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

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.

Language	:	English
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Meeting title	:	日本原子力学会2019年秋の大会
Held date	:	2019/09
Location (city)	:	富山
Location (country)	:	日本
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Keywords	:	HTTR; 3D; Neutron flux; MCNP; FMESH
Research Facility	:	HTTR(高温工学試験研究炉)
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Registration No. : BB20190439
JAEA Abstracts No. :
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