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Antariksawan, A. R.*; Hidaka, Akihide; Moriyama, Kiyofumi; Hashimoto, Kazuichiro*
JAERI-Tech 2001-011, 116 Pages, 2001/03
JAERI-Tech-2001-011.pdf:4.02MB
no abstracts in English
Takeda, Takeshi; Otsu, Iwao
no journal, ,
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.
Takeda, Takeshi; Otsu, Iwao
no journal, ,
The effectiveness of accident management measure should be confirmed in case of PWR station blackout transient with loss of primary coolant. A simulation experiment with the ROSA/LSTF was thus conducted under an assumption of nitrogen gas inflow into the primary system. After the nitrogen gas inflow, the primary depressurization rate became smaller and non-uniform flow behavior was observed among steam generator (SG) U-tubes. In addition, RELAP5/MOD3.2 code indicated remaining problems in the predictions of the primary pressure and SG U-tube liquid level after the nitrogen gas inflow through the post-test analysis.