, since the old equipment was dismantled and new equipment was not installed. 3 Visual inspection of the fuel from the core has been recently performed through the protective glass window on the biological shield of the reactor. A white layer of corrosion products and\or deposits was noticed on the fuel string taken from a fuel channel in the reactor core. Similar deposits, but at a much lower extent, could be noticed on the outer side of the fuel channel tube. Similar deposits were noticed and investigated many years ago, but no definite conclusions have been made about their origin, nor the way how to prevent their formation. Although the fuel in the reactor core has been left in the ambient temperature and humidity for more than ten years, no signs of new intensive corrosion processes were noticed, neither have any signs of fission-product contamination been observed in the reactor core and its surrounding. Some pitting corrosion could, however, take place under the layer of deposits. Whereas safety standards would require to remove the fuel from the reactor core, because of the very poor situation with the spent fuel storage pool, it was decided to leave this fuel as it is. 2.3. Reactor RA Irradiated Fuel Handling and Disposal Closely related to the problem of the future use, or eventual decommissioning, of the research reactor RA is the problem of safe and reliable disposal of the so far irradiated fuel, as well as the newly irradiated fuel if and when the reactor is operated again. Main components of the reactor RA fuel handling system are: the transporter on the central reactor body used for reactor fuelling, as well as for transferring the fuel channel with irradiated fuel into the water channel connecting the water reactor shield and the temporary spent fuel storage pool in the basement of the reactor building; the crane in the reactor hall and the crane in the spent fuel storage room; the irradiated fuel handling machine for distant automatic manipUlation of irradiated fuel, which can be observed from the neighbouring control room through a special lead glass window. The six meters deep temporary spent fuel storage pool, Figure 3, consists of four connected basins, having thick concrete walls cladded with stainless steel, and is fIlled with approximately 200 tons of stagnant ordinary water. 304 channel-type stainless steel fuel containers, Figure 4, receiving up to 18 spent fuel elements each, are placed vertically in the pool. Initially, it was planned to transfer spent fuel back to the Soviet supplier, after 4-5 years of cooling in the temporary storage pool. Since this has never happened, in order to increase the spent fuel storage capacity, some of the oldest metal uranium fuel has been taken out of the original stainless steel containers and repacked in sealed aluminium barrels, Figure 5, each containing 30 aluminium tubes receiving up to 6 irradiated fuel elements per a tube, placed in two layers in the annex of basin 4. Cadmium strips were placed in the barrels to provide the necessary subcriticality. Both the barrels and the channel-type fuel holders were filled with demineralized water, which is not supposed to mix with water from the pool. The reactor RA spent fuel inventory is the following: 6656 fuel elements with initial enrichment 2% 23SU in the temporary storage pool; 884 fuel elements with initial enrichment 80% 23SU in the temporary storage pool; 480 fuel elements with initial enrichment 80% 235U in the reactor core. The heavy nuclide inventory in the low enriched spent fuel is: 30 kg of 235U, 2400 kg of 238U, 5.20 kg of 239PU. Nuclear criticality safety studies performed for irradiated fuel at the reactor RA site proved that sufficient subcriticality was provided for all existing configurations [3]. Due to the obvious lack of attention paid to the conditions of the reactor RA spent fuel storage, presumably because it was considered only as a short term storage, it is now in very bad 4 condition [5]. Water in the pool is dirty and its chemical parameters are not maintained so to minimize corrosion. New equipment for water purification is neither operable, nor adequate. At the bottom of the pool there is a lot of sludge which even hides one spent fuel slug lost many years ago. By measuring depth of the sludge at different points of the pool, which varied from 3 to 12 cm, its total amount is estimated to be roughly 3 m3• Visual inspection showed that all steel construction elements are heavily corroded. Corrosion is also noticed on stainless steel walls of the basins and transport channels. A string of spent fuel slugs, removed from RA reactor more than 17 years ago, before the reactor was shut down because of the corrosion problems for the first time, has recently been taken out from the channel-type spent fuel holder in the storage pool and brought to the hot room in the reactor shield to be visually inspected. A thick layer of corrosion products and/or deposits, covering completely the spent fuel slugs, pointed out the high probability that fuel cladding was penetrated and that fission products leakage took place in some of the spent fuel containers. It was not possible to examine the state of the spent fuel inside aluminium barrels, but it could be presumed that the corrosion is even worse. Radiochemical analyses showed that the radioactivity content in the pool water was quite high, it contained about 3.8x107 Bq/m3 Cs-137, i.e. about 0.25 Curies for the 200 m3 of pool water. This could be attributed to the "lost" fuel slug, but suspicion can also be expressed that some contribution could also be due to corroded slugs in aluminium barrels whose leak tightness was questionable. Recent investigations show that water in some channel-type spent fuel containers, particularly those with very old fuel, is highly radioactive, what is an obvious indication that some spent fuel is leaking. 2.4. Present Activities Related to the Future Status of the RA Research Reactor and Safe Disposal of So Far Irradiated Fuel Although the present state of the 6.5 MeV thermal research reactor RA systems and components is rather poor, it could still represent a valuable facility. In the country which does not have a nuclear power program, restarting the research reactor would help to preserve expertise in reactor use and operation and bridge the time gap until the moment when the attitude of the society towards the use of nuclear energy would be changed. As the first step towards this goal, the VINCA Institute of Nuclear Sciences and the engineering organization ENERGOPROJEKT have recently started work on the Project of Restarting the RA Research Reactor, which should provide all the relevant technical and financial factors needed for making a final decision about the future status of this facility. At this moment, however, the problem of the future status of the reactor is pushed back by the situation with the reactor RA spent fuel storage pool, which has been paid little or no attention to for years and has thus become a serious safety problem. Activities are now initiated in two directions. First, to improve the present situation of the existing spent fuel storage by reducing the water turbidity, by removing the failed fuel slug from the bottom of the pool and by establishing the water chemistry control so to reduce further corrosion as much as possible. Second, to obtain an expert assessment of the conditions in the stainless steel channel holders and the aluminium barrels and to consider transferring the irradiated fuel back to the supplier or constructing an independent installation, presumably dry storage space, for long term' storage of spent fuel placed in special containers designed for this purpose. Immediate objective that these activities are 5 expected to achieve is increased safety of the research reactor facility in the sense that the possibility of uncontrolled fission products release in the reactor building, and eventually to the environment, will be reduced to the minimum possible value. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel if the reactor starts operation again. In the case that it is decided to decommission the research reactor RA, the newly built storage space will provide safe disposal of the so far irradiated fuel. 3. RESEARCH REACTOR RB 3.1. Basic Facts about the Research Reactor RB Research reactor RB at the Nuclear Engineering Laboratory of the VINCA Institute of Nuclear Sciences was the first critical facility in Yugoslavia, designed and built by Yugoslav scientists [6]. In its original form, as a natural uranium metal - heavy water critical assembly, having fixed 12 cm lattice pitch and two safety rods, it was commissioned in 1958. Later, the facility was reconstructed in order to improve its safety and flexibility. The improved reactor, with new control, safety and monitoring systems was critical again in 1962. Besides its original natural uranium metal fuel in the form of solid rods (length 2.1 m, diameter 2.54 cm, 1 mm thick Al cladding), the RA reactor fuel can be used in the RB reactor, as well. Special Al support plates make it possible to change the reactor lattice pitch (7, 8, 9, 12 or 13 cm or some of the existing multiples). Since criticality is achieved by regulating the moderator level, a large variety of core configurations with different lattice pitches and fuel arrangements have been studied over the years, making possible to perform various experiments and to obtain valuable results. The RB reactor power ranges from 10 mW to 50 W under normal operating conditions, but it can reach 10 kW or more under special conditions. The constant power level is maintained by adjusting the heavy water level. A Ra-alpha-Be neutron source of 17.5 GBq is used during the start- up procedure. The heavy water circulation system has two speeds of filling the tank (2.5 cm/min or 0.8 cm/min) and two draining speeds (11 cm/min or 1.7/cm/min). The dosimetry system comprises neutron and gama dose measuring devices in the reactor hall, the reactor control room and the reactor building. Besides being used for different kinds of experiments in rector physics and engineering, the RB research reactor is regularly used in training students studying nuclear engineering at the University of Belgrade. 3.2. An Application of the RB Research Reactor for Experimental Verification of Theoretical Models and Calculational Schemes New reactor concepts and advanced in-core fuel management schemes are the subject of increased interest from both technical and economic reasons. However, techno-economical evaluation of some advanced reactor cores requires improved calculational procedures, either because of more complex core configurations, or because of more restrictive accuracy requirements. In most of the cases, mock-up or criticality experiments for new reactor cores are practically unfeasible, while studying burn·up effects in practice would be too much money and/or 6 time consuming. Thus, experimental verification, on existing and available research reactors, of models, data and codes used for above purposes is particularly desirable and important. Here, an example of using the RB research reactor for verifying different calculational models and procedures [7] is given. The problem of treating the reactor core having different composition in axial direction was encountered in a particularly pronounced form when analyzing possibilities and effects of using a natural uranium blanket in a PWR. Since the expected effects, although important from the economic point of view, are relatively small in absolute amounts, it is important to eliminate a possibility of drawing erroneous conclusions on the basis of inadequate results. When calculating parameters of an axially inhomogeneous reactor core, procedures for determining the reactor lattice cell parameters and the overall reactor parameters can be combined in a number of ways. In order to validate different calculational approaches , a reactor core inhomogeneous in axial direction has been simulated on the RB reactor, using the fact that its fuel channel is formed of a number of fuel slugs. Three core configurations were established: (a) all fuel channels contain 80% enriched fuel only; (b) at both ends of all fuel channels there is one fuel slug containing 2 % enriched uranium metal; (c) at both ends of all fuel channels there are two segments containing 2% enriched uranium dioxide. The cores (b) and (c) simulate the situation when a number of enriched uranium pallets at the ends of a PWR fuel element are substituted by natural uranium pallets. Besides the critical heights of the reactor cores, axial and radial distributions of the fast and thermal neutron flux were measured by activating Au foils in Al or Cd cladding. Comparison of experimental and computational results suggests that axially inhomogeneous reactor cores should be treated in the following way: first, volume averaging of physical data should be performed for each axially inhomogeneous zone of an equivalent reactor lattice cell; group constants are to be determined for an axially homogenized reactor lattice cell; calculation of global parameters is to be performed supposing that the reactor is homogeneous in the axial direction. In other words, there is no justification to divide the reactor core into a number of axial zones, determine cell parameters for each of these zones and then perform global calculations, since in this case the basic concept of a reactor lattice cell and equivalent cell parameters is jeopardized. This conclusion is supposed to be true not only in the case when axial inhomogeneity originates from different initial fuel composition, but also in the most common situation when it is caused by nonuniform fuel depletion during the reactor operation. 3.3. Fast Neutron Fields at the RB Reactor When 80% enriched uranium dioxide fuel became available in 1975, investigation of fast neutron fields at the RB reactor started with constructing of an external neutron converter which transforms the thermal neutron leakage flux into the fast fission neutron flux [6]. Advantages of this converter are easy accessibility to the large experimental space and the possibility of the fast neutron spectrum down-shifting by using different screens. Its main shortcoming is the low intensity of the fast neutron flux. The intensity of the fast neutron flux was upgraded in 1982, when the inner neutron converter and the fast flux experimental fuel channel were constructed using the new highly enriched fuel. In this case, however, smaller experimental space and softer neutron spectrum are obtained. In 1983, in order to overcome problems encountered with the outer and the inner neutron converters, realization of a coupled fast - thermal system started. 7 Presently, a coupled fast-thermal system, HERBE, at the reactor RB is established, Figure 6. This system has a central zone with densely packed natural uranium fuel and 80% enriched fuel, and no moderator. In the central experimental channel the neutron flux distribution is similar to the one in the fast research reactors. During the realization of this system, complete exchange of control, safety and radiation protection equipment of the reactor RB has been performed. 4. CONCLUSION Valuable results have been obtained and considerable experience has been accumulated during many years of successful operation of the RA and RB research reactors at the VINCA Institute of Nuclear Sciences. However, having in mind problems presently encountered with the reactor RA, it can be concluded that building, running and maintaining a research nuclear facility may become a too heavy burden for a country without a well defmed nuclear research and development program, while absence or inadequacy of necessary regulatory, safety and control institutions, may finally lead to negligence of the important nuclear safety issues. International cooperation, on multilateral or bilateral basis, is important in all phases of planning, building and use of nuclear research facilities. The ongoing and forthcoming activities on solving the problem of the research reactor RA future and on improving its spent fuel storage conditions will provide an opportunity to perform a comprehensive exchange of existing experience and to develop new management practices of general interest. S. REFERENCES 1. Safety Analysis Reponfor RA Reactor, Vo1.10, VINCA internal report, November 1986 (in serbian). 2. M. V. Matausek, Research Reactor RA at VINCA Institute of Nuclear Sciences, NUKLEARNA TEHNOLOGIJA Jour., Vol. 10, 2(1995)3. 3. The Status of the Research Reactor RA at the Institute of Nuclear Sciences VINCA, Vol. 1, Elements and Criteria for Determining Future Status of the Research Reactor RA, ENERGOPROJEKT-VINCA internal report, August 1995 (in serbian). 4. M. V. Matausek, N. Marinkovic, R. Simovic, Criticality Safety ofF resh and Irradiated Fuel oft he 6.5 MW Heavy Water Research Reactor, Proc. ofInternational Symposium on Nuclear Energy SIEN'95, Bucharest, 1995, p. 27. 5. M. Kopetni, M. V. Matausek, Z. Vukadin, T. Maksin, Corrosion Problems in the Research Reactor RA Spent Fuel Storage Pool, NATO Advanced Research Workshop "Microbial Degradation Processes in Radioactive Waste Repository and in Nuclear Fuel Storage Area" , Budapest, May 1996. 6. D. Stefanovic, M. Pesic, From a Critical Assembly Heavy Water - Natural Uranium to the Fast - Thermal Research Reactor in the VINCA Institute, Monograph, Belgrade, 1995. 7. M. V. Matausek, N. Marinkovic, M. Pesic, Experimental Verification ofM ethods and Codes Used in Design Studies of New Reactor Concepts and Improved Fuel Management Schemes, Proc. Int. Symposium on the Utilization of Multipurpose Research Reactors and Related International Co-operation, IAEA-SM-3001006, IAEA, Vienna 1988. 8 • Fuel channel e Automatic rod o Experimental channel 6' Safety rod e Compensation rod e Biological channel Figure -1. Main body of the RA research reactor. Horizontal cross section. 9 d(=)mm Figure 2. Cross section of the research reactor RA fuel element 10 d(=)m WATER CONCRETE FROM THE ~~~rl REACTOR Figure 3. Scheme of the temporary spent fuel storage pool 11 -~ c-, --~.. ....!!- I. ~ I Figure 4. Stainless steel channel-type spent fuel holder 12 Figure 5. Aluminium barrel with spent fuel 13 I II o 0 0 0 o 0 o 0 o 0 o 0 o 0 o 0 0 0 o O@O 0 0 O©O 0 o 000 0 0 IV Figure 6. Horizontal cross section of the reactor RB core with the central HERBE zone (I) fuel elements with 80% enriched U~; (II) safety rods; (III) inside reflector zone; (IV) outside reflector zone; (V) reactor vessel. 14 I III IV VIII V VI Figure 7. Horizontal cross section of the HERBE fast core (I) convertor zone, 24 fuel elements with 80% enriched U~; (II) filter zone, Cd and 32 U metal fuel elements; (III) HERBE fast core, 32 U metal fuel elements; (IV) outer AI vessel; (V) intermediate AI vessel; (VI) inner AI vessel with Cd; (VII) indicator of moderator leakage; (VIII) experimental channel 15 STUDSVIK'S R2 REACTOR - REVIEW OF RECENT ACTIVITIES Mikael Grounes, Christian Grislund, Mikael Carlsson, Thomas Unger and Anders Lassing Studsvik Nuclear AB, S-611 82 Nykt)ping, Sweden • Paper to be presented at the meeting of the International Group on Research Reactors (lGORR-V), Aix-en-Provence, France, November 2-6, 1996. STUDSVIK NUCLEAR AB 1 Abstract A general description of the R2 test reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, in-pile corrosion studies, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. More detailed reviews are given of the following fuel R&D projects: DEFEX, ULTRA-RAMP, Ultra-High Burnup Fuel Irradiation Project, SUPER-RAMP IIII1Oxl0, TRANS-RAMP m, STEED I & II and the INCA facility. 1 Introduction STUDSVIK AB is partly owned by the biggest Swedish utility, Vattenfall AB, and is performing R&D work and associated activities, primarily in the nuclear energy field. STUDSVIK AB is a commercial company, active in the areas of services, supply of special equipment and systems and also consulting. STUDSVIK NUCLEAR AB, which is the largest subsidiary within the STUDSVIK group, is one of the direct offsprings of AB Atomenergi, the origin of the STUDSVIK group, which was formed in 1947. The STUDSVIK group has about 580 employees and a turnover of about 425 MSEKIyear. During the 1950's and 60's, an ambitious nuclear program was launched in Sweden. The experience and competence gained from a large number of advanced projects constitutes the basis upon which the present activities of STUDSVIK NUCLEAR are based. Among the fuel- and materials-related experiences are the design, construction and operation of a fuel manufacturing plant for uranium dioxide fuel in Stockholm, a pilot line for mixed-oxide (MOX) fuel in Studsvik:, the design and manufacture and initial operation of the Agesta pressurized heavy water reactor, and the design, manufacture and irradiation proof testing for the Marviken boiling heavy water reactor. Ambitious R&D programs for heavy water reactor super-heater fuel and fast heavy water reactor fuel and materials were also carried out. Since the 1970's, the efforts have been concentrated on light water reactor fuel and materials, and the originally domestic R&D programs have been expanded so that a large fraction is now financed by non- Swedish sponsors. t Neutron activation analysis and radioisotope production as well as beam tube experiments for basic research applications were started in the 1960's. In 1977 neutron transmutation doping of silicon began. The facilities of interest in this connection are the R2 Test Reactor, the Hot Cell Laboratory, the Lead Cell Laboratory and various other laboratories, all located at Studsvik:, 100 kilometers south of Stockholm. STUDSVIK NUCLEAR AB 2 1 The R2 Test Reactor 1.1 General Description of the Rl Test Reactor The R2 reactor is a tank-in-pool reactor, see Figure 1 and Table 1, in operation since 1960 and originally similar to the Oak Ridge Research Reactor, ORR (1). The reactor core is contained within an aluminum vessel at one end of a large open pool, which also serves as a storage for spent fuel. Light water is used as core coolant and moderator. The reactor power was increased to 50 MW(th) in 1969. In 1984-85 a new reactor vessel was installed. The R2 reactor has a high neutron flux, see Table I, and special equipment for performing sophisticated in-pile experiments. An important feature of the reactor is that it is possible to run fuel experiments up to and beyond failure of the cladding, which is not possible in a commercial power reactor. Table 1 Technical Data for the R2 Test Reactor. Power Moderator/coolant Reflector Fuel length 600mm Fuel assembly length 924mm Fuel assembly cross section 79x82~ Number of fuel plates per assembly 18 Neutron flux in experimental positions Thermal (0.3-2.5) x 1014n/(cm2·sec) Fast (>1 MeV) (0.5-2.5) x 1014n/(cm2·sec) Primary flow 1300 kg /sec Coolant temperature: Inlet :s; 40°C Outlet :s; 45°C The coolant water is circulated through the reactor vessel and flows through pipes and a large decay tank below the reactor hall to an adjoining building containing pumps and heat exchangers cooled with sea water. The present core configuration is shown in Figure 2. The components of the core are arranged in an 8xl0 lattice, typically comprising 46 fuel elements, 6 control rods, about 12 beryllium reflector assemblies and a number of in-pile loops, irradiation rigs and aluminum fillers. Rows Nos. 1 and 10 consist of beryllium reflector assemblies. The composition of the core can be altered to suit the experimental program. The R2 driver fuel assemblies are, since the beginning of 1993, of the LEU type. They have 18 curved fuel plates containing an aluminum-clad aluminum silicide matrix. The initial fuel content is 450g 23SU per fuel assembly, enriched to less than 20 %. The bumup of the spent fuel of this type reaches about 65 %. STUDSVIK NUCLEAR AB 3 The control rods consist of an upper neutron absorbing section of cadmium and a lower fuel section. They are moved vertically by drive mechanisms placed below the reactor vessel. The reactor vessel is 4.5 m high and 1.6 m in diameter. The design pressure is 3.3 bars. Some of the irradiation facilities in the R2 reactor have been descnbed in the literature (2-5), and details are given below. Most base irradiations of test fuel (irradiations at constant power, where fuel bumup is accumulated under well-defined conditions) are performed in boiling capsules (BOCA rigs). Some base irradiations and all ramp tests (irradiations under power changes) are performed in one of the two in-pile loops, which can be operated under either BWR or PWR pressure and temperature conditions. The ramp tests, simulating power transients in power reactor fuel, are achieved by the use of3He as a variable neutron absorber. Structural materials, such as samples of Zircaloy cladding, steels for pressure vessels and vessel internals and candidate materials for advanced reactors can also be irradiated in special rigs either in the loops or in special NaK-filled irradiation rigs in fuel element positions with a well-controlled irradiation temperature. Special equipment for in-pile corrosion experiments in the loops has recently been developed. The R2 core has an active fuel length of 600 mm. Most fuel rods irradiated are segments of power reactor fuel rods, so-called rodlets, with lengths in the range 300 to 600 mm. However, tests have also been performed on full-size demonstration reactor fuel rods with up to 2.5 m length. In those cases only the lower 0.6 meters were irradiated. Non-destructive examinations of fuel rodlets can be performed in the R2 pool during short pauses in the irradiation program or between various phases of an experiment, see Section 3. All the handling and all the examinations are performed with the fuel rodlets in a vertical position; this is advantageous with respect to possible movements of fuel fragments etc. Associated with the R2 reactor is the 1 MW(th) swimming pool R2-0 reactor, which is located in the same poo~ see Figure 3. The basic research performed by use of R2-0 is briefly described in Section 10. 2.2 Boiling Capsules (BOCA Rigs) The Boiling Capsule (BOCA) facility for irradiation ofBWR and PWR fuel rods was introduced J in 1973. The in-pile part of a BOCA rig consists of a bare stainless steel pressure thimble containing a shroud with flow entrance ports at the bottom and exit ports at the top. The lower part of this shroud is located in the reactor core region. A fuel test rod bundle consisting of up to 6 rodlets is located inside the shroud. The BOCA is filled with highly purified pressurized water from a special pressurization system. Figure 4 shows a simplified BOCA flow diagram. BOCA system technical and operational data are given in Table 2 (page 4). Coolant circulation and cooling is brought about by natural circulation, although no net boiling occurs. The water is heated mainly by the fuel rods. Buoyancy forces make hot water with lower density rise in the riser shroud, leave the exit port, meet the cold wall of the pressure thimble and be cooled down while flowing downwards in the annular channel between the pressure thimble wall and the riser shroud. STUDSVIK NUCLEAR AB 4 Table 1 BOCA System Technical and Operational Data. Permissible operating pressure range: - Pressurizer System BOCA A 30 to 100 bars - Pressurizer System BOCA B&C 30 to 170 bars Permissible total power generation in one BOCA rig ~90kW Permissible heat flux on test rod surface with regard to departure from nucleate boiling (DNB) 160W/cm2 Possible range of thermal neutron flux achievable by: - Positioning in the core up to 2.Ox1014 n/(cm2·s) - Internal hafuium shielding Flux reduction to ~ 50 % - Internal hafuium shielding plus Flux reduction total to 30 % hafuium shielding of the rig adapter The circulation flow rate is low, about 0.2 kg/sec, but the coolant is substantially sub cooled. When the heat flux at the fuel rod surface exceeds a certain value, between about 60 and 90 W/cm2, subcooled nucleate boiling will occur at the rod surface, which implies a surface temperature equal to or slightly higher than the saturation temperature for the actual static pressure. Power in the test rods is measured by a combination of nuclear measurements with the Delayed Neutron Detector (DNO) technique, and coolant water temperature and flow measurements. Up to five BOCA rigs can be operated simultaneously in the reactor. Two independent pressurization systems are available, each capable of supplying 3 to 5 BOCA rigs with water. t Each BOCA rig is connected to a separate outlet circuit. Each rig is constantly fed with a purging water flow in order to control the water chemistry in the self-circulating water volume of the pressure thimble. The same flow is let out to the drain. This inlet and outlet of water takes care of thermal expansion and contraction of the water in the thimble. During reactor start-up, when the water is heated up and expands, the flow is increased. The outgoing water is monitored for radioactivity (fission products), and the water chemistry is controlled. In order to make it possible to irradiate power reactor fuel with standard enrichment in the in- pile loops and BOCA rigs in the R2 reactor, it is often necessary to decrease the neutron flux. This is achieved with hafhium absorbers in the form of tubes or plates. STUDSVIK NUCLEAR AB 5 2.3 In-Pile Loops There are two in-pile pressurized light water loops used in the current LWR fuel R&D program in the R2 test reactor. They simulate realistic BWR and PWR temperature and pressure conditions. Data for the loops are shown in Table 3. Other loops, e.g. for the testing of HTR fueL are described elsewhere (5) and will not be discussed here. Each LWR fuel loop utilizes two diagonally adjacent fuel element positions in the R2 test reactor, see Figure 2. The two loops are pressurized in order that test fuel rods can be investigated at realistic operating conditions for either PWR (loop No 1) or BWR (both loops) type power reactors. • Table 3 Characteristics of the R2 In-Pile Loops . Loop No. 1 2 Type PWR/BWR BWR Pressure, bars 30-150 30-90 Coolant temperature, DC 220-325 220-285 Coolant flow rate, kg/sec 2.5-4.0 2.5-5.0 Max cooling capacity, kW 150 400 Neutron flux. 1014n/[cm2·s] Thermal 0.6-2.2 0.5-1.5 Fast (>1 MeV) 0.8-2.0 0.7-1.4 Gamma heating in stainless steel, Wig 3-12 2-9 The loops can be used for irradiation at constant power of up to 4-5 test fuel rodlets simultaneously, and for power ramp tests of single rodlets. The test rodlets to be ramp tested are installed in the loop in special capsules, which in tum are inserted in a special test rig. The loops can also be used for irradiation testing of structural materials, e.g. Zircaloy test specimens and steel specimens of various types. A technique for in-pile corrosion tests in the loops with on-line corrosion potential measurements is discussed in Section 2.8 below. , The in-pile part of the loops are of a U-tube design, taking up two core positions and thus providing two test positions for each loop in the R2 core, one of which can be used for ramp tests (Figure 5). The U-tube is isolated from the reactor primary coolant by a gas gap containing C02. Heat losses from the tube to the reactor coolant are therefore quite small, which facilitates accurate test rod power measurement. The working in-core inside diameter of these in-pile pressure tubes is 45.5 nun. The useful length in the core is 670 nun. The main features of the loops are presented in Figure 6. STUDSVIK NUCLEAR AB 6 2.4 Ramp Test Facility Ramp testing in the R2 reactor began in 1969. Originally, the tests were performed by rapid power increases of the whole reactor. This mode of operation had two disadvantages: safety restrictions limited the maximum allowable power increase and financial restrictions limited the number of tests that could be performed since other experiments, which would have been disturbed by the ramp tests, had to be unloaded from the reactor at considerable expense. In the present Ramp Test Facility, introduced in 1973, the fuel rod power during a ramp test in a loop is controlled by variation of the 3He gas pressure in a stainless steel double minitube coil screen which surrounds the fuel rod test section. The principle of operation of this system is based on the fact that 3He absorbs neutrons in proportion to its density, which can be varied as required by proper application of pressure. The axial location of the minitube coil screen in the loop U-tubes is shown in Figure 6. The efficiency of the 3He neutron absorber system makes it possible to increase test rod power by a factor of 1.8 to 2.2 (depending on the fissile content of the fuel). The 3He absorber system is designed to achieve a 100 % power increase within 90 seconds, when operating with the normal pressure variation (bellows system). In order to achieve a higher power increase than a factor of about 2, the reactor power must be increased before or simultaneously with the "3He ramping". This technique with combined ramp systems is called "double step up-ramping", Figure 7. The technique makes it possible to increase the test fuel rod power by a factor of about 3. An important advantage of the R2 Ramp Test Facility is that test rods, one at a time, can be loaded and unloaded during reactor operation. This is done by means of a lock vessel built onto an axial drive mechanism with about 3.5 m stroke. This lock vessel is bolted on top of a lock valve (ball type valve) fixed on top of the ramp rig. For BWR pressure conditions and normal rod lengths there is a 4-rod revolver lock vessel with a mechanical chain drive. For tests ofPWR fuel rods and longer than normal rods, there exists a selection of rod lock vessels and dedicated hydraulic drives. Figure 6 shows a ramp rig with a hydraulic drive. In the Ramp Test Facility ramp rates can be achieved in the range of 0.01 W/(cm·min) to about 3000 W/(cm·min). t The maximum achievable ramp terminal level depends on the neutron flux in the experimental position and on the fissile content in the test rod. Tables 4 and 5 present the estimated ramp tenninallevel of some typical rods as function of bumu p. The fissile content left in a BWR rod is highly dependent upon the axial location of the rod in the bundle, making the ramp tenninallevel estimation for BWR rods more uncertain than an estimation for PWR rods. STUDSVIK NUCLEAR AB 7 Table 4 Estimated Ramp Terminal Level ofPWR Rods. Bumup Ramp terminal level (kW/m) (MWdlkgU) 17x17 PWR (4.5 %)1 15xl5 PWR (3.4 %)2 35 70 63 55 50 40 65 40 35 Table 5 Estimated Ramp Tenninal Level ofBWR Rods. Bumup Ramp terminal level (kW1 m) (MWdlkgU) 8x8 BWR(3.1 %)3 IOxIOBWR(4.3 %)4 35 70 69 55 51 40 65 45 31 The power (linear heat generation rate in the fuel rod) is measured calorimetrically by the use of two inlet thennocouples, two outlet thennocouples, a venturi flowmeter and a pressure gage. The special calibration techniques employed have been described (3). The estimated uncertainty (± 10) is 2.3 % when the most common rod lengths (0.3-1.4 m) are used. The reproducibility obtained, when a fuel rod is irradiated several times in the same ramp rig, is ± I %. For fast ramps the discrepancy between the tenninal power aimed at and the one obtained is less than ± I kW/m. The axial thennal neutron flux distribution is measured by activation of cobalt wires in dummy rods and by gamma scanning of the ramp tested fuel rods. An axial movable SPN detector and gamma thennometer can also be used. Each test rod is mounted in a separate stainless steel "capsule" (a shroud open at both ends), primarily as a safety measure and to facilitate the removal and handling of test rods that fail in the course of a power ramp. The "capsule" with the fuel rod is connected to the actuating rod which is used to move the fuel rod axially between the rod changing device and the in-pile section of the rig. There is a small floating push-rod built in at the bottom guide plug of the • capsule. This push-rod transmits the elongation movements of the test rod to a L VDT type elongation detector built into the bottom of the ramp test rig. 1.5 Fuel Rod Failure Detection System Fuel rod failures in the loops are detected b~ a Cerenkov-type radiation sensor which monitors the activity of the loop coolant water. The Cerenkov detector is installed in a by-pass circuit in order to increase the detection system sensitivity by decreasing the background 16N activity produced in the loop coolant water. The 16N b!,.Ckground activity is decreased by the introduction of a delay time due to the fact that the Cerenkov detector is positioned in the by- pass circuit. 1 Typical17x17 PWR rod: Initial enrichment = 4.5 % 2 Typical15x15 PWR rod: Initial enrichment = 3.4 %. 3 Typical 8x8 BWR rod: Initial enrichment = 3.1 %. Void = 60 %. 4 Typical lOxiO BWR rod: Initial enrichment = 4.3 %. Void = 60 %. STUDSVIK NUCLEAR AB 8 The system detects fuel rod failure after 155 ± 10 seconds. This degree of failure detection capability has been verified experimentally. An example of system operation during a ramp test where the rod failed is shown in Figure 8. However, the moment of failure is also registered instantaneously by the rod elongation measurement system as a sudden rod contraction and also often by the power measurement system as a small thermal "spike" (3,6,7). 2.6 Noise Analysis Noise analysis can be a useful tool for monitoring changes in the fuel thermal response as well as the mechanical state during a rod irradiation. UsuaI1y the fuel thermal time constant is evaluated by analyzing the coolant temperature response and the clad elongation response to the rod power variation. The rod power is proportional to the neutron flux. Natural R2 reactor power noise (± 0.5 %) is not sufficient to permit any determination of a test fuel rods thermal time constant. It is necessary to generate an artificial reactor power noise with a power variation of about ± 4 %. Figure 9 presents the differences between natural and artificial noise; the rod elongation response to the nucleonic signal is clearly revealed. The noise measurements performed in connection with studies of the rod thermal performance have been described elsewhere (8-10). 2.7 nata Acquisition Three independent systems are used for data acquisition and storage. All three systems use the same sensors. Most reliable is a high accuracy system, with a scanning rate of 1.0 Hz, linked to the STUDSVIK Hewlett-Packard 9000 computer. During experiments a PC data sampling system is • added, and working with a scanning rate of up to 20 Hz it descn"bes the time behavior very well . The data from the PC system are calibrated towards the high accuracy system and are used both as a backup system and as a source for rapid time-dependent event studies. A chart recorder with variable paper speed also describes the change of data with time. 2.8 INCA - A Facility for Materials Testing, Water Chemistry and Corrosion Studies The experimental program at Studsvik has recently been extended to include investigations on cladding and structural materials in an in-pile corrosion rig, INCA (In-Core Autoclave). STUDSVIK NUCLEAR AB 9 A number of different conditions influence the corrosion behavior of reactor materials: Water chemistry Crud deposits Material characteristics, e.g. annealing treatment Neutron dose Hydrogen pick-up Boiling conditions. The new INCA facility has been developed in collaboration with our sister company STUDSVIK MATERIAL AB. The main feature of the new facility is the ability to control and monitor the water chemistry. Therefore the facility is of the once-through type, which means • that the rig is supported by a water supply system of its own and that the water passes the rig only once. The desired water chemistry is created by adding impurities and additives to a purified water flow, close to the test area. This technique has successfully been used out-of-pile by STUDSVIK MATERIAL AB to obtain well-characterized conditions. The facility has a flexible design and can easily be modified to suit different types of corrosion and water chemistry experiments. The corrosion test rig, which can be seen in Figure 10, is the in-pile part of the INCA facility and is installed in one of the main in-pile loops in the R2 test reactor. It consists of two major parts, the rig tube and the electrode rod. The rig tube separates the water system of the test rig from the in-pile loop main flow. The inlet water to the rig, which is degassed high purity water, is fed from a separate water supply system. It is heated by the loop water in a preheater coil and subsequently led into the rig tube. Additives and impurities (oxygen, hydrogen peroxide, zinc etc) can be added both before the rig and inside the rig just before the in-core test area in order to establish the desired water chemistry. The presently available inner diameter of the rig tube is 21 mm, but it is possible to vary the diameter or to have different diameters in different axial sections of the core. The electrode rod is installed in the rig tube and is a carrier for the test specimens, the reference electrodes etc. The tube for the injection flow is also assembled on the rod. This arrangement makes it possible to change the electrode rod from one reactor cycle to another. For the moment » the rod is bolted on top of the rig tube, but it could easily be rebuilt to be movable up and down in the core section and below the core during operation. Figure 11 is a principle flow diagram showing the feed water system and the analysis system. In order to control the water chemistry the water is analyzed before and after the rig. The oxygen content is measured by an Orbisphere instrument and the content of hydrogen is measured by a gas chromatograph. The inlet water is degassed deionized high purity water with a conductivity less than 0.08 ~S/cm. A diaphragm pump with a variable flow is used to feed the water into the rig. Hydrogen is added to the water both at the inlet and outlet side of the main circulating pump by a dosage pump or by mixing the high purity flow and water saturated with hydrogen. The dosage of additives and impurities such as oxygen, hydrogen peroxide, boron, lithium, zinc etc is done by a controllable piston pump and is mixed with the main flow before or in the rig as described before. This arrangement gives a possibility to vary the flow through the rig, and to create the desired water chemistry. STUDSVIK NUCLEAR AB 10 The INCA facility can operate under both BWR and PWR conditions. Fast (> 1 MeV) and thermal neutron fluxes up to 1.9 and 2.Oxl014n/(cm2·s), respectively, can be achieved. The facility is suitable for different kinds of experiments, for instance materials irradiations, waterside corrosion studies and in-core materials testing, all under controlled water chemistry. It has been in operation since March 1995. One of the objectives has been to develop reference electrodes for long tenn in-pile use. A radiolysis study where experimental measurements and computer model results were compared has also been perfonned (50). 1.9 Defect Fuel Degradation Studies The Ramp Test Facility is utilized with a new test technique to investigate the potential degradation due to hydriding of test fuel rodlets with primary defects. A test of this type, to be used at the R2 test reactor, has to fulfil certain experimental constraints, i.e. 1 the length of the test fuel rodlets should not exceed the active core height (600 mm) of the R2 test reactor 2 the release of fission products and fuel material from the defect fuel must be minimized in order that the contamination of the in-pile loop circuit water be tolerable. Both these requirements can be met with the unique but simple test technique adopted. Instead of an artificial primary defect (drilled hole etc) the primary defect is now simulated by a special device at the top end of the fuel rodlet (12). This device contains an enclosed small reservoir of liquid water in the Rextension plenum", Figure 12. This extra plenum is connected with the normal plenum volume of the rodlet through a tiny tube in such a manner that only water in the fonn of steam will communicate with the plenum volume and the interconnected void space of the interior of the rodlet. Heat is transferred to the enclosed water reservoir from the surrounding pressurized loop water at the start of the operation until the pressure and temperature of the whole system are in balance, simulating either BWR or PWR operating conditions. An important experimental feature of this arrangement is the possibility of post-irradiation puncture and subsequent collection of released fission gases and hydrogen from the rodlets which have been irradiated without failure of the cladding. 3 Pool-Side Examination The general appearance of the irradiated fuel rods can be studied by visual inspection in the R2 pool (3). The following phenomena can also be investigated in detail: Dimensional changes, ridge formation, rod bow and creep-down can be investigated with equip- ment for profilometry and length measurements. The existence and location of fuel rod defects can be established by means of eddy current testing. The axial distribution of certain nuclides is determined by axial gamma scanning of fuel rods or of cladding samples. Data obtained before ramp tests are used as a check on the bumup profile during the base irradiation. Data obtained after ramp testing are used to check the power profile during the R2 irradiation and for studies of the fission product redistribution. STUDSVIK NUCLEAR AB 11 Neutron radiography can be used to study the general appearance and dimensions of the fue~ the extent of filling out of pellet dishing, of center porosities and of center melting (13, 14). This type of examination also reveals the presence of special fuel cracks, hydrides in the cladding, inter-pellet gaps etc. Indications of cladding failure and of structural changes in the fuel can also be observed. In cases where there is no apparent leakage of fission products from failed fuel rods neutron radiography is an important tool since minute cladding leaks are indicated by the existence of hydrides in the cladding or by the presence of water. 4 Post IlTadiation Examination Post-irradiation examination of irradiated fuel is performed in STUDSVIK's well-equipped Hot Cell Laboratory, which has been described in separate publications (I5, 16). Post-irradiation examinations of structural and cladding materials are performed in STUDSVIK's versatile Lead Cell Laboratory, which contains equipment for mechanical testing, corrosion testing and metallography. Among the different types of post-irradiation tests performed are tensile tests, fatigue tests, low-cycle fatigue tests, creep tests, impact test, stress corrosion tests and microscopy (optical microscopy, SEMIEPMA, both with image analysis, and TEM). 5 Fuel R&D 5.1 General Information Much of STUDSVIK NUCLEAR's R&D work in the fuel area has been concentrated on fuel testing, which can be made in the R2 test reactor with high precision under realistic water reactor conditions. This type of work was started in the early 1960's. In a very general sense the purpose of fuel testing can be described as follows: Increasing of reactor availability by decreasing fuel-related operational power restrictions, defining the operational power limits. • Acquisition of experimental data for fuel-related safety considerations. Decrease of fuel costs by making increases in fuel burnup possible. The fuel testing activities can be divided into a number of well-defined steps as follows: Base irradiation, performed in a power reactor, or in STUDSVIK's R2 test reactor. Power ramping and/or other in-pile measurements, performed in STUDSVIK's R2 test reactor. STUDSVIK NUCLEAR AB 12 Non-destructive testing between different phases of an experiment. perfonned in STUDSVIK's R2 test reactor pool, or in the Hot Cell Laboratory. Destructive post-irradiation examinations, perfonned in STUDSVIK's Hot Cell Laboratory, or in the sponsor's hot cell laboratory. Fuel examination can also be perfonned on standard (full-size) fuel rods from power reactors, which can be investigated in the Hot Cell Laboratory. Ifrequired, some types of tests could also be perfonned on such fuel rods in the R2 test reactor. However, due to the rather large initiation costs, such tests have not yet been perfonned. It should be noted, however, that short fuel rodlets, suitable for ramp testing and other on-line measurements in the R2 test reactor, can now be fabricated from irradiated full-size power reactor fuel rods by the STUDFA B refabrication process. Several new hot-lab techniques have also been introduced in recent years (15, 16). The STUDF AB process for fabrication of rodlets from full-size fuel was mentioned above. Fuel ceramography can include scanning electron microscopy (SEM) and electron probe microanalysis (EPMA). Descriptions of the fuel testing facilities and the techniques used were given in Section 2 above. Several other novel testing techniques have also been introduced. A very fast ramp rate, up to 3000 W/(cm·min) can be used to obtain fast power transients and to determine the pellet-clad interaction/stress-corrosion cracking (pCI/SCC) failure boundary. The double step up-ramping technique was described in Section 2.4. On-line elongation measurements can be perfonned during ramp tests, Figure 8. Test fuel rods can be fitted with on-line pressure transducers through a refabrication process. 