MCNP6 calculation of neutron flux map in the HTTR during normal operation

Ho, H. Q.   ; Ishitsuka, Etsuo   ; Iigaki, Kazuhiko 

Detailed neutron flux distribution is important to understand the neutronic behavior during operation as well as to precise the core optimization and safety analysis of a reactor. In the literature, no calculations have been performed to show the detailed neutron flux map for the high temperature engineering test reactor (HTTR) because of the limitation of the old neutronic codes and the low performance of the computing system. The present work deals with MCNP6 Monte-Carlo calculation to determine the detailed neutron flux map in the HTTR during normal operation. At first, the calculation of neutron flux at several positions in the reactor was validated by comparing the corresponding reaction rate between the calculation and measurement. After that detailed neutron flux with the small cells of 1cm 1cm 10cm was obtained for the entire reactor core using the fmesh tally of MCNP6 code.