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Decay Heat Removal and Heat Transfer Under Normal and Accident Conditions in Gas Cooled Reactors

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Decay Heat Removal and Heat Transfer Under Normal and Accident Conditions in Gas Cooled Reactors IAEA-TECDOC-757 Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors Proceedings of a specialists meeting held in Jülich, Germany, 6-8 July 1992 INTERNATIONAL ATOMIC ENERGY AGENCY The originating Sectio f thino s documen IAEe th An i t was: Nuclear Power Technology Development Section International Atomic Energy Agency Wagramerstrasse5 P.O. Box 100 A-1400 Vienna, Austria DECAY HEAT REMOVA HEAD LAN T TRANSFER UNDER NORMAL AND ACCIDENT CONDITIONS IN GAS COOLED REACTORS IAEA, VIENNA, 1994 IAEA-TECDOC-757 ISSN 1011-4289 Printe IAEe th AustriAn i y d b a August 1994 PLEASE BE AWARE THAT ALL OF THE MISSING PAGES IN THIS DOCUMENT WERE ORIGINALLY BLANK FOREWORD The Specialists Meeting on Decay Heat Removal and Heat Transfer under Normal and Accident Conditions in Gas Cooled Reactors was held at the KFA Research Center, Jülich, Germany, 6-8 July 1992. The meeting was convened by the International Atomic Energy Agency on the recommendation IAEA'e ofth s International Working Coole s GrouGa n dpo Reactors attendes wa t .I participant y db s from China, France, Germany, Japan, Poland Russiae th , n Federation, Switzerland e Uniteth , d Kingdo Unitee th d mdan State f Americaso meetine Th . chaires g. Kugelewa K Profy d. b d Dr . an r Prof. Dr. E. Hicken, Directors of the Institute for Safety Research and Reactor Technology of the ResearcA KF h Center covered an , followinge dth : Design and licensing requirements for gas cooled reactors; Concepts for decay heat removal in modern gas cooled reactors; Analytical methods for predictions of thermal response, accuracy of predictions; Experimental data for validation of predictive methods: • Operational experience from gas cooled reactors, • Experimental data from test facilities. IAEA activitie advancen si cooles dga d reactor technology developmen conductee ar t d within the IAEA's nuclear power programme. Advance s coolega d d reactor designs currently under development are predicted to achieve a high degree of safety through reliance on innovative features and passive systems. The IAEA's activities in this field are focusing on the four technical areas which provide advanced gas cooled reactors with this high degree of safety, but which must be proven. These technical areas are: (a) The safe neutron physics behaviour of the reactor core; (b) Reliance on ceramic coated fuel particles to retain the fission products even under extreme accident conditions; abilitdesigne e th Th f y o dissipat o st ) (c e decay hea naturay b t l heat transport mechanismsd an ; (d) The safe behaviour of the fuel and reactor core under chemical attack. investigato T thermae eth l behaviou sucf ro h advanced designs, experimental investigationd san analytical studies are ongoing in several countries to confirm the ability of the reactor designs to dissipate the decay heat from the core by natural heat transport mechanisms without reaching excessive fuel temperatures. Experiments have been performed to obtain data on heat transport phenomena (e.g. natural convection heat transport both within and outside the reactor vessel). Experienc hean ei t transport under actual reactor condition bees sha n obtaine R Magnon di AG d xan reactor Unitee th n si d Kingdom, Franc HTGRJapann d i ean d Germanan ,n si USAe th d . yFuturan e efforts will focus on further development and validation of analytical tools, identifying safety margins, understanding the effect on system performance of uncertainties in key heat transport parameters, system optimization and experimental confirmation and demonstration of the predicted performance of passive systems. summaryn I , considerabl operatinR eGC g experience fielexiste decaf th do i sh y heat removal and heat transport advancer fo d an , d designs, experimenta analyticad an l l activitie o undet e sy ar rwa investigate key heat transport phenomena. A key conclusion of this Specialists meeting was that for advancee th cooles dga d reactor designs currently under development predictee th , d performancf eo passive systems for heat removal needs to be proven under experimental conditions representing realistic reactor conditions. EDITORIAL NOTE In preparing this document for press, staff of the IAEA have made up the pages from the original manuscripts submittedas authors.the viewsby The expressed necessarilynot do reflect those of the governments of the nominating Member States or of the nominating organizations. The use of particular designations of countries or territories does not imply any judgement by publisher,the legalthe IAEA,to the status as of such countries territories,or of their authoritiesand institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does implyintentionnot any infringeto proprietary rights, should construednor be it an as endorsement recommendationor parithe IAEA. ofon the The authors are responsible for having obtained the necessary permission for the IAEA to reproduce, translate materialuse or from sources already protected copyrights.by CONTENTS SUMMARY OF THE SPECIALISTS MEETING ............................ 