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AbstractPost-irradiation examination of the Molten Salt Reactor Experiment from the 1970s revealed intergranular cracking of the salt-facing material, Hastelloy-N, from the penetration of fission products, specifically tellurium (Te), into the components. Stainless steel 316H is a candidate salt-facing structural material for future molten salt reactors due to its excellent corrosion, oxidation, and neutron irradiation resistance. Thus, studies are needed to verify if Te may lead to material degradation of salt-facing components made from 316H.This work examined the behavior of stainless steel 316H in three conditions: as received, heat treated to 800°C for 100 h without Te, and with a highly concentrated Te environment. After exposure, mechanical testing was performed on all samples to reveal the loss of strength and ductility in the Te-exposed samples. Additional analysis of the Te-exposed 316H samples using scanning electron microscopy displayed intergranular embrittlement and energy-dispersive X-ray spectroscopy maps highlighted the infiltration of Te within grain boundary cracks. These results present the need for additional experiments to understand how Te weakens the structural material, especially in molten salt, and to eventually identify the driving mechanism for this observed behavior. [ABSTRACT FROM AUTHOR]

The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10−4); the splitting method could reach a precision of 10−2 level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors. [ABSTRACT FROM AUTHOR]

This work presents neutronics models of a small and large fast-spectrum molten chloride-salt reactor. The models are similar to designs being pursued by industry, and they may serve as generic preconceptual and simplified neutronics models that provide information for decision making in licensing-related areas. The two models were created using Serpent, a Monte Carlo neutron transport code, and Moltres, a neutron diffusion core simulator tool. Specifically, this study focused on exploring the applicability of diffusion theory to fast molten salt reactor (MSR) models, the capabilities of an open-source, MSR-oriented simulation tool (Moltres), and optimal energy-group structures. The proposed two-step method involves group-constant generation with Serpent and a multigroup diffusion solution by Moltres. Three energy-group structures were applied. The accuracy of the solutions was determined through comparisons between the two-step and Monte Carlo flux and multiplication factor solutions. The findings indicated diffusion theory captures neutronics with minimal error for the large MSR and yielded best results with the 27-group structure. The 27-group structure yielded an average group flux error below 2% and keff agreement between diffusion and transport solutions within 30 pcm. The accuracy of the two-step method decreased for the very small (high-leakage) fast chloride MSR, but the neutronics were captured acceptably well with the 33-group structure. In addition to exploring the capabilities of Moltres, this work contributes to the sparse literature involving open-source models of fast-spectrum MSRs. Future work is noted as expanding the capabilities of the neutronics models to incorporate thermal hydraulics. [ABSTRACT FROM AUTHOR]

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Marine sources contribute approximately 2% of global energy-related CO₂ emissions, with the shipping industry accounting for 87% of this total, making it the fifth-largest emitter globally. Environmental regulations by the International Maritime Organization (IMO), such as the MARPOL (International Convention for the Prevention of Pollution from Ships) treaty, have driven the exploration of alternative green energy solutions, including nuclear-powered ships. These ships offer advantages like long operational periods without refueling and increased cargo space, with around 200 reactors already in use on naval vessels worldwide. Among advanced reactor concepts, the molten salt reactor (MSR) is particularly suited for marine applications due to its inherent safety features, compact design, high energy density, and potential to mitigate nuclear waste and proliferation concerns. However, MSR systems face significant challenges, including tritium production, corrosion issues, and complex behavior of volatile fission products. Understanding the impact of marine-induced motion on the thermal–hydraulic behavior of MSRs is crucial, as it can lead to transient design basis accident scenarios. Furthermore, the adoption of MSR technology in the shipping industry requires overcoming regulatory hurdles and achieving global consensus on safety and environmental standards. This review assesses the current progress, challenges, and technological readiness of MSRs for marine applications, highlighting future research directions. The overall technology readiness level (TRL) of MSRs is currently at 3. Achieving TRL 6 is essential for progress, with individual components needing TRLs of 4–8 for a demonstration reactor. Community Readiness Levels (CRLs) must also be addressed, focusing on public acceptance, safety, sustainability, and alignment with decarbonization goals. [ABSTRACT FROM AUTHOR]

A molten salt reactor is one of the fourth-generation reactors and is considered to be a feasible replacement for current reactors due to their many advantages. However, there are a number of issues that remain; one of which is the corrosion of the materials. Corrosion problems in molten salt reactors have been reported since The Molten Salt Reactor Experiment at Oak Ridge National Laboratory in the 1960s. There have been many attempts to mitigate the corrosion problem, but a fundamental solution has not yet been achived. In this study, surface reaction-diffusion-coupled simulations were performed to simulate the corrosion of a Ni–Cr–Fe material, a prototype of Hastelloy N, which is being promoted as a structural material for molten salt reactors in F–Li–Na–K eutectic salts. This surface reaction-diffusion-coupled simulation framework was developed to study which corrosion reactions are dominant in molten salt environment corrosion where a large number of oxidation–reduction reactions exist, the correlation between composition of alloy and corrosion rate, and the effect of Cr depletion on corrosion. [ABSTRACT FROM AUTHOR]

