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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]

The family of Generation-IV reactor concepts comprises of multiple promising designs, one of which is the molten salt reactor. These reactors have traditionally been chosen for the possible use of the thorium-uranium fuel cycle and used 7 LiF-BeF 2 carrier salt. This particular salt choice however presents several challenges due to the cost of highly depleted 7 Li isotope for the carrier salt, tritium production, and beryllium toxicity. Additionally, lack of developed and accepted safeguards methodology for thorium fuel cycle presents a barrier. While none of these issues are insurmountable, alternatives are worth investigating. The purpose of this paper is to analyze the more cost effective and regulatory amenable fuel salt choices by using low-enriched uranium fuel in the form of UF 4 . Several eutectic mixtures are examined that avoid the use of 7 Li and Be while maintaining a melting point low enough to be compatible with standard structural materials. The optimal conditions for hexagonal lattice arrangements using nuclear graphite moderation are discussed for multiple fuel salt choices. The aim of this study is to present options of using simpler molten salt reactor alternatives focused on thermal single-fluid low-enriched uranium converter concepts. [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]

Nodal dynamic models allow rapid reduced-order simulations of complex systems. This paper presents a publicly available system dynamic model of a molten salt reactor system with a topology of the Molten Salt Reactor Experiment (MSRE), in MATLAB/Simulink, building upon a model previously published and validated with MSRE data. It adds scaling of nominal power, dynamic representation of decay heat generation, a generic decay heat removal system, improved documentation, modularity, and test cases. One hour of system transient runs in less than 10 min on a laptop. This model is intended for first order engineering calculations, developing insights into system transient behavior, equipment sizing, parameter sensitivity studies. The model contains an easy to use toolkit which can be customized and expanded to describe any molten salt reactor system. Scenarios involving off-normal transients are presented. Relation between shutdown reactivity insertions, decay heat removal, core recriticality, and safe shutdown conditions are discussed. [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]