PBF-CANDU fuel element LOCA test experiment predictions

Technical Report · Sat Oct 01 04:00:00 EDT 1983

OSTI ID:5371636

Smith, R H; Martinson, Z R

A loss-of-coolant test sponsored jointly by Ontario Hydro and Atomic Energy of Canada Limited (AECL) will be performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The test will provide both an in-reactor evaluation of CANDU fuel element thermal mechanical behavior during a postulated Loss-of-Coolant Accident and an experimental data base for evaluation of the Canadian transient fuel behavior code, ELOCA. The PBF-CANDU test train is comprised of four independently shrouded CANDU design fuel elements installed in symmetric orientations in a modified PBF test assembly. Three of the fuel elements were preirradiated in the NRX reactor to approximately 5000 MWd/t (120 MWh/kgU) at approx. 53 kW/m, and the fourth fuel element is unirradiated. RELAP5 was used to simulate the PBF-CANDU Loss-of-Coolant test. The RELAP5 calculations were used to determine the blowdown valve sequencing required to obtain the prescribed depressurization. The FRAP-T6 code was used to calculate the fuel element response to the simultaneous blowdown and power excursion during the test.

• Research Organization: Idaho National Engineering Lab., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 5371636
• Report Number(s): EGG-TFBP-6399; ON: DE84005030
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210400 -- Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
CANDU TYPE REACTORS
COMPUTER CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLUID MECHANICS
FUEL ELEMENTS
HEAT TRANSFER
HEAVY WATER MODERATED REACTORS
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PRESSURE TUBE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
STRESS ANALYSIS
STRESSES
TEMPERATURE GRADIENTS
TEST FACILITIES
THEORETICAL DATA
THERMAL REACTORS