Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient
- Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), 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:
- 383664
- Report Number(s):
- NUREG/IA-0130; ON: TI97000271; IN: CAMP002; TRN: 96:029146
- Resource Relation:
- Other Information: DN: International Agreement Report; PBD: Sep 1996
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
COMPUTERS
INFORMATION SCIENCE
MANAGEMENT
LAW
MISCELLANEOUS
22 NUCLEAR REACTOR TECHNOLOGY
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
REACTOR ACCIDENTS
R CODES
MIHAMA-2 REACTOR
STEAM GENERATORS
RUPTURES
REACTOR SAFETY EXPERIMENTS
THEORETICAL DATA
COMPARATIVE EVALUATIONS
NATURAL CONVECTION
SENSITIVITY ANALYSIS