Computational Methods and Tools for Neutronics Calculations for Molten-Salt Nuclear Reactors

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Molten-Salt Nuclear Reactors (MSRs) are one of the six Generation-IV advanced nuclear-reactor concepts being pursued by the Generation IV International Forum (GIF). Canada is a signatory of the GIF Framework Agreement and at least one Canadian company is developing an MSR concept and is seeking licensing from the Canadian Nuclear Safety Commission. MSRs use molten (liquid) salts of uranium and/or thorium which act as both fuel and coolant. The MSR core is not pressurized and molten salts have high boiling points, which make the release of radioactivity into the environment near-impossible; even following an accident. MSRs are promising as burners of transuranic elements from spent Light-Water-Reactor fuel (the most common type of power-reactor fuel in the world) thus having the potential to reduce the amount of waste from such reactors. They can provide high conversion/breeding ratios thus increasing the nuclear-fuel utilization. In MSRs, the liquid salt circulates continuously in and out of the core, which poses unique challenges for the neutronics simulations of such reactors by adding additional, “drift”, terms to the balance equations solved by neutronics codes. Consequently, neutronics codes in use for solid-fuel reactors are not adequate for liquid-fuel reactors, such as the MSRs. While neutronics computational methods and tools for solid-fuel reactors are mature, computational methods and tools for liquid-fuel reactors are in their infancy and constitute an active area of investigation. To date, no stand-alone neutronics computational tool for MSRs exists in Canada. The proposed research program is centered on the development of high-fidelity, reliable neutronics computational methods and tools for MSRs; tools that can properly account for fuel convection for three-dimensional static and time-dependent problems. Such methods require the coupling of neutronics and fluid-dynamics calculations. The research will focus on the neutronics methods and their coupling with fluid-dynamics computations, while the actual simulation of the fuel flow will be performed using an existing software package applicable to Computational Fluid Dynamics (e.g. OpenFOAM). The research will produce a stand-alone computational tool for researchers, engineers and regulators to use for the design and safety analysis of Generation-IV MSRs. Such a tool will make it possible to accurately and reliably analyze the neutronic characteristics of existing and emerging MSR designs. The training of five highly-qualified personnel will be supported by the research program. On a wider scale, the research will contribute to the development of MSRs, which promise CO2-free electricity production with minimal risk and radioactive waste. It will also support Canadian nuclear-technology companies involved in MSR research, placing them at the forefront of nuclear-technology development in the world.