The molten salt reactor (MSR) flowing-fuel simulation capability of the Griffin-Pronghorn-coupled multiphysics code system developed by Idaho National Laboratory (INL) was verified against the Center National de la Recherche Scientifique (CNRS) MSR benchmark problem. Griffin and Pronghorn, which are INL's neutronics and thermal-hydraulics codes built upon the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework, have been recently extended to handle the flowing fuel of MSRs causing the drift of delayed neutron precursors (DNP). In the Griffin-Pronghorn code system, Griffin provides the fission rate density to Pronghorn, which simulates the generation, decay, and transport of DNPs along with the fluid, and the redistributed DNP densities are fed back to Griffin. The coupling and transfers are largely automatically managed at the framework level by the powerful MultiApp system of MOOSE. The verification results against the CNRS benchmark problem demonstrate that the Griffin-Pronghorn code system can accurately simulate the unique physics phenomena of MSRs in both steady-state and transient conditions. [ABSTRACT FROM AUTHOR]

AbstractThe advanced thermal-hydraulic system code, System Analysis Module (SAM), was originally developed for the modeling of single-phase flow in advanced reactors. It has since been expanded to include a four-equation drift flux model for the modeling of two-phase flows containing a noncondensable gas. The model was expanded to support the modeling of molten salt reactor (MSR) designs in which the fuel is directly dissolved in the circulating coolant. These designs have shown that circulating gas bubbles can play an important role in the management of fission products and the operational behavior of the reactor. A drift flux model was implemented to more accurately capture the localized behavior of the void in the core and its impact on the mass transfer of fission products.A thorough assessment of the new model was performed by developing a verification and validation test suite. Verification problems were designed to test all major terms in the new governing equations. The new model converged to the correct solution at the expected order of accuracy for all verification cases. The validation cases included a wide range of flow and void conditions in different pipe geometries. Although higher void experiments show a slight underprediction of void by the drift flux model, experiments that aim to reproduce Molten Salt Reactor Experiment (MSRE) experimental conditions show good agreement with the model.The gas transport model was activated for a SAM model of the MSRE to demonstrate that it can be used in a more complex model. This gas transport model will be used along with an interfacial area transport equation being implemented in SAM for the prediction of mass transport behavior in MSR conditions. [ABSTRACT FROM AUTHOR]

AbstractIn recent years, there has been renewed interest in molten salt reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, SPECTRA and SAM are two system analysis codes that have been enhanced to include MSR-specific modeling capabilities, including delayed neutron precursor drift and modified point kinetics equations.This paper discusses the efforts taken to verify and validate these features. A standard MSR system test problem was developed to verify and demonstrate the capability of SPECTRA and SAM on the MSR transient simulation. Sixteen transients were simulated. The results obtained from SPECTRA and SAM show good agreement. The Molten Salt Reactor Experiment transient experiments were reviewed and selected to validate the SPECTRA and SAM codes. The experiments included pump startup and coastdown tests at zero power, reactivity insertion tests at different power levels, frequency tests, and a natural convection test. The simulation results from SPECTRA and SAM show good agreement with the experimental data. [ABSTRACT FROM AUTHOR]

General interest in the deployment of molten salt reactors (MSRs) is growing, while the available data on uranium-containing fuel salt candidates remains scarce. Thermophysical data are one of the key parameters for reactor design and understanding reactor operability. Hence, filling in the gap of the missing data is crucial to allow for the advancement of MSRs. This study provides novel data for two eutectic compositions within the NaF-KF-UF4 ternary system which serve as potential fuel candidates for MSRs. Experimental measurements include their melting point, density, fusion enthalpy, and vapor pressure. Additionally, their boiling point was extrapolated from the vapor pressure data, which were, at the same time, used to determine the enthalpy of vaporization. The obtained thermodynamic values were compared with available data from the literature but also with results from thermochemical equilibrium calculations using the JRCMSD database, finding a good correlation, which thus contributed to database validation. Preliminary thoughts on fluoride salt reactor operability based on the obtained results are discussed in this study. [ABSTRACT FROM AUTHOR]

Non-impregnated small-pore graphite (NSPG), which has a compact microstructure and is used in molten salt reactors (MSRs), was prepared by a novel process. The pore diameter of NSPG was reduced to ~800 nm. The irradiation evaluation of NSPG was carried out by 7 MeV Xe26+ ion irradiation. The microstructural changes of NSPG were investigated with IG-110 as a comparison. The graphitization degree of NSPG was higher than that of IG-110, though it was not subjected to an impregnation process. Under low-dose ion irradiation (<2.5 dpa), the microscopic morphology of the NSPG changes in a small magnitude, and the lamellar structure of graphite remains within the scale of more than a dozen nanometers, which exhibits a better resistance to irradiation. With the increase in irradiation dose, the accumulation of defects leads the graphite toward amorphization, which shows consistency with IG-110. This study provides an efficient and low-cost method for the preparation of graphite for MSR, and investigates the damage behavior of graphite, which is of great significance in accumulating data for the development of MSR nuclear graphite and the optimal design of graphite materials. [ABSTRACT FROM AUTHOR]

In the framework of the European project SAMOSAFER, this numerical study focuses on some thermal aspects of the Emergency Draining Tank (EDT) located underneath the core of a Molten Salt Reactor. In case of an emergency, this tank passively receives the liquid fuel salt and is designed to ensure a subcritical state. An important requirement is that the fuel does not overheat to maintain the EDT Hastelloy container integrity. The present EDT is based upon a group of hexagonal cooling assemblies arranged in a hexagonal grid and cooled down thanks to conduction through the inert salt layer up to an air flow in charge of removing the heat. This numerical thermal study relies on a conjugated heat transfer analysis coupling a Finite Element solid thermal code (SYRTHES) and two instances of a Finite Volume CFD codes (Code_Saturne). Calculations on an initial design suggest that a simple center airpipe flow is likely to not sufficiently cool the device. Alternative solutions have been evaluated. Introduction of fins to enhance the heat transfer do not bring a noticeable improvement regarding maximum temperature reached. However, a solution in which the central pipe air flow is replaced by several cooling channels located closer to the fuel is investigated and suggests a better cooling.

