Assessment of RELAP5/MOD2 computer code against the Natural Circulation Test Data from Yong-Gwang Unit 2
- Korea Electric Power Corp., Taejon (Korea, Republic of). Research Center
- Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
The results of the RELAP5/MOD2 computer code simulation for the Natural Circulation Test in Yong-Gwang Unit 2 are analyzed here and compared with the plant operation data. The result of comparison reveals that the code calculation does present well the overall macroscopic behaviors of thermalhydraulic parameters in primary and secondary system compared with the plant operating data. The sensitivity study is performed to find out the effect of steam dump flow rate on the primary temperatures and it is found that the primary temperatures are very sensitive to the steam dump flow rate during the Natural Circulation. Because of the inherent uncertainties in the plant data, the assessment work is focussed on phenomena whereby the comparison between plant data and calculated data is based more on trends than on absolute values.
- Research Organization:
- Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research; Korea Electric Power Corp., Taejon (Korea, Republic of). Research Center; Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
- Sponsoring Organization:
- NRC; Nuclear Regulatory Commission, Washington, DC (United States)
- OSTI ID:
- 7368735
- Report Number(s):
- NUREG/IA-0125; ON: TI93016058
- Country of Publication:
- United States
- Language:
- English
Similar Records
Assessment of RELAP5/MOD2 computer code against the Net Load Trip Test data from Yong-Gwang, Unit 2
Assessment of RELAP5/MOD2 computer code against the Net Load Trip Test data from Yong-Gwang, Unit 2
Related Subjects
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
COMPUTER CALCULATIONS
COMPUTER CODES
COMPUTERIZED SIMULATION
CONVECTION
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
MASS TRANSFER
MECHANICS
NATURAL CONVECTION
POWER REACTORS
PWR TYPE REACTORS
R CODES
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SENSITIVITY ANALYSIS
SIMULATION
STEAM
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS