RELAP5-3Dコードを用いた4ループPWRプラントの主蒸気管破断事故解析

摘要

The main steam line break (MSLB) accident was analyzed for several combinationsrnof reactor core and transient analysis conditions by using the RELAP5-3D code for a 17 x 17 rod bundle type four loop PWR plant. The influences of the reactor core conditions and the analytical conditions were examined upon the predicted plant behavior. In addition, using the RELAP5-3D code and the SIMULATE-3K code, analyses by the reactor core model were conducted with quantities of the state at the core inlet and outlet as boundary conditions, which were obtained from the analytical results by the RELAP5-3D code. The differences in the results between the two codes were discussed. These analyses showed that the criticality was not reached or the power hardly rose during the MSLB accident, under a little more conservative conditions than those of the actual plant. Moreover, it was also shown that the assumption for the reactor coolant pump (RCP) trip and the main steam check valve affected the analytical results, and the conservatism of the safety evaluation condition has been confirmed. The conservatism in the core thermal hydraulic model of the RELAP5-3D code and the boron reactivity coefficient etc. was clarified from the comparison between the RELAP5-3D and the SIMULATE-3K codes, showing that the predicted results by the RELAP5-3D code are still conservative. In addition, combined with the COBRA-EN code, minimum DNBR was estimated.%RELAP5-3Dコードを用いて，17×17ロッドバンドル型4ループPWRプラントを対象に，rn複数の炉心条件と過渡解析条件を組み合わせて主蒸気管破断事故解析を実施し，炉心条件および解rn析条件の違いによるプラント挙動への影響について検討した．次に，RELAP5－3Dコードによる解rn析結果から得られた過渡期間中の炉心出人口領域状態量を境界条件として設定し，RELAP5－3DコrnードとSIMULATE－3Kコードを用いて炉心解析モデルによる解析を実施し，その差異について検rn討した・解析結果から，やや保守的に設定した実機相当の条件でも臨界に至らないか，出力がほとrnんど上昇しないことを確認した．また，原子炉冷却材ポンプ（RCP）の作動状態および主蒸気避止rn弁の開閉状態の想定が解析結果に影響することを確認し，安全評価解析条件の保守性が把握できた．rnさらに，RELAP5－3DコードとSIMULATE－3Kコードによる解析結果の比較から，RELAP5－3Drnコードの炉心熱水カモデルの保守性，ボロン濃度フィードバック係数の保守性等について検討し，rnRELAP5－3Dコードの解析結果には，なお保守性が含まれることを確認した．さらに，COBRA－ENrnコードと組み合わせて最小DNBRを評価した．