核数据对球床式高温气冷堆燃料元件中子学特性的影响

球床式高温气冷堆采用了球形燃料元件,燃料区域由石墨基体和弥散在其中的包覆燃料颗粒构成,其外有与石墨基体相同材料的包壳;燃料球堆叠成填充率约为0.61的球床式堆芯活性区。在堆芯物理计算中,必须考虑其特殊的双重非均匀性结构对共振计算的影响。此外,由于石墨起到了中子慢化和结构材料的重要作用,其截面参数的准确性对共振计算和临界计算均有很大影响。本文采用蒙特卡罗中子输运计算程序SCALE/KENO-Ⅵ和Serpent-2,对比分析了ENDF/B Ⅶ.0和ENDF/B Ⅶ.1版本核数据库对不同燃料模型的有效增殖因数(keff)及反应率的影响,并进一步比较了不同双重非均匀性处理方法对计算结果的影响。结果表明,由于石墨吸收率增大,使用ENDF/B Ⅶ.1版本核数据库所得keff小于使用ENDF/B Ⅶ.0版本核数据库的结果,且计算模型中石墨材料越多,计算结果相差越大:对于包覆颗粒模型差别约为200pcm,对于燃料元件约为700pcm,对于堆芯单元约为1 600pcm。SCALE/KENO-Ⅵ程序使用DOUBLEHET模型进行多群蒙特卡罗计算所得结果与连续能量模型计算结果吻合良好,且计算效率高,对燃料球模型而言可节省约85%的计算时间。

Effect of Nuclear Data on Fuel Element Neutronic Characteristicsof Pebble-bed High Temperature Gas-cooled Reactor

The Pebble-bed High Temperature Gas-cooled Reactor utilizes unique coated fuel particle design,which is embedded within a graphite matrix and shaped into spherical element with graphite shell and coolant outside.The inherent double heterogeneity of spherical fuel element requires special simulation strategy in both modeling and resonance calculation,where graphite cross sections play significant role considering the large quantity and neutron moderation effect of graphite in pebble-bed HTR.In this study,neutronic calculations with both ENDF/B Ⅶ.0 and ENDF/B Ⅶ.1 nuclear data libraries were performed for various fuel models using SCALE/KENO-Ⅵ and Serpent-2 codes.The results show that using ENDF/B Ⅶ.0 library leads to significant under-prediction of multiplication factor when compared with results using ENDF/B Ⅶ.1 library and the deviation varies with model scale,about 200 pcm for coated particle model,about 700 pcm for fuel element and about 1 600 pcm for core unit.The results of single spherical fuel element with SCALE DOUBLEHET treatment were compared with those using explicit lattice model and it is found that the DOUBLEHET treatment produces consistent numerical results while saving about 85％ computing time.