RELAP5/MOD2 analysis of LOFT experiment L9-4

Technical Report · Tue Mar 31 23:00:00 EST 1992

OSTI ID:10146140

Keevill, M B ^[1]

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1. National Power Nuclear, Barnwood (United Kingdom)

As part of a program to validate RELAP5/MOD2 for use in the analysis of certain fault transients in the Sizewell B PWR, the code has been used to simulate experiment L9-4 carried out in the Loss-Of-Fluid Test (LOFT) facility. Experiment L9-4 simulated a Loss-Of-Offsite-Power Anticipated Transient Without Trip (LOOP ATWT) in which power is lost to the primary coolant pumps and main feed is lost to the steam generators but the control rods fail to insert in the reactor core. RELAP5/MOD2 generally predicted the transient well, although there were some differences compared to the test data. These differences are largely due to the use of power and flow as boundary conditions and because of uncertainties in the power and flow experimental data. The most noticeable difference was that the steam generator was predicted to boil down too fast. This is believed to be partly due to errors in the RELAP5 interphase drag model. The RELAP5 calculation also showed the primary pressure to be very sensitive to the primary flow rate, making the exact simulation of primary side relief valve movements difficult to reproduce.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States); National Power Nuclear, Barnwood (United Kingdom)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 10146140
• Report Number(s): NUREG/IA--0066; GD/PE-N--721; ON: TI92014260
• Country of Publication: United States
• Language: English

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
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22 GENERAL STUDIES OF NUCLEAR REACTORS
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99 GENERAL AND MISCELLANEOUS
990200
AFTER-HEAT REMOVAL
BLACKOUTS
COMPUTER CALCULATIONS
COMPUTER PROGRAM DOCUMENTATION
COMPUTERIZED SIMULATION
CONTROL ELEMENTS
FLOW RATE
HEAT TRANSFER
HYDRAULICS
INTERNATIONAL COOPERATION
LOFT REACTOR
MATHEMATICS AND COMPUTERS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PRIMARY COOLANT CIRCUITS
PUMPS
PWR TYPE REACTORS
R CODES
REACTOR SAFETY
RESEARCH PROGRAMS
SIZEWELL-B REACTOR
TRANSIENTS
VALIDATION