Assessment of RELAP5/MOD2, Cycle 36. 04 using LOFT intermediate break experiment L5-1

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

OSTI ID:5459572

Lee, E J; Chung, B D; Kim, H J ^[1]

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1. Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

This document the LOFT intermediate break experiment L5-1, which simulates 12 inch diameter ECC line break in a typical PWR, and 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 recommended to use the reflood option and not to use gap conductance option for this L5-1 type IBLOCA.

• 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: NRC; Nuclear Regulatory Commission, Washington, DC (United States)
• OSTI ID: 5459572
• Report Number(s): NUREG/IA-0069; ON: TI92013518
• Country of Publication: United States
• Language: English

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