ICAP (International Code Assessment and Applications Program) assessment of RELAP5/MOD2, Cycle 36. 05 against LOFT (Loss of Fluid Test) Small Break Experiment L3-7
- Korea Advanced Energy Research Inst., Daeduk-Danji (Republic of Korea). Korea Nuclear Safety Center
The LOFT small break (1 in-dia) experiment L3-7 has been analyzed using the reactor thermal hydraulic analysis code RELAP5/MOD2, Cycle 36.05. The base calculation (Case A) was completed and compared with the experimental data. Three types of sensitivity studies (Cases B, Cm, and D) were carried out to investigate the effects of (1) break discharge coefficient Cd, (2) pump two-phase difference multiplier and (3) High Pressure Injection System (HPIS) capacity on major thermal and hydraulic (T/H) parameters. A nodalization study (Case E) was conducted to assess the phenomena with a simplified nodalization. The results indicate that Cd of 0.9 and 0.1 fit to the single discharge flow rate of Test L3-7 best among the tried cases. The pump two-phase multiplier has little effects on the T/H parameters because of the low discharge flow rate and the early pump coast down in this smaller size SBLOCA. But HPIS capacity has a very strong influence on parameters such as pressure, flow and temperature. It is also shown that a simplified nodalization could accomodate the dominant T/H phenomena with the same degree of code accuracy and efficiency.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Korea Advanced Energy Research Inst., Daeduk-Danji (Republic of Korea). Korea Nuclear Safety Center
- Sponsoring Organization:
- USNRC
- OSTI ID:
- 6958396
- Report Number(s):
- NUREG/IA-0031; ON: TI90009897; TRN: 90-011950
- Country of Publication:
- United States
- Language:
- English
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99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOFT REACTOR
LOSS OF COOLANT
FLOW RATE
HEAT TRANSFER
HIGH PRESSURE COOLANT INJECTION
HYDRAULICS
PUMPS
R CODES
REACTOR COOLING SYSTEMS
REACTOR SAFETY
ACCIDENTS
COMPUTER CODES
COOLING SYSTEMS
ECCS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
MECHANICS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR PROTECTION SYSTEMS
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
TANK TYPE REACTORS
TEST REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
990200 - Mathematics & Computers
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Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled