Experiments on Central Reaction Rate Ratios and Fission Distributions in the FCA-XXII-1 Assembly Simulating Highly Enriched MOX–Fueled Tight Lattice LWR Cores

A series of simulated experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency to simulate a light water reactor core with a tight lattice cell containing highly enriched mixed-oxide fuel with a fissile plutonium (Pu) ratio >15%. The prediction accuracy of the neutron computation codes and nuclear data libraries in a wide range of neutron spectra was evaluated by constructing three experimental configurations of the FCA-XXII-1 assembly with different void fractions (45%, 65%, and 95%) of the moderator material (foamed polystyrene). The hydrogen-to–nuclear fuel atomic ratio was systematically varied from 0.1 to 0.8. In a previous paper, we reported the criticality and reactivity worths measured in these experiments. This technical note provides the experimental results for the central reaction rate ratios and fission distributions as follows. The fission rate ratios of uranium (U) (²³⁸U) and ²³⁹Pu relative to ²³⁵U were measured at the core centers using three calibrated fission chambers, and the ²³⁸U capture reaction rate ratio relative to ²³⁵U fission was measured using depleted U foils. The reaction rate distributions were also obtained by traversing four micro fission chambers of highly enriched U, natural U, Pu, and neptunium through each core region in the radial and axial directions. The experimental analyses were performed using detailed models of the Monte Carlo code MVP3 with the Japanese evaluated nuclear data library of JENDL-4.0. Most calculation results agreed well with the experiments, whereas those for the fission rate ratio of ²³⁹Pu to ²³⁵U were underestimated by up to 6% with the softening neutron spectrum.

本研究于日本原子能机构（Japan Atomic Energy Agency）的快临界装置（fast critical assembly, FCA）内开展系列模拟实验，用以模拟搭载易裂变钚（Pu）占比＞15%的高浓混合氧化物燃料的紧栅元轻水反应堆堆芯。通过搭建三种不同慢化剂（发泡聚苯乙烯）空泡份额（45%、65%、95%）的FCA-XXII-1组件实验配置，本研究评估了中子计算程序与核数据库在宽范围中子能谱下的预测精度。实验中系统调整氢与核燃料的原子比，取值区间为0.1至0.8。本团队此前已在一篇学术论文中报道了本次实验测得的临界特性与反应性价值。本技术简报补充给出核心反应率比与裂变分布的实验结果，具体如下：
1. 采用三台经标定的裂变室，在堆芯中心位置测得铀-238（²³⁸U）与钚-239（²³⁹Pu）相对于铀-235（²³⁵U）的裂变率比；同时采用贫化铀箔，测得铀-238俘获反应率相对于铀-235裂变反应率的比值。
2. 通过将高浓铀、天然铀、钚及镎的四台微型裂变室沿径向与轴向遍历每个堆芯区域，获取了反应率分布。
本实验分析采用搭载日本评价核数据库JENDL-4.0的蒙特卡罗程序MVP3详细模型完成。多数计算结果与实验数据吻合良好，但当中子能谱软化时，钚-239相对于铀-235的裂变率比的计算结果被低估最多达6%。