Simulation of Neutron Multiplicity Counting Based on Monte Carlo Code RMC

Abstract

Neutron multiplicity pertains to the probability distribution of the quantity of neutrons released during induced or spontaneous fission processes within fissile materials. The technology for neutron multiplicity measurement leverages temporal correlations in the emission of fission neutrons from nuclear materials. It employs mathematical tools to elucidate the processes of neutron generation, multiplication within the nuclear material, and detection of outside nuclear materials. In this paper, two multiplicity counting methods are devised building on the RMC (Reactor Monte Carlo) code. The results obtained from both methods, including singles, doubles, and triples counting rates, exhibit good agreement with MCNP. Additionally, parameters associated with the detection efficiency and decay time of the apparatus are computed. By amalgamating the acquired singles, doubles, and triples counting rates, the mass of fissile material within the sample is inversely determined using a passive method with the point model equation. Notably, the point model equation reveals that spontaneous fission neutrons and induced neutrons possess distinct energy spectra, challenging the validity of the assumption that the probability of neutrons being captured without causing fission can be disregarded. In light of these considerations, the neutron multiplicity counting equation was rederived. The accuracy of the Monte Carlo simulation results is improved using the new method.