Summary
Fissile material detection and quantification are often necessary for safeguards, nuclear security, and fuel management. Nondestructive assay, neutron, and gamma measurements are reliable means, which can facilitate the detection and estimation of the mass of fissile materials in a broad range of material matrix. Various flavours of neutron measurement are routinely used by facilities (like nuclear reactors, enrichment, and fabrication plants) to quantify fissile material mass and inventory lists. The Monte Carlo code, MCNP6, is used to model several neutron multiplicity measurements. A simulation scenario is set up in MCNP6 using the JCC71 neutron slab counter to obtain the multiplicity moments for fresh and irradiated fuel assemblies from the UMass Lowell Research Reactor (UMLRR) and Worcester Polytechnic Research Reactor (WPIRR). An MCNP6 burnup is initially performed on the fuel types under study to generate used fuel isotopic. The fresh and or used fuel isotopic is then used to produce independent SOURCES4c input tape1 files. SOURCES4c is used to generate (α, n), spontaneous fission spectrum, and the associated neutron emission rates necessary for the various fixed fuel source definitions in MCNP6 calculations. Under the comprehensive safeguards agreements, the International Atomic Energy Agency has the right and obligation to verify that no nuclear material is diverted from peaceful use to nuclear weapons or other nuclear explosive devices. Research reactors are required to be safeguarded facilities under the comprehensive safeguards. Several research efforts have studied the various flavours of neutron measurement for commercial power reactor operating at high power and long burnups; however, not nearly as many studies have been performed with neutron measurements for research reactors operating at relatively lower power and have significantly lower burnup. This work looks to establish the relevant isotopes to overall neutron source rate as well as the process involved in performing a typical neutron multiplicity measurement simulation for a research reactor fuel. The results demonstrate that the single and double moments for Worcester Polytechnic Institute (WPI) and UMLRR fuels can be measured reliably using two JCC71 slab detectors. The moment for the UMLRR and WPIRR fuel (in both fresh and used states) was estimated with a relative error below 0.031 for singles and 0.081 for doubles. The two fresh fuel types cannot be differentiated from each other on the sole basis of neutron analysis. However, fresh and irradiated fuel can be distinguished based on neutron multiplicity measurements.
