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RELAP5 analysis of OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

高出力自然循環を伴うPWR小破断LOCAに関するOECD/NEA ROSAプロジェクト実験のRELAP5コード解析

竹田 武司 ; 浅香 英明*; 中村 秀夫  

Takeda, Takeshi; Asaka, Hideaki*; Nakamura, Hideo

PWRでの1%低温側配管小破断冷却材喪失事故(LOCA)において、スクラム失敗による高出力自然循環を仮定したOECD/NEA ROSAプロジェクト実験をLSTFを用いて行った。スクラム失敗時の炉心出力は、原子力機構開発による三次元核熱結合コードSKETCH-INS/TRAC-PF1を用い、炉心部を詳細模擬したPWRのLOCA解析により決定した。実験後解析では、maximum bounding theoryに基づく二相破断流モデルを組み込んだ、原子力機構改良版RELAP5/MOD3.2.1.2コードの予測性能を検証した。LSTF実験では、二相自然循環過程において高温側配管内の流れは射流となり、水位が大きく低下するとともに、リフラックス凝縮過程において高速の蒸気流により蒸気発生器伝熱管上昇流側に蓄水が生じた。一方、改良版RELAP5コードはこれらの熱水力現象をおおむね再現したが、一次系ループ流量の低下が実験より遅くなり低温側配管内の水位を十分予測できず、二相放出過程における破断流量を過大予測した。

OECD/NEA ROSA Project experiment with the Large Scale Test Facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR analysis with coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment, while it overpredicted the break flow rate.

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