Citation:
M. Derman, “Nuclear Safeguards Feasibility Study for a Molten Salt Reactor Using MCNP Modeling and Simulations”, M.S. Thesis, Nuclear Engineering, Texas A&M University, College Station, TX (2024).
Abstract:
The technological developments within the last couple of decades have exponentially increased the demand for reliable and sustainable energy. The successful fulfillment of this energy demand shows a strong correlation with the Human Development Index, which is based on health, education,rnand income parameters. Among other energy sources, nuclear energy has become prominent in providing clean energy by protecting air quality, having a small footprint with high energy density, and being a reliable and stable option. However, considering the dual nature (peaceful andrnnon-peaceful uses) of nuclear energy, nuclear safeguards are an important international instrument to prevent nuclear material diversion for non-peaceful purposes. This work focused on developing a nuclear safeguards monitoring approach for a generic Molten Salt Reactor (MSR) designed at Texas A&M University.rnrnThis thesis includes a comprehensive neutronics modeling of the MSR using the Monte Carlo radiation transport code, MCNP®6.2. The modeled MSR has a 300 MWth power and operates in a thermal neutron spectrum at 900 K. It uses molten fluoride salt (2LiF:BeF2) as a coolant with UF4 fuel mixture with 3.5% low-enriched uranium (LEU). The reactor core design used graphite as the neutron moderator and reflector. Non-soluble fission products (FP) were extracted through gaseous extraction, and FP removal was conducted to improve the performance. A High-Purity Germaniumrn(HPGe) detector, a widely used Non-Destructive Assay (NDA) equipment was modeled for Special Nuclear Material (SNM) mass quantification. In this methodology, the first step was determining the relationship between the Pu amount and fuel burnup. In the second step, the fuel burnuprnrelationships with the radioactivity of 137Cs, the radioactivity ratios of 134Cs/137Cs, and 154Eu/137Cs were established. In the last step, these relationships were used to estimate SNM mass. The results indicate that the Pu amount relationships with 137Cs radioactivity and the ratio of 134Cs/137Cs can be utilized to quantify SNM mass at all fuel burnup levels. The ratio of 154Eu/137Cs is applicable even for very-high fuel burnup levels. However, it does not provide accurate results at ultra-high fuel burnup levels due to its saturation. The proposed safeguards monitoring approach in this thesis provides a method for estimating the Pu mass in the MSR at different fuel burnup levels so that any diversion of Pu for non-peaceful purposes can be prevented through early detection and deterrence.rn