5.2 Types of R&D Projects I Ramp tests incorporating a very fast-responding test rod power measuring system and associated on-line measurements, such as rod elongation and noise measurements for studies of the rod thermal perfonnance, are perfonned in the pressurized water loops. The ramp tests are a fonn of integral perfonnance tests where the complex interplay between the pellets and the cladding of a power reactor fuel rod is reproduced. The primary test objectives are: - Determination of the failure boundary and the failure threshold, see Figures 13 (17) and 14. - Establishing of the highest "conditioning" ramp rate that safely avoids failure occurrence. - Study of the failure initiation and progression under short time over-power transient operation beyond the failure threshold. - Proof testing of potential pellet-clad interaction (PCI) remedies. Other, more specific test objectives have also been pursued in some projects. STUDSVIK NUCLEAR AB 13 The rod overpressure exPeriments utilize the on-line measurements associated with the ramp tests combined with non-destructive examinations between reactor cycles and destructive examinations after the irradiation. When LW R fuel is used at higher and higher burnups the question of how the fuel might behave when the end-of-life rod internal pressure becomes greater than the system pressure attracts a considerable interest. On one hand end-of-life overpressure might lead to clad outward creep and an increased pellet-clad gap with consequent feedback in the form of increased fuel temperature, further fission gas release, further increases in overpressure etc. On the other hand increased fuel swelling might offset this mechanism. In connection with such considerations the Rod OverPressure Experiments (ROPE) were initiated. The defect fuel degradation experiments also utilize the on-line measurements associated with • the ramp tests and combine these with non-destructive and destructive examinations after the irradiation. Fretting type failures are predominant causes of the very few fuel failures that have occurred in recent years in LW Rs. These primary failures are sometimes followed by secondary failures which frequently cause considerably larger activity releases. In such cases the subsequent degradation of the defect fuel rods by internal hydriding of the cladding and by oxidation of the fuel are the common destructive mechanisms. In these tests an irradiation test scheme, adapted to the experimental conditions in the R2 test reactor was introduced. This scheme offers the possibility of executing comparative investigations of the process of degradation of commercial types ofLWR fuel under simulated primary defect conditions as well as of the mechanisms involved. 5.3 Sponsorship The fuel testing projects executed at Studsvik have been organized under three different types of sponsorship: International (multilateral) fuel projects Jointly sponsored internationally on a world-wide basis. Project information remains restricted to the project participants throughout the project's duration and some predetermined time after project completion. J Bilateral fuel projects Sponsored by one single organization, or a few co-operating organizations. Project information remains restricted to the sponsor, sometimes published later. In-bouse R&D work Sponsored by STUDSVIK NUCLEAR. 5.4 International R&D Programs Since 1975, a series of international fuel R&D projects addressing the PCIlSCC failure phenomenon have been conducted under the management of STUDSVIK NUCLEAR (17-20). These projects are pursued under the sponsorship of different groups of fuel vendors, nuclear power utilities, national R&D organizations and, in some cases, licensing authorities in Europe, STUDSVIK NUCLEAR AB 14 Japan and the U.S. In most of the projects the cladding failure occurrence was studied under power ramp conditions utilizing the special ramp test facilities of the R2 reactor. The recent projects have not been limited to pel/sec studies but also included other aspects of fuel performance, such as end-of-life rod overpressure (21-23) and defect fuel degradation (24-27, 52). Overviews of the projects that have been completed and those that are in progress or planned have been published (20, 28), see also Appendix 1 and 2. In most cases, the test fuel was base irradiated in commercially operating light water power reactors. In some instances the base irradiation took place in BOCA rigs in the R2 reactor. In general, the international fuel R&D projects can be divided into two main categories: - Projects aimed at decreasing the fuel costs by increasing fuel utilization and reactor availability. - Projects providing data for fuel-related safety considerations. A summary of some of the data from the former category of projects is shown in Figure 13. Reviews of the projects in the latter category have also been published (29-30). Over the past years STUDSVIK has studied the behavior of defect fuel in a series of defect fuel experiments. The international DEFEX Project, which was concluded in 1995 addressed hydriding offresh 8x8 BWR type fuel. Some results from this type of work have been published (24-27, 52). A new defect fuel degradation experiment, DEFEX n, is under discussion. The main program will include: Studies, analogous with the DEFEX Project, of the degradation process in irradiated fuel rods, but in this case with medium (around 20 MWd/kgU) bumup with simulated primary defects Testing of potential remedies against secondary failure by hydriding in order to supplement the experimental data base Modeling of defect fuel behavior t Studies of cracking behavior of the hydrided cladding. The primary defect would be simulated by a technique entailing delayed intrusion of water into the rodlet, thus simulating a case when a power rod is defected during operation. A two-year program is envisaged, starting in 1997. The primary goal is to study the behavior of BWR fuel. A PWR program is also planned but would have to be separately funded. It is assumed that the scope ofDEFEX II would be approximately the same as the DEFEX I Project, i.e. 6-8 irradiation tests would be run, and destructive and non-destructive examination would include: neutron radiography, eddy current testing, dimensional measurement, visual inspection, gap squeeze measurement, gamma scanning, SEM investigation, internal gas analysis, metallography/ceramography, fracture toughness testing of the cladding and stoichiometry determination. STUDSVIK NUCLEAR AB 15 A study of the high-burnup behavior of PWR fuel exposed to power ramps, the Ultra-High Burnup Fuel Irradiation Project, is under discussion. In this project a few fuel rodlets, previously irradiated in a power reactor to a burnup of at least about 60 GWdltU, will be irradiated further in the R2 test reactor to a total burnup of about 80 GWdltU. The properties and behavior of the fuel after different types of ramps will be studied. A new investigation of the ramp resistance, ULTRA RAMP, would address the question of the ramp behavior at high burnup (above 50 MWd/kgU). This question has been widely discussed in recent years as regards both normal and off-normal ramp rate conditions. However, only limited experimental infonnation seems to be available so far (31). The concern addressed appears to relate to the impact of changes in the physical properties of the fuel pellets at high burnup and their effects on the ramp behavior of the fuel rods. The fuel pellets seem to crack up in minor fragments and may no longer behave as solid bodies. The fission gases will be entrapped in a magnitude of small bubbles and might cause fuel rod swelling on up-ramping. Other concerns relate to the loss of thermal conductivity and the impact of the rim zone on fuel ramp behavior, The prospective ULTRA-RAMP project would constitute a combination of three groups of ramp experiments: one series concentrated on the PCI phenomena under normal operating conditions in different types of fuel and the other two series concentrated on more safety- oriented issues. Thus the ramp resistance in current fuel types would be studied both under normal operating conditions ("slow" ramps) and under off-normal operating conditions ("fast" ramps corresponding to ANSI Class IT and ill events and "ultra-fast" ramps corresponding to some ANSI Class IV events). A few recent simulated RIA experiments (Reactivity Initiated Accidents) with high-burnup fuel (55 and 65 MWdlkgU) have focused interest on Class IV events. STUDSVIK is proposing a new type of "ultra fast" ramps, faster than the fast ramps performed in earlier safety-related projects (TRI, TRII) but slower than the simulated RIA experiments. These new "ultra-fast" ramps could reach e.g. 120 kW/m during an 1 sec effective ramp time, corresponding to an enthalpy increase of30 caIIg. An MTR type reactor typically gives high neutron fluxes due to its compact design with a high • power density in the core. The thermal neutron flux can be further increased by overmoderation in a flux trap in the test rig. This method gives a test position in the reactor with a thermal flux level of more than 4xl014 nI(cm2·s), which is sufficient to give a PWR test rod (with a burnup of 45 MWdlkgU) a steady maximal LHR of over 100 kW/m. The flux trap also functions as a stabilizing tool for the test position, resulting in a reactivity balance for the rod in - out of less than 50pcm. Small test rodlets of about 10 cm length refabricated by the STUDFAB technique, are to be tested in the R2 flux trap facility. The test rod will be pushed from a non-flux position to a high- flux position by a mechanical device, achieving high power levels in the high flux position. The following measurements of interest can be performed during the experiment: Elongation measurements, internal pressure and diametral changes. The energy deposition is controlled by the time the test rod is located in the high-flux position in the flux-trap facility. Different operational parameters can be selected by choice of temperature (from cold to operational temperature), pressure (from low to operational pressure) and power (from zero to operational power). STUDSVIK NUCLEAR AB 16 Thus in the ULTRA-RAMP project a number of high burnup fuel rodlets of BWR or PWR types, or both, would be exposed to fast and slow ramp rates to preselected tenninal power levels and to "ultra-fast" ramp rates to preselected enthalpy increases, all in the R2 test reactor (31). The main objective is to identifY any adverse or inadequate fuel rod behavior as for example abnormal fuel rod swelling, fast failure of the cladding or loss of fuel integrity on clad fracturing causing dispersion offuel particles in the coolant water. Detailed non-destructive and destructive examinations (including advanced types of ceramography) would follow. A more "conventional" PCI resistance study, SupER-RAMP IIIIIOxI0, is also under discussion. In the new types of IOxlO BWR fuel the Linear Heat Rate is lower than in earlier types of fuel and the PCI resistance is presumed to be correspondingly improved. However, some utilities using zirconium liner with earlier types of fuel have raised the question whether the lower linear heat rates in lOx 10 fuel really make the added resistance against PCI failure, achieved with zirconium liner fuel, unnecessary. As far as is known no ramp tests have ever been performed on lOx 10 fuel. Thus the proposed SUPER-RAMP IIIIIOxlO project would be similar to the earlier SUPER- RAMP W9x9 project (32). It is planned to start in 1997 and be completed in 1997. A new type of test is planned where the stored energy (enthalpy) of different types of fuel rods will be investigated, the STEED project. The stored energy (the enthalpy) in fuel rods during operation depends on the fuel design (dimensions, materials), operating conditions and burnup. Experimental determinations of the stored energy provide a valuable alternative to fuel code calculations. Depending on the test technique used the accuracy can be better than the corresponding code calculations, especially with increasing burnup. The amount of stored energy is an important parameter in connection with safety considerations such as LOCA evaluations. The experimental results for unirradiated fuel can also serve as excellent bench- marking opportunities for fuel modellers. STUDSVIK's R2 test reactor is well suited for scram experiments where the thermal response of different types of fuel can be compared. The measurements for the STEED project will be performed by analyzing the heat release from a test rod after scram, thereby using the R2 test reactor's calorimetric rod power measurement system. This system is the very same as that used in ordinary ramp experiments. • A demonstration experiment, STEED-I, is in progress on unirradiated fuel rodlets. The objective is to verify the test technique and to evaluate the accuracy. An international project, STEED-II, based on tests of irradiated fuel rodlets, will be discussed in 1997. An example ofa "pre-project" demonstration test is shown in Figure 15. A safety-oriented PCI study, TRANS-RAMP m, is also under discussion. Experience from the earlier safety-oriented STUDSVIK projects, as well as from power reactor operation, shows that non-penetrating cladding cracks form during certain short-time power transients and cracks initiate already within 5-10 seconds. However, it is conceivable that these non-penetrating (incipient) cracks will not propagate further during continued operation owing to some passivation effect. The main purpose of the TRANS-RAMP IV project was to investigate experimentally the propensity for through-wall crack penetration in PWR fuel rodlets of such initially non-penetrating (incipient) PCI cracks following a second power transient (29-30,51). In the TRANS-RAMP III (TR III) project the influence ofnon-penetrating (incipient) cladding cracks in BWR fuel rodlets on the PCI failure resistance during an anticipated subsequent STUDSVIK NUCLEAR AB 17 transient occurring later in life will be studied. A project proposal will be circulated when data from the TR IV project will be available. in late 1996. Test data from the international R&D projects are often used as "benchmarking" data in the project participants' own fuel modeling work. In recent years many ramp test data have also been analyzed with the INTERPIN code, developed by STUDSVIK (33-35). INTERPIN is a fuel performance code which satisfies real-time simulation requirements when implemented on a minicomputer. 5.5 Bilateral Programs European, Japanese and U.S. fuel manufacturers and research organizations have also for many years been utilizing the R2 test reactor and the associated hot-cell laboratories for bilaterally sponsored research'. ABB Atom AB has made many series oframp tests. General Electric Co. has executed several series of ramp tests at R2, as part of the efforts to develop the zirconium barrier fuel concept. Some of the ramp techniques requested were innovative, for example the "double ramping" of the test rods. Exxon Nuclear Co. (later Advanced Nuclear Fuels Corporation, later Siemens Nuclear Power Corporation, now Siemens Power Corporation) is another major customer as well as Hitachi Ltd. and Toshiba Corporation. Mitsubishi Heavy Industries, Ltd. has performed combined power cycling and ramp tests. B&W Fuel Company (now Framatome Cogema Fuels) performed ramp tests on high-burnup PWR fuel. Tests have also been performed on behalf of other organizations but the results have not always been published. During the 1970's extensive series ofHTR fuel irradiations were performed in a special HTR gas loop system operating with on-line measurements and analyses of the released fission gas and of the fuel temperature (5). 5.6 In-House R&D Programs STUDSVIK NUCLEAR's in-house R&D work is mainly associated with improvements of test irradiation techniques, instrumentation and post-irradiation examination, all in support of ongoing or upcoming irradiation projects. Progress in these areas has made it possible to achieve • important progress in fuel research. For example, the characterization of the PCI failure progression in some recent projects was only made feasible through a combination of several new techniques. These included a very fast ramp execution using a ramp rate of up to 3 000 W/(cm·min), compared to the previous maximum of200 W/(cm·min), a prompt detection of the through-failure event using the on-line elongation detector and a subsequent special clad bore inspection technique. Another result of the in-house R&D work is the noise measurement technique mentioned above. STUDSVIK NUCLEAR has also been carrying out an in-house R&D program aimed at improving the performance of LW R fuel by the utilization of a design concept with cladding tubes which have been "rifled" on a micro-scale (36). Results ofR2 irradiations of such fuel and the associated modeling work have been published (37-44). 1 A list of available publications can be obtained upon request from the authors. STUDSVIK NUCLEAR AB 18 6 Materials R&D The R&D work in this area consists of studies of irradiation effects in structural materials. These types of studies have been concentrated on pressure vessel steels, stainless steels and nickel-base alloys (for super-heater and fast reactor cladding) during the 1960's and on zirconium alloys since the 1970's. The early pressure vessel steel work comprised investigations of different potential pressure vessel materials, of different materials variables, and of the influence of different irradiation conditions (neutron fluence, irradiation temperature etc). Recent work has been concentrated on accelerated irradiation of materials actually used in pressure vessels under as realistic conditions as possible. More recent work on stainless steels was to a large extent concentrated on fusion reactor materials within the Next European Torus (NET) program, where tensile tests, fatigue tests, and stress corrosion tests have been performed after irradiation to displacement doses of up t() 10 dpa. Some work, including post- irradiation creep and fatigue tests and crack propagation studies (including CT tests) has also recently been performed on potential FBR vessel materials. Among the in-pile tests performed stress-relaxation tests can be mentioned. In-pile corrosion tests in the R2 test reactor with on- line corrosion potential measurements were discussed in Section 2.8. The work on zirconium alloys is continuing and is being expanded in order to include in-pile corrosion studies. Originally structural materials (pressure vessel steels, stainless steels, nickel-base alloys and aluminum alloys) were to a large extent irradiated in rigs in fuel element positions either in contact with the reactor coolant water (temperature about 60°C) or in rigs where the heat- transfer conditions were closely defined and the specimen temperatures were measured but not regulated. Later rigs containing in-pile furnaces for constant temperature irradiations were introduced. Besides numerous detailed publications review papers were published on the early pressure steel work (45) and on the corresponding work on the other structural materials (46). During this period extensive work was also pursued on cladding materials for superheated reactors and for steam-cooled fast reactors. Some work was also performed on early Swedish candidates for FBR pressure vessel materials (47, 48). During the 1970's the in-pile loops in the R2 test reactor became available for irradiations of specimens of structural materials. Such specimens are now irradiated either in rigs that only allow irradiations during whole 400 hour reactor cycles or in rigs where shorter irradiations, down to less than an hour, are possible. The specimens are either in direct contact with the loop water (temperature selected in the range 230-350 0c) or in some cases specimens of pressure vessel steels have been nickel plated in order to avoid corrosion problems during longer irradiations. Later new types of in-pile rigs for fuel element positions have also been developed where the specimens are heated by gamma heating. In these rigs close temperature control (about ± 10°C in the range 250-350 °C) has been maintained by placing the specimens in specimen holders filled with a NaK alloy. The temperature is monitored by thermocouples placed in some specimens or in the NaK adjacent to the specimens. Temperature control is achieved by changes of the HeINe gas mixture in the gap between the capsule containing specimens and NaK and the rig secondary containment. Varieties of these rigs are also used up to a temperature of 550°C. Current material irradiations are partly within the frame of the ITER project and partly within the future DEMO project. For the ITER project, work is carried out mainly on stainless steel and copper joined by hot isostatic pressing (HIP). Irradiations related to the DEMO project mainly comprise a special reduced activation stainless steel, type F82H. Long time irradiations STUDSVIK NUCLEAR AB 19 up to 2.5 dpa are performed in the in-pile loops, where the specimen temperature is 290°C. High-temperature irradiations at temperatures up to 400 °C are performed in NaK capsules placed in a central position in the core. In the latter case the specimen doses will range up to 0.7 dpa. 7 Neutron Transmutation Doping of Silicon Neutron transmutation doping of silicon for industrial use in electric power components is done in a facility which is shown in Figure 16. Three shelves with silicon ingots can be irradiated simultaneously, but only the bottom shelf is shown in the figure. The silicon ingots are loaded manually onto the irradiation facility, which is situated in the R2 reactor poo~ in front of the reactor vessel. The ingots perform a horizontal helical movement on the shelves in front of the core. The neutron flux is monitored through self-powered neutron detectors and the velocity of the ingots and hence the neutron fluence is controlled by a computer. After completion of the irradiation the ingots are removed from the shelves to a conveyor which slowly transports them to the pool surface. A permanent radiation instrument monitors the dose rate and in order to avoid hand doses to the operators the ingots are lifted with a crane. They are then stored for a few days in order to let the 31Si and 32p activities decay. The decontamination is done by rinsing in demineralized water. Silicon ingots with lengths up to 600 mm and diameters from 60 to 152 mm are treated routinely. The target resistivity of the resulting conducting material usually lies in the 30-300 ohmcm range. The high uniformity and precision of the irradiation guarantees less than 5 % axial variation and a radial gradient which is better than 2 and 4 % for the minimum and maximum diameters, respectively. The day-to-day constancy of the operation of the facility is monitored by means of cobalt monitors attached to some of the silicon ingots. The irradiated material is shipped in compliance with IAEA regulations. The bulk material must have a specific activity of less than 7.4 Bq/g before it is shipped as non-radioactive material. For the nonfixed surface contamination a limit of 0.4 Bq/cm2 is maintained taking the ALARA principle into consideration. Application of these values to the process normally gives a turnover time of three weeks at Studsvik. • 8 Neutron Activation Analysis The present set-up for neutron activation analysis (NAA) permits multi-element determination by instrumental and radiochemical NAA of around 50 elements in trace concentrations. The samples which are investigated are of biological, environmental, industrial, or geological origin. The samples usually require little or no pre-treatment, and after weighing they are pneumatically transported in plastic capsules to a position close to the reactor core. Having been irradiated the samples are left to decay in a storage unit for a suitable period of time, before the gamma radiation is registered with a Ge(Li) or a high purity Ge detector. The thermal and epithermal fluxes are measured with a Zr monitor. Thus the automatic data evaluation system gives an absolute determination of the composition of the sample from the intensity of the gamma spectrum, which is characteristic of each element. STUDSVIK NUCLEAR AB 20 Neutron activation analysis in theory pennits detennination of around 70 elements. In practice the number is limited to some 30 elements when using instrumental NAA. Application of radiochemical NAA increases this number. Many elements can be determined at sub-ppb(IQ-9) levels, but high concentrations of disturbing elements may be troublesome because of spectrum interferences. Sometimes corrections have to be made for other reasons, for instance due to interfering nuclear reactions. Neither the lightest elements nor lead and sulphur can be detected. Some examples of STUDSVIK NUCLEAR projects for which instrumental NAA has proved to be an efficient method are: Multi-element studies of geological samples, with special interest in rare earth elements and iridium. Uranium, thorium and other elements in sediment. Trace elements in metallurgical products. Trace elements in food-stuff. 9 Radioisotope Production Radioisotopes can be produced over a wide range of conditions in several irradiation positions in and around the R2 vessel. The operational cycle of the reactor, however, to some degree limits the number of isotopes that are produced routinely. There are six permanent rigs in the reactor core which are used for radioisotope irradiation. One of them can be loaded and unloaded during reactor operation. The maximum flux which can be obtained for irradiation is as high as 3xl014 n/(cm2·s). The permanent rigs can be supplemented with temporary installations. 1921r is produced by irradiation in the core positions. The specific activity of the resulting product is higher than 10 TBq/g. The encapsulated isotope is used industrially for gamma radiography. 169Yb, which is also produced, has the same application. Eu is irradiated with the aim to extract 1S3Gd, which is used for sources in bone scanners. 32p and 35S are examples of isotopes produced in R2, which are mainly used for biological research. There is also a wide variety of radioisotopes being produced for medical research and therapy, • such as 85Sr, 89Sr, 86Rb, 155Cd, 11OAg, 51Cr, 59fe,45Cs, 47Cs, 9OY,I86Re and 63Ni . A few other isotopes are produced routinely. 24Na for example is delivered to the Swedish defence forces for training purposes. 10 Beam Tube Expt;riments The R2 and R2-0 reactors serve as sources of thermal neutrons for a wide variety of basic research applications. The beam tubes at the R2 reactor are used for thermal neutron scattering experiments, see Figure 17 and Table 6. The R2-0 reactor, which is mobile in the pool, is in one position used as source for a boron neutron capture radiography facility and in the other position as source for a facility for nuclear physics and nuclear chemistry experiments based on an on-line isotope separator. Researchers from the universities have easy access to the facilities through the Studsvik Neutron Research Laboratory (49).The laboratory is organized as a department at the University of Uppsala but serves users from all Swedish universities. The instruments are also available for outside users, partly through a program financed by the European Community. STUDSVIK NUCLEAR AB 21 In connection with the replacement of the reactor vessel, a large D20 box was installed on the outside of the core box. The reentrant beam tubes end at positions close to the thenna! flux peak in the D20. Two tangential beam tubes were installed through a region in the biological shield which was previously inaccessible. The new beam tubes are rectangular with height 18 em and width 8 cm. The larger vertical divergence of the beams increases the flux at the experimental positions. With these modifications substantial improvements in the thenna! flux and in the ratio of thermal to fast flux at the experimental positions are achieved. The research performed involves, for example, structure determinations in crystals and amorphous systems, studies of magnetic phenomena in condensed matter, excitations in disordered systems and determination of residual stress and texture. Table 6 Neutron Scattering Instrumentation at the R2 Reactor. Neutron Instrumentation beam tube HI Test beamline H3 REST DiftTactometer for residual stress and texture measurements H5 NPD High-resolution powder diffi'actometer H6 Vacant H7 SLAD Diffractometer for disordered materials H8 SXD Single/crystal diffractometer H9 PREFECT Polarized neutron reflectometer under construction HI0 TTOF Time-of-flight spectrometer (under construction) The above mentioned facility for boron neutron capture radiography (BNCR) is shown in Figure 18. A very pure thermal neutron field is produced by moderation of the fast neutrons from the reactor in a large D20 volume positioned immediately outside the pool liner and adjacent to the reactor core, which is located immediately inside the pool liner. At the outer edge of the D20 volume irradiations can be made in a thermal flux over a large area (30x30 cm2). The thermal flux used in recent experiments has varied between 6xlOS and 5xl09 n/(em2·s), corresponding to a reactor power of 25 to 200 kW. The facility is used extensively for biomedical research and has proved to be an efficient tool for studying the distribution of boron loaded compounds with • a specific affinity for certain tumors . A variety of nuclear physics and nuclear chemistry research programs are based on the on-line isotope separator OSIRIS at the R2-0 reactor. The main activity is aimed at studies of the properties of short-lived neutron-rich nuclides. The programs include determination of fission yields including branching ratios for gamma decay from fission products and determination of the antineutrino spectrum at a nuclear reactor. The system utilizes a novel method for plasma creation and allows higher temperatures, up to 2500 °C, and thereby shorter delay times for the released fission products. This has increased the number of nuclides available considerably and has increased the production yields of many short-lived isotopes by factors of 102-1 ()3. STUDSVIK NUCLEAR AB 23 9 OGUMA, R, BERGDAHL, B-G, SCHRIRE, D Application of a Recursive Identification Technique to Noise Analysis for Fuel Performance Study. Specialists' Meeting on Reactor Noise (SMORN-V), Munich, FRG, 12-16 October 1987. 10 SCHRIRE, D, OGUMA, R The Thermal Response of a Fuel Rod with a Small Xenon-Filled Gap. Studsvik Energiteknik AB, Sweden 1988. (STUDSVIKINF(P)-88113). 11 . MOLANDER, A, JANSON, C In Situ Corrosion Potential Monitoring in Swedish BWRs. Proc Fifth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors. Monterey, Cal, USA, 25-29 August 1991, P 118-125. 12 TOMANI,H Studsvik AB, Sweden. Patent Pending. 13 BERGENLID, U, GUSTAFSSON, I Examination of Fuel Rods by Means of Neutron Radiography. Post-Irradiation Examination. Grange-over-Sands, UK, 13-16 May 1980. Proc. of the Conf. BNES, London 1981, p 49-53. 14 BERGENLID, U, GUSTAFSSON, I Examination of Fuel Rods by Means of Neutron Radiography. Neutron Radiography. J P Barton & P von der Hardt (Eds). Brussels 1983, p 349-353. 15 FORSYTH, R S, (Ed.) The Hot Cell Laboratory - a Short Description of Programs, Facilities and Techniques. Studsvik Energiteknik AB, Sweden (1986). (STUDSVIKINF(P)-86/29) . • 16 The Hot Cell Laboratory Studsvik Nuclear AB, 1995. 17 MOGARD, H, KJAER-PEDERSEN, N A Review of STUD SVIK ,s International Power Ramp Test Projects. Studsvik Energiteknik AB, Sweden (1985). (STUDSVIK-85/6). Mm Perfonnance of Fuel and Cladding Material under Reactor Operating Conditions, G. Mulding & W. Dietz, (Eds), Kernforschungszentrum Karlsruhe, FRG, (1986), P 63-71. 18 MOGARD, H, GROUNES, M STUDSVIK's International Fuel R&D Projects - a Review. Studsvik Energiteknik AB, Sweden (1986). (STUDSVIK-86/2). STUDSVIK NUCLEAR AB 22 References 1 COLE, T E, COx, J A Design and Operation of the ORR. Peaceful Uses of Atomic Energy. Proc. Int. Conf., Geneva, 1-13 September 1958. Vol. 10. UN IAEA, New York & Vienna 1959, p 86-106. (AICONF. 15/P/420). 2 SANDKLEF, s, TOMANI, H Irradiation Facilities for LWR Fuel Testing in the Studsvik R2 Reactor. AB Atomenergi, Sweden (1973). (AE-478). 3 RONNBERG, G, BERGENLID, U, TOMANI, H Power Ramp Test Technique at Studsvik. Proc. KTGIENS/JRC Meeting on Ramping and Load Following Behaviour of Reactor Fuel, Petten, The Netherlands, 1978. (EUR 6623 EN, P 37-51). 4 SANDKLEF, S, BOOR, R Neutron Absorber Techniques Developed in the Studsvik R2 Reactor. Irradiation Facilities for Research Reactors. IAEA, Vienna, 1973 (STIlPUB/316), p 213-231. (lAEAlSM-165/37). Also AB Atomenergi, Sweden 1973. (AE-468). 5 SANDKLEF, S Irradiation Facilities for Coated Particle Fuel Testing in the Studsvik R2 Reactor. AB Atomenergi, Sweden (1973). (AE-467). 6 BERGENLID, U, MOGARD, R, RONNBERG, G Experimental Observations of the PCI Failure Occurrence on Power Ramping. IAEA Specialists' Meeting on Pellet-Cladding Interaction in Water Reactors. RisO, Denmark, • 22-26 September 1980. (lAEA IWGFPT/8, p 121-127) . A!m Res Mechanica Letters 1(1981), p 229-234. 7 BERGENLID, U, RONNBERG, G Observations of Cladding Failure During Ramp Tests. IAEA Specialists' Meeting on Power Ramping and Cycling Behaviour of Water Reactor Fuel. Petten, the Netherlands, 8-9 September 1982. (lAEA IWGFPTI14, P 55-71). Also Res Mechanica 12(1984): 4, p 297-301. 8 OGUMA, R, SCHRIRE, 0, BERGDAlfl.." B-G Fuel Rod Thermal Performance Studies Based on Noise Analysis. Studsvik Energiteknik AB, Sweden (1986). (STUDSVIK-NF(P)-86/08). STUDSVIK NUCLEAR AB 24 19 MOGARD, H, RONNBERG, G LWR fuel research at Studsvik - a review. ENC'86, Geneva, 1986, Transactions, Vol. 4 (1986), P 59-66. 20 GROUNES, M, DJURLE, S, LYSELL, G, MOGARD, H Review ofSTUDSVIK's International Fuel R&D Projects. IAEA Technical Committee Meeting on Fission-Gas Release and Fuel-Rod Chemistry Related to Extended Burnup. Pembroke, Ont., Canada, 28 April- 1 May 1992. (IAEA-TECDOC-697, P 124-130). 21 SCHRIRE, D Rod Overpressure Experiment (ROPE) - Pre-Project. Studsvik Energiteknik AB, Sweden 1987. (STUDSVIK/NF(p)-87/43). 22 SCHRIRE, D Sweden's International Fuel Rod Overpressure Experiments. Nuclear Europe Worldscan (1991): 1-2, p 60-61. 23 GROUNES, M, TOMANI, H Tying up Fuel Rod Overpressure Problems with ROPE I and II. Nuclear Engineering International 37 (1992): 451, p 30, 32. 24 MOGARD, H, GROUNES, M, TOMAN!, H Test Reactor Investigation of the Process of Degradation of Defect LW R Fuel. Jahrestagung Kerntechnik '91. INFORUM, Bonn 1991, p 339-342. 25 MOGARD, H, GROUNES, M, TOMAN!, H Experimental Studies in the R2 Test Reactor of Secondary Damage Formation in LWR Fuel Rods with Simulated Fretting Defects. Jahrestagung Kerntechnik '92. INFORUM, Bonn 1992, p. 353-356. • 26 MOGARD, H, GROUNES, M, TOMAN!, H, LYSELL, G Studies in the R2 Test Reactor of Secondary Damage Formation in LW R Fuel Rods with Simulated Defects. lAEA Technical Committee Meeting on Fuel Failure in Normal Operation of Water Reactors: Experience, Mechanisms and Management. Dimitrovgrad, Russian Federation, 26-29 May 1992. (IAEA-TECDOC-709, P 184-192). 27 MOGARD, H, GROUNES, M, LYSELL, G, TOMANI, H Experimental Studies of Post-Failure Degradation ofBWR Fuel Rodlets. Proc. 1994 International Topical Meeting On Light Water Reactor Fuel Performance. West Palm Beach, Fl, USA, 17-21 April 1994, p 423-434. 28 GROUNES, M Studsvik's Fuel R&D Projects - a Review. lAEA Technical Committee Meeting on Water Reactor Fuel Modelling at High Burnup and its Experimental Support. Windermere, u.K., 19-23 September 1994. Paper 1.11. STUDSVIK NUCLEAR AB 25 29 MOGARD, H, DJURLE, S, GROUNES, M, LYSELL, G, KJAER-PEDERSEN, N Effects of High Power Transients on the Operational Status ofLWR Fuel - Experimental and Analytical Investigations at Studsvik. Proc International ENS/ANS Conference on Thennal Reactor Safety, "NUCSAFE 88". Avignon, France, 2-7 October 1988, p 181-190. 30 MOGARD, H, DJURLE, S, GROUNES, M, LYSELL, G Effects of High Power Transients on the Operational Status ofLWR Fuel- Experimental Investigations at Studsvik. IAEA Technical Committee Meeting on Behaviour of Core Materials and Fission Product Release in Accident Conditions in LWRs. Aix-en-Provence, France, 16-20 March 1992. (lAEA-TECDOC-706, P 28-34). 31 MOGARD, H, GROUNES, M Studsvik's Experience Related to LWR Fuel Behavior at High Burnup CSNl Specialist's Meeting on Transient Behavior of High Burnup Fuel, Cadarache, France, 12-14 September 1995. 32 HOWE, T, DJURLE, S, LY SELL, G Ramp Testing of9x9 BWR fuel. Fuel for the 90's. International Topical Meeting on LWR Fuel Perfonnance. Avignon, France, 21-24 April, 1991. Vol 1, p 828-837. 33 KJAER-PEDERSEN, N A Novel Fuel Rod Perfonnance Simulation Methodology for Predictive, Interpretive and Educational Purposes. Structural Mechanics in Reactor Technology (SMiRT9), Vol C, p 53-61. Balkema, Rotterdam 1987. 34 KJAER-PEDERSEN, N A Novel Fuel Perfonnance Simulation Methodology for Predictive, Interpretive and Educational Purposes. , Res Mechanica 25 (1988), P 321-338. 35 KJAER-PEDERSEN, N Correlation Between the Potential for Thermal Feedback and Some Principal Fuel Rod State Variables. IAEA Specialists' Meeting on Water Reactor Fuel Element Perfonnance Modelling in Steady State, Transient and Accident Conditions. Preston, UK, September 19-22, 1988.(lAEA IWGFPT/32, P 302-306). 36 MOGARD,H Suppression of PC I Induced Defects by Lightly Undulating the Bore Surface of the Fuel Cladding. Studsvik Energiteknik AB, Sweden 1977. (STUDSVIK-77/2). STUDSVIK NUCLEAR AB 26 37 MOGARD, H, BERGENLID, U, DJURLE, s, LYSELL, G, RONNBERG, G Irradiation Testing of an Advanced Fuel Cladding Designed for Load-Follow and Extended Bumup Operation. IAEA International Symposium on Improvements in Water Reactor Fuel Technology and Utilization. Stockholm, Sweden, 15-19 September 1986. IAEA, Vienna 1987, p 335-351.{lAEA-SM-288115). 38 MOGARD, H, BERGENLID, U, DJURLE, S, LYSELL, G, RONNBERG, G Irradiation Testing ofan Advanced Fuel Cladding Designed for Load-Follow and Extended Bumup Operation. Studsvik Energiteknik AB, Sweden 1986. (STUDSVIK-86/1). 39 KJAER-PEDERSEN, N, KINOSHITA, MY, MOGARD, H Comparative Studies of Axial Gas Mixing in Fuel Rods with Standard and Rifled Cladding. IAEA Technical Committee Meeting on Power Ramping, Cycling and Load Following Behaviour of Water Reactor Fuel. Lyon, France, 18-21 May 1987. (lAEA IWGFPT/28, P 25-32 and lAEA-TC-624113). 40 KJAER-PEDERSEN, N, MOGARD, H In-Reactor Performance of Fuel with Rifled Cladding. ANS Topical Meeting on LW R Fuel Performance, Williamsburg, Va, USA, 17-20 April 1988, P 284. 41 MOGARD, H, DJURLE, S, L YSELL, G, OGUMA, R Continued Irradiation Testing of Fuel with Rifled Cladding: Thermal Behaviour and PCI Failure Resistance. Enlarged Halden Group Meeting, Loen, Norway, May 9-13, 1988. 42 KJAER-PEDERSEN, N, MOGARD, H Controlling Overall Fission Gas Release by Controlling Axial Gas Mixing. IAEA Specialists' Meeting on Water Reactor Fuel • Element Performance Modelling in Steady State, Transient and Accident Conditions. Preston, UK, September 19-22, 1988. (IAEA IWGFPT/32, P 269-278). 43 MOGARD, H, KJAER-PEDERSEN, N Improved PCI and FGR Performance ofLWR Fuel using Rifled Cladding. Technical Committee Meeting on Fuel Performance of High Bumup for Water Reactors. Studsvik, Sweden 5-8 June 1990. (IAEA IWGFPT/36, p 147-159). 44 MONTGOMERY, R 0, RASHID, Y R, KJAER-PEDERSEN, N Theoretical Evaluation of Rifled Cladding for LWR Fuel. Nucl. Eng & Des 132 (1992), p 309-316. STUDSVIK NUCLEAR AB 27 45 GROUNES, M Review of Swedish work on Irradiation effects in Pressure Vessel Steels and on the Significance of the Data Obtained. Effects of Radiation on Structural Metals, ASTM STP 426. Am Soc Testing Mats, 1967, p 224-259. 46 GROUNES, M Review of Swedish work on Irradiation effects in Canning and Core Support Materials. Effects of Radiation on Structural Metals, ASTM STP 426. Am Soc Testing Mats, 1967, p 200-223. 47 GROUNES, M Irradiation Effects in Pressure Vessel Materials for Steam-Cooled Fast Reactors. Irradiation Effects in Structural Alloys for Thermal and Fast Reactors, ASTM STP 457. Am Soc Testing Mats, 1969, p 156-179. 48 GROUNES, M, RAO, S New Alloy Steels for Nuclear Reactor Pressure Vessels and Vessel Internals. Trans ASM 62 (1969), P 902-914. 49 STUDSVIK NEUTRON RESEARCH LABORATORY Annual Report 1995. 50 CHRISTENSEN, H, MOLANDER, A, LASSING, A, TOMANI, H Experimental Studies of Radio lysis in an In-Core Loop in the Studsvik R2 Reactor. EUROCORR 96. Nice, France, 24-26 September 1996. 51 DJURLE, S, HOWL, D, JOSEPH, J, GROUNES, M The Influence of Non-Penetrating Cladding Cracks on Rod Behavior under Transient Operating Conditions - Data from the International TRANS-RAMP IV Project at Studsvik. • CSNIlPWG1 Specialists Meeting on Nuclear Fuel and Control Rods: Operating Experience, Fuel Evolution and Safety Aspects. Madrid, Spain, 5 - 7 November 1996. 52 GRAsLUND, C, LYSELL, G, OGATA, K, TAKEDA, T Studies of the Secondary Hydriding Process in Fresh Test Fuel with Stimulated Primary Defects. ANS 1997 International Topical Meeting on Light Water Reactor Fuel Performance. To be published. STUDSVIK NUCLEAR AB 28 Appendix 1 Overview of STUDSVIK NUCLEAR's International Fuel R&D Projects 1975-1991 Project Fuel Type Base Irradiation Research Objectives Data (duration) {No of rods) (MWdlkgU) published INTER-RAMP BWR R2 Failure threshold Yes (Ref 1,2) *) (1975-79) (20) (10-20) Failure mechanism Clad heat treatment Modeling data • OVER-RAMP PWR Obrigheim Failure threshold Yes (Ref3,4) (1977-80) (39) (10-30) Design parameters BR-3 Modeling data (15-25) DEMO-RAMP I BWR Ringha1s I PCI remedies (Annular, Yes (RefS) (1979-82) (5) (IS) niobia doped pellets) DEMO-RAMPn BWR Wdrgassen Failure threshold Yes (Ref6,7) (1980-82) (8) (25-29) PCI damage by overpower transients SUPER-RAMP BWR Wdrgassen Failure threshold Yes (Ref8) (1980-83) (16) (30-35) High burn-up effects Monticello PCI remedies (30) Safe ramp rate Gdfuel PWR Obrigheim Design parameters Yes (Ref 8) (28) (33-45) Modeling data BR-3 (28-33) SUPER-RAMP BWR Oskarshamn 2 Safe ramp rate No EXTENSION (9) (27-31) (1984-86) • PWR Obrigheim Resolve No (4) (30-35) unexplained failure resistance TRANS-RAMP I BWR Wdrgassen Failure boundary Yes (Ref6,9) (1982-84) (5) (18) Crack init. and prop. Structural changes Fission gas release Modeling data *) References, see p. 31-32 STUDSVIK. NUCLEAR AB 29 Appendix 1, cont. Project Fuel Type Base Irradiation Research Objectives Data (duration) (No of rods) (MWdlkgU) published TRANS-RAMP II PWR Zorita Failure boundaIy Yes (Ref 10) (1982-86) (7) (30) Crack init and prop. Structural changes Fission gas release Modeling data ROPE I BWR Ringhals Investigate clad creep- Yes (Ref 11) (1986-89) (4) (36) out as a function of rod overpressure SUPER-RAMP BWR Dresden PCI performance Yes (Ref12) Dl9x9 (4) (30) (1987-91) TRANS-RAMP IV PWR Gmvelines In1luence of non-penetrating Yes (Ref 13) (1988-95) (7) (20-25) cmcks on PCI failure resistance ROPE II PWR Ringhals Investigate clad creep-out as a No (1990-94) (6) Obrigheim function of rod overpressure (>40) DEFEX BWR Initially Study secondary damage formation No (1993-95) (6) unirmdiated in fuel rods with simulated fretting rodlets defects • STUDSVIK NUCLEAR AB 30 AppendiIl Overview of STUDSVIK NUCLEAR's Ongoing and Planned International Fuel R&D Projects Project Fuel Type Base Irrad Research Objectives (duration) (No of rods) (MWdlkgU) DEFEXll BWR ND<) Study secondary damage formation in fuel rods with (NS-> , 1997-99) +PWR? (10-20) simulated fretting defects ULTRA-RAMP BWR ND Study PCI performance (NS.1997- PWR >So UHBUP PWR Power Fuel behavior during ramp tests (3) Reactor to>60 thenR2 (80) SUPER-RAMP III BWR ND Study PCI performance 11Oxl0 (NO) (NS. 1996-98) TRANS-RAMP III BWR ND Influence of non-penetrating cracks on PCI failure resistance (NS. 1997-99) STEED II BWR ND Stored energy PWR (enthalpy) measurements • ·)NS = Not yet started NO = Not decided STUDSVIK NUCLEAR AB 31 Appendix 1 - References 1 THOMAS,G The Studsvik INTER-RAMP Project: An International Power Ramp Experimental Program. J Nucl Materials 87 (1979), p. 215-226 .11m Proc KTGIENS/JRC Meeting on Ramping and Load Following Behaviour of Reactor Fuel. Petten, The Netherlands, Nov 30 - Dec 1, 1978. H. R()ttger, Ed. (EUR 6623 EN, P 95-106). 2 MOGARD, H et al. The Studsvik INTER-RAMP Project - An International Power Ramp Experimental Program. ANS Topical Meeting on LWR Fuel Perfonnance. Portland, Oregon, USA, April 29 - May 3, 1979, P 284-294. (DOElET/34007-1). 3 DJURLE, S The Studsvik OVER-RAMP Project. Studsvik Energiteknik AB; Sweden. 1981. (STUDSVIK-STOR-37). 4 HOLLOWELL, T E, KNUDSEN, P, MOGARD, H The International OVER-RAMP Project at Studsvik. Proc. ANS Topical Meeting on LWR Extended Bumup - Fuel Performance and Utilization. Williamsburg, VA, USA, 4-8 April 1982, Vol. 1, p 4-5 to 4-18. 5 FRANKLIN, D G, DJURLE, S, HOWL, D Performance ofNiobia-Doped Fuel in Power-Ramp Tests. Nuclear Fuel Performance. Stratford-upon-Avon, UK, 25-29 Marsh 1985. Proc. BNES, London 1985. Vol. I, P 141-147. 6 BERGENLID, U et al. The Studsvik Power Transient Programs DEMO-RAMP II and TRANS-RAMP I. lAEA Specialists' Meeting on Pellet-Cladding Interaction in Water Reactor Fuel. Seattle, WA, USA, 3-5 October 1983. (lAEA IWGFPTI18, P 50-55). 7 MOGARD, H et al. t The International DEMO-RAMP n Project at Studsvik. Nuclear Technology 69(1985):2, p 236-242. 8 MOGARD, H, HECKERMANN, H The International SUPER-RAMP Project at Studsvik. Proc. ANS Topical Meeting on Light Water Reactor Fuel Performance. Orlando, FL, USA, 21-24 April 1985. ANS, La Grange Park, n.., 1985. (DOEINEI34130-1, Vol. 2, P 6-17 to 6-33). 9 MOGARD, H et al. The International TRANS-RAMP I Fuel Project. Fuel Rod Internal Chemistry and Fission Product Behaviour. Technical Committee Meeting, Karlsruhe, FRG, 11-15 November 1985. Proc. (IAEA IWGFPT/25, P 157-167). STUDSVIK NUCLEAR AB 32 10 MOGARD, II, HOWL, D A, GROUNES, M The International TRANS-RAMP n Fuel Project - A Study of the Effects of Rapid Power Ramping on the PCI Resistance ofPWR Fuel. ANS Topical Meeting on L WR Fuel Performance. Williamsburg, Va, USA, 17-20 April, 1988, P 232-244. 11 SCHRIRE, D, SONTHEIMER, F, LYSELL, G ROPE-I: The Studsvik BWR Rod Overpressure Experiment. Proc. 1994 International Topical Meeting On Light Water Reactor Fuel Perfonnance. West Palm Beach, F1, USA, 17-21 April, P 212-219. 12 HOWE, T, DJURLE, S, LYSELL, G Ramp Testing of9x9 BWR Fuel. Fuel for the 90's. International Topical Meeting on LWR Fuel Perfonnance, Avignon, France, 21-24 April, 1991, Vol 1, P 828-837. 13 DJURLE, S, HOWL, D, JOSEPH, J, GROUNES, M The Influence of Non-Penetrating Cladding Cracks on Rod Behavior under Transient Operating Conditions - Data from the International TRANS-RAMP IV Project at Studsvik. CSNIlPWG 1 Specialists Meeting on Nuclear Fuel and Control Rods: Operating Experience, Fuel Evolution and Safety Aspects. Madrid, Spain, 5 - 7 November 1996 . • STUDSVIK NUCLEAR AB 33 ~I ~ I:L _ J I I I I I - .. ~ ..l Figure 1 The R2 Test Reactor - General Arrangement. Hl 0 Beryllium reflector G Control fod 0 fuel element ~ Loop no H~_ H6 0 Rig or disposable position 0 Irradiation facility Figure 2 The R2 Test Reactor - Core Configuration. STUDSVIK NUCLEAR AB 34 MAGNETITE CONCllET£ ". " R£"CTOR POOL Figure 3 The R2-0 Reactor. 0.. .. ..... Ptenure C-onduc1Mf, ~llIIrdy lO. ., Oetayfl-J nevtfQl"l detector s.tt~f4td "teclor 1or'I.~ha~ Pt ..." "".t.oorI=S"'''.'''. =.. -___ Figure 4 BOCA Rig in R2 - Simplified Flow Sheet and In-Pile Part ofa Boiling Capsule. STUDSVIK NUCLEAR AB 3S Hydraulic drive One-rod magazine Lock valve Reactor cover C3 B4 Tm II , Tell A,B I In-pile tube, inlet 2 In-pile tube outlet 3 Portable shIl elded strainer 4 Main cooler 5 Strainer 6 Main circulating pumps 2 7 Flow regulating valve 8 Electricn eaters 9 Recuperator and cooler 10 lon-exchangers II Water analysis system 12 Pressurizer tank 13 FP-monitor 14 Radiation monitor 15 Feed pump 16 He-3 coils 17 Ramp capsule [,i;::' • F FIB·:. ··•• f-------.--D<-- R2 Natural Noies ------> R2 Power. Calorimetric R2 Power. Nucleonic Signal Linear Heat Rate Time Figure 9 Comparison of Signals from Random Noise and Background Noise Electrode rod --ic-I Water outlet --------1 Water inlet -.,---~~ Addition of impurities I additives __~ • Main loop flow Rig tube In-pile loop pressure tube D-Ic-1r- Heating coil Electrode Test specimens\ Reactor core Figure 10 INCA Test Rig STUDSVIK NUCLEAR AB 39 Cooling r-"""'--.,----{.+-_.--'--'----1.---<>+_----'-----L----l M-aIn -in-p1Ie_ Hz Hz Degassed high pUrity"""", Tetlng Reactotcore .TO-S' I Figure 11 INCA Feed Water System and Analysis System Figure 12 Array Simulating Primary Defects in Fuel Rods (Patent Pending). STUDSVIK NUCLEAR AB 40 F NF INTER-RAMP • o OVER-RAMP + SUPER-RAMP, W • o SUPER-RAMP BWR, KWU ... A • - -- --• '.- - - -• -o-•-•!0 'C--: C - - - - •- -W• ..-.. -jj .... -- -- - - -~.!. - - -Cl-Ua.ngGl- - --- - - h. - ~ ..... _C - -- (! @ DO 0 AA F NF OVER-RAMP, KWU/CE • 0 SUPER-RAMP, PKI ... V' SUPER-RAMP. PK2 • ¢ SUPER-RAMP, PK4 .... A SUPER-RAMP, PK6 • 0 » A o "VV' <> - - - - - -~ - - - - - ~I- - - • -;'i> ~ - -~-0= - - - - - - - - - - - - - ..... - - - --~~ - - - - -<:>- aD O~ 00 A 0 o 0 Figure 13 Summary of Some Data From the INTER-RAMP, OVER-RAMP and SUPER-RAMP Projects. The Incremental Failure Threshold as a Function of Bum-Up for Different Groups of Fuel Rods. STUDSVIK NUCLEAR AB 41 INTACT CLADDINC FAILURE THR.ESHOLD >SEC -NIN .