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This paper presents a new approach to the spent nuclear fuel (SNF) problem, which is uniquely enabled by a liquid fuel form, specifically as in the case of molten salt reactor (MSR) systems. Managing the SNF problem is critical for public acceptance of nuclear power as a climate change solution. An MSR can be refueled while operating by adding more fresh fuel salt, which grows the in-core fuel salt volume. This growth will eventually double the size of the original fuel salt, allowing to start another core with this excess fuel so long as the daughter reactor is of the same design and there is sufficient excess fuel. This study explores how such a "sourdough" strategy would work in MSRs and provides an initial calculation methodology to find the correct refueling rates to match the desired growth curve of power generation. Higher uranium enrichment levels of the refuel salt result in lower refueling rates and thus a longer doubling time. As a result, the refuel salt uranium enrichment can be tailored to match a postulated clean power generation capacity expansion. This approach allows postponing the spent nuclear fuel disposal issue using the sourdough method. Along with the MSR fuel's unique properties, it suggests a new path towards managing nuclear waste until long-term solutions become economically viable. [ABSTRACT FROM AUTHOR]

There are a wide variety of fast spectrum molten salt reactor (FSMSR) designs and corresponding fuel cycles that shown promise. These designs differ in features. Therefore, it becomes difficult to determine whether a set of parameters is applicable to all, one, or none of these designs. This paper presents a number of iterations of a FSMSR concept with a sodium chloride carrier salt. Different design characteristics and their effects on the neurotics parameters are shown. The results show that effective delayed neutron fraction, mean neutron generation time, neutron yield, and energy of neutrons causing a fission all follow trends based off the heavy metal concentration and leakage rate. This paper also gives preliminary fuel cycle calculations and the evolution of the neutronic parameters over burnup. A FSMSR initially fueled with low enriched uranium will naturally breed at a certain core volume, without a need for refueling, for decades of full power operation. [ABSTRACT FROM AUTHOR]

This paper provides an overview of the current MSR design space and lists unique features of the various designs under consideration. Some general considerations for source terms calculation for Molten Salt Reactors (MSRs) are explained. Applicability and limitations of terminology currently defined for legacy light water reactor (LWR) systems are discussed in the view of MSRs and the need for updated terminology is discussed. Calculations carried out for the Molten Salt Reactor Experiment (MSRE) are discussed with a qualitative comparison to the designs presented. The nature of the fission products (FPs) and actinides for Low enriched uranium, thorium and fast U/Pu fuel cycles employed in representative molten salt reactor systems are discussed. Computational results are obtained from a code (Serpent 2) with online reprocessing. Divergence in source terms when fission product bubbling is demonstrated. The source release for each molten salt reactor during postulated accidents is also presented. Image 1 • Discussion on terminology for molten salt source terms and accident scenarios are presented. • Burnup calculations and assumptions for source term analysis is presented. • Three molten salt reactor designs are compared against each other with respect to source term release. [ABSTRACT FROM AUTHOR]

[Display omitted] • Thorium-fueled iso-breeder design framework was developed • A specific design inspired by ORNL-4528 concept is shown • Reactivity feedback coefficients, core temperature behavior, control rod positioning and worths are discussed • Iso-breeding fuel cycle framework is developed and demonstrated This paper presents reactor physics modeling to quantify behavior of a thorium-fueled molten salt iso-breeder reactor (MSiBR) system similar to the two-fluid molten salt breeder reactor (MSBR) documented in ORNL-4528. A Python framework was developed to allow for a flexible, fully parameterized modeling strategy for analysis of the fuel cycle and core neutronics. Modeling was performed in Serpent-2 to assess the MSiBR design and performance. The framework can be used for direct evaluation of reactor design parameters such as the fuel and blanket salt fraction, lattice geometry, and reactor core size. A full core model of the MSiBR system is presented. Temperature feedback coefficients of reactivity for the fuel salt, blanket salt, and graphite are calculated. It is observed that the present model demonstrates passive and inherent self-regulation of reactivity. The developed framework is also used to demonstrate a methodology for evaluating two-fluid reactors operating with a thorium fuel cycle. This work is intended to provide a baseline for future research into the development of two-fluid MSR designs. [ABSTRACT FROM AUTHOR]