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl3 fuel and NaCl–KCl–MgCl2 coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

A series of experiments were performed to produce a fuel source for a molten salt reactor (MSR) through pyroprocessing technology. A simulated LiCl–KCl–UCl3-NdCl3 salt system was prepared, and the U element was fully recovered using a liquid cadmium cathode (LCC) by applying a constant current. As a result, the salt was purified with an UCl3 concentration lower than 100 ppm. Subsequently, the U/RE ingot was prepared by melting U and RE metals in Y2O3 crucible at 1473 K as a surrogate for RE-rich ingot product from pyroprocessing. The produced ingot was sliced and used as a working electrode in LiCl–KCl–LaCl3 salt. Only RE elements were then anodically dissolved by applying potential at −1.7 V versus Ag/AgCl reference electrode. The RE-removed ingot product was used to produce UCl3 via the reaction with NH4Cl in a sealed reactor.

In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC's input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that keff agreed within 10 pcm while the flux distribution in space and energy generally agreed within 0.1%. Larger differences were observed between the OpenMC CAD and CSG models; notably, the keff computed for the CAD model was 1.00872, which is more than 1% lower than the value for the CSG model and much closer to experiment. Several areas of the reactor that were modeled differently in the CSG and CAD models were discussed and, in several cases, their impact on keff was quantified. Lastly, we compared the computational performance and memory usage between the CAD and CSG models. Simulation of the CSG model was found to be 1.4-2.3× faster than simulation of the CAD model based on the Embree ray tracer while using 4× less memory, highlighting the need for continued improvements in the CADbased particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer. [ABSTRACT FROM AUTHOR]

AbstractThe CEA multiphysics tool combining the deterministic neutronics code APOLLO3® and the computational fluid dynamics (CFD) platform TRUST/TrioCFD is used to model the first-ever-built molten salt nuclear reactor, the Aircraft Reactor Experiment (ARE). A neutronics model and a thermal-hydraulic model of the reactor were created and coupled. Steady-state and transient simulations were performed in order to reproduce experiments realized on the ARE. The simulation results and experimental data are compared as a way of validating the multiphysics tool. The nominal state of the ARE is reproduced first; significant discrepancies were observed regarding the liquid sodium flow. Variations of $${\beta _{eff}}$$βeff with the fuel flow rate were then studied. While the simulation was in very good agreement with the experimental data for high flow rates, some discrepancies were observed at low and null flow rates. Finally, a transient simulation of a rod withdrawal was reproduced. While the beginning of the transient simulation was in good agreement with the experimental data, oscillations appeared in the second half of the simulation. [ABSTRACT FROM AUTHOR]

The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10−4); the splitting method could reach a precision of 10−2 level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors.

Marine sources contribute approximately 2% of global energy-related CO₂ emissions, with the shipping industry accounting for 87% of this total, making it the fifth-largest emitter globally. Environmental regulations by the International Maritime Organization (IMO), such as the MARPOL (International Convention for the Prevention of Pollution from Ships) treaty, have driven the exploration of alternative green energy solutions, including nuclear-powered ships. These ships offer advantages like long operational periods without refueling and increased cargo space, with around 200 reactors already in use on naval vessels worldwide. Among advanced reactor concepts, the molten salt reactor (MSR) is particularly suited for marine applications due to its inherent safety features, compact design, high energy density, and potential to mitigate nuclear waste and proliferation concerns. However, MSR systems face significant challenges, including tritium production, corrosion issues, and complex behavior of volatile fission products. Understanding the impact of marine-induced motion on the thermal–hydraulic behavior of MSRs is crucial, as it can lead to transient design basis accident scenarios. Furthermore, the adoption of MSR technology in the shipping industry requires overcoming regulatory hurdles and achieving global consensus on safety and environmental standards. This review assesses the current progress, challenges, and technological readiness of MSRs for marine applications, highlighting future research directions. The overall technology readiness level (TRL) of MSRs is currently at 3. Achieving TRL 6 is essential for progress, with individual components needing TRLs of 4–8 for a demonstration reactor. Community Readiness Levels (CRLs) must also be addressed, focusing on public acceptance, safety, sustainability, and alignment with decarbonization goals.

A molten salt reactor is one of the fourth-generation reactors and is considered to be a feasible replacement for current reactors due to their many advantages. However, there are a number of issues that remain; one of which is the corrosion of the materials. Corrosion problems in molten salt reactors have been reported since The Molten Salt Reactor Experiment at Oak Ridge National Laboratory in the 1960s. There have been many attempts to mitigate the corrosion problem, but a fundamental solution has not yet been achived. In this study, surface reaction-diffusion-coupled simulations were performed to simulate the corrosion of a Ni–Cr–Fe material, a prototype of Hastelloy N, which is being promoted as a structural material for molten salt reactors in F–Li–Na–K eutectic salts. This surface reaction-diffusion-coupled simulation framework was developed to study which corrosion reactions are dominant in molten salt environment corrosion where a large number of oxidation–reduction reactions exist, the correlation between composition of alloy and corrosion rate, and the effect of Cr depletion on corrosion.