(,HR LOG TIME Figure 14 Schematic PCI Failure Progression Diagram. STEED Test Evaluation 18 ••••••••• • Measured Power 16 • '1 R2 Gamma Heating in Rod • R2 Gamma Heating in Construction 14 + Measured Power - R2 Gamma Heating Curve Fit 12 10 • 3: ~ 8 6 4 '1 '1 '1'1'1'1'1'1'1 2 •••••••••' 1 ·iii 0 -20 -10 o 10 20 30 40 sec Figure 15 STEED Project - Example of Stored Energy Evaluation. (the top curve and the two bottom curves are measured, the second one from the top is evaluated from the others). The stored energy is obtained from the area under the derived curve. STUDSVIK NUCLEAR AB 42 \ \ REACIt)RTANK ',---:.::..:::.::=-..:~~+.---:::::::~~ tma~y REAcroR CORE Figure 16 Silicon Irradiation Facility. TEST BEAM H1 • Figure 17 PREFECT Future Layout of the Instruments at the R2 Test Reactor. Explanation see Table 6. STUDSVIK NUCLEAR AB 43 " . .. •• c '. . " ": ". " .... " ". .. .... .. ... :. • .. '. ' "j 'F"' ='= "='= "' ='" ="" --, r-------.., ( I PO,OL MOVABLE CONCRETE I CONCRETE BLOCK • • a' I BLOCK • • I LOS.OING POSITION I I I 1"1 I "I III I "•• I " ( I I •• u_. • ( I I I I t-~=rJ-- .. •" - ____ J • I •• • -~ o SCJlte 1 m Figure 18 The R2-0 Facility for Neutron Capture Radiography" MAIN EXPERIENCES IN RENOVATION OF THE DALAT NUCLEAR RESEARCH REACTOR Tran Ha Anh, Pham Van Lam, Nguyen Nhi Dien and Ngo Phu Khang Nuclear Research Institute, Dalat, Vietnam ABSTRACT The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. Some structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam- tubes and the radiation concrete shielding were retained. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. It is mainly used for research, isotope production, neutron activation analyses and training. During the recent years, the reactor control and instrumentation system has been renovated due to ageing and obsolescence of its components. Reactor general inspection and refurbishment were carried out in order to ensure and strengthen reactor safe operation. Thanks to the renovation works, DNRR has been continuously operated and effectively utilized. This paper presents main experiences acquired in the renovation of the Dalat nuclear research reactor. I. INTRODUCTION The Dalat Nuclear Research Reactor was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The latter was put into operation in 1963. In 1975 all the fuel elements were removed from the core and transferred elsewhere. During 1982-1983, the reactor was reconstructed. Some structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding were retained. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The natural convection mechanism of light water for reactor core cooling was kept unchanged. The core is loaded with VVR-M2 fuel elements with 36% enrichment. The renovated reactor reached initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. The maximum thermal neutron flux was 2.lE13 n/cm2/sec. Since then DNRR has been operated safely. It is mainly used for research, isotope production, neutron activation analyses and training. The configuration of active core was changed several times, improving the experimental condition of the reactor. During the recent years, the reactor control and instrumentation system has been renovated due to ageing and obsolescence of its 1 components. Reactor general inspection and refurbishment were carried out in order to ensure and strengthen reactor safe operation. Thanks to the renovation works, DNRR has been continuously operated and effectively utilized. This paper presents main experiences acquired in the renovation of the Dalat nuclear research reactor. 2. RENOVATION OF THE DALAT NUCLEAR RESEARCH REACTOR [1-2] 2.1 Aim of the renovation The main aim of the renovation was to redesign the reactor in keeping its old geometrical structure with an utterly different kind of fuel material, whereas requesting higher technical features such as doubling the power to 500 k W but conserving the natural convection mechanism for reactor core cooling. The most significant feature is obtaining as high neutron fluxes as possible for the given power, thus allowing the most effective use of the reactor. 2.2 Renovation works Reactor reconstruction was carried out during 19 month period. During reconstruction, some structures of the former reactor such as the aluminum reactor tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding were retained. The rather high level of radiation dose rate inside the old reactor tank had been unfavorable for the assembling process of the reactor core structures. Thus the solution of an entirely suspended structure was adopted for the reactor core. A cylindrical chemise is installed inside the aluminum tank. It has 1.90 m diameter and 3.56 m height and is hung from the top of the reactor to serve as a suspender of the core structures. The upper part of the suspender has some structures to support the control rod, the neutron detector or temperature sensor tubes, a tube for loading samples into the rotary specimen rack, a channel for transferring spent fuel bundle. At the bottom part of the suspender is a supporting base. The base has a height of 20 cm and is used for suspending the extracting well and the core. The fuel elements use 36% enriched U235 fuel made of uranium - aluminum alloy. Each element is laid out in three concentric layers to increase the heat exchange superficies. Fuel elements are of the VVR-M2 bundle type and composed of 3 concentric layers : 2 circular inner tubes and a hexagonal outer tube. The fuel layer with a thickness of 0.7 mm is wrapped between two aluminum alloy cladding layers of 0.9 mm thickness. The core is located in distilled water at a depth of 5 m. Water is used as moderator, coolant, reflector and biological protection. The neutron trap is situated at the center of the core. It is made of beryllium and has a diameter of 65 mm. The region between the active core and the old graphite reflector is also lined with beryllium. One of the most important technological systems of the reactor is the control and instrumentation system, which controls the reactor during normal operation and scrams it if a threat of accident appears. For determining the neutron flux, the reactor control system uses 9 detectors placed in dry channels at different heights and 2 positions at the outside of the graphite reflector. The detector measure neutron flux in three ranges : source, intermediate and energy ranges. Seven control rods of the reactor are located symmetrically in the core : two safety rods, four shim rods and one automatic regulating rod. The safety and shim rods are made of boron carbide and the automatic regulating rod of stainless steel. These rods are driven by a quite complex system. During reactor operation all important information from the rector technological systems is received and processed by the control and instrumentation system and finally displayed on the control console. In other to cope with higher power, the cooling efficiency was increased by the installation of an extracting well placed above the core. The cooling system of the reactor has two loops : a primary loop between the reactor tank and the heat exchanger, and a secondary loop to extract heat from the heat exchanger and release it into the atmosphere via a cooling tower. The cooling system were improved, the heat exchanger was replaced to increase the heat transfer areas, the cooling tower of the secondary cooling system was also improved to increase the cooling capacity, so finally to raise the reactor power. To ensure radiation protection for the staff working inside the reactor building, a layer of heavy concrete was added to the lateral shielding of the reactor at intermediate level to protect against gamma radiation from the reactor core, and also from strengthened upward water flow bearing activated elements produced in the reactor core. The cover of the reactor tank is made of a steel plate with thickness of 15 cm. It is used for radiation protection during reactor operation. The former bulk shielding has been modified into a storage tank for spent fuel elements. Depth of the tank is 3.7 m. The tank has covers and is filled with distilled water. It has 300 holes to contain spent fuel elements. A description of reactor components, section view is shown in figure 1. The reactor building is equipped with ventilation systems, which ensure good working conditions and allow safe gas release to the atmosphere after being purified via 40 metre-height stack. The first working configuration of the core was obtained on February 1984 with the 72 hr successful test operation of the reactor. The configuration consisted of 88 fuel elements with neutron trap in the center of the core and one wet channel in position 7-1. The irradiation facilities in this core comprised of 2 vertical pneumatic irradiation channels together with the central neutron trap and the wet irradiation channel. Surrounding the reactor core, a rotating tray containing 40 irradiation holes was arranged at the same position as the former "Lazy Suzan". The 4 horizontal neutron beam ports and the thermal column were retained from the old reactor. Refueling operations were carried out in April 1994, which consisted of adding 11 new fuel elements in the core periphery at previous beryllium element locations. This fuel reloading ensured the reactor to be exploited for 3 - 4 years before another refueling is needed. The working configuration of reactor core with 100 fuel elements is shown in figure 2. The introduction of a steel cover for the reactor tank, the suspended structure of the reactor core, the use of a reactor control system made of 7 control rods as well 3 as the additional irradiation facilities inside the reactor core make the reactor's internal structure rather complicated in comparison with the former reactor. Handling operations in the reactor core become more subtle, and so do the operations aimed at inspecting periodically the reactor tank and other internal structures. These operations are even not practicable in some areas inside the reactor tank. This is a problem of great concern, because structures kept from the old the reactor are now more than 30 years old and their periodical inspection must be performed as rigorously as possible in order to detect any unacceptable defect if exists, and thus to prevent any incident leading to the loss of integrity of the second barrier. 3. RENOVATION OF CONTROL AND INSTRUMENT AnON SYSTEM [3-4] 3.1 Aim of the renovation After ten years of reactor operation, the system has revealed some features of ageing and obsolescence. As its components and equipment were produced in the 70's, some integrated circuits (IC) and equipment were hard to fmd out on the market and as a consequence maintenance work of the system became difficult. To solve this problem the Dalat Nuclear Research Institute has asked for the assistance of the IAEA through the TC Project VIE/4/010 - "Renovation of the Dalat Reactor Control and Instrumentation System". The design obeyed some principles such as keeping unchanged the old mechanical standard and the technical specification of the old boards, but making them more reliable. Some electronic boards/blocks/sub-systems which mainly affect the system reliability will be renovated by using more modem IC and equipment. 3.2 Renovation works The instrumentation and control system of the Dalat research reactor can be divided into four main sub-systems as follows: '" Neutron Flux Control sub-System - NFCS '" Control Console Display sub-System - CCDS '" Control Logic sub-System - CLS '" Process and Instrumentation sub-System - PIS Some important boards of the NFCS were renovated. Because of low reliability, low quality of indication, the reactor control console display system was totally designed and constructed according to the project program. All electronic boards of the system were designed on the EUROCARD standard. The software of the system was written in Pascal language. Data are saved in hard disk 120 MB or in floppy-disk 1.2 MB. To increase the accuracy and the reliability of rod position indication, the indicators and related electronic board have been replaced. The indicators are of a digit and bar-graph type - INDICOMP A2000. 4 All relays in intermediate relay board and transistors in the amplifier boards of control logic block for safety rods have been replaced by items with higher quality and reliability. Decreasing currents fed to LED indication and consequently decrease in power consumption in all electronic boards has an important meaning in the reliability improvement of the CLS. Intermediate relays, power amplifier boards of control logic block for compensating rods have been renovated or replaced. Also, intermediate relays and power amplifiers of the AR control logic block have been renovated or replaced. The PIS has been totally renovated in the frame of the project VIE/4/0 10. The renovated instrumentation and control system of the Dalat reactor has been commissioned and put into operation since November 1993. The system has been working with high reliability during the three years of its operation. 4. CONCLUSION After renovation, the maximum power of DNRR was increased to 500 kW. The maximum thermal neutron flux was 2.lEI3 nlcm2/sec. Some structures of the former reactor were retained. The technological systems of the reactor were newly designed and installed. The natural convection mechanism of light water for reactor core cooling was kept unchanged. The renovated reactor attained its planned nominal power in February 1984. Since then DNRR has been operated safely. During the recent years, the reactor control and instrumentation system has been renovated due to ageing and obsolescence of its components. Thanks to the renovation works, the DNRR has been continuously operated and effectively utilized. REFERENCE [1] TRAN HA ANH et al., Main Safety Lessons from 5-year Operation of the Renovated Dalat Nuclear Research Reactor, lAEAlAECL Research Reactor Symposium, Chalk River, Oct. 1989. IAEA-SM-31 0178 [2] VU HAl LONG et aI, Dalat Nuclear Research Reactor - Safety Analysis Report, Dalat, 1989 [3] TRAN HA ANH et al., Eight-year Operation of the Dalat Nuclear Research Reactor and the Necessity of Renovation of Its Control System, IAEA Regional Nuclear Research Seminar, Bangkok, May 1992. IAEA-SR-179/15C [4] NGUYEN NHI DIEN, The Dalat Reactor Instrumentation and Control System after Renovation, IAEA technical Committee Metting on Technology and Trends for Research Reactor Instrumentation and Controls, Ljubljana, December 1995 5 Table 1. Main characteristics of the Dalat Nuclear Research Reactor Parameter NatureNalue Reactor type Swimming pool Power,kW 500 Fuel U-AI alloy Critical mass, g U-235 2781 Number of fuel elements 100 Uranium loading, g 4015 Enrichment, % U-235 36 Moderator and coolant Light water Core cooling mechanism Natural convection Excess reactivity, $ 6.5 Xenon poisoning effect, $ -1.65 Temperature and power effect, $ -0.36 Control rod worth, $ : 2.97 - Shim rod No.1 3.09 - Shim rod No.2 2.70 - Shim rod No.3 2.50 - Shim rod No.4 0.50 - Regulating rod 5.36 - Two safety rods Thermal neutron flux, nlcm2/sec : - Neutron trap 2.2E13 - Wet channel 1-4 1.28E13 - Dry channel 7-1 5.1E12 - Dry channel 13-2 4.2E12 - Rotary specimen rack 4.27E12 - Thermal column 5.5E9 - Horizontal beam tube No.3 2.5E6 - Horizontal beam tube No.4 5.8E7 Maximum surface temperature of fuel, °c 97.2 Maximum water temperature in the core, °c 54.5 6 ., ~ fh J'l 4= p;J .J eM 2 -,. 1////·· '-- r- 3 [r-l i I - ~ I' 4 J w ~ ~ f1\ '3 ~ ~~;.,:~;z:~ ~ ~ ~ ~ :.-;~ ~~ 8 = ~ ::;::?"~:'.:-: . :::;~.[~~~~~ I ;'LL-~, - 'j' --'v lr---------------------------------~------------.r--; &6M J Fig, 1 DESCRIPTION OF REACTOR CO~WONENTS. SECTION VIEW I-COVER. 2-TANK. 3-SUSPRNnING SUPPORT. 4-EXTRACTING WELL 5-TIIERMAL COLUMN. 6-COHE. 7-REFLECTOR, 8-SPENT FUEL STORAGE @ r:<3:l Wet irradiation Neutron Imp 19J channel rfJ Pneumatic Ve transfer system Safety rod Fuel Element e Shimrod Beryllium Block ~ Automatic Q regulating rod Fig. 2 WORKING CONFIGURATION OF REACTOR CORE scK~ ScK~ BRl beryllium matrix • The Be-matrix is the life-limiting component of BR2 Ongoing Refurbishment and • Phenomena: Future Utilization of .. swelling ---------> cracking BR2 .. 'He poisoning ---------> operational constraints 9 Be (n • IX) 6H e ---~-- >- 6L· ( nih. IX ) 'H I E.Koonen J3 --> '1-1 - 1996-09-09 L__ ___-__-_>__ '_H_e_ _(_n • 'H ) -1 .!!'_ ____ _ '-.i1 L-.•.._. .._.... .. " - ScK~ BRl refurbishment programme I .BRl = I~\ ,,=!UlIIll'r '-'-'-'--._-_. . II essential research tool for CEN'SCK ""'. EI)" 1 I:Jlllr-----·-·-------- ••• , = optimized design for MTR-use and safety experiments > ~"'!-Refurbishment = PLEX • Requirements: .. replacement of Be matrix .. el) S'- .. new safety case in compliance with modem standards .. availability and reliability of installation • Future use will be focused on Engineering R&D ·~41-- BrlO2 Standard c 1>" nn e t BR2 FIG. 1 ScK~ Major Refurbishment Issues ScK~ BR2 aluminium tank Safety assessments Issue: prospects for life extension • Vessel integrity • PSA -> LOCA assessments • evaluation of consequences of vessel rupture • Instrumentation and support systems • Reliability isolation systems • determination of fluences distributions • Structural and seismic requalifications • evaluation of life-limiting phenomena: • Industrial safety ~ fracture toughness Operational safety ~ low cycle fatigue • Lessons learned from past operation and incidents • Ergonomics • In-Service Inspection • Operator train ing and requalification ScK~ Fluence distribution in vessel wall o <> I .!i ! .~M=-Methods: I. Calculations - __ cak,IOCW I --""' ... ··csoIe,,. 2. Samples from first Be-matrix Ii ., 1k(l\llIb!Jk)·~.n_,.tkt .s .. , ] o qo.boIlt.cSCIl)·plc n ~Io r.: 3. Samples from shroud surrounding the vessel • :;en • llfMA :-.~~·~t·"~· __ ... 1 ____ .. _. __ • TIn ~ • lor;}. Dosimetry: Il- .! - __ ... ooIlI..t.t 27 28 I-' 28 ... thennal fluence: AI ( n , Y> AI ----> Si 1J I, .. 9 10 Be(n, Y) Be • I.--.I .. -.-I .. --I-.... ~-.-...I .-'-'-_, .I_, __ w M ~ I~~ 1M ,~ m m ~ m ~ 984 .,I_I'"'trotlll"(~'"J ... fast fluence : Be ( n , 2n > Be ----> 2 He Flr.., Admllthallll,(rlh. or tbe Ilu:rm;t1 Rttence (n ...1 ) til tbe URI ,,""I (at fb~ mar. nUJ ptIIlU:) $II mltl.I IJ9S at ult'ulntl"d, II' drduC'l!'d Irlun mt'lIJ1ut'mtftfJ nn Iht nrst (II!' nUltrb and tI' deflated ,,.om th~ 11• . ,,1111 tnU,",.~men't Fig.' ModoIedfalJguolJe ..... 1or A!I052-0 (tri'.U.-.edl C.umparisun with lieN (CII "",US £or unhnrilluled 1.5154-0 and JAERI pro(lOSCd dcsItn ItJllauc wrte rut annr#dioted At atIup looon r,--------------or--------------or--------------,---------------, ScK~ Fatigue analysis of AI tank § Areas of concern : - central cylindrical shell - inlet & outlet nozzles i !l!- Methodology : I - ASME VIII, div. 2, class I components ] - Fatigue curves for irradiated material theoretically established Conclusions : 10 I I I I ID 100 1000 100110 100000 the damage factors for the contemplated Al!cwnble numt.!r 01 qdel life extension are acceptable provided the _._ JAERI: 1t:$IsOb tmh.Dl1ura. mamly AJ051 .. 0 '" Mi06I -+- r: trend-curves - microstructural characterisation - chemical analysis Materials: - BR2 shroud material, obtained by Two successive modelling steps: electro-erosion cutting I. dislocation theory guided by transmission electron - BR2 initial control rod guide tubes microscopy - remnants from first HFR-Petten vessel 2. two complementary micromechanical models: relate Klc to the flow properties ; i n c: _~ -.-~ __ .) _0" m _ 1---,...g, . . - 0 ~ ~ o . l~ a .- ____ ~ .-~ . ~ m ;~( m 5 3 ; 3 !f ; " ~ .g. SCK~C E~ Fracture Toughness - Conclusions :} l :: -< ~ . ~ ii 3 I .3 - -,. ". -" i ~.. , n ~ _.u • <_D u OJ 0 12 MPaJm 3 ; :0 j ]) 0 3 N ~ I '" . »'" ~ i! ~ '" e- • No significant future operation risk if inspection of the vessel ffi . : [• • 3 ~ . 3 '" I ~ I Ii 5 ffi 5" does not reveal any surface-breaking crack of depth exceeding IW e-~ ('l 3 -''"" • • .c. 3 3 mm, nor any embedded flaw causing a local KI in excess of % l! ~ ('l ~ /i z 0 possibility of containment by-pass ~ remote visual inspection by means of cameras Analysis: ~ remote mechanized geometrical control .. RELAP model of BR2 primary circuit ~ penetrant testing ultrasonic testing Actions: ~ ~ Eddy current testing ~ reengineering of control & actuation systems of major automatic block valves Results: ~ reduction of closing times of those valves ~ no flaw with unacceptable dimensions .. enhanced capability of pool-communication valve .. the AI-tank has not undergone any significant ~ new fast reactor-trip on loss of pressure alteration since the 1980 inspections VeJ'.~et Primary_Loop ScK~ BR2 instrumentation & control Instrumentation: ·'~i~·~~'ll~i .. nuclear: fault-tree analysis of trip-lines -> minor modifications to eliminate single and common mode failures ·~·I-!.:~~,~1J!!j-.i::-J .. process: FMEA & common mode analysis L -> renewal of all safety related instrumentation (redundancy, diversity and testability) .. mf·t.",r.-~.;;.' ;;":"IEJ~l;::;:·:'·:·' •u.~•. . . "~:"~:': .." ~",<. -:"--~:~" •'C•• $·M- .3... ..".,,.".'j.r.':.! " ·::·'· ~: '4•••" :'ae ;::'·I:=·::*·:;i;¥:·.T::.:""'· I~·"· :L. -~-~Il:.i"r'i·;-: ': . -,~ .. r , .. Support systems: FMEA & common mode analysis . ,:)~:iH'*""'T:p:FCt~~r~~~!:::" .. 110 V dc: decentralisation & separate batteries ----_._-----_.-----. _._._._------- .. 220 V ae - vital net: increase of reliability r.3· $ .. compressed air: renewal of compressors and redesign of distribution network SCK~ BR2 seismic evaluation ScK~ Operational safety Reference earthquakes : • Ergonomic study -> new reactor control desk .. historical records most significant earthquakes max. intensity hypocentre: VII MSK • Emergency control panel outside C-8 .. safe shutdown earthquake: horizontal PGA = 0.1 g • Data acquisition system, integrating reactor vertical PGA = 2/3 * 0.1 g control, experiments and radiation-protection Methodology: ASCE4-86 standard • Lessons learned from past experience and incidents -> action plan ' Calculated structures : Reactor building, pool, vessel, isolation systems, • Accident management procedures damage by fall of cranelheavy structures, fuel storage pool, primary pipework outside C-B • Operator training and requalification programme --------------------------------- ScK~ Operation licence ScK~ Future utilization of BRl • needs only to be prolonged, not renewed Base load: internal R&D programmes • procedure: decennial safety reassessment • PWR fuel (MOX): (same procedure as Power Stations) thermal conductivity, fission gas release, peMI • basic documents : • Pressure Vessel Steel Embrittlement updated SAR, including results from studies, inspections, improvements and upgradings of RPV surveillance, modelling efforts refurbishment programme • PWR structural components • regulatory body is represented in the refurbishment embriUlemenl, IAsee review committee -------------------- -_._---- ------------- ------- ~ ----_. . ----- ScK~ Standard irradiations devices at BRl IS PWR-loop CALLISTO .. 3 IPS in standard channels .. up to 9 fuel rods in each IPS .. fresh and pre-irradiated fuel .. realistic PWR conditions • PWC capsules associated with lHe-screens tj' U'1I ..... 1\1 g. •- single-rod transient tests ~a 0.. 3 ~, .- n n n --.--;I .. ~ :>- • Reflector rigs: steel and structural materials r.:~' ~ I[ fl~r~--- c2.C 05- rn t [ • Gamma irradiation facilities a • >-l s% :~ ~I!~~ e =, 0 5 • ~g ~ II~ g" •~ ScK~ New irradiation devices for BR2 ScK~ Special capabilities of BR2 £. «% MM5f"r.,ih ''1'.. <,.-'."_'-~- .,-cc, ... «.d • Pool Side Facility: PSF BR2 is designed to accept complex experiments and • for vessel-steel & remote handling programmes large loops, as required for severe-accident simulations • space available by removal of 2 tangential beam-tubes • Power transients : • Feasibility studies: - ramping up to fuel pin failure, with possibility for post-transient irradiation ~ PWR-type integrated loop: DESTIN • in central 200 mm flux trap - non-adiabatic RIA tests, e.g. control-rod withdrawal • up to 25 fuel rods (fresh or pre-irradiated) • steady-state and transient tests • Core blockage formation ~ solid breeder blanket test-module: CEPHEID • Core debris coolability • in central or peripheral 200 mm channel • integrated loop An international context is required for these programmes ScK~ Production activities in BR2 ScK~ Conclusions • Radioisotopes • Refurbishment underway => ready early '97 - Mo-99, 1-131, Xe-133 (PRF and DGR devices) • Future utilization: - Ir-192 (up to 700 Ci/g) mainly Engineering R&D - Co-60 (300 Ci/g) • Base load: internal programmes - Sr-89, Re-186, Sm-153, Eu-154 • BR2 available for: • NTD-Si - SIDONIE irradiation device ~ scientific collaborations - mainly 5 inch ingots or wafers ~ irradiations on request RESEARCH AND SERVICES OF LVR-15 REACTOR IN REZ J. Kysela, O. Erben, V. Knoblocll, J. Burian, V. Bro! ABSTRACT Main features and characteristics of the LV R-15 research reactor are presented. Reactor is used for different purposes including material testing, radioisotope production, silicon doping, neutron beam channels and neutron capture therapy. Material testing in realised at rigs and loops which involves PWR or BWR chemistry conditions. Main radioisotopes produced at L VR-15 is for medicine. The two in-pile section is mounted in a reflector for doping of silicon where 3"crystal are irradiated. Thermal column of the reactor is used for the intensive study and development of neutron capture therapy. At the reactor nine horizontal beam channels are installed, five of them is utilised for material structure research by neutron difUaction methods. DESCRIPTION OF REACTOR LVR-IS The L VR-15 is a light-water moderated and cooled tank nuclear reactor with forced cooling. The maximum reactor power is 10 MWth. The fuel type IRT-2M enriched to 80 % or 36 % and a combined water-beryllium reflector are used. REACTOR CHARACTERISTICS maximum reactor power IOMWth maximum thermal neutron flux in the core 1,5 x 1018 nlm2s maximum fast neutron flux in the core 3 x 1018 nlm2s at the end of beam tube 1 x 1013 nlm2s irradiation channel in fuel 1 x 1018 nlm2s irradiation channel in reflector 3-5 x 1017 n/m2s The reactor core is composed of IRT-2M type fuel assemblies produced in Russia with the enrichment of 80 % 23~. This fuel has been changing for fuel with an enrichment of 36 % 23'U. The reactor core grid has a pitch of 71,S mm and 80 cells. In the basis operation configuration, 28-34 cells contain fuel cells are dedicated to channels for experimental probes and 3-6 cells in the reflector and core periphery are dedicated to vertical irradiation channels (see Fig. 1). The reactor core is situated in the reactor vessel (outer diameter 2300 mm, total height of the vessel 6 235 mm), which is made of stainless steel, the internal parts of the reactor are made of an aluminium alloy. The reactor has a forced circulation of the coolant. The generated heat is transported via three cooling circuits to Vltava river (see Fig.2). 1 The primary circuit contains 5 main circulation pumps and 2 emergency pumps which provide the circulation of the coolant through the reactor core and 2 heat exchanges. The emergency pumps are connected with an emergency battery source and a diesel generator, which secure their operation when the electrical network in the reactor site fails. The coolant flow is 2100 m3/hour. The coolant is a light demineralised water. The protection and control system ensures • Measuring of the neutron flux and of the rate of its change 10 all operational and extraordinary conditions • Start up, operation, and shutdown of the reactor • Control of the reactor with signalling of selected states • Automatic and hand control of the neutron flux The neutron flux is measured with 3 equal sets of measuring channels, realised by the electronic analog system SAKOR B. From the measuring channels; the measured values are transferred to the control panel, into the information system, and to the reactor recording devices. From excessive critical values of the neutron flux of the rate of the change of the neutron flux, the scrams are deduced in accident circuits (a choice "two out three"). Logic and accident circuits perform the emergency shut-down and control of the reactor in all operation conditions. The system is based on relays. The LVR-15 reactor has 12 control rods of equal type. By selection of the rod function in the electric circuits and by location in the active core, 3 safety rods, 8 compensation rods as a automatic regulator are determined. The absorption part of the rod is made of stainless steel filled with B1C. FUEL In the reactor LVR-15 IRT-2M type fuel assemblies are used in two modifications - 4 tubes assembly and 3 tubes assembly. The 3 tubes assembly is installed with the control rod in the centre of the fuel assembly. Until the year 1995 the with enrichment of80 % 23SU was used. From that year the use of the fuel with enrichment of 36 % 23SU been started. The main characteristics of both types of the fuel IRT-2M are: 23SU enrichment 80 % 36% total length 882mm section square - head 71,5x71,5 mm section square 67x67mm mass total of the assembly - 4 tubes 3,27 kg 3,2 kg - 3 tubes 2,64 kg 2,6 kg - 4 tubes 171 g 230 g - 3 tubes 147 g 198 g tube wall thickness 2mm cladding material A! cladding thickness 2xO,8 mm 2xmin. 0,4 mm fuel material U-A! UOrA! fuel plate thickness 0,4mm 0,64 mm active length 580mm 2 The spent fuel is stores in the 2 stores on site of the reactor. In 1995 a new high active waste storage was commissioned. All spent fuel will be transported to the new storage facility in due course. The transport route is outside the NRI area through Rei: village and back to the NRI but to a different area. For the transport the special container Skoda 1xIRT-M is used. IN-PILE MATERIAL RESEARCH Materials related problems cause a significant portion of power plant outage time. This is connected with environmental degradation processes (corrosion, mechanical and radiation effects) during power plant operation. At the same time the operators and reactor vendors are interested in plant life extension, higher fuel burnup and lower radiation fields. Towards these aims, specific investigations are performed in facilities simulating environmental parameters of light water reactors including reactor water loop performing in-pile tests in radiation environments. Experimental test facilities allowing exposition of materials to radiation environment are used. The facilities enabl to carry out research on the behaviour of materials and water chemistry of PWR and BWRs. High pressure water loop RVS-3 is operated at test research reactor. The loop was designed as a universal facility providing wide experimental possibilities. Loop is designed for a pressure of 16.7 Mpa, temperature of 334 0 C and 10 tJh flow rate. Water chemistry used corresponds to the PWRs with boric acid, pH control by potassium/lithium and hydrogen control by hydrogen gas or ammonia. BWR-l loop is designed for investigation of structural materials and water chemistry in conditions of boiling water power reactors. BWR loop has the following parameters: 10 MPa pressure, 300 0 C temperature and 2 tJh water flow rate. CHOUCA MT type irradiation rigs are used for the irradiation of reactor pressure vessel materials within the framework of reactor pressure vessel integrity, lifetime and reliability programme. Second boiling water loop BWR-2 is used for stress corrosion cracking of 50/20 CT specimens. The facility for corrosion of zirconium alloys and cladding-coolant interaction consists of two experimental sections of electrically heated fuel rods with maximum power of 100 W/cm2. One section is situated in the loop outside the reactor core (out-of-pile), the second section is situated in the irradiation channel (in-pile). The facility enables to carry out research in the following fields: • Corrosion of zirconium alloys with/without the effect of radiation fields, • Deposition of corrosion products, • Effects of water chemistry components including lithium, potassium, ammonia, hydrazine and boric acid. PIE (post Irradiation Examination) studies involve metallographic evaluation of the phase composition of alloys, degree ofhydriding, thickness of oxidic layer, and corrosion damage. The SSRT tests are used for the evaluation of irradiation assisted stress corrosion cracking (lASCC). The aim of the experiment is the determination of the effect of neutron radiation and flowing water in specimens loaded with small strain rate in the reactor water loop inside the active channel. 3 The IASCC experiment includes the following activities: • Irradiation of tensile test specimens with a diameter of 6 mm in the irradiation probe CHOUCA-MT (French production) in an inert atmosphere at a temperature of 290-300 0 C and neutron fluence up to 1x1024 n m -2. • Disassembling of the irradiation probe and removal of test specimens. Insertion of the test specimens into the carrier for SSRT. • Carrying out the SSRT in the active channel of the reactor water loop at a temperature of 290-300 0 C, pressure of 12.5 MPa and total neutron fluence up to approximately 5x1024 nm- 2 (E>1 MeV). The tests will be performed at parameters corresponding to PWRs or BWRs. RADIONUCLIDES PRODUCTION A number of reactor radionuclies can be prepared in the Nuclear Research Institute's LWR-15 reactor. These radionuclides are applied either as unsealed sources in the nuclear medicine diagnostics and therapy or as sealed sources for commercial irradiation and in the radiation oncology. Some of these nuclides have been prepared in the NRI laboratories formerly already ~o, 113mln, 12SI, 1311 and 192Ir), while the production of the other ones is the subject of current research (!52Sm, 60Co). The possibilities are being explored of producing some other nuclides with the supposed application in nuclear medicine as radiotherapeutics, namely of 89Sr, 111Ag, I~O, l~r, l~e, 188W and 1910S. SILICON TRANSMUTATION DOPING By the capture of a thermal neutron, the natural isotope 30Si can be transmuted into the unstable isotope 31Si which decays by beta emission into the stable isotope 31p (half-life Of 31p is 2.6 hr). By means of this reaction, donors can produced in a silicon crystal. The capsule with silicon crystals is held with the input device DORA from the preparation channel and inserted into the irradiation channel DONA. The DONA is placed in the outer region of the core. The maximum flux of thermal neutrons reaches 1.7xl0!3 neutrons/cm2s in the irradiation channel at the reactor power of 8 MW. During the irradiation, DONA can perform two types of motion, a rotational motion of 2 rpm around his own axis and a translational one. During the trial openltion it has been desisted from the translational motion. Neutron flux is measured by a self-powered neutron detectors in five positions at the distance of 200 mm. The position of maximum thermal neutron flux is computed from the obtained data by the least squares method. The crystal center is placed into the maximum of the fluence. The course of the thermal neutron flux enables to irradiate crystals up to 200 mm in length. The required dose is entered into the control computer. The neutron-flux detectors measure continuously the neutron flux and the moment, when the actual dose equals the required one, crystal automatically leaves the irradiation position and is pulled out to the so-called parking position. The capsule is gripped by DORA and transferred into hot cells where the crystal is removed from the capsule. One week after it is possible to handle the crystal without danger. 4 Following are produced: Typical parameters of the DONA channel: (10 MW) - max. crystal dimension: - thermal neutron flux density: 2xl017n/m2s diameter 3" (4") - fast neutron flux density: 2xI016n/m2s length 200 mm (300 mm) - typical nominal resistivity: 330, 220, 130, 60 ohms BORON NEUTRON CAPTURE THERAPY FOR CANCER The ultimate goal of cancer therapy is to achieve a degree of selectivity that would spare normal cells and may lead to recurrence, metastases and death. Nuclear reactors have been the exclusive source of neutrons for BNCT. Thermal neutron beams with energies in the region of 0,0025 eV and epithermal beams with energies of 1- 10,000 eV may be used. Thermal neutrons are attenuated rapidly in tissues, and it is difficult to obtain sufficient flux of thermal neutrons at the depth without severely irradiating the surface tissues. An epithermal neutron beam may be produced by using a filter or moderator as a spectrum shifter to slow the fast neutrons to intermediate energy range. Epithermal neutrons are relatively non-destructive, when used to irradiate tissues, provide better penetration than thermal neutrons and produce thermal neutrons at the depth in tissue because of the moderating effects of the tissue itself Such a beam of neutrons peaks the thermal neutron flux a few centimetres distance from the tissue surface. There is a need to develop alternative neutron sources to overcome the limitation of BNCT given by its current dependence on nuclear reactors as a source of neutrons. A filtered beam of epithermal neutrons at the L VR-15 reactor is used as a source for BNCT purposes. The filters are situated in the empty space of the reactor thermal column. The filter consists of cylindrical blocks of 100 cm in diameter, total length is 265 cm. The active part of the filter is composed of 55 cm of aluminium, 15 em of sulphur and 1 cm of titanium. The outer diameter of the beam of 11,5 em is adjusted by the final shutter. The central block is a graphite collimator. The techniques available to measure the neutron fluence include semiconductor Si-Li detectors, solid state nuclear track detectors, fission chambers, silicon diodes, for y-rays thermoluminescence detectors, scintillation spectrometer and for neutron spectrum determination Bonner spheres, scintillation spectrometer, and proportional hydrogen spectrometer. It is important to know the dose distribution, both neutron induced and gamma, along the patient body during the treatment. This parameter was measured using a human whole body phantom. The phantom-skeleton was filled up with water. Measuring channels are situated along the back bone and trough the head. The BNCT method offers an opportunity to treat certain types of tumours that are presently inoperable or highly metastasised. With appropriate quality of the neutron delivery system and boron compound it is now feasible that BNCT can be successfully demonstrated. The Czech RepUblic could be among a few centres where the procedure will be used for treatment as a standard one. 5 10 xxxx xxxx xxxx xxxx xxxx xxxx f####l xxxx xxxx xxxx xxxx xxx x XXXX l..!!!!J Fuel assembly DONA 9 XXXX XXXX XXXX XXXX XXXX XXXX XXXX XXXX XXXX XXXX XXXX xxx x I #eR# I Fuel assembly with a control rod #### 8 BE BE BE BE BE BE Ixx xxxRVS xxx IW ater 7 BE #### #### #### #### #### #### #### #### #### #### EJ RVS-3 loop irradiation channel 6 BE #### IG #CR# #CR# #### #CR# BE #### #### #### ##U #### EJ BWR-lloop irradiation channel 5 BE #### #CR# #CR# #eR# #CR# #### BE IC #### #### #### #### #### #### IC #### #CR# #eR# #CR# #CR# #### B CHOUCA irradiation rig 4 BE BE #### #### #### U## #### #### 3 BE #### Ie flCR# BWR ##U #flU BE o Beryllium #### #### fI### #### 2 BE #### #### #### #### #### #### xxxx G Irradiation channel #### #### #### #### #### U## XXXX 1 BE BE BE BE BE BE XXXX pp rBEl Beryllium with Irradiation channel xxxx xxxx L.!£J A B c D E F G H I DONA I Silicon irradiation channel G Pneumatic rabbit channel Fig. 1 L VR -15 reactor core configuration REACTOR LVR - 15 ::::~;:::==~~~~-------------- REACTORSOVER ,,"*---=- ROTATING STEEL LIDS CHANNEL OF FISSION CHAMBER CONTROL RODS CONCRETE 3HIELOII\IG REACTOR I/ESSEL CHANNEL OF IONISATION SHAMBER REACTOR (~ORE BEAM TUBE PRIMARY CIRCUIT TUBE. INLET --r-_ PRIMARY CIRCUIT TUBE.--+_ _ OUTLET Fig. 2 • 1 of 12 ANSTOJERI08415 HIFA R Major Shutdown Report INSPECTION OF REACTOR ALUMINIUM TANK AND REPLACEMENT OF SECONDARY COOLING WATER CIRCUIT PIPEWORK Sungjoong (Shane) Kim Australian Nuclear Science and Technology Organisation New lliawarra Road, Lucas Heights, N.S.W., 2234, Australia ABSTRACT HIFAR (High flux Australian ,Reactor) is a 100MW DIDO-class research reactor operated by the Australian Nuclear Science and Technology Organisation (ANSTO). Every four years HIFAR has a scheduled major shutdown to undertake inspection, maintenance and upgrade activities that are not possible at other times. The last major shutdown was in late 1995 and lasted 10 weeks during which, and amongst other activities, the Reactor Aluminium Tank was inspected and the major part of the Secondary Cooling Water Circuit pipework replaced. New cooling pipework was designed in accordance with Australian Standards Pressure Piping Code, employing NISA II computer software for finite element analysis where appropriate, with a view to ensuring safety under every conceivable operating and accident condition. Design, manufacture and installation activities were carried out according to ANSTO Engineering's Quality System Procedures (1S09OO1-1994 accredited) and agreed by the Nuclear Safety Bureau, an independent organisation. through staged document submissions. 1. INTRODUCTION HIFAR is a heavy water moderated, reflected (also by graphite) and cooled reactor with a maximum thermal flux of 1.4 x 1014 n cm"2 S"1 at its design power of 10 MW which makes it an ideal source of neutrons to irradiate materials for scientific, medical and industrial purposes. HIFA R is the only reactor operating in Australia. HIFAR first went critical in 1958 and has been in successful operation at full power (10 MW) since 1960. One reason for its long and safe life is the vigorous maintenance program including. but not limited to, chemical control of the cooling water which has clirect bearing on the state of the Reactor Aluminium Tank (RAT) and the cooling water circuits. Corrosion inhibitor and algaecide are added to the secondary cooling water (light water) to minimise the internal pipe corrosion and algae growth. Prior to and during the major shutdown, selected sections of the Secondary Cooling Water Circuit (SCWC) pipework were surveyed using the 'Micro-Map' system which provided thickness maps of the pipe shell to determine degrees of internal corrosion. The results showed that, after 38 years of reactor operation, some of the pipework needed to be replaced. 20f 12 ANSTOIERI08415 2. REACTOR ALUMINIUM TANK (RAT) The RAT has been regularly inspected at each major shutdown since 1970 to identify any corrosion or structural damage that could render it unfit for continued service. Previous inspections have been limited to a visual survey of the interior surfaces of the RAT, but recent improvements in the inspection equipment and procedures enabled greater precision and accuracy in identifying features during the 1995 major shutdown. Six special jigs were designed and manufactured to facilitate inspection and testing of the RAT. These enabled obtaming cast impressions of the RAT internal surfaces and measuring the surface hardness. They were also used for holding ultrasonic probes and moving them inside the RAT in the vicinity of the main welds and inside the primary cooling water pipes. The jigs were held by a manipulator inserted through appropriate vertical facility holes in the top plate of the reactor when the fuel. rigs and heavy water were temporarily removed. Their movements inside the RAT were tracked by a video camera. 2.1 Visual Survey A high resolution colour camera was connected to a Super-VHS video recorder and a laser disc image recording system This enabled individual frames to be captured to the full resolution of the camera. 2.2 Hardness Testing Remote hardness measurements were made on the internal wall and several internal facilities of the RAT to provide information regarding the current post-irradiation mechanical properties of the aluminium material The hardness tester was positioned against the RAT wall by a remote manipulator. The tester was basically equipped with an 'Equotip' head which armed and propelled the spring loaded ball to hit the wall when triggered by a pneumatic remote control device. The velocity change of the ball after the impact was used to give a measurement of hardness. The hardness tester is shown in Figure 4. More than 600 locations were tested to obtain the profile of the tank: hardness as shown in Figure 2. It was found that the hardness varied as a function of distance from the core centre line, however, the maximum hardness was measured in a region slightly away from the core centre line. The hardness measured is the result of two counteracting processes; radiation annealing of the initial cold worked structure which results in a decrease in hardness and transmutation of aluminium to silicon which results in an increase in hardness. Due to these competing processes the position of maximum hardness is not necessarily at the core centre line. Overall, the hardness of the RAT has been found to be simi1ar to that which is likely to have been the installed condition. It should be noted that the accuracy of the results and the phenomena causing these hardness changes are currently under study. Estimation of tensile properties by hardness correlation revealed that the observed hardness changes are of no significance to the structural integrity of the RAT. 3 of 12 ANSTOIERI0841S 2.3 Wall Tbickness Survey An ultrasonic device was used for point measurements of the RAT wall thickness. Demineralised water was used as the coupling medium, with the jig being positioned using the in-tank manipulator. The results showed no measurable loss of wall thickness to date. See Figure 3 for histogram showing distribution of RAT wall thickness values. 2.4 Replication of Surface Features Dental paste was employed to obtain cast impressions of the selected RAT internal surfaces. Uncured paste was pushed against the wall by a manipulator and held in position until hardened (7 minutes) to form a negative replica which was then recovered for production of a positive replica (a replica of the replica) for analysis. This method enabled true three dimensional interpretation of the surface conditions and accurate measurements of defects. e.g. deposits, tide marks. oxide flaking, peening, scuff marks and pitting. The results showed no significant defects. See Figure 5 and Figure 6 for examples of replica images. 3. SECONDARY COOLING WATER CIRCUIT (SCWC) PIPEWORK The secondary cooling water is cooled by six cooling towers and circulated at an approximate rate of 350 kg/s to remove the heat from the primary ~O cooling circuit and the rig heat exchangers inside the reactor. Water chemistry is monitored and controlled daily in order to minimise internal pipe corrosion and algae growth. Total volume of the water in the system is 332000 litres. The SCWC pipework is made of carbon steel pipes, mainly 500 mm nominal bore with other smaller pipe branches. Some sections of the pipework are buried in sand 550 mm deep (ground to the centre of the pipe). See Figure 1 for general layout of the SCWC pipework with the replaced section identified. 3.1 Old Pipe Thickness Inspection The SCWC pipework has been regularly inspected for residual wall thickness using conventional ultrasonic thickness gauges until the last major shutdown. This measured the local thickness and did not give information of the surrounding areas. An advanced ultrasonic imaging system, Micro-Map, was procured to enable more comprehensive surveys during the 1995 major shutdown. The Micro-Map contains a number of modules for performing specific ultrasonic inspections. The "See-Scan" module employs a video capture board capable of monitoring the position of a thickness gauging transducer so that readings of thickness and relative positions can be permanently recorded as the transducer is moved over the test areas until the complete area is covered. The survey results showed that the pipework was in fair condition. but a decision was made well before the major shutdown to replace it to be conservative. Figure 7 shows an example of the Micro-Map image of a test area (approx. 400 mm x 400 mm) and Figure 8 shows chemical deposits on the inside of the pipe. 4ofl::! ANSTOIHRI0841S 3.2 Design 3.2.1 Design Standards The new pipework was designed not only to comply with the current Australian Standards but also to withstand any abnormal conditions such as bushfire and earthquake experienced in Sydney in 1994 and 1989 (weak tremor only), respectively. Australian Standards for Pressure Vessels (ASI210-1989, Class 1) and Pressure Piping Code (AS4041-1992, Class 1) were adopted for the design, manufacture, installation and testing of the new pipework. These standards are comparable with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section 8. NISA n computer software for finite element analysis was employed to calculate stresses and strains in the pipework due to the hypothetical bushfire and seismic events. The original design had to be modified to satisfy both the seismic and bushfire conditions because the former required rigid pipework while the latter required the opposite. The pipe modifications included the following: • thicker pipes for rigidity and greater thermal mass - pipe temperature would not rise as much as the original pipes during bushfire, e.g. 90°C (9.5 mm thick) vs. 107°C (6.4 mm thick); • installation of new pipe supports for the buried sections to control and guide the pipe movements (due to thermal expansion and seismic vibration) in such a way that over-stress in the pipework is avoided without losing flexibility; and • reinforced pipe supports near the reactor building in order to minimise the stress in the reactor building wall penetrations (through which the SCWC pipework is connected to the circuit inside the reactor) due to the pipe movements. The design, manufacture and installation were carried out according to ANSTO Engineering's Quality System Procedures (1S09001-1994 accredited) and reviewed and agreed by the Nuclear Safety Bureau (NSB) through staged document submissions before the execution of the next stage. The NSB is an independent Australian Government organisation responsible for monitoring and reviewing the safety of the nuclear reactor operated by ANSTO. A summary of the design basis conditions and criteria is shown in Table 1. 