Assessment of improved fission-product transport models in VICTORIA against the ORNL HI and VI tests

Conference · Tue Jan 01 00:00:00 EST 1991

OSTI ID:5544181

Domagala, P; Rest, J; Zawadzki, S A

New models for the release of fission-products from fuel have recently been incorporated into a comprehensive code, VICTORIA, for the prediction of radionuclide behavior under severe reactor accident conditions as an alternative to a simple Booth diffusion calculation. It is widely known that the Booth model has severe limitations when used under conditions of changing temperature and power. A new transport model based on a two node diffusive flow formulation has been implemented into VICTORIA. In addition to the diffusive flow model, other mechanisms such as grain growth, grain boundary sweeping and intergranular bubble behavior are taken into account. These physically based models focus on fission-product behavior in intact fuel geometries. While the VICTORIA program is primarily concerned with the behavior of severely degraded fuel geometries, the capability for accurately characterizing fission-product release from intact fuel must be demonstrated before other geometries can be reliably modeled. Results of VICTORIA simulations are compared with the results of in-cell heating tests on irradiated fuel. The fission-product transport models involved in these simulations are assessed and their predictive capability examined. 10 refs., 7 figs., 2 tabs.

• Research Organization: Argonne National Lab., IL (United States)
• Sponsoring Organization: USNRC; Nuclear Regulatory Commission, Washington, DC (USA)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 5544181
• Report Number(s): ANL/CP-72066; CONF-910739-24; ON: DE91014460; TRN: 91-021354
• Resource Relation: Conference: ASME/AIChE/ANS national heat transfer conference, Minneapolis, MN (United States), 26-31 Jul 1991
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
FISSION PRODUCTS
RADIATION TRANSPORT
NUCLEAR POWER PLANTS
REACTOR ACCIDENTS
FISSION PRODUCT RELEASE
CESIUM
COMPUTERIZED SIMULATION
F CODES
FLOW MODELS
FUEL ELEMENTS
KRYPTON
MATHEMATICAL MODELS
RADIOISOTOPES
REACTOR SAFETY
T CODES
V CODES
ACCIDENTS
ALKALI METALS
COMPUTER CODES
ELEMENTS
FLUIDS
GASES
ISOTOPES
MATERIALS
METALS
NONMETALS
NUCLEAR FACILITIES
POWER PLANTS
RADIOACTIVE MATERIALS
RARE GASES
REACTOR COMPONENTS
SAFETY
SIMULATION
THERMAL POWER PLANTS
220502* - Nuclear Reactor Technology- Environmental Aspects- Radioactive Effluents
220900 - Nuclear Reactor Technology- Reactor Safety
990200 - Mathematics & Computers