Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1977
Water reactor research is summarized for the following programs: semiscale, LOFT, Thermal Fuels Behavior, Reactor Behavior, and 3-D Project. The Semiscale Program reports progress in modifying the Mod-1 facility to a new Mod-3 configuration scaled more directly to a PWR; influence of steam generator tube ruptures at start of core reflood in the Semiscale Mod-1 system is discussed. The LOFT Experimental Program presents analyses of small-scale pump coastdown during LOCA conditions; zircaloy rod cladding response to isothermal, isobaric high-temperature and high-pressure tests; and relative effects of rod cladding-mounted thermocouples on time-to-CHF for LOFT fuel rods under LOCA conditions. The Thermal Fuels Behavior Program reports on the various PBF test series (power-cooling-mismatch, gap conductance, reactivity initiated accident, irradiation effects, inlet flow blockage, loss-of-coolant accident, and the PBF-LOFT lead rod test program), on program development (including Halden fuel behavior research, postirradiation examination of commercial power reactor fuel, and a diffusion analysis of UO/sub 2/-zircaloy interaction), and on fuel model development (including a discussion of FRAP-S3 code verification). The Reactor Behavior Program describes containment analysis code development (BEACON/MOD2). The multinational 3-D Experiment Project reports completion of conceptual design for the slab core reflood experiment.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 5124192
- Report Number(s):
- TREE-NUREG-1205
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
BWR TYPE REACTORS
LOSS OF COOLANT
LOFT REACTOR
REACTOR OPERATION
SIMULATION
PWR TYPE REACTORS
BLOWDOWN
COMPUTER CALCULATIONS
ECCS
FLUID FLOW
FUEL ELEMENTS
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
MATHEMATICAL MODELS
PBF REACTOR
PERFORMANCE
PLUTONIUM DIOXIDE
POWER-COOLING-MISMATCH ACCIDENTS
PRESSURE GRADIENTS
REACTOR SAFETY
RESEARCH PROGRAMS
THERMAL CONDUCTIVITY
TWO-PHASE FLOW
URANIUM DIOXIDE
ACCIDENTS
ACTINIDE COMPOUNDS
CHALCOGENIDES
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
OPERATION
OXIDES
OXYGEN COMPOUNDS
PHYSICAL PROPERTIES
PLUTONIUM COMPOUNDS
PLUTONIUM OXIDES
PULSED REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
TANK TYPE REACTORS
TEST REACTORS
THERMODYNAMIC PROPERTIES
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
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Light-Water Moderated
Nonboiling Water Cooled
360204 - Ceramics
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