Development of a fuel depletion sensitivity calculation module for multi-cell problems in a deterministic reactor physics code system CBZ
暂无分享，去创建一个
[1] G. Chiba,et al. Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation , 2015 .
[2] G. Chiba,et al. Important fission product nuclides identification method for simplified burnup chain construction , 2015 .
[3] G. Chiba,et al. Photon transport effect on intra-subassembly thermal power distribution in fast reactor , 2014 .
[4] Go Chiba,et al. Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data , 2013 .
[5] K. Shibata,et al. JENDL-4.0: A New Library for Nuclear Science and Engineering , 2011 .
[6] W. Rooijen,et al. Diffusion coefficients for LMFBR cells calculated with MOC and Monte Carlo methods , 2011 .
[7] Toshihisa Yamamoto,et al. Optimization Study on the Diffusion Synthetic Acceleration Algorithm in Three-Dimensional Discrete Ordinate Transport , 1995 .
[8] T. Umano,et al. Burnup sensitivity analysis in a fast breeder reactor-Part I : Sensitivity calculation method with generalized perturbation theory , 1985 .
[9] Mark L. Williams,et al. Development of depletion perturbation theory for coupled neutron/nuclide fields , 1979 .
[10] H. Mitani,et al. Sensitivity coefficient of integral data and generalized perturbation method , 1972 .
[11] L. N. Usachev. Perturbation theory for the breeding ratio and for other number ratios pertaining to various reactor processes , 1964 .