The neutron transport simulations of RBMK-1500 fuel assembly were performed using both multigroup and continuous energy data libraries available within the SCALE code system in order to validate its suitability for the estimation of RBMK neutronic characteristics. The resonance processing of cross section, involved in the preparation of the multigroup data library, has a significant impact on neutron transport calculations. Standard Dancoff factors (DFs) used for the heterogeneous geometry of RBMK fuel assembly are insufficient for the accurate estimation of resonance self-shielding. Thus, the SCALE module MCDancoff was used in this study to determine location-specific DFs. The results of RBMK-1500 fuel assembly simulations using standard and user-defined DFs were compared. In addition, the continuous energy (CE) cross-section data library was applied for the benchmark calculations. The impact of different nuclear data libraries on neutron transport simulations was tested as well. It was found out that the usage of the multigroup data libraries generates some deviation from the reference simulations obtained with CE libraries. The CE library based on the estimated ENDF/B-VII.1 data proved to be the best alternative for neutron transport simulations of RBMK fuel assembly.

中文翻译：

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使用SCALE代码中的多组和连续能量数据库对RBMK燃料组件进行中子传输模拟

RBMK-1500燃料组件的中子输运模拟使用SCALE代码系统中可用的多组和连续能量数据库进行，以验证其适用于估计RBMK中子特性。多组数据库的准备工作涉及截面的共振处理，这对中子输运计算具有重大影响。用于RBMK燃料组件的异构几何体的标准Dancoff因子（DF）不足以精确估计共振自屏蔽。因此，在这项研究中使用了SCALE模块MCDancoff来确定位置特定的DF。比较了使用标准和用户定义的DF进行的RBMK-1500燃料组件仿真的结果。此外，连续能量（CE）横截面数据库被用于基准计算。还测试了不同核数据库对中子输运模拟的影响。已经发现，多组数据库的使用与使用CE库获得的参考模拟产生了一些偏差。基于ENDF / B-VII.1估计数据的CE库被证明是RBMK燃料组件中子输运模拟的最佳选择。