Transport Criticality Analysis for FBR MONJU Initial Critical Core in Whole Core Simulation by NSHEX and GMVP

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Galina Todorova; 西 裕士*; 石橋 淳一

Galina Todorova; Nishi, Hiroshi*; Ishibashi, Junichi

「もんじゅ」初臨界炉心の臨界性を、拡散近似でなく本来の輸送計算により評価するため、決定論的手法として3次元ノード法Sn輸送計算コードNSHEX、また確率論的手法としてモンテカルロ輸送計算コードGMVPを用いて解析した。結果、拡散計算に比べ、輸送計算では無視し得ないエネルギー群依存性のあることを明らかにした。特にNSHEXコードによる少数群の解析では、群定数縮約方法に検討の余地のあることを明確化した。ただし70群の計算では、NSHEXの結果とはGMVPと良く一致し、同コードの「もんじゅ」炉心への適用性を確認した。

FBR MONJU Initial Critical Core (ICC) criticality problem has been solved by deterministic and Monte Carlo transport methods by the codes NSHEX and GMVP. The analysis has been carried out in different energy-groups approximations. As a result the effect of cross-section (XS) condensation from 70 into few energy-group structures by different collapsing methods has been evaluated. The 3D discrete-ordinate code NSHEX has been applied for wide range of core simulations-from whole core, considering the fissile, fertile and shielding regions to simplified models that simulate an increased neutron leakage. It has been found that there is room for improvement in the assessment of the neutron leakage in the few energy-group approximations. The good agreement between NSHEX and GMVP results, especially without XS collapsing, is pointed out as a conformation for the applicability of the code NSHEX in FBR 3D whole core calculations. Some practical conclusions have been extracted that are important