Analysis of LOFT test L5-1 using RELAP5/MOD2

Technical Report · Sat May 01 00:00:00 EDT 1993

OSTI ID:7368576

Cooper, S ^[1]

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1. Nuclear Electric plc, Barnwood (United Kingdom)

RELAP5/MOD2 is being used by Technology Division for the calculation of certain small break loss-of coolant accidents (SBLOCA) and pressurized transients in the Sizewell 'B' PWR. To assist in validating RELAP5/MOD2 for the above application, the code is being used to model a number of small LOCA and pressurized fault simulation experiments carried out in integral test facilities. The present report describes a RELAP5/MOD2 analysis of an intermediate break LOCA test in the LOFT facility. This test discussed in this report was designed to simulate the rupture of a single 14 inch diameter accumulator injection line in a commercial PWR with a 25% break in the broken loop cold leg. Early in the transient the pumps were tripped and the HPIS injection initiated; towards the end of the transient, accumulator and LPIS injection began. RELAP5/MOD2 gave reasonably accurate predictions of the system thermal hydraulic behaviour but failed to accurately calculate the core dryout which occurred due to boil-off prior to accumulator injection. The error is due to the failure to calculate the correct core void distribution during this period of the transient. A separate calculation using the RELPIN code using hydraulic data from the RELAP5 analysis gave significantly improved predictions of the core dryout. However, the peak clad temperature was underpredicted, it is believed that the error is due to the fact that the core liquid inventory in this boildown was overpredicted in the RELAP5/MOD2 calculation.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; Nuclear Electric plc, Barnwood (United Kingdom)
• Sponsoring Organization: USNRC; Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 7368576
• Report Number(s): NUREG/IA-0118; TD/SPB/REP-0130; ON: TI93015049
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
HEAT TRANSFER
R CODES
HYDRAULICS
PWR TYPE REACTORS
LOSS OF COOLANT
COMPUTER CALCULATIONS
COMPUTER PROGRAM DOCUMENTATION
INTERNATIONAL AGREEMENTS
LOFT REACTOR
REACTOR SAFETY
TESTING
TRANSIENTS
ACCIDENTS
AGREEMENTS
COMPUTER CODES
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
MECHANICS
POWER REACTORS
REACTOR ACCIDENTS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
TANK TYPE REACTORS
TEST REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled