Severe fuel-damage scoping test performance. [PWR]

Conference · Sat Jan 01 04:00:00 EST 1983

OSTI ID:5624709

Gruen, G E; Buescher, B J

As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5624709
• Report Number(s): CONF-830702-23; ON: DE84000561
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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Light-Water Moderated
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220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ACTINIDE COMPOUNDS
ALLOYS
CHALCOGENIDES
COMPUTER CALCULATIONS
DAMAGE
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
FUEL RODS
HEAT TRANSFER
HYDRAULICS
MECHANICS
OXIDES
OXYGEN COMPOUNDS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY
REACTORS
SAFETY
STRESSES
TEMPERATURE GRADIENTS
TEST FACILITIES
THERMAL STRESSES
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM DIOXIDE
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS