[2M04-06] ライブラリ、中性子束計算

[2M06] Calculation of 3D neutron flux distribution in the HTTR using MCNP6

*Hai Quan Ho¹, Nozomu Fujimoto², Shimpei Hamamoto¹, Toshiaki Ishii¹, Satoru Nagasumi¹, Etsuo Ishitsuka¹ (1. JAEA, 2. Kyushu Univ.)

キーワード：HTTR, 3D, Neutron flux, MCNP

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.