Response calculation of a neutron detector

not registered; Takagi, Shunji*

The responses were calculated for the burnup measurement equipment placed in the spent fuel storage facility beside JOYO by using the continuous energy Monte Carlo code "MCNP-4A". In this work, the responses were calculated without cosidering the neutron multiplication in a spent fuel and were compared to the responses considering neutron multiplication to clarify the effect of neutron multiplication in a spent fuel. The calculated results show that the neutron multiplication in a spent fuel not only increases the count rates but also influence to the axial count rate distribution. This count rate distribution was flattened by the neutron multiplication.