United States Nuclear Regulatory Commission - Protecting People and the Environment

Assessment of RELAP5/MOD2 Cycle 36.04 Using LOFT Intermediate Break Experiment L5–1 (NUREG/IA-0069)

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Publication Information

Date Published: April 1992

Prepared by:
E. J. Lee, B. D. Chung, H. J. Kim

Korea Institute of Nuclear Safety
P.O. Box 16, Daeduk-Danji
Daejon, Korea

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Thermal-Hydraulic Code Assessment
and Application Program (ICAP)

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

The LOFT intermediate break experiment L5-1, which simulates 12 inch diameter ECC line break in a typical PWR, has been analyzed using the reactor thermal/hydraulic analysis code RELAP5/MOD2, Cycle 36.04. The base calculation, which modeled the core with single flow channel and two heat structures without using the options of reflood and gap conductance model, has been successfully completed and compared with experimental data. Sensitivity studies were carried out to investigate the effects of nodalization at reactor vessel and core modeling on major thermal hydraulic parameters, especially on peak cladding temperature(PCT). These sensitivity items are: single flow channel and single heat structure (Case A), two flow channel and two heat structures (Case B), reflood option added (Case C) and both reflood and gap conductance options added (Case D). The code, RELAP5/MOD2 Cycle 36.04 with the base modeling, predicted the key parameters of LOFT IBLOCA Test L5-1 better than Cases A,B,C and D. Thus, it is concluded that the single flow channel modeling for core is better than the two flow channel modeling and two heat structure is also better than single heat structure modeling to predict PCT at the central fuel rods. It is recommnended to use the reflood option and not to use gap conductance option for this L5-1 type IBLOCA.

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