Neutron fluence analyses around the reactor pressure vessel of BWR using MCNP with a heterogeneous and homogeneous mixed core model

Abstract

For reactor pressure vessel (RPV) material surveillance program, it is necessary to obtain fast neutron fluence. In this work, Monte Carlo transport code MCNP is applied to analyze it for a Boiling Water Reactor (BWR) with a heterogeneous and homogeneous mixed core model (HHMCM) and its applicability is examined. The analyses of using MCNP with HHMCM are performed to obtain the neutron flux and the reaction rate of the dosimeter wires at the inner surface of the RPV of an existing 800 MWe BWR plant in Japan. As a result, the neutron flux and the reaction rates can be estimated with an uncertainty of 8% at most. In addition, HHMCM can reduce a calculation time to 1/9 compared with a case of all bundles treated as heterogeneous.