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Several molten-salt reactor (MSR) concepts are currently being developed by private vendors and national organizations. Molten-Salt Reactor Experiment (MSRE) is the only MSR that operated for several years, for which the performance was well documented in the literature. Experimental data from MSRE serve as the standard for developing models and tools for MSR analysis. MSRE underwent several design changes from conception to construction to operation. Model validation requires the use of parameters consistent with those that existed during the various measurements. The paper therefore presents a brief discussion of the parameters that evolved over time, and details of sources for all the parameters used in the modeling. This report is part of a series of papers on modeling MSR dynamics. The goal here is to validate the modeling approach by comparing simulation results with experimental data. For this purpose, two nonlinear dynamic models are developed. The first model uses a one-region one-dimensional representation of the core. The second reproduces the nine-region core model developed by Oak Ridge National Laboratory for MSRE studies. The dynamics are analyzed by simulating transients during normal power operation. The frequency response of the models is also evaluated. Both models show excellent agreement with experimental data for both U-235 and U-233 fuels. Some apparent discrepancies in experimental data are reported and their consequences discussed. The one-region model is modified to simulate reactivity excursions resulting from operational anomalies, and the results are presented. This modeling approach can be applied to other MSR systems provided valid system parameters. [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]

• Diversion of plutonium results in changes in kinetic parameters. • Dynamic models reveal signatures of plutonium removal in an Molten Salt Reactor. • Plutonium diversion ca be seen in Molten Salt Reactor frequency response. Safeguarding a Molten Salt Reactor (MSR) is an underexplored research topic. The high temperature, radiation, and corrosive nature of the fuel salt make installing and using instrumentation difficult. This study introduces a new innovative way of using the reactor's frequency response to determine if plutonium is diverted. As a low enriched uranium fueled MSR undergoes burnup, plutonium builds up in the fuel salt and increasingly contributes to fissions. As a result, the kinetic parameters, mean neutron generation time and delayed neutron fraction, evolve with depletion. These changes are large enough to be discerned as unique patterns in the MSR system frequency response. Likewise, a specific pattern is seen if there is a removal of plutonium. Given the simplicity of establishing a frequency response of an operating MSR, this method presents a breakthrough in MSR safeguards. [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]

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]