An experimental complex was created at the NRC Kurchatov Institute for testing of structural materials developed for MSR. The complex makes it possible to implant helium and (or) hydrogen into material specimens at the U-150 cyclotron, simulating the result of long-term neutron exposure, to perform mechanical tests and microstructural studies before and after simulation exposure, and to test the compatibility of structural materials before and after exposure with fuel salt in non-isothermal dynamic conditions at temperatures up to 750°C with control of the melt redox potential. Preliminary tests with samples of nickel-molybdenum alloy KhN80MT demonstrated the efficiency of the proposed experimental technique. The concentration of helium nuclei in the alloy samples reaches ~150 ppm for approximately 10 h of irradiation. [ABSTRACT FROM AUTHOR]

The paper is concerned with the use of uranium wastes after purification of regenerated fuel from 232U in a molten salt reactor (MSR) in application to minor actinide (MA) burning. The comparison of the uranium and plutonium loadings in an MSR was performed. The results of calculations show that the uranium loading in an MSR is a possible option that has a number of potential advantages: it allows for a higher rate of MA burning during the initial period of MSR operation with approximately the same burnup efficiency and increases the effective fraction of delayed neutrons. The presented calculations are of an approximate nature and first of all show the possibility of effective use of uranium from the waste stream of the enrichment cascade as a fissile material. [ABSTRACT FROM AUTHOR]

Fuel channel flow blockage is a postulated accident scenario in graphite-moderated molten salt reactors (MSRs), caused by debris partially obstructing the flow in a hole inside a graphite moderator brick. This obstruction produces an increase in the fuel salt bulk temperatures in the blocked channel, resulting in a significant fraction of the deposited fission power being conducted through graphite from the blocked channel to the surrounding unblocked channels. In the blocked channel, a combination of forced convection and natural convection takes place, the latter being dominant at very strong flow rate reductions. At full blockage (with the fragment completely obstructing the flow), because of the poor thermal conductivity of fuel salt, correct estimation of the natural convection impact on heat transfer between the channel bulk and the walls is fundamental to determine whether boiling conditions are reached in the blocked channel. The proposed assessing approach, developed in collaboration with ThorCon in the frame of the ThorCon reactor core design and optimization, is presented in this work. Results from a case study based on the molten salt breeder reactor (MSBR) design and operating conditions are presented and discussed. These analyses show that with respect to full blockage accidents in other reactor types, the consequences of full channel blockage accidents in graphite-moderated MSRs are milder. For graphite-moderated MSRs, the main mechanical limits during the accident arise in thermal stresses in the graphite brick. Graphite thermal and mechanical properties, namely, bulk modulus, thermal expansion coefficient, ultimate strength, and thermal conductivity, are severely impacted by neutron irradiation through the life of the graphite. All these effects are taken into account in the current analysis and are discussed in the present work. [ABSTRACT FROM AUTHOR]

Real-time monitoring of uranium concentration in molten salt reactors (MSR) is crucial for reactor operation and safety. However, conventional analytical methods cannot provide accurate results when uranium concentrations exceed approximately 4 wt%. We report an electrochemical technique based on repeating chronoamperometry (RCA) for precise and accurate measurements of uranium concentrations > 10 wt% in a NaCl–MgCl2–UCl3 molten salt. Uranium concentration was measured in a molten salt containing 5–12.5 wt% of uranium chloride with high precision and a linearity close to 1. The RCA technique can be used for measuring the concentration of nuclear fuel in MSRs. [ABSTRACT FROM AUTHOR]

The power peaking factor (PPF) is an important parameter that needs to be considered to maintain the reactor's stability, safety and efficiency. The present study discusses the power density distribution in the reactor core by calculating the PPF to increase the reactor safety margin and minimize the possibility of a high power density at a certain point causing core failure in the Molten Salt Reactor (MSR), which is one of the Generation-IV reactors with fuel and coolant dissolved in molten salt. Important parameters discussed in the present study are the distribution of the reactor core region and the fuel composition in each region during the reactor operation time of 2,000 days. Analysis was performed for 3 different thermal power: 150, 450 and 1,000 MWth as a representative to determine the behavior of PPF at small to large power by taking the reference design of the FUJI U3 reactor developed by ITMSF (International Thorium Molten-Salt Forum) Japan. FliBe (Lithium and Beryllium Fluoride) is used as the coolant, and the fuel mixture is 233UF4-ThF4. The calculation was conducted using SRAC2006 with PIJ and CITATION modules that solve the neutron diffusion equation providing power distribution values and effective multiplication factors. As a result, the MSR with 4 core regions produces a more balanced power distribution, a better PPF value, and a good effective multiplication factor for 2,000 days of operation time compared to the reference design. [ABSTRACT FROM AUTHOR]

