Assessment of RELAP5/MOD2 Cycle 36.04 with LOFT large break LOCE L2-3

Technical Report · Wed Apr 01 04:00:00 EST 1992

OSTI ID:10142354

Bang, Y S; Kim, H J ^[1]; Kim, S H ^[2]

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1. Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
2. Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability to predict the thermal-hydraulic phenomena in LBLOCA of the PWR. The reactor vessel was simulated with two core channels and split downcomer modelling for a base case calculation using the frozen code. From the results of the base case calculation, deficiencies of the critical flow model and the CHF correlation at high flow rate were identified, and the severeness of the rewetting criteria were also found. Additional calculation using an updated version of RELAP5/MOD2 Cycle 36.04 including modifications of the rewet criteria shows a substantial improvement in the core thermal response.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 10142354
• Report Number(s): NUREG/IA--0070; ON: TI92013437
• Country of Publication: United States
• Language: English

Related Subjects

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
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COMPUTER CALCULATIONS
COMPUTERIZED SIMULATION
EVALUATION
FLOW MODELS
FLUID FLOW
HEAT TRANSFER
HYDRAULICS
INTERNATIONAL COOPERATION
LOFT REACTOR
LOSS OF COOLANT
MATHEMATICS AND COMPUTERS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PWR TYPE REACTORS
R CODES
REACTOR SAFETY