ROSA/LSTF test on PWR station blackout transient and RELAP5 analyses

Takeda, Takeshi ; Otsu, Iwao

An experiment on accident management (AM) measures during a PWR station blackout transient with leakage from primary coolant pump seals was conducted using the ROSA/large scale test facility (LSTF) based on the Fukushima accident. Through RELAP5/MOD3.2 code, we investigate core void fraction and surface heat transfer coefficient of the cladding. In addition, sensitivity analyses were performed with the RELAP5 code. The onset timing of SG secondary depressurization as well as the SG coolant injection flow rate were found to significantly affect the peak cladding temperature.

Language	:	Japanese
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Meeting title	:	日本原子力学会2015年秋の大会
Held date	:	2015/09
Location (city)	:	静岡
Location (country)	:	日本
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Keywords	:	PWR; ROSA/LSTF; 全電源喪失; 燃料被覆管温度; RELAP5コード
Research Facility	:	LSTF(大型非定常試験装置)
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Registration No. : BB20150763
JAEA Abstracts No. :
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