Experimental measurements were made to assess the electrical conductivity as a function of temperature and NdF3 concentration (0–20 mol %) of molten systems of (LiF–CaF2)eut–NdF3, (LiF–NaF)eut–NdF3, (NaF–CaF2)eut–NdF3 and (LiF–MgF2)eut–NdF3. The experiment used an altering current impedance spectroscopy technique with a platinum–rhodium electrode positioned in a pyrolytic BN tube and graphite a crucible/counter electrode. The conductivity of all systems under study increased with rising temperatures and decreasing NdF3 concentrations. The Arrhenius equation and linear regression have both been used to describe the experimental data. The results of the ionic conductivity for the temperature 850 °C and NdF3 concentrations 0, 10 and 15 mol %, respectively, can be compared as follows: the conductivity of the molten system of (LiF–CaF2)eut–NdF3 was determined to be 6.10, 5.95 and 5.10 S.cm−1, the results for the system (LiF–NaF)eut–NdF3 were 6.16, 5.56 and 4.13 S.cm−1, the results for the system (NaF–CaF2)eut–NdF3 were 3.78, 3.56 and 2.32 S.cm−1, and finally, the results for the system (LiF–MgF2)eut–NdF3 were determined to be for the same temperature as 5.35, 4.79 and 4.14 S.cm−1, respectively. [ABSTRACT FROM AUTHOR]

Abstract—The features and prospects of using nickel superalloys as structural materials for advanced reactor plants, in particular, molten salt reactor (MSR) are considered. The main limiting factors impeding the use of nickel superalloys under MSR operating conditions are considered. Information on new high-temperature EK198 and EK199 alloys, which have high radiation resistance and improved mechanical characteristics, is provided. [ABSTRACT FROM AUTHOR]

In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC’s input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that keff agreed within 10 pcm while the flux distribution in space and energy generally agreed within 0.1%. Larger differences were observed between the OpenMC CAD and CSG models; notably, the keff computed for the CAD model was 1.00872, which is more than 1% lower than the value for the CSG model and much closer to experiment. Several areas of the reactor that were modeled differently in the CSG and CAD models were discussed and, in several cases, their impact on keff was quantified. Lastly, we compared the computational performance and memory usage between the CAD and CSG models. Simulation of the CSG model was found to be 1.4–2.3× faster than simulation of the CAD model based on the Embree ray tracer while using 4× less memory, highlighting the need for continued improvements in the CAD-based particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer.

Molten salts such as 2LiF-BeF2 (FLiBe) have been proposed as coolants for advanced nuclear fission and fusion reactors. Critical to the design, licensing and operation of these reactors is characterization and understanding of the chemical behavior and mass transport of activation and fission products, corrosion products, and other solutes in the coolant. Electrochemical techniques are a powerful suite of tools for probing these phenomena. The design of an experimental cell for molten salt electrochemistry is described herein. As a demonstration of this design, details of the experimental methods used to conduct electrochemical experiments with molten FLiBe with addition of LiH are provided. Decommissioning of the cell is considered from the point of view of decontamination and waste generated. Main features of the cell include:•Suitable for operation up to 800 °C; suitable for operation inside and outside of a glovebox.•Enables sweep gas, gas sampling and analysis; enables addition of solid and liquid materials during operation.•Supports a variety of electrode materials and arrangements.

Renewed interest in advanced nuclear reactors, such as Molten Salt Reactors (MSRs), has spurred studies in actinide halide chemistry and property measurements. Several proposed research‐scale and commercial MSR designs incorporate uranium trichloride (UCl3) fuel. There are relatively few preparations for isolated actinide halides reported in the literature, therefore novel methods for the isolation of pure material are desired. This communication describes the thermal elimination of pyridine (py) from the coordination complex UCl3py2 to yield gram‐scale quantities of UCl3. The purity of the UCl3 product was determined through powder X‐ray Diffractometry (pXRD) and Elemental Analysis (EA). [ABSTRACT FROM AUTHOR]

BackgroundFluoride molten salt servers as a coolant and fuel carrier in molten salt reactors (MSRs). Over time of operation, fuel composition of MSR includes uranium, thorium, fission products, and carrier salt. Separating the fission products and recycling the valuable components enhances the reactor's operational efficiency and minimizes radioactive waste. Leveraging the phase equilibrium properties of multi-component mixtures and elemental migration during solidification, can effectively separate materials.PurposeThis study aims at the directional solidification technology of fuel composition of MSR with emphasis on the separation of CsF from FLiNaK molten salt.MethodsFirstly, a certain mass of CsF and FLiNaK eutectic salt were taken to prepare simulated sample of typical fuel composition by mechanical mixing, and completely melt the mixed salt by heating and waiting for the molten salt to cool naturally before using it for directional solidification separation experiments. Then, the concentration distribution of a typical fission product CsF in the FLiNaK molten salt system, subjected to directional solidification under different technological conditions, was investigated using a self-developed cold-rod directional solidification experimental device.ResultsThe experimental results indicate that, by controlling the cooling solidification rate, the Cs element concentration in the solidified salt exhibits a radial gradient distribution, decreasing from the inner to the outer layers. In comparison to the liquid phase, the Cs element concentration in the outer solidified salt shows a substantial reduction, with a maximum decrease of approximately 90%.ConclusionsThe results show that it is feasible to separate fission products via directional solidification in fuel salts. With its straightforward process and absence of by-product generation, this technology shows promise for separating fission products in fuel salts.

BackgroundThe extraction of 99Mo medical isotopes from off-gas of molten salt reactors (MSRs) has attracted significant attention. However, previous studies focused only on the estimation of 99Mo nuclide production in the reactor core, and limited studies have focused on the expansion of the entire reactor system.PurposeThis study aims to estimate 99Mo production in the off-gas of a molten salt reactor.MethodsIn this study, a 99Mo migration model with noble metal deposition and a flow effect in an MSR was established by using scientific computing software “Mathematica”, and a corresponding verification based on the deposition result of a molten salt reactor experiment (MSRE) was implemented. Subsequently, an analysis was conducted on the impact of changes in the yield, specific activity, and operational status of 99Mo crude product in MSRE off-gas on the stability of production.ResultsThe analysis results of Mo transport in the MSRE show that the production of 99Mo in the off-gas can reach 69.30 TBq with a specific activity of 0.39 PBq∙g-1 under normal operating conditions. In addition, an increase in the removal rate is conducive to the effect on the specific activity of 99Mo whilst the working conditions of bubble surface ratio bubble ratio and oxidation-reduction potential changes have little effect on the specific activity of the product.ConclusionsThe 99Mo in off-gas has high yield and specific activity, and the effects of the conventional operating conditions on the specific activity of 99Mo are beneficial, which indicates that this method can be a potential alternative for 99Mo production.

BackgroundMolten salt reactor is one of the six internationally recognized and recommended fourth generation reactors that can use liquid nuclear fuel. The production, transportation, and storage of nuclear fuel involve different processes from conventional solid-state nuclear fuel reactors.PurposeThis study aims to perform criticality safety analysis for nuclear fuel storage of molten salt reactor in compliant with requirements of nuclear fuel management and nuclear safety supervision.MethodsThe design parameters of MSRE (Molten Salt Reactor Experiment) were referred for criticality safety analysis, and the impact of different factors on nuclear fuel salt storage was analyzed. Calculation was completed by selecting storage modeling, critical parameter analysis, and Monte Carlo neutron transport software simulation for liquid fuel molten salt reactor nuclear fuel. The keff values of dry environment storage and water flooded environment storage under the design model were summarized, so did the changes in the total mass of fuel salt.Results & ConclusionsThe quality of subcritical safety control under different conditions are obtained and compared with corresponding raw material salts, intermediate products, and consideration of container walls. This study combines legal regulations and the process of nuclear material circulation for analysis and discussion, summarizes the critical safety characteristics of nuclear fuel salt, and for the first time proposes relevant supervision and evaluation points from the perspective of nuclear safety supervision.

BackgroundThe molten salt reactor (MSR) is one of the six advanced reactors identified by the Generation IV International Forum. The MSR exhibits unique characteristics, such as intrinsic safety, sustainable development, nuclear nonproliferation, natural resource protection, and economic efficiency. After a liquid-fuel molten salt reactor shuts down, residual decay heat in the reactor core is passively dissipated to the environment through a natural circulation loop of the molten salt. However, the decay heat from the molten salt in the main circuit can impact the thermal capacity of the overall system.PurposeThis study aims to determine the thermal characteristics of the passive system by establishing an analysis basis for a natural cycle model to examine the effects of natural circulation loop on the physical properties of the molten salt, the loop structure, and equipment resistance coefficient K.MethodsBased on the direct reactor auxiliary cooling system (DRACS) for MSR, the natural circulation model of the passive residual heat-removal loop of a liquid-fuel molten salt reactor was established using self-developed Python analysis program, and the temperature distribution of the molten salt in the loop was explored using the numerical model. Then, the effects of different physical properties, loop structures, and core and heat exchanger resistance coefficients K on the heat transfer and flow characteristics of the residual heat-removal system were analyzed. Finally, the effects of critical factors on the residual heat-removal capacity of the reactor core were analyzed using self-developed Python program for natural circulation calculation equations code (NCCC) and validated by using the natural cycle experimental results of the DRACS circuit in CIET1.0.ResultsThe findings indicate that the presence of decay heat from molten salt in a system loop decreases the natural circulation-driven heat transfer by the molten salt within the core. Based on the significance analysis results show that fuel salt density, specific heat capacity, and height difference between the hot and cold cores are parameters that significantly influence the natural circulation capacity. When these three parameter values are increased by 15% separately, the residual heat removal capacities increase by 26.02%, 15.00%, and 18.59%, respectively.ConclusionsResults of this study demonstrate that the molten salt properties, circuit structure, and equipment resistance coefficient all affect the natural circulation heat-removal capacity.

BackgroundMolten salt reactors (MSRs) feature high temperature, low pressure, high chemical stability, and nuclear non-proliferation, which make them promising for a wide range of applications. However, owing to the fluid nature of the fuel, changes in the core structure of MSRs are bound to affect the fuel distribution and loading, consequently affecting the physical parameters of the core. Currently, the research on the impact of structural changes of the MSR core, composed of dispersed graphite components, on reactivity is insufficient.PurposeThis study aims to explores the influence of core structure changes on reactivity.MethodsThe core design model of new solid hexagonal graphite module MSR was taken as a reference, the small geometric changes, such as deformation and displacement of the core components, caused by factors including thermal expansion, fluid erosion, core vibration, and graphite irradiation, were analyzed using MCNP code. Additionally, the impact of these structural changes in the core on reactivity was investigated in details.ResultsThe results indicate that the structural changes in the core caused by thermal expansion introduce negative reactivity. The reactivity introduced by fluid erosion and core vibration, causing graphite component displacement, fluctuates. However, the overall trend shows that a shift of the graphite components towards the center of the core introduces positive reactivity, whereas shift towards the outer periphery of the core introduces negative reactivity. Graphite irradiation-induced deformation initially decreases reactivity and then increases it. At the end of the core's lifespan, the reactivity is still lower than at the beginning of the lifespan. This reactivity change can be compensated for by online feed or control rod movement, which has a limited impact on the operation of the MSR. However, the critical issue of control should be considered when this batch of fuel is reinserted into a new core.ConclusionsResults of this study suggest the necessity of constraining the graphite components within a certain range to ensure the safe operation of the reactor, providing an important reference for the design, operation, and maintenance of MSRs.

General interest in the deployment of molten salt reactors (MSRs) is growing, while the available data on uranium-containing fuel salt candidates remains scarce. Thermophysical data are one of the key parameters for reactor design and understanding reactor operability. Hence, filling in the gap of the missing data is crucial to allow for the advancement of MSRs. This study provides novel data for two eutectic compositions within the NaF-KF-UF4 ternary system which serve as potential fuel candidates for MSRs. Experimental measurements include their melting point, density, fusion enthalpy, and vapor pressure. Additionally, their boiling point was extrapolated from the vapor pressure data, which were, at the same time, used to determine the enthalpy of vaporization. The obtained thermodynamic values were compared with available data from the literature but also with results from thermochemical equilibrium calculations using the JRCMSD database, finding a good correlation, which thus contributed to database validation. Preliminary thoughts on fluoride salt reactor operability based on the obtained results are discussed in this study.

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D2-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D2-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D2-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D2-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

Abilene Christian University (ACU) is developing a 1-MW(thermal) molten salt research reactor that will be built on the ACU campus. A conceptual reactor core model was developed to facilitate the safety analysis required for a construction permit. A series of scoping studies were performed seeking to define the reactor core design parameters subject to a variety of design requirements. A Pareto curve identifying the tradeoff between uranium and LiF-BeF2 was determined. Within this curve, at least 250 kg of uranium and 700 kg of LiF-BeF2 are needed, albeit for different reactor configurations and fuel salt compositions. The cylindrical reactor vessel associated with the best-performing fuel salt composition is ~130 cm in diameter, ~170 cm tall, and contains ~2.5 tons of graphite. The conversion ratio of the reactor is low and will require regular refueling. The shift in neutron spectrum observed with the changing fuel salt composition does not significantly impact reactivity loss with respect to burnup. [ABSTRACT FROM AUTHOR]

A four-dimensional (4D, in concentration–temperature coordinates) computer model of the isobaric phase diagram of uranium, sodium, magnesium, and plutonium chlorides, as well as four three-dimensional (3D) computer models of the phase diagrams of the ternary systems forming it, has been constructed. The technology of assembling a 4D model of 46 hypersurfaces and 17 phase regions was used in the design. The obtained 4D model of the UCl3–NaCl–MgCl2–PuCl3 phase diagram makes it possible to visualize a four-dimensional object as a whole (with all its hypersurfaces and phase regions) by any arbitrarily given 2D and 3D sections, as well as it is able to reproduce published (experimental or thermodynamically calculated) 2D sections. The scope of application of the results of the work is the development of materials for fuel components of fourth-generation molten salt reactors and pyrochemical recycling of spent fuel rods. For the first time, a comprehensive, complete description of phase diagrams composed of uranium, plutonium, sodium, and magnesium chlorides has been obtained. [ABSTRACT FROM AUTHOR]

Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. We present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term. [ABSTRACT FROM AUTHOR]

The detection of off-gassed sodium from molten sodium nitrate (NaNO3) at temperatures between 330 °C and 505 °C and off-gassed calcium from molten lithium chloride–potassium chloride eutectic (LKE) mixtures at 510 °C with laser-induced breakdown spectroscopy (LIBS) was demonstrated. NaNO3 and LKE samples were melted in a custom-built crucible that promoted the generation of off-gassed products from the molten sample. The off-gassed products were analyzed with a LIBS system designed to probe the high-temperature environment. Na D emission lines, Na(I)588.99 nm and Na(I) 589.59 nm, were detected from the NaNO3 samples after reaching a temperature threshold, which indicated the occurrence of phase change. In LKE mixtures, the detection of Ca impurities at a concentration of 78 mg/kg was possible using the emission lines Ca(II) 393.66 nm and Ca(II) 395.85 nm. This work demonstrates the real-time monitoring capabilities of LIBS in high-temperature environments that simulate the conditions of molten salt reactors. Graphical Abstract This is a visual representation of the abstract. [ABSTRACT FROM AUTHOR]

This paper presents a molten salt reactor (MSR) design parameter sensitivity study using a nodal dynamic modelling methodology with explicitly modified point kinetics equation and Mann’s model for heat transfer. Six parameters that can impact MSR safety are evaluated. A MATLAB-Simulink model inspired by Thorcon’s 550MWth MSR is used for parameter evaluations. A safety envelope was formed to encapsulate power, maximum and minimum temperature, and temperature-induced reactivity feedback. The parameters are perturbed by ±30%. The parameters were then ranked by their subsequent impact on the considered safety envelope, which ranks acceptable parameter uncertainty. The model is openly available on GitHub.