3.2.2 Reactor Containment Building (RCB) Penetration The reactor is contained in a building made of 21 mm thick carbon steel plates, 21.3 m in diameter and approximately 25 m in height. The RCB is leak tight and classified as a pressure vessel 'Class l' which requires stringent modification procedures and tests. The RCB penetrations, through which the SCWC pipework enters and exits the RCB, were reinforced after the installation of the new pipework and hermetically seal welded. Stress analysis on the penetrations was carried out to justify the design. Sofl2 ANSTO/ERI084IS 3.2.3 Corrosion Protection The external surface of the exposed pipework was painted according to the IDFAR colour scheme. The external surface of the buried pipework was coated with polyethylene for protection against corrosion. This method has been proven to be reliable for the Sydney water supply pipes for many decades. The internal surface of the pipework is not treated (bare metal) because the current water chemistry has proved to be satisfactory. Wlth the increased pipe wall thickness, which provided an additional 3 mm of inherent corrosion allowance, the new pipework is anticipated to surpass the performance of the old pipework which lasted 38 years. provided that the water chemistry is maintained. It should be noted that a working group has been recently formed to review the water chemistry and to automate the chemical dosing system. ;3.2.4 Flowmeter A new in-line electromagnetic flowmeter with 500 mm nominal bore stainless steel body was selected to replace the old 'Dall Tube' venturi type flowmeter because • • the electromagnetic meter is modern, reliable and much more accurate (±O.2%), and • the electromagnetic meter is a full bore design. hence does not restrict the flow. The principal of operation is that an electromagnetic flux is induced across the fluid when the flow inside the pipe cuts a magnetic field generated across the meter (0500 mm pipe with windings on the outside). The flow is an aggregate figure and not affected by the velocity profile of the flow across the pipe or the chemical contents in the fluid. hence yielding accurate readings. The remote transmitter is located inside the flowmeter pit outside the reactor and its signal cables (typical 4-20 rnA) are fed through the RCB penetration cartridge (socket and plug) to each of the display locations inside the reactor. The old analogue gauge in the reactor Control Room was replaced by a digital and graphic display unit for enhanced readability. 3.3 Tests The welding test specimens were mechanically and chemically tested prior to manufacture as required by the 'Class l' pressure piping standard. The weld joints in pipes were tested by radiography during manufacture and by the magnetic particle method during installation. After final installation but prior to filling the trench with sand (for buried pipes) and sealing the ground surface with asphalt. the new SCWC pipework was inspected by a sniffer dog for the presence of any explosives inside the pipework as a security measure, followed by a hydrostatic test 6ofl2 ANS1O/ERI084IS The hydrostatic test was carried out at 460 kPa (gauge), ie, 1.25 times the design pressure (which also included pressure surge due to water-hammer effect), in accordance with the relevant Australian Standards. No structural weakness was observed during the test, however, a few minor leaks through flanged joints had to be corrected by re-seating the gaskets and tightening the flange bolts to the specification. The RCB penetrations were leak tested by pressurising the whole reactor building (RCB) to 10.3 kPa (g) and holding for 24 hours. Pressure drop was used to give a measurement of the gross leak rate through all RCB penetrations and other seals. The test showed no measurable leak through the penetrations. This test is regularly carried out to ensure the leak tightness of the RCB at all times. 4. FUTURE PLAN The rest of the old SCWC pipework (mainly inside the reactor) is planned to be replaced during the next major shutdown in 1999. Seismic analysis and other relevant calculations will be carried out as part of the design. Meanwhile, the water chemistry will be changed to give better protection against internal pipe corrosion and algae growth. The chemical dosing system will be automated for more precise control of the water chemistry. S. CONCLUSION The objective of the inspection was to provide visual and other evidence of the conditions of the RAT and the SCWC pipework. The results from the RAT inspection tasks showed that the RAT is in very good physical condition. There has been no measurable corrosion or loss of wall thickness to date and no evidence of any notable change in physical condition since the previous visual inspection. The only defects observed were insignificant and limited in length and depth. The hardness of the RAT has slightly changed from what is deduced to have been the value at manufacture, however, the observed hardness changes are of no significance to the structural integrity of the RAT or on the extent of conservatism embodied in the original design. No features were observed that would preclude the operation of the RAT until the next major shutdown in 1999. The results from the SCWC pipework inspection showed that the pipework was in fair condition with some internal corrosion as expected from the 38 year old carbon steel construction even though corrosion inhibitor has been used in the cooling water. The main part of the SCWC pipework and its associated equipment (flowmeter, valves, headers and reactor building penetrations) were replaced by the upgraded components designed and manufactured with today's technology. The 1995 major shutdown was brought to a successful conclusion when the reactor returned to full power operation on 28 November 1995, one week ahead of schedule. 7ofl2 ~S11J~10841S (j. ACKNOWLEDGMENT The inspection and testing of the RAT were carried out by Advanced Materials (a division of ANSTO) with support from ANSTO Engineering (a division of ANSTO). The SCWC pipework project was managed by ANSTO Engineering. Many colleagues contributed to the successful completion of this project 7. REFERENCES 7.1 R. Finlay Examination of Corroded SCWC Pipework, R95m044, May 1995. 7.2 R. P. Harrison, C. J. Moss and D. J. Battle 1995 HIFAR Major Shutdown Task Report No. 83, Inspection of the HIFAR Reactor Aluminium Tank, February 1996. 7.3 S. Kim Design Manual for HIFAR Secondary Cooling Water Circuit Pipework, DM105914C, Engineering Job No. et93137, June 1995. 7.4 S. Kim Design Review & Verification for HIFAR Secondary Cooling Water Circuit Pipework, TNI07483, Engineering Job No. et93137, November 1995. 7.5 S. Kim Installation and Connnissioning Report for Secondary Cooling Water Circuit Pipework, ICRI07483, Engineering Job No. et93137, November 1995. 7.6 S. Kim Calculation (Pressure Vessel & Piping) for HIFAR Secondary Cooling Water Circuit Pipework, CL105890B, Engineering Job No. et93137, June 1995. 7.7 G. Peat Design Brief for Replacement of HIFAR Secondary Cooling Water Circuit Pipework, DB105518B, Engineering Job. No. et93137, August 1994 and amended in June 1995. \ 7.8 M. Walsh \ Seismic Analysis of HIFAR Secondary Cooling Water Circuit Pipework, CL105933A, \ Engineering Job No. et93137, May 1995. 7.9 M. Walsh Heat Expansion Calculation for HIFAR Secondary Cooling Water Circuit Pipework, CL105858A, Engineering Job No. et93137, May 1995. Sof 12 ANSTOIER10841S TABLE 1: SUMMARY OF DESIGN BASIS CONDmONS PARAMETER SCWC Pipework & Reactor REMARK Headers Containment Building Penetration 1. Process Fluid light water air 2. Internal Pressure gauge pressure gauge pressure • Design pressure includes the 2.1 Nonnal Oto 150kPa 0 maximum abnonnal pressure and 2.2 Abnonnal Oto 230kPa -6.9 to + 10.3 kPa pressure rise due to water 2.3 Design (0 to 365kPa)'" -7.6 to +11.3 kPa hammer effects caused by valve 2.4 Test 460kPa +1O.3kPa and pump operation. 3. Temperature + Extreme air temperatures on 3.1 Nonnal 1 to 45°C (1 to 45°C) + site (southern Sydney). 3.2 Abnonnal 1 to 90°C .'" 1 to lorC·· •• Calculated maximum pipe 3.3 Design 1 to 100°C Oto200°C temperature due to busbfire near . the reactor. It will be much lower if the water flows inside the pipe. N.B. Headers are inside the RCB and not affected by bushfire. 4. Flow Rate Electromagnetic flowmeter 4.1 Nonnal Oto 355 kg/s 0 replaced the 'Dall Tube' (venturi 4.2 Abnonnal Oto 390kR/s 0 type) flowmeter. 5. Seismic Vibration 5.1 Peak Horizontal Oto 2.3 m/s2 Ot02.3 m/s2 Analysed by theNISA II FInite 5.2 Peak Vertical ot o 1.53 m/s2 ot o 1.53 m/s2 Element Analysis computer 5.3 Dampin~ Value up to 3% of critical up to 3% of critical software. 6. Cumulative Radiation Dose Does not affect the design. 6.1 Nonnal Oto 10 Gy Up to lOGy Neoprene gaskets are suitable for 6.2 Abnonnal Oto 10000Gv UP to 10000 Gv these conditions. 7. Corrosion Externally painted. Painted Buried pipes are externally Protection Corrosion inhibitor in coated with polyethylene. cooling water. S. Material Spec. 8.1 Old system Mild steel pipe Mild steel shell, Construction of the old pipework 0500mmNB, 021 m, 21 mm thick, does not confonn to AS4041 6.4mmWT penetration hole for Class 1 standard. 0500 mm NB pipe is 8.2 New System Mild steel pipe, reinforced with rings New materials for pipes and ASTMA53B, (0790mmOD, plates confonn to AS 1210 and l 0500mmNB, 0520mmlD, AS4041 Class 1 constructions. I 9.5mmWT 2Ommthick) ~ 90112 ANSTOJERI ()8.C IS l ::! eJn ,g.., :-i :- 1 -I ce ,51 ~ :> J <.> J J .f t ~ '! Q, .i... . /""":. , ] / \ ~ ; •, '-' \•. 1 ,, , "- .. --...../ / _._--" Buried Section (m dotted line) ! ~ ' ::z: ..... -- .....•..•.••.• , I Q, , l ~ I ~ flow ~ '. Replaced sectiODS are !hOWl! in bold black 1i.De. \ I \ o 5 10m i- -Reactor _; ~..AJ. 11--1-1--1 Figure 1. General Layout of HIFAR Secondary Cooling Water Circuit 10 at 12 ANSTOJERI084" _ '0 -_-__'0 10 - __ 10 a·t-----~~~~~._------~rt~--~jr--f_----------_t----------__1 ~ _--1I1t. j ItO t-----~---+-\-f-+-_t_+____r____________1 .-..1....1....'. --.., 111 __l "Ott c::Ioo .~ 1f. ~----------_r_----------_I------_Ii_,~--~~--------_;_----------__t -....-at __ ISO M -",-140 _21. "'~-----------r----------_Ir-----------;-~~=------t----------__t -.III / - _-_-_. t .... ,. .~ ~.- ---------~0- ---------~.-..- ----------1I~H- ----------,.+..- ---------I~tI ' Distance from Core Cenue Line (rom) Figure 2. RAT Wall Hardness Profile at Various Angles 20 18 16 14 u>- •c: 12 :r:t 10 B I&. 6 4 2 0 or- C') It! 01 f'! II) gO or- .., ,... .., C') CD Q) N II) gO or- or- .o.r.-. .or- N N N or- .., or- ... ..- .... ..- M..... M.... M.... ... .....;. ..; Iri Iri Iri .- .- .- 'P' Tlckn ... Rang. (mm,. Figure 3. Histogram of RAT Wall Thickness Values (High and low values from the vertical and horizontal welds contributed the observed spread, ie, greater than 13 mm.) II of 12 ANSTOIER 10841 S Jig for housing the Equotip Head. Equotip Head with an indentor ball (0 ] rom). Figure 4. Hardness Tester - 'Equotip' Head and a Jig Deposits are 0.15 rom thick. (Magnified x 3.5) Figure S. Replica Image of Oxide Flaking on RA T Wall <0.24 mm deep) are sun·ounded by raised deposits (0.34 mm). Magnified x 3.5. Figure 6. Positive Replica of Pits on RAT Wall 12 of 12 ANSTOIERI 084 J S Green pixels (darker colour in B&W copy) indicate 6.4 mm thick - I.e., no corrosion. Yellow pixels (lighter colour) indicaJe corroded region with residual pipe thickness of 4.4mm. Figure 7. Micro-Map Image oCthe Old SCWC Pipe Thickness Figure 8. Corrosion Deposits inside the Old SCWC Pipework Modification of JRR-4 T.NAKAJIMA. MBANBA, Y.FUNAYAMA Y.HORIGUTI, M1SSHIKI Department of Research Reactor, Tokai Research Establishment, Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, lbaraki-ken, 319-11, Japan ABSTRACT' Japan Research Reactor No.4 ( JRR-4 ) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel u..c.ed and swimming pool style reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965. general investigation has been continuously succeed so as shielding experiment, fuel and material irradiation experiment, RI production, nuclear activation analyses. silicon doping. reactor school training, etc .. The modification of JRR-4 is planned about core conversion according to the framework of reduced enrichment on research reactor program. utilization facilities upgrading. and renewal of some reactor systems. The fuel will be changed from aluminized uranium to 20"10 lower enriched uranium silicide fuel with 3.8g/cc, however the reactor performances will be calculated same thermal flux level as HEU core at same thermal power. without changing of structure. size and number of fuel in the core. The utilization facilities are installed a medical irradiation facility for. BNCI' and are modified a NAA system for short life nuclides .and a large size pipe irradiation system. Furthermore to be operated long term safety and continuously. it is designed that are renewals of instrument and control system. installation of large scale fuel failure monitor system and emergency exhausting system. repairing of re.actor building for reinforced seismic design. This paper described outline of the modification of JRR-4. 1 1. INTRODUCTION In Tokai Research Establish.ment, JAERI. three research reactor as JRR-2, JRR-:3M and JRR-4 is operated over thirty years. The JRR-3M which was reconstructed in 1990 as one of the highest performance research re.actor in the world. today. is used by a great number of researchers and en",oineers. The JRR-2 which is CP-5 type research reactor and is operated for thirty-six years since 1960 will be terminated by reason of aging in December 1996. In the such situation. it was necessary that the JRR-4 was converted LEU core because of getting new driver fuels. The JRR-4 has used over 90"10 enriched uranium plate- type fuel( total 120 elements) since first criticality in 1965. amount of driver fuel of hand durability disappeared finally at January 12th. 1996. The core conversion activities in JAERI were begun in 1980's. It is already done successfully that JRR-3M is used aluminide LEU fuels. JMTR is converted to MEU in 1986. and silicide LEU in February 1994. JRR-4 has been studied to convert to the LEU silicide fuel without major change of fuel structure and core arrangement. Fig.l and Fig.2 show a bird's-eye view of the JRR-4 buildings and a view of the reactor. The other. JRR-2 has been using 33 cases of medical irradiation for Boron Neutron Capture Therapy (BNCI) since 1990. but JRR-2 will be terminated at the end of 1996. And the mission of JRR-2 for BNCf will be transferred to JRR-4 after suspension of two ya.ars, a new medical irradiation facility will be installed atJRR-4. Furthermore. some components of JRR-4 was aging cause by operating long term since first criticality. and the reactor building will be reinforced against seismic design. 2. LEU CORE DESIGN The core of JRR-4 is composed 8 x 8 array with 20 fuel elements, 7 control rods. 5 irradiation holes. a neutron source and many reflectors. The JRR-4 core arrangements is showed Fig.3. A fuel element has 15 fuel plates which is fuel meat (U3Si2) of 600x64x0.5mm with aluminum cladding. The fuel element is showed Fig.4. The design and evaluation for core conversion with LEU silicide fuel instead of LEU aluminide fuel was started in 1991. The comparison of LEU fuel with HEU fuel is showed table-I. A preliminary estimate showed that the new LEU core has been available performance for thermal flux level ( 7 x 1013 nJcm2/s ) as same as HEU core. without major change of structure and size of fuel element and core dimensions. Table 1 show the comparison of LEU fuel with HEU fuel. As a result on the neutronic design. the excess reactivity of core with fresh 20 LEU fuel elements is 11.7% 0 kIk at maximum.and the one rod stuck margin is about -2.3% 0 kIk and the reactivity coefficient is negative value in the calculated all range. The new fuel elements 2 will be able to use about six years longer than HEU fuel elements . And on thermal-hydraulic design, the core coolant flow rate is increased from 7mVmin to 8 mO/min by three main pumps because a safety margin of fuel is increased in case of power cut The fuel surface temperature is about 11l"C which value is not exceeded 125"C as ONB(onset of nucleate boiling) temperature. The fuel meat temperature is about l1S"C at rated power 3.5MW operation. The minimum DNBR(departure from nucleate boiling ratio) on the normal operation and transient condition are about 3.1 and 1.6 respectively, and those are not below 1.5 safety limit The dynamics-characteristic of core is keep on enough safety with reactor control system. Table 2 show the characteristic of LEU core. Next, the safety evaluation according with the National Safety Guideline for research reactors showed that the fuel should not failed at any events in both abnormal transient conditions and accidents. But in the site evaluation events, two cases which are one fuel element failure and all fission product release from the core were evaluated, the results showed that JRR-4 site evaluation satisfied the judgment criteria of above guideline under the condition of installation of emergency exhausting system. 3. UPGRADING OF UTILIZATION FACILITY In JAERI, the utilization of test and research reactors is contributing successfully so that JMTR can be mainly used for irradiation, JRR-3M can be mainly used for beam experiments and multi-purpose irradiation and JRR-4 is mainly used for simple and special experiments and irradiation, respectively. The utilization facility of JRR-4 are two big pool and dry shielding room for reactor shielding experiments, in-core irradiation holes for silicon doping, nucleate activation analyses. RI production, thermal column for special experiments, N-16 gamma ray irradiation system. In addition, a medical irradiation facility will be installed. the activation analyses system is modified, and a large pipe irradiation system is changed. The medical irradiation research in JAERI is conducted thirty-three cases for boron neutron capture therapy against brain cancer at JRR-2 since 1990. But after JRR-2 is terminated in December, 1996, the mission of BNCI' is transferred to JRR-4 in reply to the request of medical researchers in national hospitals and universities. The general aIT'.mgements of the facility is shown in Fig.5. The irradiation room and the medical treatment room are located on the basement The cross section of the medical irradiation facility which is shown in Fig.S is composed with a heavy water tank and a beam experimental hole. The heavy water tank plays role of neutron beam filter with aluminum and heavy water. The heavy water layers can be changed for optimized performance both thermal and epi-thermal be.ams according to the need of e.ach treatment. The beam experimental hole is composed bismuth and cadmium filters. lead and graphite reflectors. LiF and B-IC covers, LiF collimeter. and polyethylene and 3 lead for shielding materials. As a result of beam design and analysis. the beam performance of facility will be more than two times of JRR-2 to short treatment for thermal neutron beam, furthermore the facility will be use epi-thermal beam for treatment without surgery operation, as is shown Table 3. A neutron activation analysis system is used to analysis in widely for research of environment, geology, biology and so on, and the system is requested analysis for more shorter life( approximately 1 minute) nuclide recently. Then a new system with automatic analysis by repeating irradiation method is installed. A large pipe irradiation system is mainly used for silicon doping, the diameter of pipe will be changed from 4 inches to 5.5 inches for a large sample irradiation. The plan of utilization of new JRR-4 is proposed many experiments, example are a leaser beam instrumentation system, a in-situ me.asurement irradiation experiment, a neutron radio-graphy experiments, special reactor training. 4. WORKS AGAINST FOR SYSTEM AGING, SEISMIC DESIGN JRR-4 is constantly operated longer than thirty years, and is used successfully. In those term, some reactor components and systems are remodeled as instrumentation system, thermal heat exchanger, secondary cooling system, ventilation system, but many components have been to keep safety by inspection and repairing for maintenance since the time of first criticality. In the modification works, it will be performed for long continuously operation for future that detailed inspections of core tank, two reactor pools and cooling systems , remodeling of instrument and control system including control rod drive mechanism, installation of emergency exhausting system, duplication of emergency power and soon. Furthermore the JRR-4 reactor building was designed according to old guideline of the Building Standard Low of Japan in 1960's. Then as a result of new seismic design according to the Examination Guide for Seismic Design of Nuclear Power Reactor Facilities, it was necessary for seismic safety that the remodeling of roof structure, reinforced of some pillars and walls of the building. 5. CONCLUSION The fuel will be changed a silicide fuel with 3.8 U-g/ee, the new core performance is similar with BEU core its without major change of structure and size of BEU fuel. And the activities of reduced enrichment works will be finished on research reactors in JAERI . The medical irradiation facility will be had higher performance than that in JRR-2, and will be able to use epi-thermal neutron beam too. In addition, the reactor system will be remodeling for safety operation and steady utilization for long term. The licensing of the LEU core is already permitted by Japan government on September 1996. The JRR-4 modification work will be performed for about two years. After then, the new JRR-4 will be start to use many experiments as a boron neutron capture therapy etc. in 1998. In the future, JRR-4 is expected to use for a wide variety utilization by many researchers as one of important reactors in JAERI. 6. ACKOWLEDGMENT This paper represents a summarizing report on the JRR-4 LEU Project which many colleagues and coworkers from various departments have contributed to. Especially the authors wish to thank lVIr. E.Shirai and Mr. N.Ohonishi who are former directors of department of research reactors for good direction of this project REFERENCE 1) K Koba, et .al., "Construction of JRR-4·'. (In Japanese) JAERI-1141(1967) 2) M. Takayanagi, et. al., "The Reduced Enrichment Program for JRR-4 ", Proceeding of the Asian Symposium on Research Reactor, Sanpia Hitachi. Japan, JAERI-M 92- 028(1992) 3) Y. Nakano, H. Ichikawa, T. Nakajima, "Comparison of JRR-4 core Neutronic Performance between Silicide Fuel and TRIGA Fuel", Proceedings of the 16th International Meeting on RERTR, Oarai. Japan, JAERI-M 94-042(1994) 4) T. Nakajima, N. Ohnishi, E. Shirai, "STATUS OF CORE CONVERSION WITH LEU SII1CIDE FUEL IN JRR-4" , in proceedings of the 1994 International Meeting on RERTR, Williamsburg, USA, September 18-23,1994, in press. 5) K Yokoo, T.Yamada. T.Nakajima, eta!. "A new medical irradiation facility at JRR-4~, in proceeding of the Seventh International Symposium on Neutron Capture Therapy for Cancer, Zurich, Switzerland, September 4-7, 1996.in press. 5 Table 1 Comparison of LEU fuel with HEU fuel Item I LEU fuel I BEU fuel I Enrichment, % I 19.75 I 93 I Uranium density, g/cc(outer plateJ I 3.811.9~ I 0.661.0.331 J Specific content of U-235 per element, g I 204 1 166 I Number of Fuel plate per element I 15 I Fuel meat material I U~Si2-Al I UAlx I Cladding material I Aluminum-allqy I Maximum burn-up ratio, % I 50 I 20 I Size of fuel element, mm I 80 x 80 x 1025 I Table 2 Characteristic of LEU core calculated Item value I remark Excess reactivity, %.6 k/k max. 11.7 I initial core lOne rod stuck margin, %.6 klk min. -2.3 I initial core Temperature of plate surface,oC max. 111 I 3.SMWt Temperature of fuel meat, °C max. 113 I 3.SMvVt lONE temperature at hot spot,OC max ..p...- ~ I 3.SM\Vt IrvIinimum DNBR 3.1 I 3.SM\Vt 6 Table 3 OBJECTIVES OF BEAM DESIGN FOR BNCT 1. Thcrmal flux; ~ 1 X 109 n/cm2/s (Thcrmal bcam Modc) 2. Epi-thcrmul flux; ~ 1 X 109 n/crn2/s (Epi-thcrmul bcarn Modc) 3. Gamma dosc contaminatioll; =::::3 x 10-13Gy/n-cm2 4. Fast neutron contamination; =:::: 5 x lO-I3Gy/n-cm2 S. Sizc of bcam port; 200 x 200 mm Fig_ 1 BIRD'S.EYE VIEW OF THE JRR-4 nUILDlNGS 6. Estimatcd Irradiation timc; =:::: 2 h 7 8 I; .. I II~FYIO~ .. 1I ~.~~!-=!fi£:: ":~FLEcrOR-':<'~D • CONTROL. ' .R OD .- • IRRADIATION ~ HOLE • NEUTRON 1&1 SOURCE Fig. 3 JRR·:j CORE ARRANGEMENTS Fig. 5 GENERAL ARRANGEMENTS OF A NEW MEDICAL IRRADIATION FACILITY J~~~t·JAEru ~~ r' I \ ~r~=;;.!~:=r-="=~J = -=-::= ;f~M':.-. :·. . · .... ·fl~CCfEt!H~1~ -~1 1 ~-~~.l::, I : L ~ u:r ! .~:W!. t;.:'-- .........." J c ; Irradiation roum ·!~~~::~·::;I::--:. .. i~ ~.~, j JTI~iJIn L' l±ill §-ii~ I 1U!E!IH'n: ':!-'4-i!"' -0:: ::-..... r-==---r--;:;-;::--I ~~ lmnllurtL= :':!..J'-..!: .1._J ! .... ~ ~ 1': ... ",t- :,,'::': Fig. Ii CROSS SECTION OF Fig. 4 JRR·4 FUEL ELEMENT ~g&~.:j,-".=~,,, .. ..... . MEDI.Ci\ L .. !.~~.R.I\.~~~~~(),N..~~~! I:ITY 9 10 1 OVERVIEW ON THE MAIN ENGINEERING WORKS PERFORMED ON FRENCH RESEARCH REACTORS THESE LAST YEARS P. ROUSSELLE (Technicatome) - G. de SAINT OURS (Technicatome) J. GUIDEZ & C. JOLY (CEA-OSIRIS) - M. MAZIERE (CEA-ORPHEE) H. GUYON (CEA-SILOE) Introduction From the beginning of eighties, design, construction and commissioning of new neutron sources like research reactors have severely slowed down in western developped countries as well as worldwide. On the contrary several faciIities have been shut down and some of them decommissioned. In France, disregarding reactors dedicated to safety studies and very low powers reactor (critical assemblies and training reactors) only four faciIities are still operated, namely OSIRIS and SILOE for technological irradiations and multipurposes and HFR and ORPHEE for fundamental research. Abroad, the situation is not better : new projects are rare nowadays and subject to well known difficulties surch as lack of budget, green opposition, enrichment offueI... In this general framework what are the remaining activities for engineering companies specialized in the nuclear field ? The purpose of this paper is to review the main engineering works performed in France on research reactors during these last years in strong cooperation between operators (CEA, ILL) and engineering companies (Te chnicatome). RRX018/2034 DAI/96-O 1023 2 Main factors involving engineering works in the field of RR When facilities are under operation, operation teams have enough well qualified and trained staff to perfonn the operation of the reactor, itself its current maintenance as well as minor refurbishing and adaptations/modifications of limited importance. But some factors necessitate more important works and the, operators have to call to engineering companies and cooperate strongly with them for the preparation, the execution and the tests of the interventions to be done on the facility. The main factors involving such engineering help to main modifications/refurbishment are : • ageing of main equipment and materials, • the obsolescence of equipment and the apparition of new technologies (mainly 1& C equipment), • the evolution of safety considerations and rules, which may involve some adaptations of the facility, • new projects and modifications necessitated by the evolution of research programs and last but not least : • the decommissioning questions. RRX018l2034 DAIl96-01023 3 The evolution of safety considerations and rules As seen hereabove the main research reactors still operated in France have been designed from the early sixties (Sn.OE, OSIRIS) to the end of seventies (RHF, ORPHEE). From this time safety considerations have moved towards more and more strictness and rigorism resulting that for each one of this reactors some design feature would not be nowadays accepted by French safety authorities. For example: At Siloe, the first design and the execution (1963) of the reactor pool and of the working pool which were ceramic lined did not allow the complete monitoring of the second barrier : checking eventual leakage through the bottom of pools was not possible. This point was put right with the reactor stainless-steel pool erected in 1988 and then, on the working pool which was stainless-steel coated in 1995. Consequently some adaptations andlor modifications are made on the facilities taking the opportunity of major refurbishments or works where there is no actual urgency, in order to improve the safety level and to be in accordance with modern safety considerations. There are, in France, only few written rules that are related to research reactors ; howerer these documents are worth to be quoted here because there were progressively put in applications in existing facilities these last years and involved some works : • L'arrete du 10 Aout 1984 relatif a la qualite de la conception (Design phase), de la construction (Construction phase) et de l'exploitation (operation phase) des installations nucleaires de base (basic nuclear installations), • Fundamental safety rule on the installations on ventilation filters RRI (August 4th, 1986), • Fundamental safety rule on fire protection in research reactors RR2 (July 2nd, 1991). RRXOl8/2034 DAIl96-01023 4 Conclusions To conclude this general overview of some of the more significative works performed these last years on the French research reactors the following can be underlined : In France, the strong cooperation, and the mutual assistance between reactors operators and engineering staff allows : for one part to provides operators of CEA the necessary assistance for important refurbishment and renovation works which overpass the job of reactor current operation and ordinary maintenance. On the other side the engineering staff takes advantages of these many and strong contacts with operation teams to incorporate operating experience feedback in new projects. The number and the variety of interventions on French operating research reactors allow to maintain them in very good conditions, to adapt them to the evolutive research programs and needs and, last but not least, to improve their safety level, even for a reactor that is to be shut-down in a near future (like SILOE at the end of 1997). So, the challenge of the future French RJH research reactor (design, construction, operation) can be considered with a high confidence, relying on qualified teams. RRX018/2034 DAIl96-O 1023 5 References • OSIRIS 2000 : a young man of thirty years I (C. Ioly, I. Guidez, C. Thiercelin, AAiberman and M. Roche Proceedings of an ENS Class 1. Topical meeting - Brussels 4.6 Iune 1996 • OSIRIS the first MTR with a new instrumentation and control system based on digital logic of vote (C. Ioly, C. Thiercelin, 1. Corre, I.F. Dubois, G. de Contenson) 3rd lGORR conference - Iapan - September 30 - October 1, 1993 • OSIRIS Refurbishment and management of ageing effects (C. Ioly, J. Guidez, G. de Contenson, J.P. Marin) IAEA Meeting - Geesthacht - Germany - May 8-12, 1995 • Upgrade of the experimental facilities of the Orphee reactor (B. Farnoux, P. Breant) 3rd lGORR conference - Japan - September 30 - October 1, 1993 • ORP:a:EE reactor. Upgrade of the installation (B. Farnoux, M. Maziere) 4th lGORR conference - Gatlinburg, Tennessee, May 23-25, 1995 • INSTITUT LAUE LANGEVIN Refurbishment programme of the reactor and progress of work (I.M. Astruc - ILL) 2nd lGORR conference - Saclay - France - 18,19 May 1992 • SIRIUS 2 : A versatile medium power research reactor (p. ROUSSELLE) 2nd IGORR conference - Saclay - France - 18,19 May 1992 RRX01812034 DAIl96"()1023 liiltedilicatome CE3Cl MUUNFACTORSUNVOLVING ENGINEERING WORKS • AGEING OF MAIN EQUIPMENT & MATERIAL • OBSOLESCENCE OF EQUPMENT, NEW TECHNOLOGIES • EVOLUTION OF SAFETY CONSIDERATIONS • NEW PROJECTS • DECOMMISSIONING RRX01812034 DAIJ96-01023 CE3Cl MAIN CHARACTERISTICS SILOE(CEA) OSIRIS (CEA) HFR(ILL) ORPHEE (CEA) First criticality 1963 1966 1971 1980 Power 35MW 70MW 57MW 14MW I Max. Flux th 4.4 4 15 (reflector) 3 (reflector) (10 14 n.em ol.sol) f 4.5 4.5 5 Fuel U.AI (93%) U3 Si2 (19.75%.) U.AI (93%) U.AI (93%) Moderator H2O H2O D20 H2O-Be Reflector H20, Be H20,Be D20 D20 Coolant H2O H2O D20 H2O Utilisation Polyvalent: Polyvalent: Fundamental research Fundamental research • Technological irra- • Technological irra- • 17 beam tubes • 9 neutrons beam diation diation • 2 cold sources tubes (20 beams) • RI production • RI production • 1 hot source • 2 cold sources • Silicon doping • Silicon doping • 1 hot source • 4 neutron beams • Neutron radioe:raphy • Silicon dopin2 Scheduled Shut-Down End 1997 RRXO I 812034 DAIl96-O 1023 I!ltecIDcatome CE::C1 AGEING OF MAIN EQUIPMENT SILOE OSIRIS HFR(ILL) ORPHEE 1963 1966 1971 1980 • Reactor pool (Ex: 1986/88) • Decay tanks • Reactor block changing • Zircaloy core housing Ceramic liner ~ SS liner (ex: 1994/95) (and all in-pool primary Dismantling of the Pool cooling and core circuit equipment: heavy I oldlone cooling (painted C.S.) water collectors, (St 1996) • Neutron beams connecting sleeves• ••• ) (Ex: 1986/88) • expertise (Ex: 1991) Erection ofa new one • fabrication and (St 1997) erection (Ex: 1992) • Auxiliary Pool • PrimarY coolinlt! core • Auxiliary water circuit (Ex: 1993/94) exit main pipe (Ex: 1996) Ceramic liner ~ SS liner (Ex: 1994) - arrangements Repairing - new air-coolers • Heat exchangers • Auxiliary canals (Ex: 1992/93) (St: 1995) Replacement • Cellular core structure , Replacement (Ex: 1996/97) Ex : ExecutIOn St: Study RRXOl812034 DAIl96~1023 liiI t...I!ICI. *mI:m ... ~ EVOLUTION (IMPROVEMENT) OF SAFETY ,SILOE OSIRIS HFR (ILL) ORPHEE 1963 1966 1971 1980 • New stainless steel reac- Ventilation tor pool Physical separation of Borax resistant with different sections of low • Improvements of installa- shock absorbers at the voltage tion of iodine traps on nu- bottom clear ventilation exhaust circuit • Nuclear measurement: • Installation of a test Geographical separation equipment for the period of redundant cable trays control of iodine traps • Logic of vote (RPS) ImprOVed performances • Fire protection • Fire protection • Fire protection Expt;rtise (Expertise: 1992) • Siloe (1993) • Siloette (1995) • Emergency control room (Ex: 1996) Ex : Execution St: Study RRX018/2034 OAll96"() 1023 I!!tectncatome ce::l OBSOLESCENCE OF EQUIPMENT NEW TECHNOLOGIES SILOE OSIRIS ORPHEE 1963 1966 1980 I & C Equipment I & C Equipment Health physics equipment : St: 1994 • Reactor control processing system • Reactor control processing system (Digital equipment) (Digital equipment) Ex: 1992/93 Ex: 1992 • Nuclear measurements • Nuclear measurements (SIREX (Digital technology: SIREX type) type) and RPS system Ex: 1992/93 Ex: 1992 - New logic of vote and RPS • Electrical low voltage supply - New cold neutron source (CNS) St: 1995 network Cylindrical type Ex: 1996 Ex : 1989-1996 . • Electrical low voltage supply I. Mixed bed resins without regene-/- Upgrading of neutron guides network rations Supermirrors Ex: 1989194 Ex: 1994 • Water Purification system Ex: 1992/93 Ex : Execution St: Study RRX018/2034 DAIl96-01023 ~ NEW PROJECTS ISIS Upgrading of power and feasibility of adaptation of the reactor for BNCT Application (1990) ORPHEE+ Implementation of new cold guides on the beam port 4F : increase by about 300/'0 the number of beam positions SIRIDS Preliminary design of multipurpose reactors derived from SILOE. SIRIUS 2 : 15t030MW SIRIUS 3 : 5 to 10 MW RES New reactor for training of crews for submarines/air-craft carrier· REACTOR New French Research Reactor for Irradiation JULES HOROWITZ RRXOl8/2034 DAIl96"() 1023 · Ii!technIcatome ~ DECOMMISSIONING (RESEARCH AND EXPERIMENTAL REACTORS) PEGASE (CADARACHE) Engineering study & following up of works 1975-80 EL3 (SACLAy) Engineering study & following up of works 1979-80 TRITON & (FONTENA Y AUX Engineering study & following up of works 1982-89 NEREIDE ROSES) RAPSODIE (CADARACHE) Engineering studies 1982-90 EL4 (BRENNILIS) General studies 1985-87 execution 1995-96 MELUSINE (GRENOBLE) General studies 1994-95 SILOE (GRENOBLE) Preliminary studies 199,0 others reactors: Gl, G2/GJ, VANDELLOS, SNLE RRX018/2034 DAIl96~1023 UTILIZATION OF THE BUDAPEST RESEARCH REACTOR ISTVAN VIDOVSZKY KFKI Atomic Energy Research Institute H-1525 Budapest 114, P.O.Box 49, Hungary E-mail: vidov@sunserv.kfki.hu ABSTRACT The Budapest Research Reactor, theflrst nuclear facility ofH ungary wasflrst put into operation in 1959. The reactor operated for 27 years without any safety problem; it provided the country with neutron research possibility, supplied it with radioactive isotopes and served as a basis for training as well. After a major reconstruction and upgrading the start-up procedure began in 1992. The upgraded reactor serves for: basic and applied research, technological and commercial applications, education and training: (e.g. involving the JAEA). The reliability oft he reactor is very good as e.g. in the 1995 operation period only two unexpected shut downs occurred. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The neutron research will get a much improved tool, when the cold neutron source will be put into operation. The start-up of the cold neutron source is scheduled for late 1998. Background The research reactor in Budapest was in operation from 1959 to 1986. No incident occurred during the 27 years of reactor operation. In this period the reactor played an essential role in establishing nuclear research and technology in Hungary. It served as a basic facility for neutron scattering, nuclear and particle physics, radiochemistry, shielding investigations; for establishing nuclear medical applications providing radioisotopes; for performing pressure vessel surveillance programme for reactor safety studies; and it was an important school of university and postgraduate training. The full scale reconstruction and upgrading project started in 1986 1, aiming the substitution of aged components, the enhancement of reactor safety, the increase of reactor power to 10 MW. The reactor vessel and the primary piping turned out to be much less corroded than previously assumed. Reactor safety has been enhanced by adding some new safety systems and thus satisfying the "Defense in Depth" concept and also by applying more up-to-date and reliable systems. A new safety analysis report taking into account all relevant recommendations of IAEA has been prepared. The increase of reactor power was facilitated mainly by building new cooling towers, permitting a reactor power of 20 MW. However, the characteristics of the VVR-SM fuel do not allow a reactor power higher than 10 MW and the higher power can be achieved only by applying a new type of fuel elements. 1 Short Technical Description The Budapest Research Reactor 2.3 is a tank type reactor, moderated and cooled by light water. The reactor is in a cylindrical reactor tank, made of a special aluminium alloy. The diameter of the tank is 2300 mm, the height is 5685 mm. The heavy concrete reactor block is situated in a rectangular semi-hermetically sealed reactor hall. The area of the reactor hall is approximately 600 m2• It is ventilated individually. The fuel of the research reactor is of the VVR-SM type (Russian product). It is an alloy of aluminium and uranium-aluminium eutectic with aluminium cladding. The uranium enrichment is 36%, the average U-235 content is 39 g/fuel element. The fuel elements contain three fuel tubes, the outer tubes are of hexagonal shape, while the two inner ones are cylindrical. The active length of fuel elements is 600 mm. The equilibrium core consists of 223 fuel elements, with a lattice pitch of 35 mm. The core is surrounded radially by a solid beryllium reflector. The reactor is equipped with boron carbide safety and shim rods. There is a stainless steel rod for the purpose of automatic power control. The reactor can be characterized by the following main technical data: thermal power: 10 MW mean power density: 61.2 kW/litre approx. maximal thermal flux: 2.5 x 1014 n/cm2s approx. maximal fast flux: 1.0 x 1014 n/cm2s cooling water inlet temperature: 54°C maximum cooling water outlet temperature: 6QOC The reactor has 10 horizontal beam tubes (8 radial and 2 tangential). Irradiations may be carried out by inserting samples into the 51 special vertical channels. The reactor staff has a long experience in assisting physical experiments and radioisotope production. The reactor is supplied with a reasonable amount of fresh fuel for at least 6 - 7 years (depending on the future demands for annual operational hours), so this problem is solved up till the first years of the next century. Operation The reactor was operated according to the pre-determined timetable throughout its operation period. These timetables have always been revised and modified according to the changes in the demand. The timetable was last modified for the year 1995. One reactor operation period was II days, which meant 270 hours of continuous operation and one cycle consists of 4 - 6 such periods. One cycle was 44-66 effective days, which was followed by a refuelling period of two weeks. The 1995 timetable was planned for 4476 operational hours for the year and this amount was exactly fulfilled. The total energy production was 1883 MW days. The reliability 2 of the reactor was very good as in the 1995 operation period only two unexpected shut downs occurred. The equilibrium core size was reached by the end of the year 1995. The timetable for 1996 has the same structure as that for 1995. The Use of the Reactor Use in research The main goal of the reactor is to serve neutron research. Investigations based on neutron scattering are essential in studying the condensed matter structural properties. These studies will get a much improved tool, when the cold neutron source will be put into operation (in 1998). Investigations of biological objects (models or in vivo) affected complex irradiation from the reactor, dosimetric calibrations, development of chemical dosimeters, etc. Besides the basic research some problems for possible applications are foreseen, two of them are mentioned here briefly: It is of great interest to investigate segregation processes, e.g. determining crystallite orientations by texture analyses and carrying out internal stress analysis by high resolution lattice parameter measurements. A very practical use can be the aging calibration of turbine blades by small angel neutron scattering and neutron diffraction. The use of the reactor can not be considered as a fmished matter, further ideas are welcome, a limited of excess space is still available. Use for practical purposes A lot of technical problems can be solved by means of the reactor 4• These problems are radioisotope production, neutron radiography, activation analyses, and pressure vessel surveillance studies. Some other applications, as e.g. silicon doping, might be of some interest as well. The production of radioactive isotopes can be considered, as one of the main applications of the reactor. The production restarted in 1993, 1994 was the first full year of isotope production. The quantity and variety of the isotopes produced increased during 1995, but a slight progress is still expected. Neutron radiography is a well accepted method. Collimate neutron beam is used to investigate objects in closed volumes. In dynamic radiography moving objects or processes can be recorded. A new channel was constructed and put into operation in 1995, devoted to 3 static radiography, where the resolution is better, than in the dynamic one, so the two radiographic procedures can now complete each other, providing the users with a full- possibility radiography. The current use of radiography is mainly related to refrigerators and fire extinguishers, but other fields, e.g. combustion engine investigations are foreseen as well. Two horizontal channels are used for neutron radiography, one for the dynamic investigations and the other for static pictures. Activation by reactor neutrons is a very sensitive analytical method. About 70 various chemical elements can be detected in an extremely wide range of content, i.e. from percents to 10-10 gig concentrations. A pneumatic rabbit system has been constructed to serve the laboratory of activation analysis. The laboratory is capable to analyze 100 - 150 samples per week. The moderate demand of users (e.g. medical industry) has been satisfied by neutron activation analyses; marketing activity was started, it may increase the use of this method significantly. An extended national programme for the surveillance of the power plant's pressure vessels is going on. As regions, where the flux is higher, than the flux at which the pressure vessel of NPPs is exposed, can easily be found in the research reactor, the neutron induced embrittIement of 20 - 30 - 40 years can be studied after a few month irradiation. The possibilities will be extended by a heatable irradiation channel, providing with better simulation of the conditions of a power reactor pressure vessel. This channel will be put into operation in 1996. Cold neutron source The installation of a cold neutron source equipment is foreseen at one of the tangential beam tubes. The project of the cold neutron source is partially sponsored by the !ABA. Financing problems seem to be solved now, so there is real hope, that the cold neutron source may start to operate in September 1998. The cold neutron source will extend the use of the reactor, especially in the scientific field. The cold neutron source will be the liquid hydrogen type. The moderator cell is relatively small, i.e about half litre volume. The relatively low estimated heat release (about l00W) makes feasible the direct cooling of the condensed hydrogen in a double walled moderator cell by helium gas (12 - 14 K). The preliminary safety report of the cold neutron source is being prepared now. If this report will be accepted by the Hungarian regulatory body, construction works may start early 1997 and could be finished by August 1998. The scheduled start of the operation of the source is September 1998. Organisation The Budapest Research Reactor is operated by the KFKI Atomic Energy Institute (AEKI), which is responsible for reactor safety and utilisation as well. The institute is prepared for any reasonable cooperation. 4 For the utilization of the reactor in the field of basic research the Budapest Neutron Centre (BNC) has been set up, by three research institutes: KFKI Atomic Energy Research Institute, KFKI Research Institute for Solid State Physics and the Isotope Research Institute of the Hungarian Academy of Sciences. The BNC has an international scientific advisory council. The utilization of the reactor for practical applications is organized by AEKI as well, it can be considered in international cooperation. International Relations The Budapest Research Reactor is considered as Centre of Excellence by the Central European Initiative. The international user system for neutron research will be started in the near future. The co-operation between the operator of the Budapest Research Reactor (AEKI) and IAEA is complex and important. lAEA has sponsored the reconstruction (providing the Beryllium) and sponsors the construction of the cold neutron source. The safeguards is naturally controlled by the staff of the IAEA. The co-operation might be extended: e.g. training courses organized by IAEA could be performed at the reactor. Personal contacts between the management of the Budapest Research Reactor and of other research reactors in Europe (Rez, Seibersdorf, Ljubljana, Petten, RiSIiJ, etc.) are excellent and beneficial. Many co-operations are organized on bilateral basis, e.g. some pressure vessel surveillance investigations are performed in a co-operation between PSI (Switzerland) and AEKI. References 1 T. Hargitai: Dismantling and Reconstruction of the Budapest Research Reactor, lAEA Topical Seminar on Management of Ageing of Research Reactors, Geesthacht, Germany, May 1995 2 I. Vidovszlcy: Research Reactor in Budapest, ENS Regional Meeting: Nuclear Energy in Central Europe: Present and Perspectives, June 13 - 16, 1993, Portoroz, Slovenia, Proceedings p2 3 F. Gillemot, T. Hargitai, I. Vidovszky: The Budapest Research Reactor, JAEAISR- 183123, (1993) 4 J. Gad6, I. Vidovszlcy: The Budapest Research Reactor: Tasks Ahead, NUCLEAR EUROPE WORlDSCAN, No. 9/10 September/October 1995, p45 5 Design Modification of HANARO Reflector Cooling System J. S. WU, S. Y. HWANG, Y. K. KIM HANARO Center Korea Atomic Energy Research Institute P.O.Box 105, Yusung Taejon, KOREA ABSTRACT HANARO, 30MWth IS a light-water-cooled and heavy-water-reflected research reactor. The Reflector Cooling SystemCRCS) circulates the heavy water from the reflector vessel through the heat exchanger then back to the bottom of the vessel. Two different features are provided to protect the overpressure of the reflector vessel. One is the concept of the relief valve for air vent activated by a relatively small amount of pressure build-up during normal operation. The other is a rupture disc for large volumn release in case of the accident condition that the system pressure abnormally exceeds the predefined limit During the system function test, an unexpected transient peak pressure was detected when the operating pump was stopped. It was found out that this pressure peaking came from the water hammering phenomena in the piping and the overpressure protection mechanisms could not act, as intended against the transient peak pressure. To resolve this problem, we introduced an additional extension pipe at the position of the rupture disc to give more room for absorbing the peak pressure from the water hammering. The test result for the modified system indicates that the transient peak pressure was almost absorbed into the air-cushion inside the extension pipe newly added. This paper concentrates on the basis of the design modification and the system behaviors before and after the implementation of the new mechanism. 