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Assessment of RELAP5/MOD2 using semiscale intermediate break loss-of-coolant experiment S-IB-3

Technical Report ·
OSTI ID:10156389
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This report presents the results of the RELAP5/MOD2 assessment utilizing a Semiscale intermediate break of loss-of-coolant experiment S-IB-3. Comprehensive analysis with RELAP5/MOD2 is performed to predict the transient thermal-hydraulic responses of the experiment. Test S-IB-3 is a 21.7%, communicative cold leg break LOCA experiment using Semiscale Mod-2A facility in 1982, for the principal objective to provide reference data for comparison of Semiscale test results to LOBI facility B-R1M test results. Through extensive comparison between test data and best-estimate RELAP5 calculations, the capabilities of RELAP5/MOD2 or predict the intermediate break LOCA accident were assessed. Emphasis was located on the capability of the code to calculate core level depression and break flow rate during system blowdown, pump suction liquid seals phenomena, and temperature excursions behavior etc., throughout the whole experiment. Besides, some sensitivity studies involving the effect of steam generator secondary side pressure boundary, adjustment of two-phase discharge coefficient, intact loop pump coastdown behavior, and some interesting studies regarding break flow etc., were also investigated in this report.
Research Organization:
Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; Institute of Nuclear Energy Research, Lung-Tan (Taiwan, Province of China)
Sponsoring Organization:
Nuclear Regulatory Commission, Washington, DC (United States)
OSTI ID:
10156389
Report Number(s):
NUREG/IA--0080; ON: TI92016318
Country of Publication:
United States
Language:
English

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Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900
99 GENERAL AND MISCELLANEOUS
990200
BLOWDOWN
COMPUTER CALCULATIONS
EVALUATION
FLOW RATE
HEAT TRANSFER
HYDRAULICS
INTERNATIONAL COOPERATION
LOSS OF COOLANT
MATHEMATICS AND COMPUTERS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PWR TYPE REACTORS
R CODES
REACTOR SAFETY
STEAM GENERATORS
TEST FACILITIES