FORSCHUNGSZENTRUM JÜLICH GmbH KFAJ Institut für Sicherheitsforschung und Reaktortechnik Methods and Data for HTGR Fuel Performance and Radionuclide Release Modeling during Normal operation and Accidents for Safety Analysis K. Verfondern R. C. Martin R. Moormann Berichte ties Forschungszentrums Jülich ; 2721 ISSN 0366-0885 Institut für Sicherheitsforschung und Reaktortechnik Jül-2721 Zu beziehen durch : Forschungszentrum Mich GmbH - Zentralblbllothek Postfach 1513 - D-5170 Mich - Bundesrepublik Deutschland Telefon: 02461f61-6102 - Telefax : 02461161-6103 - Telex; 833556-70 kfa d Methods and Data for HTGR Fuel Performance and Radionuclide Release Modeling during Normal Operation and Accidents for Safety Analysis K. Verfondern') R. C. Martin 2) R. Moormann') Research Center Jülich 2) Oak Ridge National Laboratory Research Center Jülich GmbH - 1SR Jül - 2721 January 1993 METHODS AND DATA FOR HTGR FUEL PERFORMANCE AND RADIONUCLIDE RELEASE MODELING DURING NORMAL OPERATION AND ACCIDENTS FOR SAFETY ANALYSES by K. Verfondern (Research Center Jülich) R. C. Martin (Oak Ridge National Laboratory) R. A7oortnann (Research Center Jülich) ABSTRACT The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of know- ledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying da- tabase for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed . Forschungszentrum Jülich GmbH - ISR Jül - 2721 Januar 1993 RECHENMETHODEN UND DATEN ZUM HTR-BRENNSTOFFVERHALTEN UND ZUR. SPALTPRODUKTFREISETZUNG IM NORMALBETRIEB UND STÖRFALL IM RAHMEN VON SICHERHEITSANALYSEN voll K. Verfondern (Forschungszentrum Jülich) R. C. Martin (Oak Ridge National Laboratory) R. Moormann (Forschungszentrum Jülich) KURZFASSUNG Der im Jahre 19137 erschienene Statusbericht mit der Beschreibung einer Referenz-Datenbasis und Rechenmodellen ist auf den neuesten Stand gebracht worden und beschreibt den gegenwärtigen state-of-the-art im deutschen HTR-Programm . Der Inhalt des Statusberichts ist erweitert worden um Informationen aus den HTR-Programmen anderer Nationen, um einen Vergleich der Analysemethoden sowie der zugrunde liegenden Datenbasis zur Beschreibung des Brennstoffverhaltens und des Spaltprodukttransports zu ermöglichen . Der Bericht umfaßt die Bereiche Freisetzung und Transport von Spaltprodukten während des Normalbetriebs sowie im Verlaufe von Unfallszenarien von Kcrnauflieizung, Wasser- und Lufteinbruch und Druckentlastung. Table of Contents 1 .0 INTRODUCTION . .I 2.0 HTGR FUEL BEHAVIOR DURING NORMAL OPERATION . 3 2.1 FUEL DESIGN . .3 2.1 .1 Coated Particle . 3 2.1 .2 Fuel Element . 10 2.1 .3 HTGR Fuel Quality . 11 2.1 .4 Fission Product Inventories . 13 2 .2 COATED PARTICLE IRRADIATION PERFORMANCE . 13 2.2.1 Kernel Migration (Amoeba Effect) . , . , . 17 2.2.2 Fission Product Interaction With Silicon Carbide . 18 2 .2.3 Pressure Vessel Failure . 19 2 .3 RADIONUCLIDE RELEASE DURING NORMAL OPERATING CONDI- TIONS . .19 2 .3 .1 Metallic Fission Product Release . 22 2.3.2 Uptake of Fission Product Metals by SiC , . 24 2 .3 .3 Fission Gas Release . 25 2.3 .4 Fission Product Activity Distribution in the HTGR Primary Circuit at Acci- dent Initiation . 32 3.0 HTGR FUEL BEHAVIOR DURING CORE HEATUP ACCIDENTS . 41 3 .1 THERMODYNAMICAL BOUNDARY CONDITIONS . 41 3.2 COATED PARTICLE PERFORMANCE UNDER ACCIDENT CONDI- TIONS . .42 3 .3 RADIONUCLIDE RELEASE FROM COATED PARTICLES UNDER AC- CIDENT TEMPERATURE CONDITIONS . 47 3.3 .1 Metallic Fission Product Release . 47 3.3.2 Fission Gas and Iodine Release . 51 3.3.3 Particle Failure Model Discussion . 53 3.4 FISSION PRODUCT TRANSPORT WITHIN THE CORE CAVERN . 54 4.0 HTGR FUEL BEHAVIOR DURING WATER AND AIR INGRESS kCCI- DENTS . .61 4.1 Fission Product Release from Defective Coated Particles . 61 4.2 Fission Product Release from Graphite . 63 4.3 Massive Long Term Air Ingress . 66 4.4 Release of Fission Products Plated-Out on Metal Surfaces . 68 4.4.1 Mobilization by Liquid Water . 68 Table of Contents iii 4.4.2 Mobilization by Steam or Air Attack . 69 4.4.3 Exemplary Source Term Contributions Caused by Mobilization of Plateout Activity . .73 4.5 Fast Reactivity Transients in Combination with Water Ingress . 74 5.0 HTGR FUEL BEHAVIOR DURING DEPRESSURIZATION ACCIDENTS . 75 5.1 Desozption of Plateout Activity due to Pressure Drop . 75 5.2 Liftoff of Dust-Borne Activity . 75 5.3 Source Terms in Depressurization Events . 80 6.0 ACKNOWLEDGEMENT . 83 7.0 REFERENCES . 85 Appendix A. TRANSPORT DATA FOR DIFFUSION MODEL . 105 A. I FUEL KERNEL . 107 A.2 PYROCARBO'N . l08 A.3 SILICON CARBIDE . 