7 DESIGN AND LICENSING REQUIREMENTS FOR GCRs (Session I) Principles of decay heat removal in reactor technology — Present status and future prospects ............................................... 15 KugelerK. Safety criteria and provisions for the evacuation of residual heat from graphite gas cooled reactors ................................................ 29 R. Lheureux, A. Aguilera Bases for the MHTGR source term and containment concepts .................... 41 P.M. Williams ultimate Th e HTR-Module safetth f yo e during hypothetical accidents ..............7 4 . G.H. Lohnen CONCEPT DECAR SFO Y HEAT REMOVA MODERN LI N GCRs (Sessio) nII Distribution of the decay heat in various MODUL HTRs and influence on peak fuel temperatures ............................................... 53 E. Teuchert, K.A. Haas, A. van Reek, P.R. Kasten Afterheat removal for HTR-10 test module under accident conditions ............... 63 Gao Zuying, Shuyan,He ThongMin Flow scheme desigd san n feature HTGf o s R residual heat removal systems ...........2 7 . V.F. Golovko, A.I. Kiryuschin, N.G. Kuzavkov Passive deca residuad yan l heat remova MHTGRe th n i l .......................5 7 . D.A. Billing, S.K. Chose, J.M. Berkoe, S.A. Caspersson, G.C. Bramblett Impact of increasing MHTGR power on passive heat removal .................... 83 T.D. Dünn, A.A. Schwanz, P.A. Silady Trends in safety criteria for future reactor plants ............................ 89 HickenE. ANALYTICAL METHODS FOR PREDICTIONS OF THERMAL RESPONSE, ACCURACY OF PREDICTIONS (Session III) Analytical and experimental investigations of the passive heat transport in HTRs under severe accident conditions ......................................... 95 W. Rehm, H. Barthels, W. John, J. Cleveland, M. Ishihara Presentation of decay heat removal computer codes used for gas cooled reactors ......... 104 G. Carvallo, Dobremelle,M. MejaneA. Modelin analysid gan heaf so t transfer fro MHTGe mth R core throug hsteea l reactor vessel to the reactor cavity cooling system ................................... Ill D.A. Dilling, J.M. Berkoe, S.K. Ghose, T.D. Dunn, S.A. Caspersson Analysi f afterheaso t removal from modular HTGRs during accidents concepa d an r , fo t f sphericao e us l fuel element LWRn si s ...............................7 11 . 7.5. Mosevitskij, A.O. Goltsev, P.V. Mikhailov, V.F. Tsibulskij, V.D. Davidenko, V.S. Popov, Yu.N. Udyanskij EXPERIMENTAL DAT VALIDATIOR AFO PREDICTIVF NO E METHODS— OPERATIONAL EXPERIENCE FROM GCRs, EXPERIMENTAL DATA FROM TEST FACILITIES (Session IV) Heat transfe uppee th n HTTe ri r parth f Ro t pressure vessel during los forcef so d5 coolin12 . g Y. Shiina, M. Hishida Development of an inactive heat removal system for high temperature reactors ......... 131 K. Kugeler, M. Sappok, B. Beine, L. Wolf Passive heat removal experiment advancen a r sfo d HTR-module reactor pressure vessel and cavity design .............................................. 139 L. Wolf, A. Kneer, R. Schulz, A. Giannikos, W. Hafner Test apparatus of cooling panel system for MHTGR .......................... 147 Takada,S. Suzuki,K. Inagaki,Y. MiyamotoY. SANA experiments related to self-operating removal of decay heat ................. 151 H.F. Nießen, M. G. Lange Lis Participantf o t s ..............................................1 16 . SUMMARY OF THE SPECIALISTS MEETING The requirements for decay heat removal are met at current reactors by active heat removal systems. While these systems have prove highle b o nt y reliable, their failur lean coro eca dt e melt accidents. Research and development activities worldwide are now showing that design of the reactor and plant for self-acting decay heat removal to avoid core melt accidents and guarantee retention of fission products insid fuee eth l elements, activl eveal f ni e heat removal systems failactualln ca , e yb realized. Self-acting decay heat remova lapproac w e elemensystemne on a e nucleao n ht i ar st r engineering which is termed 'catastrophe-free' nuclear engineering. By this it is meant that major accidents in category 7 (catastrophic accident, e.g. Chernobyl) of the INIS scale are deterministically excluded. Experimental demonstration performance prooe th th f f o fo s f self-actino e g systeme sar required. In Japan the gas cooled reactor activities are focused on the High Temperature Engineering Test Reactor (HTTR) project. Constructio HTTe Oarae th th f nt Ro a i Research Establishmen Japae th f no t Atomic Energy Research Institute has been under way since March 1991 and first criticality is expecte 1997n di . Thi MW(th0 s3 ) reactor will produce core outlet temperature 850°f so t rateCa d operation and 950°C at high temperature test operation. It will be the first nuclear reactor in the worlconnectee b o dt higa o dt h temperature process heat utilization system reactoe Th . r wile b l utilized to establish basic technologies for advanced HTGRs, to demonstrate nuclear process heat application o servt d en irradiatio an a ,als s a o n test facilit r researcfo y n higi h h temperature technologies.
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