Applying the continuous-energy Monte Carlo calculation code, MVP3, to analysis of kinetic parameters measured for light-water moderated UO₂ and MOX cores of the TCA and EOLE critical facilities

In order to evaluate the accuracies of the kinetics parameters predicted with the continuous-energy Monte Carlo calculation code, MVP3, core analysis was performed for light-water moderated UO₂ and mixed oxide (MOX) fuel cores for which the ratios of effective delayed-neutron fraction (<i>β_eff</i>) to prompt-neutron life time (<i>ℓ</i>) – represented hereafter by <i>β_eff</i>/<i>ℓ –</i> and <i>β_eff</i> values were measured. The results obtained with the JENDL-4.0-based neutron library showed that (the calculation values (C)/the measurement values (E)−1) ranged from −1.0% to 4.0% for the <i>β_eff</i>/<i>ℓ</i> values with an average of 1.0% and a standard deviation of 1.1% for the 17 cores (the UO₂, and UO₂-MOX mixed cores) tested in the Tank-Type Critical Assembly (TCA). With respect to the <i>β_eff</i> of one UO₂ core in TCA, C/E − 1 was 1.2%. For the <i>β_eff</i> values of the UO₂ and MOX cores tested in the EOLE critical facility, C/E − 1 was −1.5% for the former and −4.6% for the latter. The calculated <i>β_eff</i> value of the MOX core using the JEFF-3.2-based neutron library was larger by 4% than that calculated using the JENDL-4.0-based neutron library and showed better agreement to the measurement.

为评估基于连续能量蒙特卡罗计算程序MVP3所预测的动力学参数的精度，本研究针对轻水慢化二氧化铀（UO₂）与混合氧化物（MOX）燃料堆芯开展堆芯分析。此类堆芯的有效缓发中子份额（effective delayed-neutron fraction，β_eff）与瞬发中子寿命（prompt-neutron life time，ℓ）的比值（下文以β_eff/ℓ表示）以及β_eff数值均已完成实测。基于JENDL-4.0中子库的计算结果显示，在槽式临界装置（Tank-Type Critical Assembly，TCA）中测试的17座堆芯（含二氧化铀堆芯及二氧化铀-MOX混合堆芯）中，β_eff/ℓ的（计算值C/实测值E − 1）分布区间为−1.0%至4.0%，平均值为1.0%，标准差为1.1%。针对TCA内一座二氧化铀堆芯的β_eff参数，其（C/E − 1）为1.2%。对于在EOLE临界设施中测试的二氧化铀与MOX堆芯的β_eff数值，前者的（C/E − 1）为−1.5%，后者为−4.6%。采用JEFF-3.2中子库计算得到的MOX堆芯β_eff数值，较基于JENDL-4.0中子库的计算结果高出4%，且与实测值的吻合度更优。

搜集汇总

背景与挑战

背景概述

该数据集聚焦于核反应堆动力学参数分析，使用连续能量蒙特卡洛计算代码MVP3评估轻水慢化UO2和MOX燃料堆芯的有效延迟中子分数与瞬发中子寿命比值的预测准确性。研究基于JENDL-4.0和JEFF-3.2中子库，在TCA和EOLE临界设施中进行实验验证，结果显示计算值与测量值偏差在-4.6%到4.0%之间，为核反应堆安全分析提供了关键数据支持。