110 A.4 GRAPHITE . 113 A.4.1 Matrix Graphite . 113 A.4.2 Structural Graphite . 116 A.4.3 Concentration Dependence of the Diffusion Coefficient . 117 A.5 ZIRCONIUM CARBIDE . 120 Appendix B. INPUT DATA FOR PARTICLE FAILURE MODELS UNDER AC- CIDENT CONDITIONS . 141 Appendix C. SORPTION ISOTHERMS OF FISSION PRODUCTS OVER GRAPHITIC SURFACES . 149 LIST OF ABBREVIATIONS AGR Advanced Gas-Cooled Reactor AVR Arbeitsgemeinschaft Versuchs-Reaktor BISO Buffer Isotropic Coating (Buffer and Pyrocarbon Layers) BOL Beginning-of-Life CAGR Commercial Advanced Gas-Cooled Reactor CEGB Central Electricity Generating Board EFPD Equivalent Full Power Day EOL End-of-Life FIMA Fissions per Initial Metal Atoms FRG Federal Republic of Germany FZ Forschungszentrum (Rossendorf) GA General Atomics HFR High Flux Reactor (Fetten) HOBEG Hochtemperaturreaktor- Brennelement GmbH HRB Hochtemperatur-Reaktorbau GmbH HTGR High-Temperature Gas-Cooled Reactor HTI High-Temperature Isotropic HTTR High-Temperature Engineering Test Reactor IAEA International Atomic Energy Agency INET Institute for Nuclear Energy Technology JAERI Japanese Atomic Energy Research Institute KFA Forschungszentrum (Jülich) KfiFA Kühlfinger-Versuchsapparatur LEU Low-Enriched Uranium LTI Low-Temperature Isotropic MHTGR Modular High-Temperature Gas-Cooled Reactor MIT Massachusetts Institute of Technology MTR Material Test Reactor SEM Scanning Electron Microscopy THTR Thorium-Hochtemperaturreaktor TRISO Tristructural Isotropic Coating (Buffer, SiC, PyC) VGM Modular HTGR Design of the Russian Federation VHTR Very High Temperature Reactor LIST OF ABBREVIATIONS v vi 1.0 INTRODUCTION Much experience has been gained since the 1960s in the operation of high- temperature, gas-cooled reactors (HTGRs), beginning with the experimental Peach Bottom reactor in the United States (US), continuing with the Dragon re- actor project in the United Kingdom (UK) and the AVR project in the Federal Republic of Germany (FRG), and resulting in the construction and operation of the prototype reactors Fort St. Vrain in the US and the Thorium High- Temperature Reactor (THTR-300) in Germany for the production of electricity. Unfortunately, for a combination of technical, political, and economic reasons, these reactors are no longer in operation and no more are planned in the near future in Germany. The experience with these reactors has consistently demonstrated the safety mar- gins inherent in HTGR design, even under adverse circumstances and accident scenarios. The safety features and the economic viability of compact reactor de- signs continue to attract international interest in the HTGR concept. Although the German HTGR program has a long and productive history, the HTGR as a national priority has been deemphasized for the future. The US maintain an active research program in the modular-HTGR (MHTGR) design, but a long- term commitment to the construction of new HTGRs has not been made at the present time. In contrast, other nations have expressed a serious near-term commitment to construction of small demonstration HTGRs. Construction of the 30 MW(th) High-Temperature Engineering Test Reactor (HTTR) is currently underway in Japan . The People's Republic of China is moving ahead with plans for a 10 MW(th) Test Module HTGR to be constructed later this decade. HTGR-related experimentation and research continues in other countries, most notably the Russian Federation, the UK, and France. The changing national priorities in HTGR development suggest that interna- tional cooperation in documenting the existing design database and analysis of existing models and codes used in safety studies could enhance the prospects for future licensing of HTGR reactors around the world. Each national HTGR program has specific strengths and emphases, and comparison of the current international state of knowledge can be a cost-effective approach to defining pri- 1.0 INTRODUCTION ority data and analysis needs as well as providing a cooperative forum for vali- dation of predictive safety codes using data from present or future experiments. In 1987, a status report on reference data and models for fission product trans- port to be used in German HTGR safety analyses was issued (Ref. 1). This sta- tus report summarized the German experience in the development and quality assurance of fuel, to be used as the basis for validating calculational models which can reproduce experimental results and predict the fission product release behavior under normal operating and accident conditions.