Kabach Ouadie, Chetaine Abdelouahed, Benchrif Abdelfettah
Mohammed V University, Faculty of Science, Nuclear Reactor and Nuclear Security Group Energy Centre, Physics Department, 4 Avenue Ibn Battouta B.P. 1014 RP, Rabat, 10000, Morocco.
Mohammed V University, Faculty of Science, Nuclear Reactor and Nuclear Security Group Energy Centre, Physics Department, 4 Avenue Ibn Battouta B.P. 1014 RP, Rabat, 10000, Morocco.
Appl Radiat Isot. 2019 Aug;150:146-156. doi: 10.1016/j.apradiso.2019.05.015. Epub 2019 May 21.
A comparative study has been performed with the latest evaluated nuclear data libraries JEFF-3.3 and ENDF/B-VIII.0. The study has been conducted through the benchmark calculations for 120 criticality problems and the TRIGA Mark II research reactor with the libraries processed using NJOY21 for MCNPX Monte Carlo transport code. The criticality benchmarks assemblies, taken from the ICSBEP benchmark, cover Uranium (highly enriched uranium, intermediate enriched uranium, low enriched uranium, andU) and Plutonium fuel systems in a various metal forms, and under a various spectral conditions. The Moroccan TRIGA Mark II research reactor calculation is used to look into the predictive capability of those nuclear data libraries and then to compare the accuracy of the predicted results with the experimental data published elsewhere. Actually, the purpose of this study is to investigate some neutronic and kinetic parameters of those benchmarks for both libraries. The former consist of effective multiplication factor, heat distribution, neutron flux distribution, effective delayed neutron fraction (β), prompt removal lifetime (τ) and the mean neutron generation time (Λ). The results show that the calculated effective multiplication factor, heat distribution, neutron flux distribution, and the kinetic parameters are in good agreement with references. However, it is found that the computed values are strongly depending on the nuclear data set used in calculations.
已对最新评估的核数据库JEFF - 3.3和ENDF/B - VIII.0进行了一项比较研究。该研究通过对120个临界问题以及TRIGA Mark II研究堆进行基准计算来开展，使用NJOY21对这些数据库进行处理，用于MCNPX蒙特卡罗输运代码。取自国际临界安全基准评价项目（ICSBEP）基准的临界基准组件涵盖了各种金属形式以及各种谱条件下的铀（高浓铀、中浓铀、低浓铀和天然铀）和钚燃料系统。摩洛哥TRIGA Mark II研究堆的计算用于考察这些核数据库的预测能力，然后将预测结果的准确性与其他地方公布的实验数据进行比较。实际上，本研究的目的是研究这两个数据库中这些基准的一些中子学和动力学参数。前者包括有效增殖系数、热分布、中子通量分布、有效缓发中子份额（β）、瞬发中子泄漏寿命（τ）和平均中子产生时间（Λ）。结果表明，计算得到的有效增殖系数、热分布、中子通量分布和动力学参数与参考文献吻合良好。然而，发现计算值强烈依赖于计算中使用的核数据集。