1. Introduction The purpose of the Reflector Cooling SystemCRCS) IS to cool and maintain the - 1 - purity of the heavy water which acts as the reflector in the annular cylindrical tank. The ReS consists of two 100% circulating pumps. a heat exchanger, two ion exchanger columns and all necessary piping and instrumentation as shown in Figure 1. Relief Valve Air Rupture r-'"l--l>'O-- Supply Disc Reodor Pool Secondary Cooling System Hd Exchangel Reflector ...... --t..., Vessel Figure 1. Flow Diagram for Reflector Cooling System The system circulates the heavy water(D20) at a normal flow of about 43.2 Vs dissipating up to 2.5 MW of heat to the heat exchanger. A small bypass flow 0.1 Vs is taken through one of two ion exchange columns for purification. An expansion tank is provided to control the system pressure coping with the volume change of DzO and to purge Dz gas from the heavy water. The top area of the expansion tank is filled with the controlled cover gas supplied by compressed air system. The internal pressure of the reflector vessel should be maintained below the design limit even during the anticipated transients such as loss of reflector circulation, loss of secondary cooling system, etc. The overpressure protection for reflector vessel is provided by two mechanisms, a relief valve and a rupture disc. A relief valve is installed at the top of the expansion tank. It is an 1 1/2 inch glove type and the pressure set point is 35 kP8.(g). A 6 inch rupture disc is installed at the pump suction line. According to the process design, the internal pressure of the reflector vessel is 134.9 kPa(g) in normal operation. The reflector vessel was tested over this pressure at the - 2 - manufacturing shop. Then the maximum allowable pressure in the vessel is 245.1 kPa.cg) considering the hydrostatic pressure of pool water, 110.2 kPa(g). The rupture disc limits the internal pressure of the reflector vessel below the maximum allowable pressure and its setpoint for burst was decided to 180 kPa(g) through RELAP5/KMRR simulation which is a computer code to analyze the thermal hydraulic behavior of RCS under the anticipated and postulated transients. During the system function test, an unexpected transient peak pressure was detected when the operating pump was started and stopped. It was found out that the overpressure protection mechanisms could not absorb the transient peak pressure from the water hammering phenomena during pump stopping and starting. In design stage the RCS pump was selected with bellows mechanism to reduce DzO leakage to the minimum. The RCS· pump has a characteristic of a very small inertia It means that the pump stops immediately and then acts like a check valve when the electric power is failed. The maximum pressure rise at the pump suction depends on the rate of decrease of the flow and the pressure wave reflections. The RCS volume change was identified due to deflection of the heat exchanger plates. The maximum peak pressure was over 180 kPa(g) of setting level of rupture disc burst Therefore, it is needed to provide any scheme for absorbing the transient peak pressure. At that time, the field works for example, pipe cutting and welding, were very limited to maintain the cleanness of the RCS and reflector vessel. Finally, we selected an additional stand pipe at the point of rupture disc with flange connection and carried out the design modification and the experiment. 2. Design Modification The pressure variations were observed on pump start and stop operation. With the RCS pump operating, the RCS pressure of 344 kPaCg) at the heat exchanger(HX) exceeds the Secondary Cooling SystemCSCS) pressure of 263 kPa(g). With the RCS pump stopped, the RCS pressure at the HX of about 24 kPa(g) is lower than that of the SCS. When the RCS pump stops, the higher SCS pressure deflects the plates to reduce the DzO volume in HX. The rapid volume change cannot be accommodated quickly enough when pump operation mode changed, because the small pipe to the expansion tank has high flow resistance. The cause of pressure transients is identified as water hammer wave with rapid RCS volume changes due to deflection of the heat exchanger plates when the pressure differential between the ReS and SCS is changed. - 3 - In order to relieve a maximum pressure rise on reflector cooling system, we added a length of 6 inch pipe to the existing rupture disc flange which raises the rupture disc location 5.275m above its previous location. The 6 inch standpipe will act as additional expansion tank to provide a fast source of M supply on pump startup and to remove D20 quickly from the existing RCS loop on pump stop. The Figure 2 shows outline of the design modification on reflector cooling system. The cover gas system has a function to purge the D2 gas concentration from radiolysis except giving a space to allow volumetric increase of heavy water from the temperature variation of RCS. The D2 gas concentration in the standpipe cannot be measured by the spectrometer measurement of the expansion tank. The original system design for purging D2 gas in the expansion tank can not be applied to the new standpipe. Therefore, an air connection to the standpipe for purging D2 gas concentration should be provided. It was done by connecting the air supply and exhaust line to the stand pipe as shown on the Figure 2. NEWLY ADDED O.R iGIN ~L PIPING DESiGN Rupture 1 1/2' .:"-., : R~lief Valve Disc ~ V€rli 1I. &1. 1 IJ" 'l L- -~---r+- Air S pply 6' ,,- Stand " Pipe I----- -- 010 - '" - ~ "'"- - Expansion '" Tonk '" "- I I 3/4" ~ ....... ... -................... ........ . .=n I I I 6" I Reflector I ~ Vessel Pump Figure 2. Overview of Design Modification The purge air enters the standpipe and flows through the Biz inch pipe which is provided from the top of standpipe to the expansion tank. . Then it is exhausted to the ventilation duct Since the flow goes through both the standpipe and the expansion tank, all of the air space can be purged. - 4 - 3. Measurements And Results 3.1 Pressure ~easurements A 6 inch rupture disc was originally installed at the pump suction line between the reflector vessel and the branch of the expansion tank as shown in the previous drawings. Later on, 5 meter length of 6 inch upright standpipe was newly installed at the previous location of the rupture disc. Accordingly, the rupture disc moved to the higher position as much as the length of the standpipe. The pressure measurements were achieved under two different conditions-before and after the addition of the standpipe. A high resolution pressure transducer and a high speed X- Y recorder were used for .signal recording. The pressure transducer was connected to the pressure tap provided at the pipe cap which was temporarily installed at the previous position of the rupture disc. After the installation of the standpipe, the pressure transducer was mounted on the same elevation as the rupture disc was originally installed. The pressure measurements were carried out in the following manner : 1) As a normal operating condition, two Secondary Cooling System(SCS) pumps are being operated during these measurements. 2) One out of two RCS pumps, pump No.1 starts. 3) After the stabilization of the system flow, the pump No.1 is intentionally turned off to simulate a pump failure. (Pump No.2 is at standby position.) 4) A few second later, the standby pump then starts automatically by the transfer control logic. 5) Finally, the pump No.2 stops. 3.2 Results and Analysis The Figure 3 shows the pressure transient behaviors at the moment of the pump startup and stop in the sequences previously mentioned. For comparision purposes, two separate measurements-before and after the design modification are plotted on the same graph. The dotted line indicates the pressure measurement before the system modification, while the solid line means the measurement after the addition of the standpipe. As shown on the graph, on each pump start the pressure at the rupture disc location, which is pump suction side is abruptly decreased to -84 kPa(g) then recovered to the steady state in about 10 seconds. The cause of this negative pressure peak has been identified as rapid RCS volume changes. Before the RCS pump is started, the SCS pressure exceeds. the RCS pressure and the heat exchanger plates are deflected to reduce the RCS volume. As soon as the pump starts the RCS pressure now increases above the SCS pressure so deflects - 5 - 200 t,: !: \\ :: : ' 150 : 'l :: : : 5- ~ ~ i ~ Peak=198kPa(g) .I.I.: 100 0 z 1 0 50 Pi No ftart :i~ :7 -~ E ~ ~., :t( 0 !. ", .. ,. .............../ . ..... ! >. ............ '\. ..... -.......... . .. III III . . , W :! \ II: II. -50 ! Pump NO 1 sop : / Pump No.2 sop -100 <---- <, Pe.k=-84kPa(g) Pump No2 auto stB" -20 o 20 40 60 80 100 120 140 160 180 200 220 TIME [SEC) Figure 3. Pressure Transients Behaviors on ReS Pump Startup and Stop the heat exchanger plates in the direction of increasing the RCS volume. This generates the negative pressure peak which lasts until the additional heavy water to accommodate the volume changes has been transferred from the expansion tank. Although the heat exchanger plates are thin and flexible and can be deflected within the first second by the pressure differentials, the small size and relatively long length of the connecting pipe to the expasion tank brought that it took more than 10 seconds for the required transfer of the heavy water to occur. On the contrary, when the RCS pump stops, the higher SCS pressure deflects the plates to reduce the RCS volume rapidly, which results in the positive pressure peak at the position of the rupture disc. According to the Figure 3, the measured peak pressure reached about 198 kPa(g) which is higher than the setpoint for bursting the rupture disc. Within few seconds, the standby pump starts by the automatic transfer control logic. Again, the heat exchanger plates deflect in the direction of increasing the RCS volume similarly to the previous case. It is concluded that the change in sign of pressure differentials across the plates of the heat exchanger created very fast volume changes in the RCS. Since the absorption of the volume changes by flowing into or out of the expansion tank is - 6 - much slower than the heat exchanger plate can deflect, this causes the pressure decreases of up to -84 kPa{g) on RCS pump startup and the pressure increases of up to 198 kPa{g) at the position of the rupture disc. After the implementation of the 6 inch standpipe at the previous rupture disc location, the pressure measurements were carried out in the same manner as we did before the design change. As indicated in the Figure 3, the magnitude of the pressure transients on pump startup and stop was significantly reduced. The positive peak is just about 42 kPa{g) and the negative peak is almost disappeared. This is achieved by permitting rapid inventory changes into or out of the standpipe instead of the expansion tank. That is, the 6 inch standpipe acts as another expansion tank to provide fast heavy water supply from the standpipe to the RCS loop on pump startup and to remove heavy water quickly from the RCS loop to the standpipe. This flow path has large area and low flowing resistance to reduce the pressure fluctuations due to the rapid volume changes that could not be managed by the existing long, small size connection line to the expansion tank. 4. Conclusion Based on the test result with modified system of Fig. 3, the maximum peak pressure at the rupture disk position is 42 kPa(g) instead of the 198 kPa{g) before modification. In conclusion it is verified that the additional expansion pipe can accommodate most transient peak pressure from water hammer wave and volume change due to pump stop and start and that the rupture disc is not blasted under transients condition on normal operation. References [I] Design Manual of KMRR Reflector Cooling System, KAERVKOPEC, KM-321-DM-POOI(I992) [2] I.C.Lim, Analysis of TIH Behavior of Reflector System, KAERI, KM-033-RT-K020(1994) - 7 - IAEA ACTMTIES ON RESEARCH REACTOR SAFETY I F. Alca1a-Ruiz Division of Nuclear Installation Safety International Atomic Energy Agency Vienna, Austria 1. INTRODUCTION IAEA activities on research reactor safety are determined by its statutory functions and responsibilities and on the current situation of research reactors around the world. In particular, they respond to the current situation of the 274 research reactors in operation in 58 countries around the world as per August 1996. General objectives of these activities are: (a) To complete a set of safety publications, including Safety Standards, Safety Guides and Safety Practices2, that would cover all safety aspects related to reactor siting, design, commissioning, operation, utilization and modification, and which would provide guidance to reactor management and regulatory supervision. (b) To provide IAEA Member States with assistance, including training and to implement the IAEA Safety Standards and Safety Guides for research reactors to foster and enhance their safe operation. (c) To collect and assess information on operational experience in research reactors by creating and maintaining a database on safety-related unusual events and to disseminate relevant information to Member States. In accordance with the above objectives, these activities are grouped in the following three projects within the IAEA Programme and Budget for 1995-963: (1) Development of Safety Guidance for Research Reactors (2) Integrated Safety Assessment of Research Reactors (INSARR) Services (3) Incident Reporting System for Research Reactors (lRSRR): Collection of Data on and Systematic Analysis of Safety Relevant Events. ITo be presented at the International Group on Research Reactors (IGORR) 5 Conference, Aix-en-Provence, France, 4-6 November 1996. 2This category may be ditontinued in the future and replaced with Safety Reports in the new IAEA Safety Report Series. 3For the forthcoming two-year period 1997-98, these activities will be re-grouped in the following three projects: (I) Research Reactor Regulatory Issues and Safety Assessments; (2) Research Reactor Operational Safety; and .. (3) Research Reactor Safety Advisory Services. The research reactor statistics shows the following: (a) The total number of operating research reactors in the world is decreasing, but the number is remaining constant for the group of developing countries; (b) The number of developing countries operating research reactors is double that of industrialized countries; ( c) Ageing of research reactors can be considered an important concern; (d) Many activities (safety reviews, in-service inspections, modification projects) are carried out to extend the life of research reactors; and (e) Some new research reactors are planned or under construction. These features determine to some extent the above progranune, whose basic characteristics are development of documents, provision of technical advice and assistance, and fostering the exchange of operational experience. 2. DEVELOPIv1ENT OF SAFETY GUIDANCE FOR RESEARCH REACTORS In accordance with its statutory function of establishing safety standards and providing for their application,4 the IAEA has developed a significant number of standards publications on nuclear safety. The former Safety Series No. 35, Safe Operation of Research Reactors and Critical Assemblies, 1984 edition, while establishing general requirements and providing practical guidance on safe operation, did not deal with many other aspects which arose during the design, commissioning, licensing or modification of research reactors and which influence safety. To remedy this, basic principles and requirements for the safety of research reactors and critical assemblies were compiled in two Safety Standards, one on design and one on operation (Safety Series No. 35-S1 and No. 35-S2). These standards, which supersede Safety Series No. 35 of 1984, also include basic requirements for siting, quality assurance and regulatory control of research reactors. These two Safety Standards accompany the lAEA Safety Fundamentals, the Safety of Nuclear Installations (Safety Series No. 110), and are further developed by a number of Safety Guides which provide recommendations and guidelines, based on international experience, in relation to the fulfilment of the basic requirements, and Safety Practices, which give practical examples that can be used for the fulfilment of the requirements and recommendations of the standards and guides. An overview of the present status of finalization of the set of documents in the RRSP is shown in Fig. 1. At present, there are drafts of all the documents in different status of finalization. Because of the recent changes in the procedures for preparation and review of documents of the IAEA Safety Series, the publication of various Safety Guides (on Commissioning, on Operational Limits and Condition, on Maintenance) and Safety Practices (on Operating Procedures, on Provision of Radiation Protection Services) may suffer some delays. 4 IAEA Statues, Art III, A.6. 3 New documents on research reactor safety, such as a Safety Practice on safety in core management and fuel handling or a TECDOC on selected safety assessment issues, have been included in the IAEA Programme and Budget for 1997-98. 3. INTEGRATED SAFETY ASSESSMENT OF RESEARCH REACTORS (INSARR) SERVICES Activities under this group respond primarily to functions and responsibilities of the Agency related to its projects (with Member States) to which the Agency's Safety Standards and Safety Guides apply. An example is the performance of safety missions to research reactors which are covered by a Project and Supply Agreement with the Agency. An account of the safety missions conducted so far is shown in Table I. The Agency has extended such missions to any type of research reactor. The objectives of these missions may be related to the design, the commissioning, the operation or the refurbishment of the research reactor. Many of the missions included in the table have been conducted following the procedures developed for the so-called Integrated Safety Assessment of Research Reactors (INSARR) missions. These missions are offered as a safety service of the IAEA to all Member States operating research reactors. Other activities in this group are those related to IAEA responsibilities such as encouraging and assisting nuclear research and fostering the exchange of technical - information, scientists and experts and their training. In this regard, the IAEA has organized the following types of activities in relation to research reactor safety: (a) Co-ordinated research programmes (CRP); (b) International symposia (SM), inter-regional and regional seminars (SR); and (c) Interregional and regional training courses (TC). The above activities are normally organized in co-operation between the Agency's Departments. A CRP on Applications of Non-destructive Testing and In-service Inspections to Research Reactors is currently going on. An international SM on Management on Ageing of Research Reactors was held in may 1995 in Geesthacht, Germany, an inter-regional training course on Safety in the Operation of Nuclear Research Reactors took place in Chalk River, Ontario, Canada and Argonne, Illinois, USA, from 8 May to 2 June 1995, and a regional training course on Safety Documentation for Research Reactors was held in Cairo from 9 to 20 March 1996. An important type of activity related to research reactors is organized by the Department of Technical Assistance and Co-operation. At present, there are on-going Technical Co-operation projects on research reactor safety in the following countries: Algeria, Bangladesh, Egypt, Ghana, Iran, Kazakstan, Nigeria,Morocco, Syria, Thailand and Zaire. 4 4. INCIDENT REPORTING SYSTEM FOR RESEARCH REACTORS (IRSRR) Based on its statutory responsibilities of exchange of technical information and on Member States' requests, the IAEA has included in its current programme on research reactor safety the establishment of an Incident Reporting System for Research Reactors (IRSRR). The objective of the IRSRR is to improve the safety of operating research reactors through the exchange of safety-related information. The exchange of information on unusual events with safety significance is considered beneficial for the improvement of the operational safety. The IRSRR will collect, maintain and disseminate reports on unusual events which are received from Member States participating in the system. The IRSRR makes use of the existing IRS for nuclear power plants to the extent possible. The IAEA has convened various meetings in the past years to prepare a basic working document for the establishment of IRSRR. The lAEA held a TCM in October 1996 to clearly define the types of unusual events which should be reported and to make recommendations on the implementation and operation of the proposed IRSRR. The TCM recommended the further implementation of the IRSRR and requested the IAEA to ask the Member States officially for their participation so that the IRSRR is operational by January 1997 for a trial period of two years. The IRSRR makes use of the existing resources of the IRS for NPPs and has a similar organization. In particular, the nomination of the national co-ordinators should be the responsibility of the Member States. These national co-ordinators should preferably be officers of the regulatory bodies with technical background on research reactors. FIG. 1 IAEA RESEARCH REACTOR SAFETY PUBLICATIONS (Schematic) Safety Series Categories: Safety (*) ~ (*) Standards Code on Design ~ ,"." Code on Operation ,--I SS No. 35-S1 SS No. 35-S2 I ., , ., I ~ w If ~ y Safety (*) * it Safety Assessment Commissioning Utilization & (it Emergency (*) Guides & SAR Modification Planning Decommissioning SS No. 35-81 SS No. 35-G4 SS No. 35·G2 SS No. 74 , ~ ., r ~ r ~ r ~ r ., ., r Safety • • Instrumentation Radiation Maintenance & Operational limits Operating • Training and • Control Qualification of Practices & Prot. SelVice Periodic T-esting & Co.. nditions Procedures ... -.,. , Operating Personnel -.-~ SS No. 35-P SS No. 35-P SS No. 35-G7 SS No. 35-66 SS No. 35-PS SS No. 35-P J J " J I' J J ~ J ~ J ~ JI' J ~ J J ~ '" '" " TECDOCs (*) MTR (*) • Core Conv. (it) (*) Safety • Experience wI Siting Earthquake (*) Ageing Reviews Resistant Fuels Guidebook (lNSARR) Accidents (IRS) No. 403 No. 348 No. 467 Nos. 233/643 No. 792 Instrumentation (*) PSA (*) • Status as of April J 998: Source -Draft of relateD publication exists & Control Studies Term (*) Published No. 448 Nos. 400/517 636· A1.CAlA04. CORW TABLE I: INSARR MISSIONS 1972-1976 1977-1981 1982-1986 1987-1991 I 1992-1996 (INSARR missions pursuant to project or 5upply aj:reements ""ith the Agency) Ir---------------- Argentina 1973 (2) 1978 (2) 1992(2) Chile 1973 1977 1986 1991 Finland 1976 1981 1987 Greece 1972. 1976 198~. 1980 1993 lndoJ1esiH 197~. 1974 1975. 1979 I Peru 1978. 1981 1987 1992 Philippine. 1972. 1973. 1975 1978 1983 Romania 1983 1992 Spain 1982. 1986 Thailand 1974 1978 1982 1987 Turkey 1977 1986 1992 Uruguay 1974 1978. 1979 1984 Viet Nam 1985 1989 1995 Venezuda 1975 1979 1984 1988 YUj:oslavia (Slovenia) 197() 1985 1992 Zaire 1979 1984 199b (INSARR mis.inns at tI,e re'lue.1 of Memher States) Bangladesh 1995 Bulgaria 1990 Brn;1 1993 1977 1991 Chile 1991 Colombia 1977 1983 1987 Egypt 1985 Hunga!), 1983 1989 1993 Indone.i. 1979 1982. 1986 Iraq 1988(2) Korea. Rep. of 1976 1982(2) 1988(2) Peru 1992 Ponugal 1992 Turkey ()tr~" "- . 1998 YII,.iRi ... SSA' 1991 USSR 1990(2) YUIl,,,lavia (Semi.) 198~ Kazatstan 1993 Uzbeltistan 1993 No. of miuiona in period 25 27 32 21 19 CHAIRMAN : B • FARNOUX SESSIONIA HFIR UPGRADE PROPOSALS (Colin West) Question from Kir Konoplev of PNPI : About the choice and configuration of the reflector. A : OK. Did everyone hear the question? rm only asking that to give myself time to think! I proposed replacing the beryllium reflector, this outer - the larger piece of the beryllium . reflector here with a tank of heavy water, which is a better kind of reflector as we all know, for neutron scattering work. The consensus at the laboratory - there was a long debate - but the outcome of the debate was that that was risky. Not technically risky, but risky from a regulatory point of view. If we made such a major change to the reactor, then we would be into very many safety reviews and regulatory reviews. and it might take years - three years, four years, who knows? Whereas simply replacing the beryllium reflector with one with a slightly larger hole in it was a much easier process. So that is why we did not adopt the heavy water reflector. Now the other question I have forgotten! Oh. yes - the beam tubes penetrate essentially all the way through this outer reflector and into a small indent in the next layer of the reflector. So they are, in fact, closer to the thermal neutron peak than the beam tube tips at, for example, RHFIILL Grenoble or, I think, ORPHEE. By putting the cold source very close like that, we can get lots of neutrons, but also lots of heat, and Doug will tell you how we tackle that. Question from Hans-Joachim Roegler of Siemens: There are two reasons for stopping the ANS, one was the cost and the other was the enrichment. Now, with your statement that you will operate this year up to the year 2030, will you not have to justify that, for that long of a period, you will still use a high enriched uranium grade material in the reactor, and will you have similar studies as you had in the past with the ANS. Secondly, is there any evaluation of the budget you will have available safely for upgrading that and do your technical proposals not have to check about whether they fit into that budget? A : Yes, I have no doubt that we will have to debate the use of uranium with lower enrichment, although, at the recent RERTR meeting in Korea, the paper from Jim Matthaus from Argon said that at present they do not have a fuel that is suitable for the conversion of HFIR, and that is true. But, no doubt, they will keep reviewing that. As to the budget, I do have a few copies only of a little newsletter that discusses the budget. The budget that we need is about 70 million dollars, and of that, almost 10 million is to change the berylliwn, which we have to do anyway to keep operating. So the 60 million difference is for the upgrade. Of that 60 million, we already have spent or have available, we believe, about 10 million. We've got money already available for the cold source and we are working on it. So, we need about 50 million additional dollars over the next 4 years, which is not so very much. I mean it's much less than the operating budget, for example, of the reactor. So, I am hopeful that we will be able to get that, although the budget situation is very, very, very tight in the United States, and also here, I know; and so ifs possible that we will have delays, but we can't have too many delays because the berylliwn will need to be changed anyway regardless of Congress. Question from Horst Hassel of Jecta Consulting: I have two questions. The first is, is there any intentions of producing Mo-99, and the second has been partially answered already, the fuel element you use - is there any change besides the question of using the REU, any change of the existing fuel element design and, in this connection, if you start with REU, of course, is there any intention, later on, to change if there is a very high density fuel available without any change in the fuel element design, then will you use the LEU maybe? A : The Mo-99 production: we did propose producing Mo-99 at Oak Ridge and we could do so, but that is not the US Department of Energy's plan at present. Obviously, we have enough flux and space to do that. The upgrade itself does not include any change to the fuel element, although I have proposed at the laboratory that we consider increasing the fuel loading, using existing fuel, maybe even U308, in order to increase the core life and save on operating costs. It's the US government's policy that if the suitable fuel is available to convert reactors to LEU then they should be converted when the fuel and the money is available. That has not been the case so far, for example for the NIST reactor or the University of Missouri reactor either. Question from Edgar Koonen of CEN/SCK : Concerning the berylliwn reflector, what are the criteria for replacing it, what fluences have been reached, and is there a surveillance program? A : There is a surveillance program - we have to look at the berylliwn, I mean inspect it. We know from past experience that there has never been any problem at less than, I believe it's 279,000 megawatt days, and so, administratively, we say that we are required to replace the beryllium to operate after 279,000 megawatt days. It develops cracks. We have never had any problem from those cracks because we always change it out before we get a problem. Anonymous question: Have you ever lost any material? A: Yes, but not enough to cause problems anywhere else. I mean, not big pieces! Question from Hans-Joachim Roegler of Siemens: On the second side, the last sentence you didn't read - I don't know why - it says that you have minor modifications only to operate the reactor only up to the year 2030. It surprised me a little bit - how can you assess that because you have nearly 40 years to go. Does the history of this reactor really demonstrate that these 40 years will be accomplished without bigger changes? A: We think so. The major problem that we have is the embrittlement of the pressure vessel. It's a carbon steel vessel with a stainless steel coating. As you irradiate it, at first it embrittles rapidly, and as time goes by less and less rapidly because all the embrittlement you can do is almost done. Our most recent results and calculations, and surveillance results, imply that at least over that time period the embrittlement will not be enough to pose a safety hazard. The only other things that seem like they might be problems are that the beryllium, which we have to change out anyway, and all of the other components like the pumps and the motors and the control rod drives, and that's minor routine maintenance - it's not so minor in money, but its a thing you would have to do anyway. That's why we believe that. OPERATION EXPERIENCE AND CURRENT STATUS OF HANARO (Kye-Hong Lee) Question from Jean-Luc Minguet of Technicatome : What is the reason for having chosen an outer core and what is the technology of this outer fuel elements? Is it the same than for the inner core? A: It's the same. The fuel assemblies are the same. Only the difference between the cores is we had different shapes, one circular with 18 elements, and the other fuel elements with 36 elements, and those fuel assemblies are the same in the inner core and the outer core. We have eight holes in the outer core: four of them are used for fuel loading, the other four we use for the material we test. Question from Jean-Luc Minguet of Technicatome : Is there a special dedicated refueling program for these outer elements ? A : No, during refueling, we refuel during the same refueling period. Question from Klaus Boning of TU Munchen : You mentioned a Low Flow Critical Heat Flux experiment. Could you say a few more words about this? What are the details of these experiments ? A : We got this low flow, low pressure critical heat flux data from the AECL; and they performed their experiment, but their point wasn't that many to satisfy the regulatory body, so we conditionally had put some more time together, some more data points in that low flow low pressure critical heat flux data, so one of our colleagues in the Hanaro Center is setting up the test loop - I think it's already done - and in December they are going to run the test experiment so soon we'll get some data out of it. But I can't go more into the details! Question from Hans-Joachim Roegler of Siemens: So far we have only heard rumors about the money you paid to AECL for that project, but we have never heard the total cost figure. Is there any internal assessment at KAERI of what the total plant has finally cost, meaning internal costs of KAERI's supplies from Korean industry plus money you pay to AECL? Can you at least mention an oral figure, if you don't print it? A : I'm sorry I didn't catch his question very well ... Can you rephrase it? ru try to repeat it slowly. The total cost of the plant, because it has contributions from KAERI internal efforts, from Korean industrial efforts, from imported things and from costs you pay to AECL. Can you give a total cost figure for your plant? A : Mr. Chae, can you answer that question? Answer by Mr. Chae : I didn't have the detailed data for that - if I had expected this kind of question I would have prepared. Anyway, l, as project manager, I control the whole budget and the schedule, I can say the simplest our total investment in the Hanaro project. except the site purchase is estimated at 130 million US dollars When we purchased a significant part from the AECL, as you may know, for the extra structure like Zircalloy for the reflector tank and fuels, we had to purchase extra components, more material. The States didn't want to sell any more material to our COWltry, so they set some limitations for us to do this kind of business. We had some collaboration with the AECL for this kind of significant great components so we purchased these components for about 16 million dollars from Canada. (just a public loan from the EDC of Canada). All the other parts were done with local companies. Does that answer your question? CURRENT STATUS OF RESEARCH REACTORS RA AND RB AT THE VINCA INSTITUTE OF NUCLEAR SCIENCES (Miroslav Kopecni) Question from Johannes Wolters of JUlich Research Center: What is the reason that the pressure increase in your fuel barrels? Can you explain it? A : Yes. This is only, of course, estimated pressure, and the first step we are going to do is to measure exactly the pressure which might be present in the barrels. The reason is hydrogen buildup due to the corrosion process. And then, if you consider the results which are available in the literature on the corrosion of aluminum in a spent fuel pool solvent storage; you can estimate roughly what the pressure could be. Question from Klaus Boning of TV MUnchen : If I understood you correctly, you mentioned that you used graphite in the heavy-water cooled reactor RA. What is the reason for using graphite in a heavy water reactor? Do you think it's better or what? A: This is our inner core of the reactor which is situated in a stainless steel vessel and this is a reflector as far as I know. rm here for the chemistry of the pool, so rm not quite sure that I can give you a correct answer to that question. STUDSVIK'S R2 REACTOR - REVIEW OF RECENT ACTIVITIES (Mikael Grouues) Questiou from Hans-Joachim Roegler of Siemens: R2 is obviously widely used by these experiments, so of course the financing is also dependent on these experiments. Maybe I didn't get it - who is behind these experiments, who finances the experiments and who will use the results for improving the power reactors? Is that the power stations in Sweden or all over Europe or even more? A : No, Sweden plays a rather minor part. When these experiments were started in the 70s, they were mainly financed by Swedish parties, but since then the international cooperation has increased very much. So I can say that there are three types of experiments here : the most widely known, and the ones I have been speaking about all the time are the international programs. This means that, for each individual program here that I have described, a sort of consortium is formed, internationally - from different countries - to a large extent the fuel vendors, power reactor fuel vendors, to some extant also power reactor utilities in many countries, in some cases, safety organizations. But we also have a much wider volume that I have not talked about at all, with about the same types of experiments, which are bilateral - that is, an individual company, very often a fuel vendor, approaches us and says, "We have a new type of fuel we would like to test but we don't want our competitors to know about it unless the results are very good." So, in the international programs, Sweden usually participates but only to a small extent, some bilateral programs are of course made for the Swedish utilities and for the Swedish fuel vendors, but the very large majority of them are made for organizations in Europe, in the United States and in Japan. Question from Jong-Sup Wu ofKAERI : You mentioned damages in the primary cooling system due to debris. I would like to know what kind of debris and what was the damage - I mean damage to the primary component or fuel? A : Well, you see, what happens is that during maintenance, and in the western world one perhaps does not always have the discipline among the workers as you have in your countries, so this means that some swarf that should have been removed is left in the power reactor and since the water is circulating at very great speed, small pieces of metal can get stuck in the fuel- you know, power reactor fuel has spacers, and in the spacers, this debris can get stuck and then it sort of vibrates against one part of the fuel and finally it wears a small hole, which can only be a half a millimeter by one millimeter, and out of the hole one does not get very much fuel, so that is not a safety problem in itself, but if the water that intrudes into the fuel gives rise to cracks that are half a meter long, then so much fuel might come out and there have been cases where even the bottom part of the fuel rod has broken or there was one very well publicized event in the United States when the fuel was taken out, half of the fuel rod was left in the reactor because it had broken. So these are the problems. It's not an acute safety problem, but in the long run one does not want to have a buildup of radioactivity in the power reactor. Question from Francis Merchie of CEA : I understand that the last INCA once-through loop is installed inside the core and the core vessel and has to cross the lead of the vessel and the bottom of the vessel, right? A : Well, not the bottom of the vessel. It is installed in one of the in pile loops. That is, we have used one of the existing in pile loops, which is a hairpin loop - it looks like a U-tube. It takes two of the fuel element positions. So there is no penetration in the bottom of the reactor and what we have done is that we have put this device into one of the loops. It can be removed and we can have different types, but that's the way it operates. Q : What is the diameter of the loop? A : The inner diameter, in the present experiment, I think is 24 millimeters. That is, we can have samples that are 24 millimeters but it can be wider if there should be an interest need for that. Question from Johannes Wolters of Jiilich Research Center: When you do the fast-ramp tests, is there any impact on the reactor? Is there any impact on the power of the reactor? A : Are you speaking of the standard ramp test or are you speaking about this new ultra-fast ramp test? Q : The fast test. A: Well, the new ultra-fast ramps - the first one will be made very soon, it hasn't been made yet, but according to all the safety reviews and the calculations we have made, there will not be any impact that will be disturbing. That's one of the reasons why we have such a small piece of fuel for these experiments. Question from Adrian Verkooijen ofIRI, Delft T.U. : I would like to ask you with respect to you ultra-high burn-up program - did I understand this correctly, that you use the standard industrial fuel which has a burn-up of more than 20 megawatt days, and then extend that to 60 'til 80. A : Yes, we will take standard fuel that has been irradiated to 60. We will then refabricate short fuel rodlets, because, since the R2 reactor has a core of 60, this means that we can handle. .. the most convenient is to use so-called "rodlets", which are the fuel set lengths of perhaps 400 or 500 or so. So we would refabricate them. We have a remote refabrication capability. And then we would irradiate in a fuel element position. We have something called "boiling capsules" where several fuel rods can be irradiated together, just as a sort of continued base irradiation. We would then irradiate them up to 80. And then we would make ramp tests with them. Question from JJ. Verdeau ofTechnicatome: When do you intend to begin this ultra ramp project? A : Well. we have always had the policy that we are not going to start a national program with a technique that we have not demonstrated in full scale, and for this ultra-fast ramps - the ones that will take one second - the first tests will be made in January or February of next year. I don't remember. So this means that shortly after that time we are going to invite prospective participants to a first pre-project meeting in Studsvik. We have on severnl occasions invited people to other pre-project meetings, and then we have also had the infonnation meeting discussing this type of experiment, but we didn't want to start the actual attempts to get people committed to these experiments until we have shown that we really can make them in the way that we have intended to. So, the ultra ramp consists of three parts. one can say. Two of them are ramps of a type that we have been pursuing for many years - that could have been started a long time ago. But since many people were interested in these "ultra-fast ramps", we didn't want to start the program until we had proven that the ultra-fast ramps could be made in the way that we had intended to. And that will take place early next year. MAIN EXPERIENCES IN RENOVATION OF THE DALAT NUCLEAR RESEARCH REACTOR (pham Van Lam) Question from Dedi Ben Kraiem of CNSTN : Could you give some indication please of the rwming cost for operation, the time of operation of the reactor and the kind of users of the reactor? A : Now, our reactor is mainly used for radioisotope production. activation analysis and research and training. The main mode of operation now is one cycle of one hundred hours of continuous operation once in four weeks. We use the rest of the time for maintenance and some sorts of operations. And the running costs? Actually, I have not kept track of running costs at the time. Excuse me. Question from Wolfgang Knop of GKSS : What is the core size and where have you measured the neutron flux of 2.11013 neutrons per centimeter and second? In the neutron trap? Yes, the high neutron flux is received in the neutron trap in the center. The diameter of the core is 40 cm. And the height is 60 cm. Question from Francis Merchie of CEA : What is the use of the thermal column? A : At that time we put a pneumatic transfer system in the thermal column beside two pneumatic transfer systems in the positions 71 and 32 and in the thermal column we also put a pneumatic transfer system and from 4 horizontal channels we now use channel number three, it is a tangential channel, and channel number four - now we use two horizontal channels. UPDATE ON THE BRl REFURBISHMENT (Edgar Koonen) Question from Kir Konoplev of PNPI : Is it possible to use the concept of the "leak before failure" for your vessel? A : No, no, this was not accepted Question from Yang Tsing-Tyan of INER : You mentioned that you will implement the ISCC program on the BWR stressed components, but now the ISCC of BWR is more severe than the PWR. Why don't you extend the program forthePWR? A : rm not sure I understand your question. It's mostly BWR and PWR - it's what you say. Q: For example many cracks have been found in the BWR core shroud. A : There are two aspects to this, first the internal program has to be for domestic power stations and we only have PWR reactors, so the internal program is focused on our reactors. Now, we discussed having some done nationally, but in that case we will build a new loop. As far as it's based only on our net program, we will do it for some PWR conditions. So it depends on whether the negotiations come to an end or not. Question from Chen shih-Kuei ofINER: You mentioned the PSA study for the reactor. My question is, first, why do you want to do a PSA study for that? Is that a regulatory requirement? The second is, are there any criteria for your results? Why do you want to perform modifications for your LOCA situation? A : We started this PSA in '91. At that time it was not requested by our authorities, but our authorities have changed since then - we had another authority which specifically concerned with us, it was not the same people as those who deal with power stations. So we did the PSA and the aim was to prioritize items for the refurbishment program, so it was an internal effort. Afterwards, when we had this new authority AV N they had already made some PSA studies for all our PWRs, so they were very happy that we had already done it! We had to make a presentation and they were quite satisfied with the way we had done it. They are not really looking at the figures. They are looking at relative importance of items. Now, the LOCA is one of the reference accidents for the PWRs. Of course, that sort of study had not previously been done for BR2 so we had no relap model before that. The relap model was triggered by the PSA study. And there is, of course, some aging in the design of BR2 - part of the primary circuit is outside of the containment building. So, even in an improbable case of core damage, we must show that we are absolutely certain that we can keep any activity inside this containment and that this primary circuit is somehow a containment bypass. Question from Albert Lee of AECL : Edgar, how did you solve the problem with RELAP5 being unable to model situations at very low pressure and low flow? A : I didn't solve it, first of all! rm afraid we would have to talk to the specialists. We started the model several times, changing the calculation step. rm unable to give you the real details, but there is definitely a problem at low pressures. Sometimes the calculation stops. Also the values that you have in the tables don't go so low. I wasn't the one who made those calculations, this is some dedicated person who does only that. Q : Without valid steam property tables and water property tables at low temperature and low pressure, RELAP5 gets into numerical instabilities and you can alter the time steps to try and adjust for it but you have to be very careful about matching up the answers between successive cases. There can be some unphysical discontinuities that appear. A : We have, and are still busy by making some benchmark calculations using the 1963 thermohydraulic perturbations. Those were not really LOCA , but there were loss of flow, loss of pressure, and we still don't have all of the data, but most of the data, that RELAP would require. So we try to recreate these conditions. One of the problems that we have is that we don't have the really detailed time delays from those days of valve closing and how the scram was actually operating and so on. These are of course factors that you really have to know, so the course closing of the valves. But we are using these 1963 perturbations to benchmark the calculations. It's the best we can do. Question from Jean-Luc Minguet of Technicatome : I have just two questions. Would you remind me of the level of pressure of the primary circuit of your reactor? And then what are the scheduled research and irradiation programs after the startup of your reactor next year? A : The pressure is between 10 and 12 bars. And I said that we'd try to start in April, or maybe if this doesn't work it will be June. Next year we will probably be starting three cycles and the next year 1997 will be five cycles. I can't really tell you exactly. Mainly, the standard cycle is 21 days, but sometimes we fWl very short cycles for transients, which are not counted there. So 25 times 5 is 125 - that's what I said about 100 days CHAIRMAN : C. DESANDRE SESSION 1B RESEARCH AND SERVICES OF LVR-15 REACTOR IN REZ (Jan Kysela) Question from Johannes Wolters of Jiilich Research Center: You mentioned that you operate the reactor with a staff of 32 people. What does this include? Radiation protection people and maintenance people? What is included? A : answer not recorded Question from Hans-Joachim Roegler of Siemens: You mentioned that you convert