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Calculation of 3D neutron flux distribution in the HTTR using MCNP6

Ho, H. Q.   ; 藤本 望*; 濱本 真平  ; 石井 俊晃 ; 長住 達; 石塚 悦男   

Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Ishii, Toshiaki; Nagasumi, Satoru; Ishitsuka, Etsuo

In this study, a detail 3D thermal/fast neutron flux in the HTTR core was calculated using the Monte-Carlo MCNP6 code with FMESH tally. The results is useful for understanding the neutronic characteristic as well as for the core optimization and safety analyses of the HTTR.

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