基于LR-0基准题的CENDL-TMSR-V1数据库验证

为验证氟盐冷却先进堆型的物理特性，在捷克LR-0装置上开展了关于石墨和FLiNa盐的中子物理实验，形成了满足国际临界安全分析评价标准的基准题。基于上述基准题，利用MCNP和SCALE程序，对中国核数据中心研制的钍-铀循环专用核数据库CENDL-TMSR-V1进行了验证。结果表明，CENDL-TMSR-V1计算得到的石墨和FLiNa盐样品组临界实验keff、能谱和中子通量均与实验结果符合。临界计算最大差异为-0.001 87，在实验不确定度范围内。相较于ENDF/B-Ⅶ.0的计算结果，CENDL-TMSR-V1计算值与实验结果更接近。不确定度分析表明，CENDL-TMSR-V1计算得到的石墨和FLiNa盐核数据不确定度明显小于SCALE6.1自带协方差数据库的计算结果。

Validation of CENDL-TMSR-V1 against LR-0 Benchmark

Neutron physics experiments on graphite and FLiNa salt were carried out on LR-0 of the Czech Republic to verify the physics characteristics of advanced reactor cooled by fluorine salt. Benchmarks meeting the international critical safety analysis and evaluation standard were established. CENDL-TMSR-V1, developed by China Nuclear Data Center for Th-U fuel cycle analysis, was verified by using MCNP and SCALE program against LR-0 benchmarks. The results show that the keff, energy spectrum and neutron flux in the graphite and FLiNa salt of LR-0 calculated by CENDL-TMSR-V1 agree well with the experimental results. The maximum difference of critical calculation is -0.001 87, which is within the range of experimental uncertainty. Compared with the calculation results of ENDF/B Ⅶ.0, the calculated values obtained by CENDL-TMSR-V1 are more close to the experimental results. The uncertainty analysis shows that the uncertainty caused by graphite and FLiNa salt data obtained from CENDL-TMSR-V1 covariance data library is significantly less than that obtained from SCALE6.1 covariance data library.