the reactor from 80% to 36% - this is still HEU as we all know of course. Is it not possible for you to stick to the 80% until the 20% is ready, or do the Russians no longer supply the 80%? Is that the reason why you had to convert? A : answer not recorded Anonymous question: Regarding the BNCT - what is your tentative schedule for completely implementing this BNCT facility? A : answer not recorded HIFA R MAJOR SHUTDOWN REPORT (Shane Kim) Question from Doug Selby of ORNL: I have two questions: Is the vessel pure aluminum or an aluminum alloy? And secondly, what was the pH quality of the water in the place where you saw flaking? A : It's pure aluminum, 1000 series. I think its 1050, rm not sure. As for the pH quality, how can I answer that? The water chemistry in the heavy water is monitored every day. I think it's about 5.5, something like that. But it varies. Question from Kir Konoplev ofPNPI: What is the fluence of the aluminum tank? A : 1.4 x 1014 that's the flux. Is that too high? Question from Francis Merchie of CEA : Some time ago, I mean two or three years ago, ANSTO was considering the replacement of HIFA R by a new reactor. What is the situation now? A : We're just waiting for it. In about three year's time we should know, but we have to make sure our Greenpeace is happy about it. Question from Edgar Koonen of CEN/SCK : Can you explain why, when you made the ultra-sonic inspection it was just used for wall thickness measurements and not for flaw detection or other indications? A : Actually we used two different types of ultra-sonic measurements, one actually is to measure the thickness of the aluminum tank to make sure that thickness didn't change. Between the wall and the probe there is about a one millimeter gap filled with de-mineralized water. Q : So you could have detected some flaws? A : We are not finding any imperfections in the wall. To find those imperfections, we have time-of-light ultra-sonic measurement which is angled at 90%, and we did that along the weld joints. Q : Is there any limit set for the life of this vessel? A : That's what we tried to find out, and so far we haven't seen any reason why we couldn't continue operating the reactor with that aluminum tank. But we'll keep watching. Q : You do this every four or five years. I just wonder what the dog does in between time? What does the dog do when he's not doing this? A: Getting fed! Question from C. Desandre : I think that the reactor in the range of power of 10 megawatts is certainly the oldest reactor in operation, isn't it? Do you know of any other reactors? It was critical in '58, so that's 38 years of operation. I think it's the oldest, isn't it? Anonymous answer: No, there's one in Denmark. MODIFICATION OF JRR-4 (Teruo Nakajima) Question from Edgar Koonen of CEN/SCK : What was the intensity of the earthquake that you had to consider in your calculation? A : The study considered a level of 0.72 g. In Japan, the examination guide for seismic hazards in nuclear plants is 0.72 g. Question from Johannes Wolters of Jillich Research Center: Do you need a new license for the modifications and what are the costs? A: Yes and we estimated about 35 million dollars. Question from Wu Jong-Sup ofKAERI: You considered a larger bore size of the silicon doping system. Do you know the diameter of that? A: Yes, we changed to a 5.5 inch round pipe. Now they are 4 inches and we are changing to 5.5 inches. Question from Jean-Luc Minguet of Technicatome : When do you intend to start the BNCT treatments for patients. A: Maybe when the modification work is over in 1998 - at the end of 1998, in December. OVERVIEW ON THE MAIN ENGINEERING WORKS PERFORMED ON FRENCH RESEARCH REACTORS THESE LAST YEARS (pascal Rousselle) Question from Hans-Joachim Roegler of Siemens: Does the new porous concrete which you use below the bottom of the pool allow detection of where the leakage is or just does it just detect that there is a leakage? A : You can detect leakage by conducting a potential leakage to sumps. But, of course we have to make a specific examination after to localize it. Question from Klaus BlSning of TU MUnchen : You mentioned that you planned to exchange the pressure bolUldary around the core very soon, and my question is: is this the first exchange of the core shroud for ORPHEE? A : Yes, it will be the first one. Q : After 15 years? A : Yes, after 15 years of operation. Q : And what technique are you going to use? Do you have to fill the whole heavy water system with light water? A : Yes, we have to fill it. Q : No problems with pollution by contamination of the heavy water? A : We have done this operation many times, when we remove the tubes for instance. Q : Do you have an on-line cleaning system? A: We clean the heavy water tank before putting the water inside. So that is very easy to do. There are no problems - we've done it many times. UTILIZATION OF THE BUDAPEST RESEARCH REACTOR (Istvan Vidovszky) Question from Bernard Farnoux of CEA I DSM : What kind of material do you plan to use for the cell of the liquid hydrogen cold source? A: Aluminum Question from C. Desandre : Is it a vertical type, is it a thermo-siphon type? A : Liquid hydrogen loop. Question from Jean-Luc Minguet of Technicatome : Who performed the basic design of this CNS? A : This is complicated. It was mainly done by PNPI in St. Petersburg, but this was in cooperation with many other partners. And some of the design work will be done by a Hungarian design company and even some tasks will done by members of foreign subsidiaries. It's complicated. Q : After the startup of your reactor, how many years do you plan to operate it? A : This question is very hard to answer. Of course, it depends mainly on financial problems,. So we have the fuel for six more years, maybe seven, and whether we will be able to buy fuel again is an open question now. Of course, we hope that from the technical point of view the reactor could be operated 30 more years, but whether it will be financed or not I can't say. CHAIRMAN: K. KONOPLEV SESSION Ie DESIGN MODIFICATION OF BANARO REFLECTOR COOLING SYSTEM (Jong-Sup Wu) Question from Shieh Der-Jhy of INER : In your design, the reflector cooling system has a higher pressure compared with the secondary cooling system, right? In this case, do you have any concern about the diffusion of tritium through the cooling system and into the cooling tower? A: We have a radiation monitoring system in the secondary cooling system to detect if there is some leakage from the primary or D20. Question from Hans-Joachim Roegler of Siemens: If you had to design this as a new system, instead of improving an existing system, would you have taken another heat exchanger like a pipe-type heat exchanger instead of the plate type? A : In our D20 boundary, the space was very small, there was a limit to the operator heat exchanger type. But I think the tube and shell type is larger than the plate-type heat exchanger. The plate-type heat exchanger has the advantage of a higher efficiency and is highly compact, but it has the disadvantage of the plate heat deflections. So we would be very prudent when using this type of exchanger. Comment from Guy Gistau of Air Liquide : rm not used to water heat exchangers, I mainly deal with cryogenic heat exchangers. But we also use plate heat exchangers and there are some possibilities to avoid this volume change even when the pressure changes. A : But in our case, the reflector cooling system is very sensitive to limit the pressure in the system, because we have an outside shell of the core which is very sensitive to high pressure from the reflector system, so it is limited. IAEA ACTIVITIES ON RESEARCH REACfOR SAFETY (Francisco Alcala-Ruiz) Question from Hans-Joacbim Roegler of Siemens: Assuming that a COWltry does not have its own rules code and standards on research reactors, would you then consider the schemes that IAEA has implemented is sufficient to rWl a licensing process for a nuclear plant or for a research reactor? A : Well, the scope of the documents prepared by the agency for nuclear reactors is also limited. These documents are only intended for some research reactors, not all. We couldn't establish a clear limit for the scope of the documents. We say that they should be applied to research reactors up to several tens of megawatts. For high-power reactors, I think another standard protocol should be used. But just up to this distribution of power by a research reactor, I think that many, many reactors come Wlder the potential application of these documents. But not all, of course. SESSION 2 RESEARCH REACTORS IN DESIGN OR CONSTRUCTION PAPERS STATUS OF THE TRR-II PROJECT Li-Fu L!n, Wei-Min Chia,Chao-Yie Yang, . Heng-Shiung Sheu, Cheng-Chung Wang, Der-Jhy Shieh Institute of Nuclear Energy Research 1000, Wenhua Road, Chiaan Village, Lungtan, Taoyuan, Taiwan, 325, R.O.C, FAX: 886-3-4711443 ABSTRACT The 40 MW Taiwan Research Reactor (TRR) operated by the Institute of Nuclear Energy Research (INER) went critical in 1973 but was pennanently shut down in 1988. After the shutdown ofTRR, remodeling of the reactor into a light water pool type research reactor, i.e., TRR-II, has been considered by INER. TRR-II project is currently working on environmental impact study, reactor dismantling and construction planning study, and some conceptual design. The design goal is to have a reactor with thennal as well as fast flux around 1014 cm'2 sec·1 and a nucl~ar thermal power of less than 20 MW to meet the utilization requirements. 1. Introduction The Taiwan Research Reactor (TRR) was a natural uranium metal fueled, heavy water moqerated and light water cooled reactor with a thermal power output of 40 MW. It first went critical in January 1973 and, after a little more than 15 years in operation, was permanently shutdown in 1988. An evaluation program was then launched to figure out: (1) how to deal with (i.g., dismantle) the TRR, (2) do we need a new research reactor, and (3) what type of the reactor is the most proper one in consideration of the techriology localization and the domestic need of neutron sources for research, industrial and medical applications. Initially, a proposal of dismantling the original core block by one piece removal and procuring a multipurpose research reactor through turn- key project has been discussed for years and revised several times. However, most of the reviewers had suggested that instead of purchasing a research reactor from foreign vendors, INER should build one by the institute itself in order to promote the local technical capabilities. In 1995, a new proposal based on comments from reviewers got ~ approval from Atomic Energy Council (AEC). Then, the proposal was submitted to the National Science Committee (NSC) of Executive Yuan. Due to the lack of consensus on environmental impact assessment ofTRR-II and on the priority for TRR-II construction, NSC reviewers only approved a limited amount of budget for the following studies in 1996: (1) environmental impact assessment, (2) basic engineering design and safety analysis, (3) project schedule and detailed cost estimation, (4) cost benefit analysis, and (5) user training program. After completing the studies, the TRR-II Project will be submitted to Executive Yuan for an approval again. 2. Status ofTRR-II Project 2.1 Objective After the shutdown ofTRR, not a neutron source with flux higher than 1014 cm-2sec-1 is available in Taiwan. For the purpose of continuing and promoting the peaceful atomic applications, by taking a careful survey of needs and considering the constraint conditions, it is decided"to build a multipurpose research reactor with a thermal flux around 1014cm-2sec-1 Survey of need was conducted through numerous face-to-face discussions with domestic universities, research organizations, hospitals and industries as well as considering the experimental requirements of INER itself. It was realized that the local utilization requirements of interest to use TRR-I1 as a neutron source include the followings: • neutron transmutation doping (NTD), • neutron beam experiments (NBE), • nuclear fuel and material development (NFMD), • neutron activation analysis (NAA), • neutron radiography (NR), • boron neutron capture therapy (BNCT), • water and radio-chemistry (WRC), and • radioisotope research and production (RIPP). Among these applications, INER and other institutes in Taiwan already have some experiences in isotopes production, neutron activation analysis, neutron radiography, and BNCT. The rest actively worldwide studied applications, such as NTD, fuel and material tests, and neutron beam applications, will be somewhat new to Taiwan. TRR-II will contribute greatly in promoting these new applications. Currently, seven users' planning groups were formed and are working actively in recruiting potential users. 2.2 TRR-II Design Because a thermal as well as fast neutron flux around lOI4cm-2sec-l is sufficient for majority of users and INER technical strength is V02 type of fuel, therefore, V02 fuel,is selected as the first choice for TRR-II core. In the future, if a higher neutron flux will ever be in demand, the core can be readily converted with other types of fuel. For core design, two core configurations are under study. One design is a rectangular core and the other design is a circular core. All fuel rods will contain low enrichment V02 pellets. The core will use light water as moderator and coolant, and the core is surrounded by beryllium reflectors and a heavy water tank. The reactor core is located at the bottom region of the reactor pool. Control rod drive mechanisms (CRDMs) are installed below the reactor core. Fig.l shows the cutaway view of the TRR-II. For the reactor cooling and shutdown systems, the key features are as followings: (l) The core is submerged in a pool. The depth of water is about 8 meter. The pool is covered with top shield to reduce the radiation level. (2) The coolant flow is downward for the purpose of reduction of the radiation dose of l~. (3) The primary coolant pipes penetrates the pool wall at a level higher than the core to avoid core uncover caused by coolant pipe break. . (4) The primary coolant loop is equipped with siphon break valves to avoid loss of pool water due to siph~n phenomenon. (5) The primary coolant pumps are backed up with emergency auxiliary pumps to assure decay heat removal after reactor shutdown. (6) The lower plenum of the core is equipped with two flap valves so that decay heat can be removed with natural circulation for long term cooling after shutdown. (7) The pool water clean up and heat removal system wiIl play the role of ultimate heat sink when the main coolant systems shutdown. (8) In addition to the control rod shutdown system, the heavy water dump system will be used as a backup shutdown system in case of anti.cipated transient without scram (ATWS). All of the control rods will drop into the core by gravity, also is the dump of heavy water. 2.3 Dismantling of the TRR core The new reactor wil! be constructed after dismantling the old one. For the sake of minimizing radwaste generated from dismantling of the old reactor, the possibility of keeping the biological shields and other usable equipment is still under study and the TRR reactor building and stack will be retained. The conceptual study of the piece by piece dismantling work for the TRR core is elaborated below. Fig.2 shows the general arrangement of the old TRR reactor. The zircaloy calandria is the central component of the reactor. The calandria is a cylindrical vessel. It is supported by the lower axial thermal shields. The vessel is 106 inches diameter (inside) and approximately 129 inches high. It is penetrated by 199 vertical tubes arranged on a hexagonal lattice. The calandria is surrounded by the graphite reflector, and then the thermal shields and the biological shields. Before developing the idea ofTRR piece by piece dismantling method, specific activities of the major components have been calculated using ORIGEN-II and QAD-CG codes. The weight, volume and material for the major components are shown in Table 1. Table 2 shows the radioactivities of these components. Referring to these results and according to ~e strategy ofTRR remodeling project, the piece by piece dismantling method was developed. We compared the dry process and wet process for dismantling by taking lots of factors into account such as cost, schedule, waste production, etc. Finally, it is dicided to use wet process. The dismantling work will be done under the water. To ensure no leakage will occur, all the openings of reactor will be sealed, and a layer of cylindrical cover plats surrounding the biological shield will be installed to act as a second barrier. The principal dismantling step are shown in Fig.3. Since no final disposal place is available in Taiwan, on site storage for high radiation level waste generated from the dismantling work is the preferred choice. 2.4 Environmental impact assessment A fonnal environmental impact assessment for research reactors is not a regulatory requirement in Taiwan now. But today, people are getting more concerned with th~ quality of environment and a construction project of nuclear facility surely will attract great attention of the environmental group. Therefore, following the suggestion of National Science Committee of Executive Yuan, a complete environmental impact study similar to the one for a nuclear power plant is conducted. The task is awarded to the Environmental Department ofE&C Engineering Corporation. Because strong protest from the surrounding inhabitants is foreseeable if the LPZ is beyond the boundary ofINER, one of the TRR-II policy is to design a reactor with the Low Population Zone (LPZ) inside the boundary ofINER, 3. Concluding Remarks TRR-II Project team is currently working on environmental impact study, reactor dismantling and construction planning study, and some conceptual design. In the design process, the most difficult problem comes from the uncertainty of regulation for research reactors. In our study of the design of research reactors around the world, we find that there are lack of consensus with regard to some design principles, such as: • the postulated accident for LPZ calculation, which is varied from the assumption of a single fuel bundle handling accident to the BORAX type accident, • the necessity of a secondary shutdown system, • the extent of redundancy and physical separation, • the applicability of the standards of nuclear power plants, such as AS:ME Section ill, to research reactors, etc. In our design, the safety codes ofIAEA(t)(2)(3) are followed. However, many subjects still are remained to be discussed with the regulatory authority, i.e., ROCAEC. A good research reactor is an essential tool for peaceful atomic applications in Taiwan. .T herefore, we look forward to receiving a final approval for construction of the reactor in 1997. 4. References (1) IAEA, Code on the Safety of Nuclear Research Reactor: Design, No. 35-S1, 1992. (2) IAEA, Code on the Safety of Nuclear Research Reactor: Operation, No. 35-S2, 1992. (3) IAEA, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, No. 35-Gl, 1994. Table 1. Weight, volume and material of reactor components Pan Weight (t) Material Volume 1. ReIIloval biological shields 70.86 . concrete 3.81l1 X O.5t(m) 2. Fixed biological shields outside 61.24 concrete removal biological shields 3. Upper thermal shields 32.80 cast iron 4. Embedded thermal shields - 56.47 cast iron 5. Lower thermal shields 71.22 cast iron Sill X 0.42t(m) 6. Floor plate 24.49 cast iron 7. Side thermal shields inner 138.49 mild steel 62 X250 X36(cm) outer 8. Reflector inner 60.67 Graphite outer 9. Thermal column 19.59 Graphite 10. Biological shield 1574.42 Concrete 11. Calandria 3.12 Zircaloy 12. Revolving floor 18.45 Cast iron 13. WShapes 5.75 14. Master plate -- Total of 1-14 2137.57 15. Floor concrete block 588.86 Total 2726.43 Table 2. Calculation result of the activation of the major components in reactor core Structure 2yr Syr 10 yr 30yr Total Weight Specific Activity Components (2yr) Reactor Calandrla 3.921E2 I.2S6El 1.356El l.57SEl 3.12 4 p.Cilg Graphite Reflector 1.726El 1.660El I.S73El 1.390El 60.67 0.3 p.Cilg Upper Thermal 1.8S9E6 8.3S1ES 2.208ES 1.16SE3 32.80+S6.47 Shield 9E3 Ii Cilg Lower Thermal 4.038E6 1.814E6 4.798ES 2.S31E3 71.22+24.49 2E4 p.Cilg .. Shield Iron Thermal 2.395E5 1.07SES 2.835E4 !.392E2 138.49 7E2 p.Cilg Shield +3.510E3 +1.602E3 +4.204E2 +2.046EO Thermal Column 6. 12 7E-5 6. 125E-S 6.121E-5 6.106E-S 19.59 3E6 p.Cilg Reactor Biological 1.466E-l 1.0S7E-2 1.057E-2 1.0S7E-2 IS74.42 lE-S Jl. Cilg Shield 1.466E-l 1.0S7E-2 1.057E-2 1.057E-2 Removable Biological Shield 1.04SE-l 1.08 1E "2 1.080E-2 1.080E-2 70.68 SE-5 Jl. Cilg ·----r---f-Wa-r1Sdng~ P~co l I .. '-" . __to primaty cooling system -+-1-1-+----1 .- Decay TanI!. Basement V VLH. t Fig. 1 A vertical cross section view of TRR- IT • 12 1 2 :, '. 3 ", '.. . 4 14 16 10 LO\VER HE.-illER ROOM 11 l-CENTRAL T1IBIBLE 9-BOTTOM TIIEIDL\L SHIELDS 2-~XPERIMENT,\L LOOP POSITIONS lO-SIDE TllERMAL SHIELDS 3-J-ROD POSITION ll-AIR EXllAUST 4-TOP TIIERM,\L Sill ELD 12-INLET llEADER 5-GRAPlIlTE REFLECTOR 13-BIOLOG!CJ\L SHIELDS 6-REMOVABLE PL~G 14-EXPERIMENTAL HOLES 7-REMOVABLE GRAP[JITE BLOCKS 15-10N C[JA!IBER [JOUS 1N GS 8-LEAD DOOR 16-FLooR PLATE Fig. 2 Rea.ctor general arrangement Preparation work Removing 4 layers of Biological shield Installation of the modified fueling ' ________- +{ machine and center-hole jacks ' dismantling and Removing of 2 layers of upper thermal shield Installation of core boring machine and disk saw Cutting of the top fixed biological shield Dismantling of Radial thermal shield Installation of Special milling Cutting tool Dismantling of reactor calandria Installation of multi-a.'(is 1+----------11 Installation. of I underwater plasma . vacuum devices . Dismantling of reflector Dismantling of I st ring ofiron shield Dismantling of 2nd ring of iron shield Dismantling of Lower thermal shield Settlement of reactor crane Removing of Lower plate Final Refurbishment ofbiologica1 shield Restoration Work Fig. 3 Principal steps of TRR dismantling __- ------------------------------- 1 UNIVERSITY OF BOCHUM Department for Nuclear and New Energy Systems Prof. Dr.-Ing. H. Unger Thermohydraulic and Mechanical Analysis of a Scale FRM-II Core Dummy J. Adamek, H. Unger IGORR 5 Meeting Aix-en-Provence, November, 4. - 6. 1996 Contents D Characteristics of the FRM-II, D Objectives of the Experiments and the Test Program, D Test Facility for the Qualification of the FRM-II Fuel Element, D First Experimental Results: • Vibration Tests, • Pressure Drop Measurements, • Start-Up and Shutdown Tests of the Main Pump, D Summary. 2 View of the FRM-II Reactor Building Primary Heat Exchanger r-Control Rod Primary Pump Fuel Element 3 Partial View of the Inner Section of the FRM-II Control Rod Shutdown Rod Drive Unit (5X) Central Channel 4 Fuel Element KKE-7 of the FRM-II Spacer Fuel Plate Sieve 720 1326 <= Flow Direction Central Channel Tube Fuel Element Outer Tube Cooling Gap Fuel Element Inner Tube 243 ~----------------------------5 Objectives of the Experiments and the Test Program Performed D Objectives of the experiments: • Verification of the hydraulic design of the FRM-II fuel element, • verification of the feasibility of the core cooling in case of a partly ob- structed sieve, • verification of the fatigue strength of the fuel element by means of an en- durance test. D Test program: The tests are performed with a 1 : 1 scale dummy fuel element, with deple- ted uranium. • Vibration behaviour of the fuel ele- ment in dry and operating conditions, • pressure drop measurements at diffe- rent flow rates, • start-up and shutdown tests of the main pump, __- --------------------------6 Objectives of the Experiments and the Test Program Performed (Continued) D Test program (Continued): • Flow division in the region of the fuel element, • flow profile at the fuel element outlet in case of free sieve, • flow profile at the fuel element outlet in case of partly obstructed sieve, • endurance test of the fuel element, la- sting 60 days (equivalent to 1.2 opera- ting cycles). ,---------------7 The Test Facility for the Qualification of the FRM-II Dummy Fuel Element D A test facility was built at the Ruhr Uni- versity of Bochum (Germany). D The test facility mocks up the central region of the reactor coolant system in a 1 : 1 scale. D The two pump units enable flow rates from 0 lis to 60 lis (small pump) and from 200 lis up to 425 lis (large pump). D Presently tests are performed at a flow rate of 60 lis (emergency core cooling), 300 lis (nominal flow rate of the FRM-II) and 425 lis (double hydraulic load rela- tive to the nominal flow rate). 8 Block Diagram of the FRM-II Fuel Element Test Facility Pump Loop 1 ON 150 PN 16 ON 150 PN 16 P"25kW n = 2950 llmin Cooling Loop Pump Loop 2 ~. ... = 425 I/s ON 100 T P P.. ... = 595 kW P PN 10 ON 350 PN 16 T p. ... = 729 kW T n. ... = 1475 l/mln ____ ....T. ... Bypass 1 v. .. 3 ~-------------~------------------L---= 10 m (Flow through the Gap Between the ~ Fuel Element and the Central -----I Channel Pipe) ~ Bypass 2 v = 31 I/s ~---~~4-------------~------------------L--4 (Flaw through the Gap Between the P = 18.5 kW Fuel Element and the Control Rod) P ON 350 PN 10 T P=22kW n = 2950 1/mln ..-~-- (.) ca c LL .o - U eCnD ...., c: CI) E -CI) -eo o w ."--C - .c De.o., CI) c en.CeDn ::J -- LL :E• a: LL .C.I.) .c II- o ~ .C-I) > __- -------------10 Vibration Measurements in Operating Conditions o Preliminary experimental modal analy- sis has proofed: Vibration measurement of the fuel ele- ment is possible at the outside of the central channel tube also. o The vibration measurements in opera- ting conditions are performed with ac- celerometers placed on different locati- ons on the outside of the central chan- nel tube. o These tests are performed at flow rates up to 360 lIs. ~---------------11 Locations of the Accelerometers for the Vibration Measurement Position of Accelerometer at the Spherical Seat Position of Accelerometer at the Central Channel Tube .F- ---------------------------------------------12 Acceleration at the Spherical Seat Flow Rate 200-360 lIs, Uranium Dummy Fuel Element 130 120 r--.. m ""0 '--" Q.) > 11 0 Q.) - c 0 +J 0 L. Q.) 100 - Q.) U 0 200 lis «u 9all~1~ fs 250 lIs • 300 lis ..-- 360 lis 80 ~I! __~ ____~ __~ __~ ~ __~ __~ ____~ __J __ _~ ___ _ o 500 1000 1500 2000 2500 Ref.: 3.1 4E-7m/s2 Frequence (Hz) ~---------------------------------------------13 Amplitudes at the Spherical Seat Flow Rate 200-360 lis, Uranium Dummy Fuel Element 60 50 40 e 200 lis -A-- 250 lis r--. en 30 .•. 300 lis ""0 360 lis '--" - 20 Q) > Q) 1 0 .- Q) ""0 .:J 0 ..-.. ... 0.. -10 «E -20 -30 -40 -50 0 500 1000 1500 2000 2500 Ref.:5E-8m Frequence (Hz) --------------~14 Pressure Drop Measurements D Measurements of the pressure drop within the fuel element, with particular interest in: • Pressure drop at the sieve, • pressure profiles within the plate zone of the fuel element. D The tests are performed at flow rates up to 360 lis. __ -----------------------------------------15 Locations of the Pressure Transducers for the Pressure Drop Measurements o 10 10 ,.. • 10 ....... ,..... ..... 1...0.. O,.....,l. "~'" "'" ,~., 10 "' <".'"" D <.D ,1..0.. co <.D co o co Ol ~ Ol o ..... 0 ~ .~... ...,...,...... ..... ~ 16 Pressure Drop Within the Uranium Dummy Fuel Element Flow Rate: 200 - 360 lIs Top of Fuel Element: Transducer 1 Sieve: Transducer 2 - 3 Plates Zone: Transducer 5 - 1 0 Bottom of Fuel Element: Transducer 12 ,C..). ~CIO -~-=. Cb co 5 f/) ~ "t Q:: (\I - ,..... _ 4'0 # .s>oo ~ ~<$»o o~- '9'01;>.., ~6"o "" ~.:6' _ "b ;- <5>4;s.,.1 ~ -Q.~ ~ __- --------------------------17 Start-Up and Shutdown Tests of the Main Pump D Acceleration measurements at the sphe- rical seat of the fuel element and its bot- tom region during pump speed run-up and shutdown. D Test program: • Main pump speed run-down from a flow rate of 300 lis simulating the run- down of the primary pumps of the FRM-II, • shutdown of the main pump at a flow rate of 300 lIs simulating the blockage of one of the four primary pumps of the FRM-II, • starting of main pump, test of transient phase from idling conditions (200 lIs) up to the nominal flow rate of 300 lis. ~-----------------------------------------------18 Run-Down from 300 lis to 200 lis Acceleration at the Spherical Seat 1,25r-------~--------~--------~--------------------------~--------~------~ 450 1,00 •• __ • _._. ___ + ___ * __ *_~. * .L __ ~~ ._~~ ___ • ________ • __ • _. . ... --: .. -- ,- . 400 0,75 ·············--·······-i:· ·····--········-------··- . .;. i······ --- ---t-·- 350 i Flow Rate 0,50 300 ........ ::r::n:. 5 '-' 0,25 250 Q) 1\1 ~ 0:: Q) 0,00 200 ~ 0 ~ u:: -0,25 150 -0,50 100 Acceleration .. ., -0,75 -1--•.. -.-.. --.•. -•. ..·-·-f·-·······-··············· .. !.. 00_._ - ----. ---. --. -··-r-· 1- . -~.- ... _--_ .. -·-~-·i .---.. --- .. .... --+.-..... ---.-.. -. l 50 -1,00 I I 0 o 1 2 3 4 5 678 Time [s] ~---------------------19 Main Pump Turn Off at a Flow Rate of 300 lis Acceleration at the Spherical Seat 1,25 450 1,00 400 0,75 +:-------------------------il-- ---------. - .. - .- .•.. - -. ----. ---------- .. -.. ~. .· ·-1-- --_. ------ ___ .0_ ------ + ••• - --_ ••• ----••• __ o •• ~ ........ _- .0_ --- 350 N" 1 0,50 300 .~.... . I .~1 .... c 0,25 ! 250 -CD 0 cu ~ 0: CD 0,00 200 G) ~ 8 LL .- -0- CD "c . :E::J O:::ltn • • • Pre-project Phase: 95/11/01-97/03/31 • Advance the Project Schedule • Reduce Risk to Cost and Schedule by * Advancing the Analysis * Advancing the Testing * Advancing Fuel Development * Developing the Concept * Starting the Licensing Process IRF Pre-Project Activities Advance Development of System Concepts for IRF • identifying technical feasibility issues and developing feasible system concepts • confirming feasibility for manufacturing components • improving definition of project work scope • developing concepts for novel systems (e.g., fuel handling for test loops) • updating requirements for experimental facilities • advancing system design • investigating alternative IRF building concepts IRF Pre-Project Activities Up Front Licensing Activities • prepare IRF PrQject QA Program - Complete • prepare Licensing Plan - AECB Review • prepare Licensing Basis Document - AECB Review • prepare Safety Analysis Program • prepare Safety Design Guides - 75% Complete • define requirements for environmental assessment • prepare software validation documents • develop criteria for containment design • prepare preliminary assessment of fuel disposal options - Complete Up Front Licensing Objectives • Familiarize AECB staff with IRF concept • Reach agreement with AECB on IRF QA Program • Identify major safety analysis, design and licensing issues • Reach agreement with AECB on licensing basis • Define requirements for environmental assessment • Define design criteria for containment A~ IRF Pre-Project Activities Analysis'and Testing • Review thermal hydraulics and physics methods and actions added to IRF Pre-project plan - Complete • Building detailed thermalhydraulics and physics models - 75% Complete • Performing calculations to support safety analysis and design • Assessing requirements and code capabilities for 3-D , kinetics . • Preparing validation of thermal hydraulics and physics codes - Technical Basis Document in review -fos/ IRF Pre-Project Activities Analysis and Testing • Performing heat transfer measurements to improve correlations for CATHENA - Tests complete • Performing bundle CHF tests for CATHENA validation • Measuring thermal conductivity for irradiated fuel • Measuring oxide layer on irradiated fuel • Developing techniques for fabricating U3Si2-AI fuel rods with burnable poison • Assessing need for physics parameter measurements using ZED-2 - Complete, experiments not required B C 1-------------«) 35000-------------l ~iB C 0..-H..._..". """ Test Fcu:ilUy. ol' Sleam SoaIri:ng Loop LDopB Room 89m" ~B "?C To/From Operations Bldg. - - '" • ~ i I ~., Performance • Operating Power, Nominal 40 MW • Horizontal Test Sections - Test Section Power-1300 to 2000 kW - CANDU Bundle Power -670 to 1200 kW - Outer Element Power -53 to 77 kW/m • Beam Tubes - Unperturbed Flux 2.7 to 3.8 X1018 n/m2/s - Perturbed Flux at Cold Source Chamber 2.2 X1018 n/m2/s Beam TUbe (8) r ':Old Neutron Souru (2) ~"f-_ Materials Test Site (4) DrivCf Fuel In Split Core Irndiation Research Facility om Reactor Plan VICW Major Features • Maple Reactor, Open Tank in Pool, Vertical Core • Two Vertical Core Segments • LEU U3Si2 in AL Fuel, AI Clad • Light Water Coolant, Upward Flow • Heavy Water Reflector • Hafnium Metal Control Absorbers • Shutdown System 1 -Absorber Drop • Shutdown system 2 -Reflector Dump • Reactor Building ~; Update on IRF Concept • Revise IRF Complex layout and concepts • Reactor based on a MAPLE-type reactor assembly • Design experimental facilities for CANDU support • Design experimental facilities for beam research programs • Revise cost and schedule estimates - reference cost estimate of $500 million - reference schedule of 87 months - risk of schedule delays of 12-24 months Validation Methodology Steps 3, 4 & 5 are code dependent • Validation Plan - Demonstrate that code version· accurately represents governing phenomena • Validation Exercises - Compare model predictions with selected data sets • Validation Manual - Summarize. code accuracy, sensitivities, and uncertainties for specific applications #J;/ Validation of Computer Codes Used in Safety Analysis for IRF • Adopted validation-matrix methodology • Based validation framework on NEAlOECD work for LWR community andCANDU industry work • Three major disciplines into which physical phenomena can be grouped -physics - thermal hydrauliCS - safety-related analyses (e.g., containment behaviour, fission product release & transport) • Five major steps in validation A; Validation Methodology Steps 1 & 2 are code independent • Technical Basis Document - Relate safety concerns, behaviours of plant subsystems and main phenomena during specific accidents • Validation Matrices - Relate all relevant phenomena to accidents and data sets 0::: o o ---.J LLO o f- .. (J)---.J o:::W LL w~ ~~ ......... ......... ......... ......... ......... ......... 0::::: o o --.l LL o Z o U w (f) alL alL 8lL I II cr: o o --1 LL o cr: I I--- , , ... ~ j ( / j ......... , / ......... / ......... ......... / ......... J I ,~ r-----'- rr--ll « I « z o I- U W (f) I Preliminary Study of Core Characteristics for TRR-II . J. T. Yang, L. S. Kao H. M. Hsieh, D. Y. Yang, J. A. Jing, and S. K. Chen Institute of Nuclear Energy Research . P. O. Box 3-3 Lung-Tan, Taiwan, R. O. C. Fax 886-3-4711404 TRR-II is the Taiwan Research Reactor Remodeling Project. The proposed multipurpose reactor is a water-moderated and cooled pool type of 10 MW to 20 MW design. Enough irradiation space will be provided for radioisotope production, silicon transmutation doping, material and fuel tests, etc. Neutronic and thermal-hydraulic performances were studied for a core using typical PWR D02 rod fuel configurations. The core is surrounded by a concentric cylindrical aluminum heavy water tank as the reflector. A large fraction of the fission neutrons will leak out the core surface and slow down in the reflector to form a high flux peak with very pure spectrum of thermal neutrons. The coolant is designed to be down . flow. Calculation results show that the maximum unperturbed thermal flux in heavy water reflector is about 1 X 1014n/cm2 • s. The design concept of the core is that it has great flexibility of using alternative fuel designs for core (i.e., neutron flux) upgrade. 1. Introduction T-he old Taiwan Research Reactor (TRR) was a 40 MW(th), heavy water moderated and cooled, graphite reflected r~search reactor. It was permanently shutdown in 1988 and will be remodeled, designed and constructed domestically as a mUlti-purpose research reactor to meet national demand in the fields of radioisotope production, basic research in neutron applications, and industrial and medical uses. To satisfy the above utilization goals, the design of Taiwan Research Reactor Remodeling (TRR-ll) has to meet certain performance requirements as well as economic aspect. The design principles are as follows. a. The maximum available thermal neutron flux in the flux trap should be at least 1 X 1014 n/cm2.sec b. Low enriched uranium fuel is chosen by taking into consideration the easiness of obtaining. c. The aluminum alloy is chosen for the core components to reduce neutron absorption and radiation dose under the maintenance work. d. A conventional swimming pool design is chosen and control rods are driven by the control rod driving mechanism installed below the core for easy access to the core. e. The coolant flow is planned to be downward for the purpose of reduction of the radiation dose cause of 16N which is produced by the (n,p) reaction in 160, and a 16N decay tank is installed in the primary cooling system. The main goal of this report is to introduce the current progress in the preliminary study of core characteristics for TRR-II. The brief description of reactor is given in section 2. Section 3 dipicts the calculation models regarding to nuclear design, thermal-hydraulic design and accident analysis. Section 4 summarizes the nuclear and thermal-hydraulic characteristics of TRR-II. The conclusions in this study is given in section 6. 2. Reactor Description 2.1 Fuel One of the unique feature ofTRR-ll is to use domestic made D02 rod. It is very well known for their extensive use in L WR. Their physical properties, in pile behavior, manufacturing features etc, are well mastered. The high density 10.3g/cm3 ofU02 matrix leads to a higher uranium content, making it possible to reduce the enrichment about 3 to 10% 0 The fuel assembly of TRR-II consists of 6 X 6 U02 fuel rods. The diameter of the fuel rod is 9.5mm with zircloy as cladding. 2.2 Core The core is cylindrical in shape, about 0.6m in diameter and about 0.6m in height. It is composed of fuel assemblies, control rods, irradiation elements and beryllium reflectors. Its average power density is 110 kwll at the rated power operation. A configuration of the core is shown in Figure 1. The control rods are made of hafuium, and connected to the aluminum follower elements. They are driven by the control rod driving mechanism installed beneath the core. The irradiation element is similar to a fuel assembly in its outer dimension and has an irradiation hole of 60mm diameter at the center. There are five elements in the core. 1111 Be block D fuel assembly [J cantrol assembly • test channel •.0 20 reIIectcr Figure I. TRR-II core configuration · Beryllium reflectors are installed between the fuel region and the inner wall 2.4 Core 6. TCC) 5 TinCC) 40 Inlet pressure{psia) 23.0 5. Conclusion The design of TRR-II is still under its feasibility study stage. The inverse neutron trap and the surrounding D20 reflector were employed as core concept in order to obtain high pure thermal neutron flux and to secure enough irradiation space. The design flexibility of using the advanced, silicide uranium dispersion fuel were to be studied, so as to improve the reactor performance in the future. 6. References (1) M. A. Edenius, "CASMO-3 Benchmarking", Trans. Am. Nucl. Soc., .56 (1988) (2) D. R. Vondy, "VENTURE-A code block for solving mUlti-group neutronics problems by applying the fmite difference diffusion theory approximation to neutron transport", ORNL-5602 (1975) (3) J. F. Briesmeister, "MCNP-A general Monte Carlo code for neutron and photon transport." LA-7396-M, Rev. 2 (1986) (4) Jason Chao, "COBRAIIICIRERTR A Thermal-Hydraulic Sub channel Code with Low Pressure Capability", CP-0326, Argonne National Laboratory, Dec. (1980). (5) K. E. Carlson, etal, "RELAP5fM:OD3 Code Manual," NUREG/CR- 5535, (1990) Construction of the HTIR and its Irradiation Program Masahiro ISHIHARA·I, Takayuki KIKUCHI·1, Jun AIHARA*t, Haruyoshi MOGtl , Taketoshi ARAI·2 and Toshiyuki TANAKA·1 Department of HTIR Project·1 Japan Atomic Energy Research Institute Oarai-machi,Higashiibaraki-gun,Ibaraki-ken,Japan Depertment of Advanced Nuclear Technology·2 Japan Atomic Energy Reaearch Institute Tokai-mura,Naka-gun,lbaraki-ken,Japan Abstract The HTTR (High-Temperature engineering Test Reactor) is a high- temperature gas-cooled reactor with a maximum helium coolant temperature of 950°C at the reactor outlet. The construction of the HTTR started in March 1991 aiming at the first criticality in 1997 after the commissioning test. The main components in the HTTR, such as a reactor pressure vessel, an intermediate heat exchanger, hot gas piping and graphite core structures, had been installed in the reactor building. At present the functional test of the HTTR has been carried out, and fuel elements are being fabricated at works with their loading targeted in 1997. The HTTR project is intended to establish and upgrade the technology basis necessary for HTGR developments. At the same time materials and fuel irradiation tests are also planned for an innovative basic research after safety demonstration tests in the HTTR. Preliminary tests on selected research subjects which should be performed as irradiation tests, such as new development of semiconductors, superconductors and optical fibers, have been carried out at high temperatures with/without irradiation. This paper describes present status of the construction, major features of its irradiation condition and prospects on irradiation test programs using the HTTR. 1. Introduction It is essentially important in Japan, which has limited amount of natural resources, to make efforts to obtain more reliable and safety energy supply by extended use of nuclear powers including high temperature heat from nuclear reactors. Hence efforts are to be continuously devoted to establish and upgrade HTGR technologies and to make much of human resources accumulated so far. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. The HTIR has been, therefore, so designed as to be an engineering test reactor which aims to establish and upgrade the technological basis necessary for advanced HTGR develo~ment, and to conduct various irradiation tests for innovative basic researches{l . . The reactor building, which has two stories aboveground and three stories underground, is a size of about 48mX50m in the plane. A reactor pressure vessel, an intermediate heat exchanger and other heat exchangers in the cooling system are installed in the containment vessel, as illustrated in Fig.!. The HTTR is a helium-gas-cooled and graphite-moderated reactor with 30MW thermal power and outlet coolant temperature of 950°C for high temperature test operation. The major specification of the HTTR is shown in Table 1. The reactor pressure vessel made of 2 1/4 Cr-lMo steel, which contains core and core support structures, is 13.2m in height and 5.5m in diameter. The core structure is composed of fuel blocks, control rod guide blocks, and replaceable reflector blocks, and the core support structure is composed of large graphite components, carbon components and metallic components as illustrated in Fig. 2. The reactor cooling system is composed of a main cooling system (MCS), an auxiliary cooling system (ACS) and two reactor vessel cooling system (YCS ) as shown in Fig. 3. The ACS is in the stand-by condition during the normal reactor operation and is operated to remove the residual heat from the core when the reactor is scrammed. The VCSs are operated during the normal operation in order to cool biological shielding concretes. It is also used as a residual heat removal system when the forced circulation in the primary cooling circuit is no longer available. The construction of the HTTR is proceeding in the Oarai Research Establishment, JAERI. The first criticality will be attained in the end of 1997, and the reactor performance test will follow to achieve its full power operation in FY1998. Various kinds of operation and irradiation tests are also planed, such as experimental researches for the development of advanced fuels and materials and demonstration of the nuclear process heat utilization. The operation and test plan in the HTTR is shown in Table 2. 2. Construction of the HTTR The installation permit of the HTTR was issued from the Government in November 1990. The design and construction methods of the HTTR was approved from the Science and Technology Agency in January 1991 and the construction work of the HTTR was started in March 1991. The excavation of ground was completed in August 1991 and the construction of concrete base- matt was also completed in May 1992. The reactor containment vessel was installed and the pressure-proof and leak tests were successfully conducted in November 1992. The reactor pressure vessel, the intermediate heat exchanger, primary helium circulators and the primary pressurized water cooler were installed in the reactor containment vessel in 1994. The first pressure-proof test for the primary cooling system was carried out successfully in October 1995. The reactor building was completed in December 1995 by closing the temporary opening for the large components. The JAERI obtained the uranium material for the first loading fuel in September 1994 and the fuel fabrication started in Jun 1995. The first criticality will be attained in the end of 1997. The HTTR construction schedule is shown in Table 3. 3. Irradiation Test Program of the HTTR 3.1 Irradiation Characteristics of the RITR The HTTR has the capability not only to conduct the R&D on the advanced technologies for the HTGRs but also to irradiate large sized samples at a high temperature with a uniform neutron flux. The core of the HTTR consists of prismatic fuel elements of hexagonal blocks, with 580mm in height and 360mm in across flats, and therefore the maximum size of an irradiation sample is 500mm in height and 300mm in diameter. The samples can be irradiated at high temperatures ranging from 400aC to HOOaC at maximum. Various irradiation tests are possible to use five irradiation regions as shown in Fig. 4. The irradiation conditions for these regions are summarized in Table 4. Irradiation tests at a high temperature (400 to HOODC) and high neutron flux (about 2x1017n/m2'S (E>0.18 MeV)) can be performed in the center column region. Central four columns, the center column region and the test fuel loading region, are permitted to load fuel elements. Irradiation tests at a lower temperature (400 to 6000 ) and lower neutron flux (about 8xlO14 n/m2'S (E>0.18MeV)) can be performed in the permanent reflector region. 3.2 Preliminary Design of Irradiation Facility A Preliminary design of a new irradiation capsule (Rotation drive type capsule,RDC) has been carried out for use in the material irradiation test at high temperature with neutron fluence controlled uniformly. The RDC capsule is to - Development of in-core neutron and gamma ray detectors in service at high temperature up to 950°C. - Development of heat-and radiation-resistant optical fiber for an instrumentation of high temperature reactors. (5) Other researches - Study on radiation shielding - Development of fuels for an advanced reactor Some research themes are of great significance in view of irradiation test candidates for the IITTR but need to be subjected to preliminary tests (out-of pile or in-pile) to assess the scientific effectiveness. Preliminary out-of pile experiments are being undertaken currently for the following themes. - Neutron transmutation doping of low atomic number high temperature semiconductor - Improvement of high temperature oxide superconductors by irradiation - Pecuriality chemical reaction at high temperature irradiation condition - Development of heat- and radiation-resistant optical fiber - Development of in-core neutron and gamma ray detectors for use at very high temperature Here, typical preliminary test result is shown in Fig.6(2). This figure shows a test result for optical fiber. Two kinds of Si02-based optical fibers, the high purified and OH doped fibers, were tested at 800DC in an air condition; measuring wavelength was 850nm. Both fibers show stable optical intensity during about 300 hours heat treatment, furthermore the high purified optical fiber shows a lower optical intensity. From this preliminary test high purified optical fiber was selected as a good heat-resistant one. At present an atmosphere effect, e.g. contained H2, O2, etc. elements effect, on the optical intensity is investigating at high temperature using this optical fiber. 4. Concluding Remarks The HTTR is a high temperature gas cooled reactor which has various aims and operational modes. The construction of the HTTR has progressed on schedule and its first criticality is foreseen at the end of 1997. The various tests planned in the HTTR will make a great contribution to confirm salient characteristics of HTGRs and reliable supply of heat as high as 950°C, Moreover the IITTR has a unique and superior capability for carrying out high temperature irradiation tests for innovative basic researches. be installed in the replaceable reflector region A in Fig.4. This capsule is schematically shown in Fig. 5. The ROC is composed of capsule units, drive gear, rotation shaft and so on. Irradiation test specimens are inserted into the capsule unit made of graphite. The maximum specimen size is about 100 mm in height and 80 mm in diameter. The capsule unit is moved to certain longitudinal position in the core by the drive gear, and then rotated at the same position by the rotation shaft until a specified neutron fluence. Capsule units are turned at both ends of the ROC by the drive gear. Two types of capsule units are considered in the preliminary design; the one is an irradiation capsule unit in which the irradiation specimen is contained, the other is a dummy capsule unit made of graphite for the purpose of a neutron shielding. Both temperature and neutron fluence are measured nearby irradiation position in the capsule. 