Power-cooling-mismatch test series. Test PCM-4. Postirradiation examination. [PWR]

Technical Report · Sat Jul 01 00:00:00 EDT 1978

OSTI ID:6631373

Seiffert, S L; Cook, T F

Test PCM-4, conducted in a pressurized water reactor (PWR) test loop in the Power Burst Facility at the Idaho National Engineering Laboratory, used four 1.01-m-long, unirradiated PWR-type zircaloy-clad fuel rods. The four rods were subjected to a preconditioning period, followed by four separate flow-reduction, departure from nucleate boiling test cycles performed at linear rod peak powers up to an average of 69 kW/m and at minimum coolant mass fluxes of 600 kg/(s . m/sup 2/). All four rods were operated in film boiling conditions during the fourth (final) test cycle for times varying up to 140 s prior to manual reactor shutdown. The overall conditions and changes in the fuel and cladding that occurred as a consequence of film boiling operation are discussed. Cladding temperatures estimated for various circumferential and axial positions along the film boiling region of each test rod are presented. Cladding deformation from collapse and bowing are reported. Cladding embrittlement associated with high temperature cladding oxidation reaction with the coolant and the fuel is evaluated and comparisons with fuel failure criteria are presented. Effects of film boiling on fuel restructuring are also presented.

• Research Organization: Idaho National Engineering Lab., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 6631373
• Report Number(s): NUREG/CR-0238; TREE-1230; TRN: 78-018755
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
POWER-COOLING-MISMATCH ACCIDENTS
HEAT TRANSFER
HYDRAULICS
METALLURGICAL EFFECTS
PWR TYPE REACTORS
DEFORMATION
FLOW RATE
FUEL CANS
FUEL RODS
PERFORMANCE TESTING
POWER DENSITY
POWER DISTRIBUTION
TEMPERATURE GRADIENTS
TEST FACILITIES
URANIUM DIOXIDE
ZIRCALOY 4
ACCIDENTS
ACTINIDE COMPOUNDS
ALLOYS
CHALCOGENIDES
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
MECHANICS
OXIDES
OXYGEN COMPOUNDS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
TESTING
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled