Comparison of the single and double count using MCNP6 and ONMS Geant4 software. (March 2020)

• Record Type: Journal Article
• Authors: Dim, O.U.; Aghara, S.K.; Kütt, M.
• Is Part Of: Progress in nuclear energy. Volume 121(2020)
• Publication Date: 2020-03
• Subjects: Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 333.7924
• DOI: 10.1016/j.pnucene.2020.103240
• Languages: English
• ISSNs: 0149-1970

Abstract

Abstract: The quantification and detection of fissile material are valuable for safeguards, nuclear security, and fuel management. Non-Destructive Assay (NDA) (i.e. using neutron and gamma measurements) provide reliable means to detect and estimate the fissile content (i.e. mass) in a vast range of material in different forms found across various facilities in the nuclear fuel cycle. Neutron measurements are used either for total counting (i.e. pure metal fissile samples) or coincidences/multiplicity counting (i.e. impure fissile samples). These various flavors of neutron measurement are routinely used to quantify fissile material mass and inventory lists. Monte Carlo codes have evolved considerably and are used routinely for modeling and simulation of diverse physics. MCNP6 (Monte Carlo N Particle) and ONMS-Geant4 (Open Neutron Multiplicity Software) are used in this work to generate single and double count (i.e. based on the accidental and real plus accidental histograms) for two fresh fuel sources with the assumption that the neutron emission process is solely due to spontaneous fission. Previous work has shown that for these fresh fuel sources, about 93% of the neutrons emitted is due to spontaneous fission. ONMS was designed to provide an open-source platform that will facilitate researchers in the nuclear community to perform neutron counting simulations. ONMS has been shown to give good results for detectors with high efficiencies, strong (α, n) sources, strong spontaneous fission sources and sources baring plutonium. Detailed model simulations result from ONMS and MCNP6 are compared for a weak californium source and uranium-bearing fresh fuel assembly using a detector set up with relatively low efficiency. The detector response is characterized to compare the single and double counts obtained from both code packages. For the weak californium source, the neutron triplets are also evaluated from both code packages.