Monte Carlo N-Particle eXtended (MCNPX) Simulation for Passive Neutron Measurement of Fuel Debris at Fukushima Daiichi Nuclear Power Plants

To quantify the nuclear materials in the fuel debris at Fukushima Daiichi Nuclear Power Plants (1F), we, Plutonium Fuel Development Center of JAEA, are considering applying passive neutron techniques, such as Neutron Multiplicity, Differential Die-away Self-Interrogation and Passive Neutron Albedo Reactivity. In order to evaluate the applicability of passive neutron techniques to the fuel debris measurement, we investigated the neutron behavior in the fuel debris by using Monte Carlo N-Particle eXtended transport (MCNPX) simulation code. Because the physical and chemical properties of the fuel debris at 1F are not clear at this moment, source term data used for simulation were prepared by referring to Three Mile Island data such as density of fuel debris, fill volume, and canister for fuel debris. The results show that passive neutron technique supplemented with gamma measurement technologies and burn-up calculation code might have enough applicability to quantify the nuclear material in fuel debris. This paper provides the results of MCNPX simulation for the fuel debris measurement at 1F with passive neutron techniques.