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This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the...

A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted...

我が国の商業用軽水炉は,その優れた実績を背景に一層の安全性向上のみならず,廃棄物低減,経済性向上,原子燃料サイクルの確立といった多様な要求を満たすべく高度化の道を歩みつつあるそうした変化に対応するため,炉心設計技術も徐々に進化してきているが,本稿では,その基礎となる核計算手法の近年の高度化の動向について述べる。

The improved coarse mesh method, which was originally derived by Askew and applied to fast reactor core calculations by one of the authors, has been applied to boiling water reactor core calculations. Since direct application of the method leads to negative fluxes in the reflector region, a procedure has been developed in which the correction factor for coarse meshes is not utilized in the reflector. The results of the improved coarse mesh calculations were compared with those of fine mesh calculations. The negative fluxes were eliminated and good agreement was obtained for the neutron multiplication factor and power distributions in fuel assemblies near control rods, though the accuracy of power distributions near the reflector was as that of conventional coarse mesh calculations.