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

Hai Quan Ho

Presentation information

Oral presentation

II. Radiation, Accelerator, Beam and Medical Technologies » 201-1 Nuclear Physics, Nuclear Data Measurement/Evaluation/Validation, Nuclear Reaction Technology

[2M04-06] Library, Neutron Flux Calculation

Thu. Sep 12, 2019 10:20 AM - 11:10 AM Room M (Common Education Bildg. 3F A31)

Chair:Masahide Harada(JAEA)

10:50 AM - 11:05 AM

[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.)

Keywords: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.