Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1976
Light water reactor safety research performed April through June 1976 is discussed. Results of a reflood heat transfer test series were analyzed to determine the reflood heat transfer characteristics of the Semiscale Mod-1 electrically heated core. A special test to investigate the sensitivity of departure from nucleate boiling to a change in core behavior was conducted with the Semiscale Mod-1 facility. Two loss-of-coolant experiments, including an investigation of hot-wall delay effects, were conducted in LOFT. Preliminary results for the PCM-2 and IE-1 Tests in the Power Burst Facility and postirradiation examination results from the IE-ST-1 and IE-ST-2 Tests are given. Model development activities include continued development of RELAP4/MOD6, development of a two-velocity thermal equilibrium flow model, and development of BEACON code mixed-dimensional multiregion coupling and equation-of-state routine for water. LOCA analysis verification activities consisted of studies of WHAM code modeling and uncertainty analysis methods for a LOCA analysis computer code. Fuel behavior verification efforts concentrated on a UO/sub 2/ and mixed-oxide swelling model, uncertainties in the expressions for cladding expansion and specific heat, modification of the zircaloy plastic deformation model, and analysis of results from steady state axial gas flow experiments utilizing PWR rods with argon and helium gases. As part of the industry cooperative effort, RELAP4 calculations were compared with separate effects jet pump data from the General Electric Company two-loop test apparatus.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 7116778
- Report Number(s):
- TREE-NUREG-1004
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
LOSS OF COOLANT
LOFT REACTOR
REACTOR OPERATION
SIMULATION
PWR TYPE REACTORS
BLOWDOWN
ECCS
FLUID FLOW
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
MATHEMATICAL MODELS
PERFORMANCE
PLUTONIUM DIOXIDE
PRESSURE GRADIENTS
REACTOR SAFETY
RESEARCH PROGRAMS
URANIUM DIOXIDE
ACCIDENTS
ACTINIDE COMPOUNDS
CHALCOGENIDES
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
OPERATION
OXIDES
OXYGEN COMPOUNDS
PLUTONIUM COMPOUNDS
PLUTONIUM OXIDES
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
TANK TYPE REACTORS
TEST REACTORS
TRANSURANIUM COMPOUNDS
URANIUM COMPOUNDS
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled