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RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS

PWR非常用炉心冷却系単一故障条件下の17%低温側配管中破断冷却材喪失事故に関するROSA/LSTF実験データを用いたRELAP5コードによる不確かさ評価

竹田 武司; 大津 巌

Takeda, Takeshi; Otsu, Iwao

An experiment was conducted for the OECD/NEA ROSA-2 Project using LSTF, which simulated a cold leg intermediate-break loss-of-coolant accident with 17% break in a PWR. Assumptions were made such as single-failure of high-pressure and low-pressure injection systems. In the LSTF test, core dryout took place because of rapid drop in the core liquid level. Liquid was accumulated in upper plenum, SG U-tube upflow-side and inlet plena because of counter-current flow limiting (CCFL). The post-test analysis by RELAP5/MOD3.3 code revealed that peak cladding temperature (PCT) was overpredicted because of underprediction of the core liquid level due to inadequate prediction of accumulator flow rate. We found the combination of multiple uncertain parameters including the Wallis CCFL correlation at the upper core plate, core decay power, and steam convective heat transfer coefficient in the core within the defined uncertain ranges largely affected the PCT.

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パーセンタイル:64.68

分野:Nuclear Science & Technology

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