3.3 Preliminary Study on Innovative Basic Research Sixty eight themes have been proposed as the innovative basic research at a high temperature under neutron irradiation conditions until March 1994; 34 themes for new material development, 5 themes for radiation chemistry research, 8 themes for nuclear fusion research, 8 themes for high temperature irradiation technique development and 13 themes for other researches. Typical research themes proposed are as follows: (1) New material development The proposals involve of development of the new material by a high temperature irradiation and explication of the material behavior under a high temperature irradiation condition. As for the new material development, the following research themes were proposed. - Development of high temperature semiconductors - Improvement of high temperature superconductors (2) Radiation chemistry research - Research for the decomposition reaction of a high polymer - Research for the peculiar reaction of radiation chemistry - Research for radiation assisted synthesis of fullerenes containing actinoids (3) Nuclear fusion research The proposals in nuclear fusion research consist of irradiation tests of several types of blanket components for high temperature application and development of the first wall materials for a prototype/commercial fusion reactor. (4) High temperature irradiation technique development REFERENCE (1) S.Saito et al.,Design of the High Temperature Engineering Test Reactor(HTTR),JAEI 1332(1994). (2) T.Shikama,private communication. Tab,le 1 Major specification of the HTTR. Thermal power 30MW Outlet coolant temperature 850/950°C Inlet coolant temperature 395°C Fuel Low-enriched U02 Fuel element type Prismatic block Direction of coolant-flow Downward. flow Pressure vessel Steel Number of main cooling loop 1 Intermediate heat exchanger. Heat removal Pressurized water cooler Primary coolant pressure 4MPa Containment type Steel containment Plant lifetime 20years Table 2 HTTR operation and test plan. I~tem r 1997 1998 1999 2000 2001 2002 2003 T A • A 1. Establishment criticality V 1\ Initial core Advanced core of HTGR 'I v • technologies VReactor/ 30MW /850°C--950°C J performance tests /aOMW /950°C 2. Upgrading of / Irradiation tests HTGR J 1 technologies I I Safety demonstrati11 Construction and demonstration tests / tests of nuclear heat application system 3. Innovative /~ Development of very high temperature and basic heat resistant materials technology • Tritium production and recovery tests, etc. _1 Table 3 HTTR construction schedule. Fiscal Year 1990 1991 1992 1993 1994 1995 1996 1997 1998 Fueling MILESTONE 'V \IV Construe ion start Critl ::ality Approval of design and lsl 2nd construction methods 0 0 t p~ ~ P Site renovation I Excavation for reactor building I J Reactor building I I Reactor containment vessel ~ Cooling system Reactor pressure vessel and core internal I I Fuel I Criticality and power up test L - - - - - L- ___ --- Table 4 Irradiation condition in the HTrR. Irradiation position Fuel regions @Replaceable @)Replaceable @Permanent 0.18Mev) (n/m2 • s) Themal neutron flux -7X 1017 -5X 1017 -4X 1017 -4XIOI7 -3X 1017 «2.38eV) (n/m2 • s) I Intensity of gamma ray -1 x 102 -1 X 102 -3X 101 -3X101 -3 (Gy/s) Temperature rC) 400 to 1100 400 to 1100 400 to 800 400 to 800 400 to 600 Temperature instrument, Neutron flux instrument, Neutron fluence instrument, In-core instrumentation Fuel failure detection system L..... Refueling machine Spent fuel storage pool Intermediate heat exchanger Pressurized water Reactor cooler contaip.ment vessel Fig. 1 Bird's-eye view of the HTIR reactor building. Stand pipe Permanent reflector Replaceable reflector Core restraint mechanism t--- Fuel element Hot plenum block ~l l~I~~~ll= Core support post Lower plenum block ~l..\.tHRFil~ Carbon block 1~:::I{;i'-'~~- Bottom block Core support plate Core support grid Auxiliary coolant outlet pipe Main outlet coolant pipe Fig.2 Bird's-eye view of the pressure vessel and the core of the HTTR. IHX : Intermediate heat exchanger PPWC: Primary pressurized Reactor containment vessel SPWC water cooler Vessel cooling panel PGC : Primary gas circulator == SPWC: Seconda,ry pressurized ~) I water cooler \.! SGC SGC : Secondary gas circulator AHX : Auxiliary heat exchanger AGC : Auxiliary gas circulator Auxiliary IHX water AHX Pressurized air cooler water pump PPWC J L, Pressurized water = = air cooler Reactor I Auxiliary ~ water ~ pump Auxiliary cooling system Main cooling system Fig.3 Cooling system in the HTTR. (DCenter column region @Replaceable reflector (1 column) region A (3 columns) ®Test fuel loading region (3 columns) @Replaceable reflector @The permanent reflector region B (12 columns) region (4 holes) Fig.4 Irradiation location in the HTTR. Reactor pressure l:Capsule unit vessel 2:Drive gear 3:Rotation shaft Irradiation capsule unit Dummy capsule unit Fig.5 Rotation drive type capsule in the HTTR. 5 ! lr-----~----~----~----~~ : • High purified optical fiber ~ ! 4 ._ ....... _ .... ;. .. _. ....... _._ ... ; a Of-! doped optical fiber I I I I . . . .. 3 · .. ····. . _·····.,._···_. . _-· .. ·• .. t···. . _···· .. · .. · ...... · ...... "1' .. ••••••••. . •• ...... ••• ..t .... • ....• •. . ••••. . _ .. • .... ··t .. ••·•·••• .. ·••. ... ·····t._-_ ............. . 2 ~ ~ i ! ~ : : : : : ! t ~ ~ i 1 ....................... ;. ....................... i. .................... ~. ........................ i. .................. -.; ........................... i. .................... . I I I ! I I ~ ~ ~ f i 1 . ....... . .: !. !. • o 50 100 150 200 250 300 350 Time (hour) Fig. 6 Preliminary test result for the optical fiber. (This figure shows a optical intensity change at 800De heat treatment condition; measuring wavelength 850 nm) SIEMENS November 8, 1996 KWU NLFM I Didier 5th meeting of the International Group on Research Reactors (IGORR V) November 4-5-6,1996, Aix-En-Provence - France Agenda 1. Introduction 2. Status of the Project Barrier remoyal Installations PrQject management. tools Action against the first partial license 3. Outlook 4. Technical Highlights Research Reactor Munich, FRM-II , Projectstatus November 1996 SIEMENS November 8, 1996 KWU NLFM I Didier 5th meeting of the International Group on Research Reactors (IGORR V) November 4-5-6, 1996, Aix-En-Provence - France 1. Introduction Since our last IGORR-IV meeting in Gathingburg (May 1995) the "high flux neutron source" FRM-II has developed rapidly. As we have reported at our last meeting this project was still in the design and licensing phase. It had to overcome some important barriers before start of civil construction. Those barriers have been removed I On August 1s t, 1996, the first sod procedure has taken place under the attendance of representatives from • the German Federal Republic (BMBF) • the local municipalities Fig. 1 I • Science and Industry • Licensy authorities and experts Fig. 2 I • and finally the Bavarian State Government headed by the Prime Minister Dr. Edmund Stoiber Research Reactor Munich, FRM-II, Projectstatus November 1996 -2- 2. Status of the Project - Barrier remoyal • Project funding is settled, 720 Mio. OM for the whole project incl. 15 % tax • First partial license has been issued April 9th, 1996 • Environmental impact examination has been concluded successfully • Use of fuel with high enriched uranium (HEU) was accepted by the licensing authority • General suppliers contract, TUM-Siemens, has become effective April 9th, 1996 Fig. 3 1 • Site preparation is concluded. including protection measures to Fig. 4 1 control the permission of staff to enter on the site • Foundation of reactor building in the excavation (8 m deep and almost 2000 m2 wide) - Installations • The metallic cladding (stainless-steel) of the reactor pool is ordered, the supplier is preparing to manifacture this liner in such a way that it can be used as nlost form" in spring 1997 • Two dummy fuel elements have been manufactured (one aluminium, one depleted uranium) which are being zested in a hydraulic loop at the University Bochum Research Reactor Munich, FRM-II J Projectstatus November 1996 -3- - PrQject management. tools Fig. 5 I • TUM projectgroup is revalorized to a central organisation reporting directly to the president • The Siemens FRM-II project-team has been settled and set up for operation • Electronic management tools have been implemented for: - time scheduling and follow up - cost planning and follow up - data management system - 3 dimensional computer aided design system linked with a logistic system for procurement and installation - reporting • Implementation of a professional public relation group exclusively operating for this project. - Action against the first partial license • 1. action for the restriction of suspensive effect • 2. action against substantial matters of licenses • Main complaints: - liability of fuel element - radiation exposure in case of a beyond design accident to high - provisions to avoid a spontaneous rupture of primary piping • 8 actions by communities and individuals • Withdrawal of action by 4 parties because of low hope to succeed Research Reactor Munich, FRM-II , Projectstatus November 1996 -4- 3. Outlook Fig. 6 I Timetable 4. Technical Highlights F•I g. 7 • 8 I • Moderator tank • Overall building sections Research Reactor Munich, FRM-II , Projectstatus November 1996 VI Z w ~ -w VI ... .... .-..... ' .' .. :.- \ ,.; ..~ . ':( , ::.,""l ~'~-:~"'~'p" ' ;.t:~" r ., ' 'i'':: '$~~, .~ .:::::- .. ,, ~;;~:~~, ~t:~ . ' \ ,) -~ ~.t ,.,··} -"' 'I'' :. ~,~.:, ~ t~ . ;" ~?!> .. ., , -,: , , ', ,)of: , ,,-tC :;, .. : .. ~~ ~ ., .. ,.,.!'. ~.,.~~. i"" . "': r-, :'\: :" ~ . ' ~. • J . .c .. .,~ .. 'f ..: ' ~ . ....... " .' 7- .,( , .,., - .... ~.\;.", . ,~'4} " ~,,': _. J'. ~ ~ . ,.. ..ir:-~ ""t. 1• • 1:.' . ~o,t . -1 .. , ;# '. ~ 01 " " ~·"ill ".,, ' / ' t ~ .~. ~, , ' ~.. . ..". ~ ~ ;~. >:~:-: ~-.> ~ ' >< . I: t.·-·F, , ; , "• .t r T~_~_ ~~ " - ,-. \. ~ '." . ' ,' I• • .), # ' ,. - l ,,' " t ~"jb J ~:~. J :~1f , j":"~ SIEMENS Organisational Chart of FRM-II-Project President Technical University Munich (TUM) I Board of Directors J Overall Project Management TUM Research Experimental Administration and Public Re!atlon Reactor Facilities Control Group Siemens TUM TUM TUM I I I Project Managementll Suppliers Lurgi 1\ I Commercia l A ffairs I Licensing and Safety Ir ...T,e.c.h-nic-s_ . .....I. Neutron PhySiCS; I Thermohydraulics Process Enginr.ering J Civil Construdion I Mechanical H Construdion I Power H Supply I Instrumenlallon J and Control Physica l Proledion I Radialian Protedlon I Materials J Quailly Assurance I FRM-II KWU NLFMlDidier 30.10.1996 SIEMENS FRM -II Neutron Source Munich - Overall time Schedule 19 13 1994 19" lB 1~ 17 199B 19119 20011 lQ 11 IN. 1 2 314 1 I 2 I ;, I 4 1 2 I 3 4 1 2 "; ) 1 _2 . 3 " 1 2 ;) 4 1 iii . ~ _4 , 2 3 4 1 Z 3 " 1 Concept, Licencing documentation " for 1st Partial licence ~ - 1-- Fulfillment of Authority comments - - 1- - 1st partial licence, release for ., , - f- -- -- - -- ~ reactor building erection Order of pool liner, preparation of erection area & reactor building ~ l- .. - . . - Detailed design phase, licencing documentation 2nd partial licence ~ Erection of reactor building .............. 2nd partial licence for complete , equipment & building ~ I-- Ordering & manufacurlng of complete equipment ~ Installation of systems & equipment 1- - -I--1- ~ Cold commissioning & acceptance phase - . I . ~ Licencing documentation I for 3d partial licence - -- ........ -- 3d pattlaillcence for hot ., , commissioning & operation . ._ - - .-. Hot c:ommlsslonlng phase t ~_... • . 10. Neutronenquelle FRM II The R.J.H. reactor THE JULES HOROWITZ REACTOR (R.J.H.) THE C.E.A. FUTUR.E TOOL FOR TECHNOLOGICAL IRRADIATIONS The R.J.H. reactor o GENERAL BACKGROUND • EXISTING EXPERIMENTAL REACTORS DEDICATED TO TECHNOLOGICAL IRRADIATIONS • SILOE • OSIRIS • European Reactors All these reactors were designed and constructed in the 1960's o NEED FOR A NEW TOOL DEDICATED TO TECHNOLOGICAL IRRADIATIONS IN THE NEXT CENTURY The R.J.H. reactor • GENERAL BACKGROUND Fhat tIL (E14 ."".2..1) 1U:iiQ:m,J ~~ t~oc1 40 [ I I 35 [ ==;-=" I T ~ r[---------r--------1-~ 2S tt I IBI~R-,!'-",'M j II~~~JI-----l • 10 rl- ---_+_ 1000 1600 2IlCXI 2SOO Pv(kWn......, o ~ dMiII" III reclum:luJlO1/tIanIenttik _ dmnai1l8 d'61ud8 The R.J.H. reactor o GENERAL OBJECTIVES • FUEL QUALIFICATION > THERMAL NEUTRON REACTORS > FAST NEUTRON REACTORS > POWER TRANSIENTS > OTHERS • NUCLEAR MATERIAL QUALIFICATION • Radio-isotope production The R.J.H. reactor o TECHNICAL REQUIREMENTS • HIGH FLUX > 5.E14 cm-2.s-1 in the thermal range «0.625 eV) > 1.E15 cm-2.s-1 in the fast range (>1 MeV) • WIDE USE SPECTRUM (various, evolutive and changing needs) • ADEQUATE EXPERIMENTAL VOLUME • EXPLOITATION FLEXIBILITY (accessibility,adequate cycle length ... ) ... Cl) E E :::s U) c ... .-.c .C-) 0 .c: ~ •• (.) WCl) ca J:'" Cl) O.d ~ til ~c"u'C-l) • I- J: -z .-. «Cl» ...,• :w:J «C E.!!! ~ Cl).~ • I- ...J ol-E "'-' 0:: til ....... U) z ...J W Cl)a-W Cl) 0 ::J _I -~Clt)il .J: 0 LL til I- 0 • • The R.J.H. reactor o SAFETY OBJECTIVES • Risks for environment and peoples will be as low as possible during normal operating conditions and accidents • Radiation doses will be as low as possible for reactor workers > CIPR 60 > ALARA principle The R.J.H. reactor o EXPERIMENTAL OEVICES • E1 = "capsules" (0=30 to 80 mm) in core • E2 = "small loop" (0=80 mm) in reflector • E3 = "loop in displacement box" (0=80 mm) in reflector • E4 = "large loop" (0=150 mm) in reflector • E5 = "central loop" (0=80 to 200 mm) in core center The R.J.H. reactor o OPERATING CONDITIONS • "S" MODE =W ithout penetrating central loop (Nominal Experimental Charge) • "B" MODE = With penetrating central loop (FBR irradiations - PWR irradiations and power transients) • "M" MODE =M ock up conditions (pre-irradiation neutron measurements) ~ Co!) ~ -< ~ U i-J -< f-c L. ,0.., Z ~ N ~ (.) ~ 0 ca ~ l""'f ~ Cl) ~ + L. ~ l""'f ~ • ~ 0 .J..:, i-J l""'f • -< • Z I:t: ~ ~ CI) 0 .c z F ..•.. c o :p jg U) c Q) Q) ~ E .c ::::s ::::s ::::s o J:; I- ::::s - L... UJ +oJ C C\1 U) .- c ~ o E o .::-::s o .c- · Q) ::::s E ::::s ... U. « ~ ,o.., m o :E ca w en ~ ~ • c .J.:,• ~ Fig.S. Neutron spectrum incide heavy water reflector at place· of cold neutron source ( p=ioo w ) \ . elF/dE (n / c:m2 s eV) . 1E+11 ~---------'---------r---------r------~ 1E+10-d------4--4--------t-r---------r-------~ 1E+9 --+-------+-------t---I------4--------I 1E+8~--------+-------~--------~r_------~ 1E-4 1E-3 1E-2 1E-1 1E+O· E ( eV ) Fig.4. Neutron spectrum incide cold neutron source (P=100 W) · . Fig.5. Thermal neutron flux distribution, 109 nlcm2s , P=lOO\>l) elF/dE (n / cm2 s eV) 1E+8~~--------.---------~---------r--------~ 1E+7 --t--- 1E+6~----------r---------4-~------~------~~ 1E-4 1E-3 1E-2 1E-1 1E+o E ( eV ) Fig.5. Neutron spectrum at exit of DEC-3 (95 cm from channel bottom,P=100 W) 7 .~. , '\, J~ , , .".€~ ,Em 2,t;7m Fig.7. Neutron time- flight installation 1- eNS 5- Chopper 2- Experimental channel 6- Motor 3- Neutron collimator 7- ~e-detectors 4- Chancheable filter (Pb or Be) 8- Neutron shielding , . Low temperature . moderator circuit I- VacUlDD Cooling unit and Nitrogen system cryogenic pipes !- - - supply system Deuterium preparation I- system I Helium preparation system Fig.8. Diagram of cryogenic supply Paper to be presented by K. COBRECHT at the ICORR V Conference Nov. 4 - 6. 1996 In A1s-en-Provence : THE COLD NEUTRON SOURCE AND OTHER IN-PILE EXPERIMENTAL FACILfI1ES OF THE NEW RESEARCH REACTOR FRM-II IN GARCHING Klaus GOBRECHT and Erich STEICHELE TU-MOnchen, FRM-ll, Reaktorstation, D-85747 Garching Abstract : The new high flux research reactor of the Technical University Munich will have 13 beam tubes and several in-pile irradiation facilities : 5 thermal beam tubes. one for medical appli- cations (converter for fast neutrons), one for a fission product accelerator, one for a positron source, one viewing the hot neutron source (HNS) and 4 on the cold neutron source (CNS), including one for very cold and ultra-cold neutrons. The cold neutron beam-tubes will house all together 10 neutron guides, 6 of which will feed about 12 high performance neutron spectrometers in a modem neutron guide haIl. Among other irradiation facilities a silicon doping loop and two rabbits will be installed. We introduce briefly these facilities and report in more detail on the CNS. The highest possible flux of low energy neutrons will be obtained by placing the CNS very close to the reactor core in the thermal neutron flux maximum. This can only be achieved if the nearly 5000 W developed by the nuclear heating in the secondary moderator at 25 K are evacuated by liquid deuterium in a thermal siphon. A 1:1 model of the thermal siphon is studied around room temperature with freon. The CNS will go into operation in the year 2000. The basic design, as well as heat load and cold neutron flux calculations will be presented. page 2 1. Introduction A modem research reactor is a source of neutrons for both neutron scattering experiments in condensed matter research and for production of radioactive materials for many practical and medical applications. For most efficient use the thermal neutron field in the moderator around the fuel element has to be either modified by "spectral shifters" or used by special irradiation devices. For use of the neutrons outside the reactor pool beam tubes and neutron guides have to be installed in an optimum geometrical arrangements with respect to a minimum effect on the core reactivity. All these devices are installed in the immediate neighbourhood of the fuel element and are therefore called ,,in-pile-facilities". The FRM-ll (Forschungsreaktor Milnchen-ll) near Garching is a 20 MW reactor with H20 cooling, D20 moderation and a relatively small fuel element with highly enriched uranium (BEU 93 % enrichment). With such an optimised design an "undisturbed" thermal neutron flux of 8 x 1014 nlcm2s, comparable to the characteristics of the best research reactors in the world, can be obtained with low power and low amount of radioactive waste. The FRM-ll should become operational in the year 2002 and will then be the strongest neutron source in Germany and one of the most attractive in Europe. The FRM-II will be equipped with a cold neutron source (CNS) 111, a hot neutron source (ENS) 111, a "converter" /21 for fast fission neutrons, 10 horizontal, two inclined and one verti- cal beam tubes and a series of special irradiation devices, which will be presented in the following chapters. The fact of a high flux-to-power-ratio allows us to locate the cold neutron source quite near the thermal flux maximum. We therefrom expect intensities in the cold neutron beams comparable to those of the high-flux reactor at the ILL in Grenoble (typically 5 x 109 cm-2s"1) 14/ and we therefore want to present the design of the large cold neutron source in more detail in the present paper. 2. Experimental Devices in the Moderator Tank The experimental devices in the moderator tank (2.5 m diameter and 2.5 m high) are organised in a way, that beam tubes are mostly in the horizontal plain and that secondary sources (CNS, HNS, converter) and irradiation devices are vertical with operation from the upper reactor floor. A horizontal cut through the reactor pool is shown in Fig.l. Beam tubes #1, #2 and #4 are looking onto the cold source. In front of beam tubes #5 and #7 an optional vertical cold source can be installed. The big beam tube #5 is designed for future installation of a super- thermal source for ultra-cold neutrons, where voluminous shielding of a liquid helium vessel against radiation from the core have to be installed in the beam tube. Beam tube #9 looks onto the hot source. Beam tube #6 is a double-ended hole, dedicated to experiments with charged fission products. These are produced in a U-235 target (about 1 gram) in the middle of the tube and can be extracted and accelerated outside the reactor hall for heavy ion reactions. Beam tube # 1 will take up the principal system of guide tubes going into the neutron guide hall. Beam tube #10 is oriented to the converter, from which a beam of unmoderated fission neutrons can be extracted for medical tumour therapy and for non-destructive material testing by computer tomography with fast neutrons. The large diameter of all beam holes allows for the extraction of two beams with a maximum angle of 80 in between or the installation of additional guide tubes or a horizontal cold source. A vertical neutron guide will be installed into the D2 tube of the cold source and deliver very cold neutrons into an ultra cold neutron turbine. The two inclined beam tubes are in discussion to be used for neutron scattering, production of a positron beam and an ultra-fast pneumatic irradiation device. page 3 The hot source is a cylindrical graphite block 300 mm in diameter and 300 mm high. which is heated up to about 2200 °C by nuclear radiation from the core, thermally insulated by carbon felt and contained in a zircaloy vessel. The energy range of the neutrons from the HNS extends from about 50 meV to 1 eV. The thermal-te-fast neutron converter is a 2 mm thick U-235 plate of 142 mm x 220 mm in a thermal flux of about 1 x 1014 em-2s-1 providing a fast and intermediate (ener~ > 1 keV) neutron beam of 20 x 28 cor area outside the reactor with an intensity of2 x 10 em-2s-l • Several irradiation devices will be installed in the FRM-ll. For very homogeneous phosphor doping of semiconductor silicon two vertical tubes for maximum 4" and 8" diameter crystals are installed in a thermal neutron flux of about 1 x 1014 em-2s-1 • Three pneumatic and one hydraulic rabbit systems for irradiation times between seconds and weeks in thermal fluxes between 5 x 1012 cm-2s-1 and 4 x 1014 em-2s-1 of samples between 1 em' and 12 em' cover most of the demands of isotope production and activation analysis. 3. The Cold Neutron Source The FRM-ll will be equipped with two CNS, a vertical one right from the beginning, the other one later. and may be horizontal. The vertical CNS will work with boiling deuterium (D2) at 25 K and 150 kPa, or with a mixture of deuterium and hydrogen «10%) at the same pressure. The typical lay-out is shown in Fig.2. The moderator fluid serves itself as the heat transfer medium, taking the heat away from the moderator volume to a heat exchanger in a two-phase flow driven by natural convection. The cold moderator vessel is a 300 mm diam. cylinder with elliptical bottoms, made of aluminium. It contains - for neutron-optical reasons - a displacer, which fills itself with deuterium vapour and forms a re-entrant hole in the liquid. The liquid deuterium content is about 16 litres. The axis of the vessel is 400 mm away from the reactor core axis, and the closest point of the vessel lies only 110 mm away from the core. The heat load from nuclear radiation emanating from the core onto the vessel and from neutron activation is estimated to 5 kW (250 W per MW of reactor power), leading to a D2- evaporation rate of nearly 20 gls. More characteristic data of the planned CNS can be found in Tab. 1. In order to limit the heat load on the plumbing, a single, "heat-pipe"-Iike transfer tube links the moderator vessel to the heat-exchanger condenser 4 m above. The tube is inclined by about 10°. The condenser is cooled by a 6 kW Brayton cycle helium refrigerator with a 400 kW screw compressor. The cold box is equipped with heat exchangers for a second CNS of about 3 kW at 24 K and the plumbing for an extra twbo-expander. Together with the second CNS a second screw compressor will be installed in the compressor building. 4. The ffigh Temperature Model The single tube heat-pipe is a novelty in CNS design : the liquid deuterium flows downward along the wall of a tube from the condenser to the moderator cell, where it vaporises and streams upwards as vapour in the free section of the same tube. In order to optimise the flow conditions we built a 1: 1 scale model working above room temperature with freon RII, a fluid for which the relevant hydromechanical properties can be adjusted to closely match those of D2 (similarity criteria). The experiments take place at the same density ratio Pliquid/PYapour and at the same Wallis-number, which is the relevant scaling number for our case. It is known that the counter current flow in a pipe has its limitations (CCFL = counter-current-flow limitation), also called flooding in the literature 13/. In the CNS single tube heat pipe the flooding takes page 4 place above a critical heat load. which depends on the geometry of the CNS. Our experiments have so far not reached the CCFL, which means that the geometrical form of the thermal siphon is satisfactory. 5. The Gas Handling The deuterium is completely sealed in a double-walled stainless steel circuit. When it is not liquid, it will be stored as a metaI-deuteride in two tanks containing about 400 kg of a special hydride-forming alloy (e.g. LaC03Ni2). Then the residual pressure in the vessel and in the plumbing is <10 Pa. For the condensation of the D2 into the CNS the tanks have to be heated to desorb the gas from the metal-deuteride. In the neutron flux: deuterium becomes radioactive by neutron capture (deuterium + neutron => tritium, and tritium => 3He + 13-). After several years of operation the storage tanks will be used to ship the radioactive metal deuteride to a reprocessing plant with-out any risk of contamination. The out-of-pile parts of the 02 circuit are protected from 02-air explosion risk by an inert gas liner (nitrogen) pressurised to slightly above atmospheric pressure. 6. The Cold Neutron Flux Monte-Carlo-simulations have been made with MCNP (version 4A) with cross-sections for W2 and LH2 supplied by IKE Stuttgart (Bemnat et al. lSI). The 27A l cross-section has been modified to include the gamma rays produced in the decay of Al-28 after n-capture. The simulations give the spectral distribution of the neutron flux, the brilliance of the CNS, and the heat load on the CNS. The calculations show that the liquid 02 vessel, as designed, is too small to give the optimum moderation for wavelengths above 4 A. By adding a few percent of hydrogen to the 02 one can adapt the neutron mean free path to the small vessel dimensions (Ageron Ill). A neutron flux increase of30 % from 6 to 20 A is expected (Fig.4). 8. References 111 AgeroD, P. "Special Neutron Sources" in Symp. on Neutron Scattering in the Nineties IAEA-JUlich 1985 121 Wagner, F.M., et aI. in Nuclear Science and Engineering, 110, 32-37 (1992) 131 Boning, K. "FRM-II, Germany's new neutron source" in nuclear engineering dec. 1994 141 Gobrecht. K. "The ILL Cold Sources" in Proc. Int. Workshop on Cold Neutron Sources, Los Alamos 1990 (Conference Report LA-12146-C) 151 Bernnat. W. et al. "Evaluation and Validation ofN eutron Cross-Sections for Liquid Hydrogen and Deuterium ... " same Workshop as in 141 page 5 Tab.l FRM-ll Cold Neutron Source: essential characteristic data FRM-ll ILL vertical CNS (when different) units Nominal reactor power 20 57 MW Integral neutron flux in CNS H014 cm-2S_1 Diameter of the moderator vessel 300 360 mm Mean wall thickness 1,75 mm Volume of the moderator vessel 20 24 litre Mass of0 2 in the moderator vessel 2,1 3 kg Distance from core (axis to axis) 400 760 mm Temperature of the cold moderator 25 K Pressure in the cold moderator 150 kPa Pressure in the warm 02-system ....() 300 kPa Expected refrigeration power 5.5 kW Hydride forming time (for 95 % 02) 6 n.a. min Volume of the buffer 10 18 tn3 Number of tubes in the thermal siphon 1 3 Material of the in-pile vacuum thimble Zircaloy Mean wall thickness 4 6 mm Vertical beam tubes for VCNIUCN 1 Horizontal beam tubes 3 1 Horizontal cold guides or collimators in-pile 10 5 D.i -hydride storage O2 -buffer vertical beam lube He-refrigerato ~,I\ ~ I ~ 1 -cold ""","',,, fuel- . ~~. . beam tube experiments element (neutron guides) bol7ing 02 • I 02 -moderator vessel D2 0 Fig.! moderator tank Cold Neulron Source -schemolic- " heat exchanger Va1 SG-7 neutron- ..!Ifl---H- guide light woter H 0 ... 2 '"I -', ~ ~ ~l c B I J _.. _ .. .. 00 Q)< pageS Fig. 4 ... 450 i 400 I!! 350 3..:. 300 CD 1a. ... 250 (2 00 c: 150 .5 'S 100 i 50 0 0 1 2 3 4 5 6 7 8 9 10 Lambda (A) POST IRRADIATION EXAMINATION OF Z6 NCT 25 STAINLESS STEEL FROM SF2 COLD SOURCE CELL OF THE REACTOR ORPHEE M.MAZIERE Service d'Exploitation du Reacteur ORPHEE CEA-Sac/ay 91191 Gij-sur-Yvette Cedex, FRANCE ABSTRACT At the time of the design, by the end of the seventies, the ORPHEE reactor was a second generation source type, devoted only to neutron beam use for fundamental and applied research. For this purpose, it has been equipped with a set of up-to-date experimental such as hot and cold sources and cold neutron guides. In the original design of the reactor, the choice of liquid hydrogen for the moderator of cold sources has led two use two « flat» cells in order to illuminate the beam tubes looking at the cold sources called SF1 and SF2. The cells were made of stainless steel in order to offer the possibility of staying in the reactor, under neutrons irradiation, without cooling by the moderator. As the behavior of the stainless steel cells was bad known, a surveillance programm of the cells has been proposed in 1981. This programm foresaw an examination of the cells. These examinations have been made in 1987 (the results have been reported in a previous IGORR meeting) and 1991. The SF2 cell has been taken out of the reactor in july 1990 and the examinations have started in 1991. The post irradiation examination includes helium measurements, (on each face of the cell) and correlation with the neutron flux estimation. Tensile tests have been performed following a programm of 20 samples at various temperature. Comparaison with the result of SF1 cell has been made. -1- 1. INTRODUCTION The ORPHEE reactor is a medium flux neutron source for fundamental and applied research using neutron beams. In the design, special features of a modem reactor like the High Flux Reactor in Grenoble were used : pool type reactor block of compact size, tangential beam tubes, hod and cold sources, neutron guides[!), The reactor has 9 horizontal beam tubes tangential to the core carrying 20 neutron beams (two beams per standard channel and four for a double channel). The front ends of each tubes are in the reflector close to the core. Three tubes are directed at two « cold sources» called SFI and SF2 (see Figure 1). The moderator selected for the cold sources was hydrogen. The two identical sources are located in the reflector tank, 400 mm away from the core centerline. The active part of the sources is a flask shaped cell (205 mm high, 130 mm wide, 50 rom thick at center) made of 0.8 mm stainless steel. The cell hangs from a concentric supply line (with the liquid in the inner tube). The flask-shaped cell is located within an evacuated tube (150 rom O.D., 5 mm wall thickness) made of aluminium alloy. The cells were made of stainless steel in order to offer the possibility of staying in the reactor without cooling by the moderator. In such a case, the cell was filled with helium and the external tube of aluminium too. The design shows that the temperature of stainless steel grows up to 425°C. As the behavior of the stainless steel cells was bad known, specially with the risk of embrittlement by hydrogen and helium formation, a surveillance programm of the cells has been proposed in 1981 when the reactor started. This programm foresaw an examination of the cells when the quantity of helium due to (n,a) reactions would reach 1100 ppm. These examinations have been made in 1987 and the results have been reported previously. (The main results are recalled in the §2)[2). The results have permitted to extend the life of the second cold cel1 (SF2) up to 1500 ppm of helium in the stainless steel. The SF2 cell has been taken out of the reactor in july 1990 and the examination have started in 1991. -2- 2. RESULTS OF THE FIRST POST IRRADIATION EXAMTNA nON (1987) Three series of 5 tensile test samples have been cut up in the cell by electro erosion. Helium measurements have shown quantity of helium quietly different on each face of the cell : • 907 ppm for the front face (looking at the core) and • 543 ppm for the back face (looking at the beam tube). That difference has been explained by the disturbance in the flux of neutrons due to the beam tube SF looking at the back face of the cell and the normal gradient of the flux in the reactor tank. In the same range, the swelling ofalIoy was respectively 0,66 % and 0,41 %. For the tensile tests, the yield strenght (at 0,2 %) remains the same that the original alloy for the samples heat treated at 425°C during 10 days. Elongation decreases from 15 % to 9 %. 3. POST TRRADTA nON EXAMlNA nON OF THE SF2 CELL (1992) The stainless steel used for the flat cell was Z6 NCT 25 stainless steel from CREUSOT LOIRE. His theoretical composition is : • 25%Ni • 15 %Cr • 1,5 %Mo .2%Ti • - 56,S % iron. His mechanical properties are: • R> 915 MPa • Eo.2 > 600 MPa • A> 15 % As shown on Figure 1 the cold cell SF2 illuminates two beam tubes (4F and 9F) placed symmetrically on each side of the flat cell. -3 - 3.1 Post irradiation examination programm a) The post examination includes helium measurements (on each face of the cell). Tensile tests have been performed following a programm of 20 samples at various temperature: 12 samples have been tested after heat treatment at 425°C during 240 h and 8 samples without heat treatement. The tensile tests have been performed at - 170°C, 25°C, 425°C and 650°C. They have been realised with a constant speed of3.1004 S·l. The detailed programm is given below. ORPHEE COLD CELL SFl TENSILE TEST 4FB5 no - 170°C 3.1004/s 9FBI no - 170°C 3.1004/s 4FB4 no 25°C 3.1004/s 9FB2 no 25°C 3.1004/s 4FB3 no 425°C 3.1004/s 9FB3 no 425°C 3.10-4/s 4FB2 no 650°C 3.1004/s 9FB4 no 650°C 4FHl 425°C, 240 h -170°C 3.1004/s 9FHl 425°C, 240 h -170°C 3.1004/s 4FH2 425°C, 240 h 25°C 3.1004/s 4FM2 425°C, 240 h 25°C 3.1004/s 9FH2 425°C, 240 h 25°C 3.1~/s 9FM3 425°C, 240 h 25DC 3.1004/s 4FID 425°C, 240 h 425°C 3.1004/s 4FM4 425°C, 240 h 425°C 3.1004/s 9FID 425°C, 240 h 425°C 3.10-4/s 9FM2 425°C, 240 h 425°C 3.1004/s 4FH4 650°C, 24 h 650°C 3.1004/s 9FH4 650°C. 24 h 650°C 3.1004/s ·4· 3.2 SF2 Tensile tests 0) Non Ireat treated samples Figures 2 to 4 show the results of tensile tests for the 4F and 9F samples without heat treatment. Measurements have been made at - 170°C, 25°C, 425°C and 650DC, for yield strenght (0,2 %), ultimate strenght and total elongation. b) Heat treated samples Figures 5 to 7 show the results of tensile tests for the 4F and 9F samples after heat treatment. c) Discussion Except for the temperature of - 170DC, the results show a good accordance between the two faces of the cell. Influence of heat treatment : on the face 4F, the yield and ultimate strenght decrease after heat treatment specially at low temperature. In the same time, total elongation increases, and remains at 5 % for the non heat treated sample at 25DC. On the face 9F, the yield strenght decreases after heat treatment which have no effect on the ultimate strenght. Total elongation increases after heat treatment and remains at 5 % for the non heat treated sample at 25°C. 3.3 Comparison between SFt and SF2 Figures 8 to 10 show the comparison between SFI and SF2, for the non heat treated samples. Comparison is made for yield strenght, tensile strenght and total elongation. For SF2 the values used for the curve are obtained from the average of 4F and 9F values. For SF} the values used for the curve are those of the face looking at the beam tube 8F. -5- Discussion The yield and ultimate strenght are equivalent except for the measure of yield strenght at 25°C. The material keeps a good ductility at low temperature (- l70°C). At 25°C, the total elongation decreases for SF2 (6 %). 4. CONCLUSION The post examination programm developped for the SF2 cell of the reactor ORPHEE after 1668 EFPD has confirmed the results obtained for SF! after 1000 EFPD. The material keeps good mechanical properties, for all the working conditions. After heat treatment (425°C, 240 h) the embrittlement at 25°C seems to be reduced. -6- REFERENCES [1] P. BREANT, « The ORPHEE Reactor, Current Status and Proposed Enhancement of Experimental Capabilities », looRR I, Knoxwille (USA), Feb.28-March 2, 1990 [2] P. BREANT, B. FARNOux, 1. VERDIER, «Cold Sources of the ORPHEE Reactor », International Workshop and Cold Neutron Sources, Los Alamos (USA), March 5-8, 1990. -7- ORPHEE REACTOR 6T 4F 2T 21.. hCeDaRv! y __ 3. hIgh-le...,.".,wra saun:e 4. 1aw-temperBlure saun:e &'paaI 6. paaI imler_I 7. annular 8PIII:e 8. paaI outer wall 9. radicH!;aape p!IIduction c:IIanneI 10. _lube 11. _ irradialion channel Figure 1 -8- RpO,2(Mpa) ORPHEE - SF2 - COLD CELL - TENSILE TESTS 1100 \ I· RnO "4Fnt 1000 RpO,2J9Fnt 900 .. '" "- " ..... --. 800 "---~- !o.... 700 -.......: ~ ~ 600 ~ t-- soo "- ')l 400 ·200 o 200 400 600 800 Tempemrure ("C) Figure 2 : Yield strenght evolution. Comparison between face 4F and 9F (non heat treated samples). Rm(MPa) 1200 I 1100 III Rm/4Fnt ~ Rml9Fnt 1000 900 ~ 800 "~ - r- ~ 700 b.... 600 ~ ~ 500 ·200 o 200 400 600 800 Ternpe:a1Ure ("C) Figure 3 : Ultimate strenght evolution. Comparison between face 4F and 9F (non heat treated samples). .9· At(%) ORPHEE - SF2 - COLD CELL - TENSILE TESTS 30 . Atl4Fnt 2S 99 %). The performance of the novel spherical annulus design has been verified, and shown to provide good gains to long wavelengths. Detailed comparison of the performance (cold neutron production and heating) to calculations provides a rigorous test of the codes and models used. These comparisons show that the heating rate calculated for the moderator and chamber are 25 % too high. Interpretation of the neutron spectra are dependent upon the assumed ortha/para hydrogen ratios - assumption of an ortho hydrogen fraction between 65 and 75 % provides similar accuracy for neutron intensities, and excellent agreement for spectrum shape, while assumption of a 100% para hydrogen fraction are in complete disagreement with the data. Improvements to the design to enhance the source performance are now being planned for installation at a future date. REFERENCES 1. Williams, R.E., ''Upgrade and Modernization of the NBSR," Proceedings of the Fourth Meeting of the International Group of Research Reactors (IGORR-IV), May 23-25, 1995, Gatlinburg, Tennessee. 2. Farnoux, B. and Breant, P., ''Upgrade of the Experimental Facilities of the Orphee Reactor," Proceedings of the Third Meeting of the International Group on Research Reactors (IGORR-Ill), September 30 - October 1, 1993, Naka, Japan. 3. "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A, II LA- 12625, J. F. Briesmeister, Ed., Los Alamos National Laboratory (1993). 4. Fessler, T. E. and Blue, J. W., "Radiation-Induced Conversion of Liquid Hydrogen," Phys. Rev. Let. 14, 811 (1965). 5. Nelms, L. W. and Carter, H. G., "NERVA Irradiation Program: GTR Test 21, vol. 2, Nuclear Radiation-Induced Conversion of Para Hydrogen to Ortho Hydrogen," U. S. Atomic Energy Commission, FZK-351-2 (1968). 6. Williams, R. E., Rowe, J. M., and Blau, M., "Benchmark of the Nuclear Heat Deposition in the NIST Liquid Hydrogen Cold Source", presented at the Ninth Symposium on Reactor Dosimetry, September 2-6, 1996, Prague, Czech Republic (in press). 7. GalImeier, F. X. and Gehin, J. C., "MCNP--Cross Section Modifications for Reactor Physics Calculations," Private Communication (1996), to be published. Figure Captions Figure 1. Schematic of the NIST liquid hydrogen cold source. Figure 2. Plan view of the cryogenic beam port showing the relative positions of the cold source, fuel, and the in-pile sections of the neutron guides. Figure 3. Gains for the NIST liquid hydrogen cold source as a function of wavelength, measured at the NG-7 SANS. Figure 4. Cold neutron spectrum: measured vs. calculations. Figure 5. Calculated nuclear heat load. I HYDROGEN COLD SOURCE THERMOSIPHONI He In l Ballast Tank Hydrogen Condenser Fig. 1. Schematic of the liquid hydrogen cold neutr9~ source. Each component is completely surrounded by a helium containment, not shown. ,-INSULATING VAOUUM FUEl. El.EMENT 4 • • $' A, • i •• " '4 ~,' 4' '",, , • 4, G 0.82 MeV 1.3 x 1012 Thermal E<0.625eV 1.7 x 1014 Thermal Flux at Tube Nose 9.7 x 1013 Table 3 Gamma Fluxes at HANARO Energy Level(eV) Gamma fluxes( r Icm2·sec) 1.0 x 10"' to 1.0 x 10' 5.818 1013 X . 1.0 x 105 to 5.1xlO' 4.452 x 1013 5.1 x 10' to 6.0 x 10' 5.610 X 1011 6.0 x 10' to 1.3 106 X 3.993 12 X 10 1.3 106 X to 1.3 6 12 X 10 5.287 X 10 3.0 106 to 7.5 106 1.286 1012 X X X 7.5 6 X 10 to 1.4 107 4.544 1010 X X 2.2 Design requirement The design criteria for HANARO CNS is maximum increase and stable supply of cold neutron flux, and safety with regard to personnel and the reactor. High technology for cryogenics and gas explosion and reactor safety engineering is required in design, construction and operation for safety of eNS system and the reactor. Especially. the measure to control the hazards related to the hydrogen use is very important in design. The design requirements are summarized in the table 4. 4 Table 4 Design Requirements for HANARO CNS Item Requirement Gain factor More than 20 for 5 A Cold neutron flux More than 5 x 10· nlcm2·sec at the end of cold neutron J!;Uide tube Life time consideration 30years Applications • Crystallographic studies • Neutron non-destructive evaluation • Small angle neutron scattering • Neutron reflectometry • Neutron inelastic for chemical analysis • Neutron and nuclear physics Safety consideration • Explosion due to hydrogen-air contact • The effect on reactor by installation of CNS(Reactivity, Pressure etc.) Installation consideration • To be installed in the existing vertical hole( rp 16 cm) • Interfaces with the existing facilities in the reactor pool • Space limit for installation near reactor pool Other • Design of ultra-cold neutron facility 2.3 Heat release The heat release in CNS includes heating due to neutrons(slowing down and absorption) and prompt r -rays and {J -radiation because of capture of thermal neutrons. Heat release for CNS is not calculated in detail yet, but some rough estimation can be done. This is important because it is necessary for choosing an efficient way for the heat removal. The estimations are shown in Table 5. This data will be defined more accurately again in the near future. Table 5 Heat release data Fast and epi- r -rays Total thermal {J -rays Specific nuclear heating 1.3 WIg 0.4 Wig - 1.7 Wig in hydrogen s Specific nuclear heating 0.2 WIg 0.2 WIg - 0.4 WIg in deuterium Specific nuclear heating - 0.2 WIg ~ 0.2 WIg in zircalloy Specific nuclear heating - 0.2 WIg 0.3 WIg 0.5 WIg inAl 2.4 Source Chamber It would be reasonable to use the maximal volume of the moderator and to optimize its efficiency by varying the concentration of hydrogen in deuterium. The source chamber will be made of cylindrical type having 140 mm diameter and 210 mm height, and the available volume will be about 3 liters. The cylindrical chamber with elliptic bottoms and with thickness of Al walls of 2 mm has the weight about 650 g. Taking into account the weight of chamber tubes in the field of irradiation, the total weight is estimated to be about 1.1 kg and the heat load due to the nuclear heat release in the material is about 550 W. The nuclear heat release in hydrogen is estimated 450 W or in deuterium 240 W. It is expected that optimal concentration of the hydrogen in deuterium-hydrogen mixture is about 50 % or less. Therefore the more realistic estimation of the heat load for the deuterium-hydrogen mixture is about 350 W. Thus rough estimation of the total heat load for the CNS is about 900 W. 2.5 Heat Removal The way for heat removal for HANARO CNS is a subcooled liquid hydrogen thermosiphon. No bubbles, self-regulation, safety are the benefit features of this way. There is a natural circulation of the subcooled liquid hydrogen between the chamber and the heat exchanger. The minimum helium temperature is higher than hydrogen freezing point. The cold helium removes the heat from liquid in the heat exchanger. A counter- flow heat exchanger is the preferable one. The thermosiphon loop is placed entirely in a vacuum containment. The loop is filled by liquid moderator(hydrogen-deuterium mixture) without vapor. The chamber contains about 3 liters of moderator and is made of aluminum. The chamber weight is about 1 kg. The total radiation heat load is expected to be about 1.0 kW. 6 2.6 Review of interferences with the existing facility For installation of CNS at HANARO, some technical review and analysis should be taken to reveal physical interfaces and problems. Vacuum system, helium containment and number of the diverging cold neutron guides from the nozzle of reflector wall had not been considered technically when the reactor structures and embedded parts were being designed and constructed. Really, removing the light water from the installation hole in the reflector tank to improve the gain becomes an issue because the wall of the reflector tank is a little bent due to heat for welding. Therefore, we consider extra helium layer between the reflector tank wall and vacuum containment as shown in Fig 2. . Fig 2. Installation of the cold neutron source This helium environment can function as the helium safety blanket at the same time. The pressure of helium in the blanket is equal to hydrostatic pressure in the pool at the depth 11.3 m. The plug at the bottom of CNS hole is not necessary, and the water is taken out of the hole by means of the helium bubble. 7 3. Safety The CNS facility might be hazardous and demands special approaches for safety though it can be used for various experiments of neutron physics. It is necessary to apply technological and organization approaches in the design to escape dangerous influence on people. reactor and surrounding atmosphere. CNS safety is considered in view of nuclear activity. radiation and hydrogen explosion. • Cryogenic system failure In case of cryogenic system failure, all liquid vaporizes and the moderator cell may be overheated. The vacuum containment will be filled by helium automatically to transfer heat from the cell to the containment wall. • Moderator cell rupture In case of cell rupture a vacuum disappears in the containment, and cryogenic cooling loses. Moderator vaporizes. The final pressure in the moderator tank will be less than the pressure in normal warm conditions. The moderator should be evacuated after shut- down. • Failure of helium blanket around vacuum source containment In this case. only cold neutron flux may decrease due to incoming of water. • Vacuum failure In case of vacuum failure. the moderator vaporizes. The vacuum containment should be filled by helium additionally. • Reactor shut-down Reactor shut-down does not cause any trouble. Moderator will be in the loop at a lower(a little higher than freezing temperature). 4. Neutron Guide System The task of the neutron guide system is to transport neutrons produced in the cold neutron source to experimental devices which is intended to perform scientific investigations with neutrons. Transportation of neutrons is realized on the base of the 8 mirror reflection of cold neutrons from the walls of the neutron guide. The curved neutron guide channels can be used to cut off fast neutrons and gamma radiation because their total reflection coefficients are negligible in comparison with that of cold neutrons. The scheme of the neutron guide system is presented in Fig 3. The neutron guides are channels with rectangular cross section, prepared by using mirrors with a glass substrate and S8Ni reflecting coating. The internal cross section is of 40 mm width and 140 mm height. There are two main neutron guide. The possibility of insta1lation of the third channel will be reviewed. CNG in the figure is the converging guides made of NtIfi super-mirror. Separation of the neutron beams formed with these channels in the horizontal plane is defined by an appropriate angle between the channels. The value of the angle 2a 1 = 120 X 10-3 radians is restricted by dimensions of the channel tube containing the neutron guides as shown in Fig 4. Characteristic wave lengths for different neutron guides. and cross sections of the guides are shown in the Table 6. Table 6 Characteristic wave length Characteristic it Cross section Neutron Guide Coating (A) axhmm 1 NG 6 20x60 '8Ni 2NG 2.5 40x60 '8Ni 3NG 3.5 20x60 ,aNi 4NG 8 40x60 5aNi NG 40 x 140 5~ CNG Super NiITi 9 NG Fig 3. The scheme of the neutron guide system Fig 4. Neutron guide installation in beam tube 10 5. General scheme The general scheme for cold neutron source at HANARO reactor is shown in Fig 5. The main parts of the scheme are liquid hydrogen-deuterium loop. hydrogen-deuterium supply system. vacuum system. helium refrigerator. eNS loop has a connection with hydrogen-deuterium supply tank of volume 3 m3• The hydrogen system is closed and surrounded by helium blanket. Helium pressure in blanket is 0.2 bar. which can be controlled. The helium refrigerator maintains the moderator at liquid state in the loop a nominal reactor power. The refrigerator includes cold box, helium compressor and helium Teceiver for containing necessary volume of helium . -- '.