Pretest predictions of the Fast Flux Test Facility Passive Safety Test Phase IIB transients using United States derived computer codes and methods

Conference · Wed Jul 01 00:00:00 EDT 1992

OSTI ID:7281319

Heard, F J; Harris, R A; Padilla, A

The SASSYS/SAS4A systems analysis code was used to simulate a series of unprotected loss of flow (ULOF) tests planned at the Fast Flux Test Facility (FFTF). The subject tests were designed to investigate the transient performance of the FFTF during various ULOF scenarios for two different loading patterns designed to produce extremes in the assembly load pad clearance and the direction of the initial assembly bows. The tests are part of an international program designed to extend the existing data base on the performance of liquid metal reactors (LMR). The analyses demonstrate that a wide range of power-to-flow ratios can be reached during the transients and, therefore, will yield valuable data on the dynamic character of the structural feedbacks in LMRS. These analyses will be repeated once the actual FFTF core loadings for the tests are available. These predictions, similar ones obtained by other international participants in the FFTF program, and post-test analyses will be used to upgrade and further verify the computer codes used to predict the behavior of LMRS.

• Research Organization: Westinghouse Hanford Co., Richland, WA (United States)
• Sponsoring Organization: USDOE; USDOE, Washington, DC (United States)
• DOE Contract Number: AC06-87RL10930
• OSTI ID: 7281319
• Report Number(s): WHC-SA-1483; CONF-921003-4; ON: DE92019669
• Resource Relation: Conference: International conference on design and safety of advanced nuclear power plants, Tokyo (Japan), 25-28 Oct 1992
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
LMFBR TYPE REACTORS
LOSS OF FLOW
EXPERIMENT PLANNING
FFTF REACTOR
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
REACTOR CHANNELS
REACTOR CORES
REACTOR FUELING
REACTOR SAFETY
S CODES
STRESS ANALYSIS
TRANSIENTS
ACCIDENTS
BREEDER REACTORS
COMPUTER CODES
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
LIQUID METAL COOLED REACTORS
MECHANICS
PLANNING
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
SAFETY
SODIUM COOLED REACTORS
TEST REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
Breeding
990200 - Mathematics & Computers