-- . _. - 0 ..' "-.' ... -. 0..: -~ -:-- -- -,'- Fig S. The general scheme for CNS at HANARO 11 6. Experimental instruments The cold neutron experimental instruments at HANARO will be utilized for studies in the broad category as follows: - Crystallography - Chemical Physics of materials - Magnetism and superconductivity - Surface and interfacial studies - Macro-molecular and microstructure studies - Residual stress, Texture and radiography - Analytical chemistry - Neutron metrology and dosimetry - Irradiation For this purpose, about 10 experimental facilities will be installed in the CN experimental hall. The experimental facilities to be installed will comprise a 30m-SANS, a 8m-SANS, a ret1ectometer, 3 triple axis spectrometer, 2 time of flights, a backscattering spectrometer, a SPINS spectrometer, a prompt-gamma activation analysis spectrometer and 2 facilities for fundamental neutron physics. 7. Conclusion Cold neutrons has been used extensively for the study of the structure and dynamics of materials in some advanced countries during the last decades or so. The cold neutron source at HANARO will be designed from the end of 1996 and the installation will be completed by the end of 2001. The research activity using cold neutron in Korea stays behind the other advanced countries. With the completion of the CNS facility, the study of the material science, solid physics, chemistry and biology etc. will be going on more vigorously centering around the HANARO. CNS facility is becoming important to develop the state-of-the-art technology by itselfin many fields with the recent remarkable growth of the Korean industry. Because potential cold neutron users are recently increasing more and more in Korea, the installation of cold neutron source is essential to provide neutron beams for more users. 12 References [1] J. M Rowe, nThe NBS Cold Neutron Research FacilityR, NBS Special publication 711, NlST(1986) [2] H. 1. Prask, J. M Rowe et al, "The NlST Cold Neutron Research Facility", J of Research of the NIST, Vol. 98(1993) [3] JAERI, "Handbook of Utilization of Research Reactors",(1995) [4] A. F. Schebetov, N.K. Pleshanov et al., "Construction and testing of a multi-channel polariser for thermal neutrons", Nuclear Instruments and Methods in Physics Research(1994),575-580 [5] A. P. Serebrov, V. A. Mityukhlyaevet al., "Experimental Study ofa Solid-deuterium Source of Ultrac old Neutrons", JEPT Lett., Vol. 11, No. 59, 1994 P.728-733 [6] A. P. Serebrov, "High Flux Reactor PIK and the Associated Research Program", Nuclear Instruments and Methods in Physics Research A284(l989) 212-215 [7] A. P. Serebrov, A. A. Zakharov et al., " An Universal Liquid Hydrogen Source of Polarized Cold and Ultracold Neutrons at the LNPI WWR-M Reactor", SOy. Phys.- HEPT Lett. Vol. 44, No.6, 1986, P. 344-348 [8] A. P. Serebrov, I. S. Altrev et al.,"Cold and Ultracold Neutron Sources in Gatchina, Russia", Neutron Research V. I, No.4, 1993, P 71-77 13 , COLD NEUTRON CROSS SECTIONS HISTORICAL REVIEW presented to the 5th meeting of the International Group on Research Reactors D. L. Selby November 1996 COLD NEUTRON CROSS SECTIONS HISTORICAL REVIEW (1) • First major slow neutron scattering data for molecular hydrogen and deuterium was published by Young and Koppel in 1964. • Compared with measured data these neutron scattering cross sections were good down to about 3 me V. • Model was based on gas model. overpredicted hydrogen with respect to measured values by almost a factor of 2 at energies below 2 me V and overestimated gain factor by almost 50%. overpredicted deuterium with respect to measured values by as much as a factor of8 at very low energies « 1 meV). • Cross sections have been used to develop JUlich, ISIS, and FRG-I cold sources. ~ COLD NEUTRON CROSS SECTIONS HISTORICAL REVIEW (2) • In 1977, while at ILL, Utsuro modified the Young and Koppel model for deuterium to account for the liquid state at the lower temperatures. • Greatly improved the agreement between measured and theoretical deuterium scattering cross sections, but still overpredicted with respect to measured values by almost a factor of 2 at very low energI• es. • The Utsuro model was used to perform the neutronic design for the upgrade of the vertical cold source at ILL and has also been used by the Research Reactor Institute at Kyoto University. COLD NEUTRON CROSS SECTIONS HISTORICAL REVIEW (3) • In 1988 Bernnat at IKE used a somewhat different approach to account for the liquid state to develop a scattering cross section model for ortho- and para- hydrogen and deuterium. • Agreement with measured data was excellent for hydrogen. • Although measured deuterium data was still overpredicted by about 25%, it was an improvement over previous models. • This model has been used for the ANS project, the FRM-2 cold source design, and the NIST cold source upgrade. II' , " J 1 • ... -..... kOPP£L.lOtnfG , I k f lIOUID !I o SElffE.' ' 11970l fl' ",'!;'";i~';-'··~·"-' '.'", • ': -, ~'1.r· " • 'J'" , : "",:~if!!:~'f :.:,~I ::i=, :I,L~·: ..'" ,·.·."." ,;~ ..' [!..:.:.:• . ;,,' :i: ':"! , ,f.' -~~~.~·:t~ .' : .7.··. .. . i ~~ ,." 'J -, ;,• .\: ,"i~' .,.I ,;_~ ,' .... 10, I ~ ' .. :.,.~.\. .~., .: -.: .. ~' • t ", ', 11111 .. " , , 1 ! .. -..... '0 , • -"'- .. - ". .'. ......... .. " 1:' I •... , : , I I , t , '-"', ;"--,, -f "", •• I, •'.• - • • ''':'._' ""pO ..... ~~... ••• " -.·,.~-.._ ~.-1..·.'..~.' .t::;.!;'.;r..t .' l1itpG...ef.' !" .• ' '" " ',~;.~~~~~~~.' ::.,. . :, ; .' .' ~-.~'. .... :: ...... -',;. FIg.';, 4: Totallleutron'e~*,'i(jriJ]i I' - , " • ~. ' '.' . '" .oJ: .: " .. , " ; ... . ' ~ \ ":,' .(-1" \0 J. 1 " rho = 33 kg/m3 and for Tout =21.5 K-> rho = 71 kglm3 ). This value is much larger compared to the density difference of the en- forced circulation with A tho = 5 kglm3. The value of the large difference in altitude and density are the reason of the stability of the natural convection. Furthennore, the peak in the heat capacity at T = 34 K aggravates the increasement of the hydrogen temperature above 35 K, because the hydrogen absorbs a lot of heat (see fig. 5 ). ! . -. g I I I \ I I / \.... I I I I I 1-,....... I 1 I I ..... tt' ',. .,' c " • • • • M • • • T-M2 P 1 "liitI eeh n"l catome FIG 5 SILOE REACTOR HALL photo : MFP I""t eeh n.l eatome FIG 6 ORPHEE REACfOR photo: CEA I'-"t eeh m. eatome FIG 7 PARTIAL VIEW OF ORPHEE REACTOR HALL photo : ANPHI FIGS a_ +".50 !5O ., 85 .1 • • I • • • • t • • to. SIRIUS 3 - REACfOR BUILDING CROSS SECTION TA·SOIUv.l I-Itechnlcatome FIG 9 CLIMATIC ~ CONDITIONS 'l-r1Yff' 1,, ,, , HANDLING OF CO, MPONENTS , r, ------------.. --. : ----.. -.. ~ .. ---------.. ----,.-~ t :" : EXPERIMENTAL !,, t ,, ,, NEUTRON ,, GUIDE t BUILDING IlEAtTlfOR 1IU1UJINfJ OP1IlU.TIONAL R1IUTllD DItS1QN BEflUlB1iII8N'lS ( NORIIAI. CONJJnIONS ) REACTOR BUILDING OPERATIONAL RELATED DESIGN REQUIREMENTS (NORMAL CONDITIONS) TA· so Rev. I I""t eeh n.l eatome FIG 10 AiR CRAFT ~ CHASH ~ / FXPLOSIO:fS /" S 'OW - - I ~ WI .-D - - r r , , ~ , C \ " ~ f i WATIW HAM\fElt , OVER ~ r- PRESSlifdl " r .... r I AlP, f I -'CONTAMINATION r I FIP.F. - - --= -'z-' Q , o · o .-..l. ACCES::i / I / I , ,r f t !.... PHYSICAL SEISM UNCONTROLLED PROTECT[QN LEAKS lUlACTOR BUILDING SAnf'Y BaATBD DBSIGN R8qUIRlfJlBNTS ( IlAIN llAZARDS ) REACTOR BUILDING SAFETY RELATED DESIGN REQUIREMENTS (MAIN HAZARDS) TA - SOR."I CHAIRMAN : R. WILLIAMS SESSION 4 RESULTS OF A SURVEY ON THE DESIGN BASIS FOR RESEARCH REACTOR CONTAINMENT I CONFINEMENT BUILDINGS (Albert Lee) Question from Klaus Boning ofTU Miinchen : And here we should mention that we take into consideration a full core melt under water, which is important, so that there would not be any evacuation of the population outside of the facility fence necessary, so no emigration. But, we have not taken into consideration a severe increase in internal pressure, as for the BR2 , we have to discuss the figures. Another point that we also should designate to be taken into consideration is the airplane crash, from a military or high-velocity jet into the building, which is unique. We have 1.8 meters of thick walls around the building. OK. that is a brief report. One question : I wonder why you did not consider terrorists? A : OK, it's true. That is a question I didn't ask. I consider terrorist events and external humanly-initiated events to be something that is beyond the ability and control of a reactor designer to try and design. I think that has to be covered in the realm of the establishment of security on the site rather than defending it at the reactor building. Maybe it's not totally true, but one has to look at in terms of site considerations, and I agree that it's much more difficult at Munich than it would be at, say, Serpong in Indonesia. OK? So how one treats it I think, really has to be dealt with within the context of where the reactor is located, and what kind of surrounding site is provided. I also don't think that the issue of the vulnerability of the reactor building to a terrorist event is something that we really want to publicize with anybody. Question from Clifford Hickman of Technicatome : rd like to talk about aircraft crash risks. You say they are site dependent, which is obviously true. But national regulations - can you say something about that? They are not necessarily site dependent. A : Well, I think again it varies. In North America, at least, I know that at the Oak Ridge National Laboratory and at the Argone National Engineering Laboratory, and also at Los Alamos, like in Canada at Chalk River and the Whiteshell Laboratories, we have an exclusive air exclusion corridor where aircraft are not permitted to fly. That air exclusion corridor, at least at Chalk River, is 20 kilometers on either side of the site. I can't rule out every event. All it does is establish the probability you have to defend against for an aircraft crash. This is where one always gets into probabilistic assessments on it. rd never rule out pilot error or a deliberate event like they did at the White House about a year ago when somebody flew a small plane into the back yard of the White House. There is no exclusion corridor around the White House, but you can't prevent it. You have to look at it and decide what is the most sensible what are the most likely events you have to defend against within your regulatory regime. Question from Hans-Joachin Roegler of Siemens: Don't you share my view that, when we start discussing such events like aircmft cmshes as potential events that we have to protect our towns against, then it will come up more and more and it will raise the costs more and more. So we should really try to prevent making that an aspect to consider. Although Munich is a bad example, and when reporting on this as the first reactor having this protection GA even claim that their Texas Airbase TRIGA was the first research reactor protected against aircraft crashes. I think we should stop describing this as a positive feature. A : I agree with you. I don't think it's an issue to mise automatically for seveml reasons - it does drive up costs, it detmcts from the real issue of safety that should be in the design of the building. And one has to keep in mind that there are two issues that appear on the surface to be related but are not always related, and you have to keep this in mind. One should be ensuring, as the operator of a reactor and as the designer of a reactor, that we specify systems to ensure safety. Safety first. The regulator's responsibility is to review your safety case to determine whether or not in their opinion it's licensable. Licensable arguments, licensability, can be based upon judgment criteria that are not always related to safety. They will look at events that could be very highly improbable in your judgment, and the challenge is how far you have to go to defend against very improbable events. One has to try and draw the line somewhere, otherwise costs of these reactor facilities will be totally uncontained and I could easily take the ANS and add some arguments for very improbable events and take the cost of the ANS up to the cost of the superconducting supercollider. But I don't want to be that perverse! One may not, but I could re-evaluate an existing reactor that has opemted very safely for close to 30 years and apply inappropriate arguments and convince somebody that it's not licensable today. And that's not the case. Question from Guy Gistau of Air Liquide : As I pilot, I can tell you that, in the Rhone Valley, the minimum height above a reactor is 1000 feet, 300 meters. However, above the site at Cadarache it is a restricted area which is not allowed for flights. A: Just as a closing remark, the building for the Maple 1 reactor is complete. We completed it in 1992. We stopped the project in 1993 for various reasons. We are about to restart that project. But when we completed the building design, the reactor hall that surrounds the top of the pool is constructed out of heavy concrete with about 60% reinforcing steel in it. The walls are 30 cm thick, the roof is 30 em thick with a I-inch steel plate on top. OK? The worst case accident we looked at was a single channel flow blockage with internal pressurization of 5 PSI. The building will withstand your avemge 1500-kg vehicle tmveling at 180 km per hour 5 meters above the ground, impacting it. FRENCH RESEARCH REACTORS DESIGN OF REACTOR BUILDING IN ACCORDANCE WITH SAFETY APPROACH AND IAEA RECOMMENDATIONS (Jean-Luc Minguet) Question from Edgar Koonen of CEN - SEK : Is there a specific leak rate established, a value of, say, 3 x 10-4 of the air volwne? A : I don't have the values in mind, but considering the different reactors I mentioned before, we have for each of them specific maximal allowed leak rate is to respect and we have to remain under these specified values. These tests are performed by under-pressure tests, and we observe the loss of pressure during a specified period of time. These values are specified at the issue of the safety analysis. Q : Is it the result of the borax analysis? A : It is based on the analysis of the potential consequences outside and these values are linked to the potential acceptable consequences outside the reactor building. Question from Johannes Wolters of Jiilich Research Center: In Germany, when we speak about design basis accidents, when we consider them, we have to show that we keep below certain planning values concerning the radiation exposure outside the reactor. Is it the same in France and what are the values? A : In general terms, it's quite the same : for the accidents considered for the design, the wholebody dose for the public has to remain below acceptable value (0,5 REM). However, for DBA, this point has to be further discussed in the frame of a specific project. Question from Guy Gistau of Air Liquide : This is not a comment related to the topic of safety, but could you show the picture of the pool of ORPHEE. As a cryogenist, I would like to show here something which is not necessary to do, or something to be avoided. OK - rm looking for it. Is this the view you wanted to see. Yes. You see here, on the left side, these two big vertical capacities which keep heliwn. This is equipment which can be put anywhere inside a reactor, but I think not near the pool where the area is quite expensive. You can put that anywhere else in the reactor, but not there. Question from Jong-Sup Wu ofKAERI: I think the definition of the confinement and containment are different. It is very difficult to verify the allowable leakage rate during operation because it depends on environmental conditions such as wind and external temperature. What kinds of measurement methods are applied to French reactors? A : Confinement is a building in which, by the ventilation system, you keep a depression in relation to the outside so as to avoid and control any leakage directly from the reactor building to the outside. A leak-tight containment building is a building in which it is possible to maintain radioactive products, even if there is a level of pressure higher than outside. For these containment building, as I said beforre, we perform under pressure tests. Q : Another question is the method of measurement of the leak rate during operation. In the case of containment, I think it is easy, but for confinement. At HANARO, it is confinement, but in our experience it is very difficult to verify the leak rate during operation. What is the case for you for leak testing in confinement? A : In fact, we have containments, but for confinements it is possible to observe how the pressure increases when you have a sudden stop of the ventilation. Comment from Francisco Alcala-Ruiz ofIAEA : I understood that the question was related to buildings which would not be able to withstand any overpressure in that case. Q : When we tested according to time, it depends on the outside environment, for example the wind, the atmosphere. It is very sensitive. That is my question - how do we verify the exact leak rate during operation ? Comment from Edgar Koonen of SCK-CEN : I just want to add some information - when we measure the leak rate, we even have to take into account dilatation of the building in the data. This is a significant correction that we make. If we don't do it, we might even measure a negative leak rate! The building can have quite an important dilatation on top; it's 40 meters high and it's quite hot there. Right, of course. We have to take into account the possible temperature evolutions during the tests. Comment from Colin West ofORNL : Similarly, you have to be careful at what height you measure the pressure. If you have a very tall containment building, the delta P between the basement and the roof is more than the leak rate of delta P you're looking for. Now, if the humidity changes outside more than it changes inside, you have to worry about that, too. That's just a guess! We still have time left for any other comments you might want to make. Question from Hans-Jfirgen Didier of Siemens : I only want to make a comment thafs a question., too. Isn't it a dogma which you use as a regulation., the Borax. I think it might be very dangerous to have a dogma instead of a regulation, because I don't find any mechanistic reason why you have to handle a Borax event in a research reactor. That was our problem in Germany. Answer: Jean-Luc Minguet : I think that it's a rather old story. In the past, when deciding to choose an envelope event, Borax was considered as the worst severe possible accident based on the experience and the studies performed in the US on uranium aluminum fuel. As I said, it could be certainly of a great interest as we start new safety analysis in France for instance for the RJH, to put all these things on the table to see and possibly to define another more relevant approach and to define the new basis for a major severe accident to be considered in the design of a research reactor. Question from Jean-Jacques Verdeau of Techn icat ome : May I add something? For SaOE, we did a refurbishment in 1987 and we had to add a liner to the pool because the pool was not leaktight and in order to prevent the borax we only had this stainless steel vessel around the core. And you shouldn't think that take into account the borax is very expensive. It's not very expensive in fact. Question from Hans-Joachim Roegler of Siemens : I say that the French have mainly made a fashion out of that and ifs not considered as an accident in full with all the consequences that are potentially there. There have been no experiments to demonstrate that, in all the reactors you have here, borax is really overcome by this small structure from buffering the pressure at the water pool. So I don't think that is a consideration we would have in Germany if we had the same assumption for borax in our licensing. Comment from Kir Konoplev of PNPI : I would like to mention that leakage is not a figure that is representative enough for containment. After the international experts of the reactor PIK, we tried to make an investigation of our situation. We compared our situation with the containment of other reactors. ~ You see, the leakage of the different isotopes of course has a different influence on the radiation on the environment. For example, any containment can help you with this cesium. If it comes into the containment, it will come into the environment. I It is the contrary for the short life isotopes. So it depends on the leakage and the volume of the containment itself. For our examinations, we melt the core to release all the fission products that exist in the fuel and look to see in the leakage, what is the presence of every isotope, and the dose at the boundary of the site. Comment from Francis Merchie of CEA : Just a comment about the role of the confmement or containment building. From a safety point of view, we have to control the three safety ftmctions of the reactor. We have to control the reactivity in normal and abnormal situations, we have to control the heat dissipation and thirdly we have to confine the radioactive products. The building is the third barrier for the confinement and, by applying the defense and death principle, the third barrier must be the last barrier. That means that the first barrier, the cladding, and the second barrier, the primary circuit and the pool, must be inside the third barrier. This is very important and is the reason why in the French reactor design, the primary circuit is included in the building and inside the concrete reactor block. Coming back to the borax. I think that borax is only a name - the problem is what type of reactivity insertion we have take into account in the design of a reactor. Of course, we know the experience with borax at SPERT, the final experience at BORAX delivered an energy of 135 megajoules. At the SPERT experiment, the energy was higher than 135. In France, we performed many experiments at CABRI in the '60s, and we came to the conclusion that even with lower insertion of reactivity, I mean for example 2% of delta kIk, the period is very low and the energy liberation is always very important, creating mechanical damage to the core. So, taking the so-called borax accident as a DBA is like a credit in release of energy in the case of reactivity insertion. We have also performed some simulations using explosive powders in mock-ups of a scale of 113 or 1/5 for different swimming pool type reactors, and we saw that there was in fact some mechanical damage to the pool and on the beam tubes. So the beam tubes are very important from that point of view. This is why we have installed what we call safety valves at the outlets of the beam tubes to be sure that in case of any type of leakage on the pool, there will be no leakage of radioactive water outside the concrete block. The End SESSION 5 WORKSHOP ON R &- D NEEDS AND RESULTS WORKSHOP ON R&D NEEDS AND RESULTS Klaus Boning Technical University of Munich 1. Introduction The goal of this R&D workshop has been - as in the IGORR meetings Nr. 1 to 4 - to iden- tify research and development problems which IGORR groups of research reactor opera- tors and designers might have and to stimulate contacts with other IGORR groups who plan to perform or have already performed R&D investigations on problems they have. The former IGORR meetings have demonstrated that there is, indeed, considerable over- lap in the needs of various groups in the same R&D topics. So this workshop has develo- ped to something like an open market for the exchange of information between the various groups on - R&D work needed (without having plans for own work) - R&D work announced (i.e. work needed and planned) - R&D work just peformed (so that preliminary or final results could be made available). 2. Old R&D Topics (from IGORR 1 to 4) In the report on the R&D workshop of the last IGORR meeting (see IGORR-IV Procee- dings, page 391 ff) all the R&D topics which have been brought up, but not yet addressed so far, have been identified with an alphanumeric key: the leading number giving the IGORR meeting (Nr. 1- 4) at which this topic has been raised, and the following letter helping to put the various topics in consecutive order. Many of these topics have been discussed in the R&D workshop of the present IGORR-5 meeting: Nr. 2a: "Multidimensional kinetic analysis for small cores". Christoph Doderlein, who is now at CEA Cadarache, gave a brief report on what he has done as a Ph.D. student for the FRM-II compact core reactor. He has written a short summary which is included somewhere in these Proceedings. Nr.3c: "Chemical and other energy release from core melt events". Albert Lee, AECL told the auditory, that Michael Corradini from the University of Wisconsin, USA, has developed a computer code for predicting aluminum fuel to water coolant interactions. The code which is based on experimental data is called TEXAS-III. The documentation is available from M. Corradini. Nr.4e: "Cold neutron beam tube guides size and geometry optimization studies". Doug Selby, ORNL, reported that work on thermal (not cold) neutron beam tubes has been done at Oak Ridge, but has not been published, whence interested groups should contact him directly. Nr. 4f: "HANS-3 fuel capsule irradiation (in HFIR) evaluation". Doug Selby, ORNL, gave a summary of these very interesting results; his transparencies are reproduced in Appendix A of this report. Doug Selby also reported briefly on the following R&D topics for which results have been obtained from the ANS closeout activities: Nr. 4f: (as above) Nr. 4g: "Meat fabrication with spherical powder fuel". Nr. 4h: "Centering of fuel in plate fabrication". Nr. 4i: "Effect of flow blockage shape". Nr. 4j: "Reduced ph effects on aluminum corrosion". Nr. 4k: "Final summation of the ANS thermal hydraulic test program". Nr. 41: "Final results from the HANSAL aluminum irradiation tests and examination of BNL data". Nr. 4m:"lrradiation creep in aluminum". The references and further information on all these topics can be found in Appendix B of this report. Finally, Doug also submitted a list of references refering to fuel plate stability studies for the ANS, see Appendix C. This topic had already been brought up in the R&D workshop of IGORR-1 and has been further discussed after the presentation of Jurgen Adamek at this meeting. Other R&D topics from IGORR-4 were addressed in special contributions to this IGORR-5 meeting: so, concerning Nr. 4a ("Requirements for the design of containment") and Nr. 4b ("Cold neutron nuclear data") separate papers of Albert Lee/AECL and Doug SelbylORNL can be found in these Proceedings. If we further drop all those topics which have been discussed at IGORR-4, we end up with the Table 1 of R&D problems from previous IGORR meetings for which answers have not been given yet: Table 1 R&D needs identified at IGORR-1 to IGORR-4 which have not been addressed so far: 2b Burnable poison irradiation 3f Test of cryogenic circulators for single-phase forced-convection cold sources 4c Thermal-hydraulic data (flow instability) on the heat transfer in fuel elements (up to 500 W/cm2) for high cooling water velocities (around 17 m/s) but low system pressure (below 10 bars) 4d Method to calculate the decay time required after full power operation before the core is safe in air. 2 worb0496 The cancellation of "old" R&D topics does, of course, not mean that the answers given were sufficient in all cases to really solve the particular R&D problems; so it might well be that some of these items could come up again in the future. 3. New R&D Topics (from IGORR-5) In the second part of this workshop transparencies were made available to the audience which could be completed by interested participants. These forms covered the three areas "R&D work needed lor announcedl or performed". What followed was a most lively activity of many participants coming up to give their presentations. The various R&D topics are listed in Tables 2 and 3, labelled 5a to 5h according to the order in which they were presented. Table 2 New R&D work announced at IGORR-5 Nr. AffilationlName Topic 5a Klaus BoninglTUM Irradiation tests of highly enriched silicide fuel up to high fission densities in the silicide particles 5b H.J. DidierlSiemens Afterheat removal from highly loaded fuel elements cooled by natural convection, including flow reversal 5c Albert Lee/AECL Probabilistic safety analysis work to support major research reactor refurbishments 5d Kir Konoplev/PNPI Cold neutron source benchmark experiment on an existing critical facility 5e Edgar KoonenlSCK-CEN Embrittlement, fracture toughness and fatigue crack growth on irradiated series 5000 and 6000 aluminum alloys 5f Guy GistaU/Air Liquide Survey of operating cold neutron sources and their comparison based on indentical evaluation criteria 5g Jean-Luc Minguet/ Comparison of regulations for research reactors Technicatome in various participating countries Table 3 New R&D work just performed Nr. AffilationlName Topic 5h Kir Konoplev/PNPI Detritiation of 0 20 in connection with waste detritiation 3 worbM96 It might well be that we will hear more about several of these new R&D topics on the next IGORR meeting. 4. Conclusion This has been the fifth R&D workshop in the fifth IGORR meeting. The active discussions which developed and the numerous presentations which were given demonstrated once again that this "open market" for: the exchange of information seems to meet the demands and interests of the IGORR members, indeed. 4 worb0496 RESULTS OF ANS U3Si2 FUEL STUDIES presented to the 5th meeting of the International Group on Research Reactors D. L. Selby November 1996 » '0 '0 CD ::s U\ Q, )c' » The primary goal of the HANS-3 test was to compare specimens from hot-rolled plates to previous 'results for powder mixtures • HANS-l and -2 tests were cold pressed mixtures of fuel and Al powders • HANS-3 specimens were from hot-rolled plates with typical fuel! Al contact and interaction zones • Since the fuel performance model was based on these results, it was important to compare the hot-rolled plate results with the cold pressed mixture specimens • Most specimens were low volume fraction, HEU, U3Siz, U3Si, U30 S' or Ua~ • Some high volume fraction specimens of LEU and MEU were included to test lower fission rates and bumups 6" Microstructural examination confirmed the fuel performance model for the low volume fraction fuels • Diffusion of Al into the fuel particles was enhanced by the improved contact obtained from the hot-rolled plate specimens • Conclusions remain the same as for the earlier tests for U3Si2 and the backup fuels • U3Si2 expected to perform well at ANS conditions at temperatures up to 400°C at volume fractions up to 20% ;.J • High U3Si1 volume fraction MEU and LEU specimens faI•l ed at temperatures a..t°'lle. tlMlud about 250°C • Specimens were from existing plates at fuel volume fractions of 0.41 to 0.44 • Fission rates and temperatures for the failures were significantly above those of any existing AI-plate-type test reactor • Failure occurred only after complete depletion of the Al matrix due to particle growth and AI diffusion into the fuel 00 Appendix B R&D Topics from ANS Closeout of R&D Activities Doug Selby/Oak Ridge National Laboratory ORNL 4f: HANS-3 fuel capsule irradiation (in HFIR) evaluation a) G.L. Copeland, G.L. Hofman, and J.L. Snelgrove, "Postirradiation Examination of HANS-3 Capsule", ORNUM-4864, Memorandum to D.L. Selby, September 29, 1995. b) G.L. Hofman, J. Rest, and J.L. Snelgrove, "Aluminum-U3Si2 Interdiffusion and its Implications for the Performance of Highly Loaded Fuel Operating at Higher Temperatures and Fission Rates" (A Preliminary Assessment), Paper presented at the RERTR Conference in Seoul, Korea, October 1995. 4g: Meat fabrication with spherical powder fuel A paper was presented and published as part of the RERTR Conference held in Seoul, Korea, October 1995. General conclusions of this fabrication work were that somewhat better homogeneity was obtained for higher density fuels when the spherical powder fuel was used. 4h: Centering of fuel in plate fabrication Main report on this issue was never published. The purpose of this work was to examine the feasibility of fabricating the fuel plate so that the fuel meat was always centered in the plate even when the fuel was graded (I.e. the thickness of the fuel varied along the span of the plate). This has significant thermal-hydraulic advantages for very high power densi- ty systems, but is not considered important for most existing research reactors. The con- clusion of this study was that approximately centered fuel plates could be fabricated, but the process would add significant cost to the fabrication process. 4i: Effect of flow blockage shape a) T. Stovall et aI., "Flow Blockage Analysis for the Advanced Neutron Source Reactor", ORNL-6860, January 1996. b) JA Crabtree, ''The Effect of Alternate Inlet Flow Blockage Shapes on Heat Transfer and Flow Behavior in Rectangular Channels", Master's Thesis, the University of Tennessee, Knoxville. c) OK Felde, TK Stovall, and JA Crabtree, "Experimental Investigation of Flow Blockage Phenomena in the ANS Core", American Nuclear Society 1995 Annual Meeting, Philadelphia, June 25, 1995. 4j: Reduced pH effects on aluminum corrosion a) S.J. Pawel, D.K Felde, and R.E. Pawel, "Influence of Coolant pH on Corrosion of 6061 Aluminum Under Reactor Heat Transfer Conditions", ORNUTM-13083, October 1995. 4k: Final summation of the ANS thermal hydraulic test program a) G.L. Yoder et aI., "Update to Advanced Neutron Source Steady-State Thermal- Hydraulic Report", ORNUTM-12398/R1, November 1995 (published May 1996). b) N.C.J. Chen, M.W. Wendel, and G.L. Yoder, ''Transition to Natural Circulation With and Without Depressurization for the Advanced Neutron Source Reactor", Proceedings 1994 American Society of Mechanical Engineers (ASME) Winter Annual Meeting, Chicago, Nov. 6-11, 1994. c) G.E. Giles, "Advanced Neutron Source Reactor Thermal Analysis of Fuel Plate Defects", ORNUTM-13072, August 1995. d) M. Siman-Tov, D.K Felde, J.L. McDuffee, and G.L. Yoder, "Experimental Study of Static Flow Instability in Subcooled Flow Boiling in Parallel Channels", 4th American Society of Mechanical Engineers/Japanese Society of Mechanical Engineers, Maui, Hawaii, January 1995. e) M. Siman-Tov et aI., ''Thermal-Hydraulic Correlations and Experimental Database for the Advanced Neutron Source Reactor - Closing Report", ORNLlTM-13081. f) M. Simon-Tov et aI., "FY 1995 Progress Report on the ANS Thermal-Hydraulic Test Loop Operation and Results", ORNLITM-12972, November 1996. g) M.W. Wendel, N.C. Chen, and G.L. Yoder, "Updated Pipe Break Analysis for Advanced Neutron Source Reactor Conceptual Design", Proceedings 1994 American Society of Mechanical Engineers (ASME) Winter Annual Meeting, Chicago, Nov. 6-11, 1994. 41: Final Results from the HAN SAL Aluminum Irradiation Tests and Examination of BNL Data a) K Farrell, "Assessment of Aluminum Structural Materials for Service Within the ANS Reflector Vessel", ORNUTM-13049, August 1995. b) D.J. Alexander is in the process of preparing a final report on the HAN SAL aluminum irradiation studies performed for the ANS project. A draft report has been written and it is presently being reviewed. This report is expected to be ORNLlTM-13084 and should be published early in 1997. 4m: Irradiation Creep in Aluminum a) "Experimental Simulation of Radiation Creep in the ANS Core Pressure Boun- dary Tube", ORNLlTM-13063, August 1995 Appendix C ANS Fuel Plate Stability References Doug Selby/Oak Ridge National' Laboratory ORNL 1. W.F. Swinson, RL. Battiste, and G.T. Yahr, "An Experimental Investigation of the Interaction of Primary and Secondary Stresses in Fuel Plates", PVP-Vol. 338, Pressure Vessels and Piping Codes and Standards, Volume 1, ASME, 1996. 2. W.F. Swinson, RL. Battiste, L.R Luttrell, and G.T. Yahr, "An Experimental Investigation of the Structural Response of Reactor Fuel Plates", Experimental Mechanics, Volume 35-Number 3, September 1995. 3. W.F. Swinson, RL. Battiste, and G.T. Yahr, "Structural Thermal Tests on Advanced Neutron Source Reactor Fuel Plates", ORNLlTM-13062, August 1995. 4. W.F. Swinson, RL. Battiste, and G.T. Yahr, "Circular Arc Fuel Plate Stability Experi- ments and Analyses for the Advanced Neutron Source", ORNLlTM-12977, August 1995. 5. W.F. Swinson, L.R Luttrell, and G.T. Yahr, "An Examination of the Elastic Structural Response of the Advanced Neutron Source Fuel Plates", ORNLlTM-12712, 1994. 6. W.F. Swinson, RL. Battiste, L.R Luttrell, and G.T. Yahr, "Follow-up Fuel Plate Stability Experiments and Analyses for the Advanced Neutron Source", ORNLITM-12629, November 1993. 7. W.F. Swinson et aI., "Fuel Plate Stability Experiments and Analyses for the Advanced Neutron Source", Journal Pressure Vessel Technology, American Society of Mechanical Engineers, July 1993. 8. W.F. Swinson et aI., "Structural Response of Reactor Fuel Plates to Coolant Flow", Pressure Vessels and Piping, Vol. 258, Flow-Induced Vibration and Fluid-Structure Interaction, 21-33, American Society of Mechanical Engineers, 1993. 9. W.F. Swinson, RL. Battiste, L.R Luttrell, and G.T. Yahr, "Fuel Plate Stability Experi- ments and Analyses for the Advanced Neutron Source", ORNLlTM-12353, May 1993. 10. W.K Sartory, "Nonlinear Analysis of Hydraulic Buckling Instability of ANS Involute Fuel Plates", ORNLlTM-12319, March 1993. The above list provides the major ANS reports dealing with plate stability issues. The Space-Time Kinetics of the FRM-II Reactor C. OOderlein 1 The importance of multidimensional or space-time kinetics in large power reactor-cores is a well-known fact. The size of these cores and the weak neutronic coupling of their different parts limit the applicability of the conventional point kinetics in the analysis of localized reactivity insertions (eg. control rod ejection). Yet, the analysis of the very compact, 020-reflected cores of modern research reactors too reveal a space-time kinetics phenomenon. This effect, which has its cause in the small core size and the long lifetime of thermal neutrons in heavy water (020), has been studied at the instance of the FRM-ll's compact core KKE7 by means of analytical and Monte Carlo neutronics calculations. Eventually, a new method of kinetics calculations, which covers this phenomenon, has been developped. Transient calculations with this method allowed to quantify the effect's contribution ot the inherent safety of the FRM-II reactor. 1) The physical basis of the reflected compact core kinetics phenomenon The compact core concept, as it is going to be realized in the FRM-II reactor near Munich (Germany) [1], consists in a small annular-cylindrical core of 0.7 m height and 0.2 m diameter, which is placed in the center of a 020-filled moderator tank of 2 m diameter. This arrangement, called "inverse flux trap", leads to the formation of a maximum of the thermal neutron flux outside the core, which is essential for the reactor's vocation as beam tube neutron source. By choosing light water (H20) as in-core moderator and coolant, the core size could be further minimized for the sake of performance, thus yielding an unperturbed flux maximum of about 8.10 '4 cm-2 S-1 with a reactor power of 20 MW. Figure 1 shows schematically the average histories of the neutrons in the chain reaction cycle, detailled in energy (vertical dimension, with high energy at the top) and space (horizontal dimension, simplified in core- and moderator tank-region). Of 100 fission neutrons, about 73 escape from the core with high energy into the moderator tank. Whereas about 25 of these neutrons are almost immediately reflected back into the core, the remaining are moderated in the heavy water, whereupon a part (about 18) re-enter the core by diffusion. These moderator tank neutrons, moderated in the 020, account for about 30% of the fissions that give rise to the next generation of fission neutrons. There are hence two distinct moderation environments, with very contrasting neutronic properties: inside the core, with its light water and high enriched uranium, that limit the average prompt neutron lifetime to some 15 microseconds, and the moderator tank, the 020 of which permits the thermal neutrons to diffuse about two milliseconds before re-entering the core. 1 current address: CEA Cadarache, DRS/SEA, F-13108 Saint-Paul-Lez-Durance, France 1 reactor-core moderator-tank 100 fast - - - - - - - - - energy range - thermal fission generated neutrons = 100 Figure 1: Schematlcal depletion of the neutron life cycle in the KKE7. Decreasing bar- width symbolizes absorption- and leackage-Iosses, circle sections represent the contributions to the production of the next generation of fission neutrons; the thermal energy limit is 0.6 eV. The part "reactor-core" comprises the control rod region. Therefore, a part of the 34.3 neutrons moderated in this path, serv for reactor control, i.e. are absorbed In the absorber section of the control rod. At steady state power, these two paths of moderation are in equilibrium and the "lateness" of the moderator tank neutrons has no effect. In a power rise, however, the re-flux of thermal neutrons from the moderator tank will follow the time evolution of the neutron population in the core with a certain delay. Given the importance of the moderator tank neutrons for criticality (30% of all fissions!), this delay will have an impact on the reactor's kinetics in form of an additional "inertia", opposed to rapid changes of reactor power. As this phenomenon is not covered by the conventional point kinetics, a new calculation scheme had to be developped to quantify the significance of this effect for the reactor's kinetic properties. 2 2) The two points-two groups kinetics model .. .tl.S.Sl.o.n .n.e.ut.ro.n.s. . v :<~~1. fast lei core group Function :~~c~. thermal Ic2 core group Figure 2: Schematical Illustration of the two points-two groups kinetics model with the continous Reflector Response Function G(t}. The arrows represent the neutronical coupling; the k-symbols designate coupling effiCiencies, the J..symbols stand for transfer delays. Initially based on a proposal of Difilippo et.a!. [2], a specialized model of the compact core kinetics has been developped [1] (fig 2). The problem is separated in two spatial regions or "points", namely the core and the moderator tank. The neutron population of the core region is represented in two energy groups, fast and thermal, whereas the kinetic properties of the reflector tank are described by the so-called " Reflector-Response-Function" (RRF). This time dependent function Gte), first proposed by Shinkawa et.a!. [3] and depicted in figure 3 for the case of the FRM-II, relates the re-flux of thermal neutrons r (t) from the tank into the core at time t to the outflow j + (t) of fast neutrons at an earlier time tJt: co F(t) = Jr (t-t)·G(t) dt . o In other words: a B-pulse of neutrons, which leave the core with high energy at the instance t, will return moderated into the core later at time t+'t with an intensity proportional to G('t). Moderation and diffusion in the moderator tank stretch the B-pulse to the form depicted by the RRF. The first moment G1 of the normalized response function 00 G] = J-c ·G(-c)d-c o gives the average delay of the thermal moderator tank neutrons, which equals to G1= 1.8·10-3 s for the FRM-II. 3 - Reflector-Response-Function 10'" 10 1 time 1 [S) Figure 3: Unear-Iog plot the the Reflector Response Function G(t) of the FRM-II, without experimental installations in the moderator tank. The function has been determined by Monte Carlo calculations and is normalized to 00 JG ('t) d't = 1. o The dotted lines designate the statistical 1 a-error Intervall of the Monte Carlo calculation. The distinct elements of this model (Le. points and groups) are linked by coupling factors so as to represent the actual flow of neutrons (cf. fig. 1 and 2). The coupling factors consist of coupling efficiencies (k in fig. 2) and time constants (~, which designate an average delay associated with the transfer of neutrons from one point or group to another. The model can be understood as a variant of the points kinetics method, enhanced by a time-varying core spectrum and a delayed source term2• It can be formulated by a system of coupled differential equations, which lend itself easily to numerical integration. The reflector response function G('e) is thereby represented by its moments Gn, with an order n up to four. The values of these moments, as well as those of the coupling factors, can be determined by means of Monte Carlo calculations and a statistical analysis of the neutron histories simulated. In order to validate the factors obtained in this way, an analytical model of time-dependent core neutronics in cyclindrical geometry has been developed. Its results, obtained with minimum computational effort, correspond in first and second order to those of the expensive Monte Carlo calculations. 2 In Diflippo's original proposal, the moderator tank was, like the core, represented by two coupled points. This model, however, proved to be unable to reproduce the rather complex time structure of the real response function. The same objection applies to the model of Ott, who proposed to treat the delayed moderator tank neutrons like the conventional delayed neutrons, that is by families of "pseudo delayed neutrons." 4 3) Transient calculation of the FRM-II klntlc behaviour The concept of the FRM-II reactor as an university research tool and its location near to an urban area imposed a rigorous safety philosophy from the very beginning of the design process. This philosophy favoured a simple and inherently safe lay-out, as for example an unpressurized primary coolant loop without active control valves. This attitude has led, as well, to an elimination of all effects that might result in an uncontrolled fast injection of reactivity. The assumptions for the reactivity insertion transients, calculated with the two points-two groups model and presented here as an example of its application, were therefore completly hypothetical. The calculations show that in the case of subprompt reactivity injections3, the conventional delayed neutrons dominate the excursion. The "inertial" effect of the delayed moderator tank neutrons appears late in the transient, at a point where the active shutdown devices would already have made subcritical the reactor. For superprompt over-reactivites, however, the space-time kinetic phenomenon is visible from the beginning, manifesting itself in a slowing-down of the power excursion (compared to a simple point kinetics calculation). The use of the new calculation model is therefore recommended for the analysis of this kind of transient. The case of the theoretical maximum reactivity insertion corresponds to an instantaneous displacement of the central control rod from the highest to the lowest efficiency position. The reactor periode resulting from this unrealistic assumption, as calculated with the two points-two groups model, lies well above the critical value of 4 ms. This critical value is believed to be, for this kind of reactor, a threshold for the BORAX phenomenon. By exceeding this threshold, the compact core KKE7 is physically unable to give rise to a power excursion fast enough to produce the BORAX effect. The two points-two groups method hence demonstrates the inherent safety of the FRM-II reactor for this extreme and virtually impossible assumption. References: [1] Chr. DOderlein: "Dynamics and kinetics of a reflected research reactor core," Dissertation at the Technical University Munich, 1994 (in German) [2] F. Difilippo, M. Abu-Shehadeh, R. Perez: "Two-Point and Two-Energy Group Kinetics Model of the ANS Reactor," Trans.Am.Nucl.Soc., 59, 347 (1989) [3] M. Shinkawa, Y. Yamane, K. Nishina, H. Tamagawa: "Theoretical Analysis fa Coupled-Core Reactors with the Method of the Moderator Region Response Function," Nucl.Sci.&Eng., 67, 19-33 (1978) 3 injected reactivity p < 1 $ 5 SESSION 6 BUSINESS MEETING AND CLOSING SESSION BUSINESS SESSION Colin West chaired the Business Session of 100RR-V. He resigned as Chairman of IOORR (after six years of service) and conveyed the resignation of the Technical Program Coordinator. Kathy Rosenbalm. who was not present at the meeting. Professor Klaus Bijning. a leader of the FRM-ll Project, was elected as the new Chairman by unanimous acclamation. A vote of thanks was ~de to the fonner officers. The sixth IGORR meeting. according to the agreed-upon rotation (America, Europe. Asia) should be held in Asia, and the new chairman will investigate the possible host organizations. Grateful thanks were expressed to the French organizers of lGORR-V and to Technicatome for their support. Colin D. West LIST OF PARTICIPANTS LIST OF PARTICIPANTS November 04-06, 1996 ARGENTINA INVAP DENMARK RISOE NATIONAL LABORATORY LOLJCHJos6 NIELSEN Kristen H. Moreno 1089 Reactor DR 3-214 P.O. Box 49 8400 Bariloche DK-4000 Roskilde Phone: 54.944.22121 Phone: 45.46.77.43.07 Telefax: 54.944.23051 Telefax : 45.46.75.50.52 AUSTRALIA ANSTO FRANCE FERRY Roger KIM Sangjoong (Sbane) Villa nO 8 - Petite Arcadie, Chemin de Banon Lucas Heights Research Laboratories 13100 AIX EN PROVENCE New lllawara Road Phone: 33.4.42.23.07.27 Lucas Heights, NSW Telefax : None Private Mail Bag 1 Menai, NSW 2234 FRANCE AIRLIQUIDE Phone: 61.2.9717.3181 GISTAUGuy Telefax: 61.2.97179269 BP15 38360 SASSENAGE AUSTRIA IAEA Phone: 33.4.76.43.60.36 ALCALA-RUIZ Francisco Telefax: 33.4.76.43.61.31 Wargramerstrasse 5, P.O. 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