The molten salt reactor, utilizing molten fluoride salt as fuel solvent and coolant, is one of the generation-Ⅳ nuclear reactors for advanced nuclear energy. The generated fluoride nuclear waste during the operation of molten salt reactor and reprocessing of spent fuel should be immobilized in a stable matrix before disposal. At present, vitrification is still the only technology in the world that could industrially immobilize highlevel waste. However, such fluoride waste cannot be directly vitrified into borosilicate glass since it contains a large amount of F, which could lead to oversaturation in the glass. In order to provide an effective solution for the safe treatment of molten fluoride salt waste, H2C2O4 was used as a defluorination agent to defluorinate the simulated salt waste mainly consisting of alkali fluorides. The mixed sample of H2C2O4 and simulated fluoride salt waste was characterized with TG-DSC-MS and XRD to analyze the endothermic and exothermic behaviors, released gases, and phase transitions at elevated temperatures. Afterward, the effects of thermal treatment temperature and the molar ratio of H2C2O4 to F on the fluorine removal efficiency were investigated, and the optimal process parameters of defluorination were finally determined. The results show that H2C2O4 could react with alkali fluoride salt to release HF gas besides being decomposed into H2O, CO and CO2 at temperatures between 100 ℃ and 300 ℃, while the alkali fluorides turned to alkali oxalates, which could decompose to alkali carbonates at temperatures of 500 ℃. The fluorine removal efficiency would reach up to 93% when the molar ratio of H2C2O4 to F was 2 and the thermal treatment temperature was 300 ℃. The defluorinated waste thus obtained was then immobilized in a borosilicate glass waste form at about 1 200 °C with a waste loading of 25%, and the normalized elemental (B, Li, Na, K, Cs, Sr, and Ce) releases of the waste glass conducted with 7-day product consistency test were lower than 2.0 g/m2, showcasing acceptable durability for nuclear waste glass. The above results indicate that the proposed approach which defluorination with H2C2O4 in the first step at temperatures of below 300 °C and vitrifying the remaining waste into a borosilicate glass in the second step, would provide a practical way to safely treat the molten fluoride salt nuclear waste.

Non-impregnated small-pore graphite (NSPG), which has a compact microstructure and is used in molten salt reactors (MSRs), was prepared by a novel process. The pore diameter of NSPG was reduced to ~800 nm. The irradiation evaluation of NSPG was carried out by 7 MeV Xe26+ ion irradiation. The microstructural changes of NSPG were investigated with IG-110 as a comparison. The graphitization degree of NSPG was higher than that of IG-110, though it was not subjected to an impregnation process. Under low-dose ion irradiation (

The French National Center for Scientific Research (CNRS) is carrying out design studies on a nuclear electric propulsion (NEP) engine based on a molten salt reactor (MSR). A NEP engine based on liquid nuclear fuel could allow developing a core design with relatively high power densities and temperatures while using simple reactivity control systems and keeping low pressure and temperature gradients in the fuel. Nevertheless, the design work of such an engine poses significant technical challenges and requires the use of advanced numerical simulation tools. Different MSRs for space are currently being studied. In this work, a MSR concept using a fast neutron spectrum is investigated using a multiphysics tool based on a numerical coupling between the OpenFOAM (computational fluid dynamics) and SERPENT 2 (Monte Carlo neutronics) codes. The analysis of this paper is focused on the reactor core coupled neutronic and thermal-hydraulic phenomena. Steady state full-power conditions are calculated for two different fast MSR designs using low-enriched uranium (LEU) and highly enriched uranium. The results show that the proposed core layout and materials allow obtaining a satisfactory temperature distribution in the core (maximal values and gradients) without significant penalization of the reactor operating conditions. A reactivity control strategy excluding the use of control rods is studied for the LEU concept. Transient and safety studies are also performed and show acceptable performance. [ABSTRACT FROM AUTHOR]

Technitium-99m (99mTc), a widely used radioisotope, is used in tens of millions of medical diagnostic procedures annually. However, it is hard to store and must be immediately used upon production due to its short half-life (i.e., 6 h); thus, it is currently produced from 99Mo, which itself is a result of 235U fission. The majority of 99Mo supplies to U.S. patients are currently provided by foreign producers and produced using highly enriched uranium (HEU). In order to minimize the proliferation risks of HEU-based medical isotope production, the U.S. Department of Energy's National Nuclear Security Administration has funded a program to accelerate the development of technologies to produce 99Mo without the use of HEU. Today, the global supply of 99Mo depends on a limited number of nuclear reactors, and production has been interrupted unexpectedly since 2009 due to the fleet's advanced age. Alternative options for 99Mo production are discussed in this paper, and one potential option is to obtain 99mTc from molten salt reactors (MSRs). A MSR is a nuclear fission reactor that can operate at or close to atmospheric pressure with liquid fuel, which allows for producing isotopes in a timely manner. In this paper, the past and current production of 99Mo via nuclear reactors is described, and the future of 99Mo production by MSRs is discussed. The behavior and chemical properties of molybdenum in fluoride salts in MSRs and the possible extraction methods are also examined in addition to the limitation of current studies. [ABSTRACT FROM AUTHOR]

Four Ni-xW-6Cr alloys with W content of 10, 20, 30 and 35 wt% were prepared, and the effects of W content on the irradiation resistance of the alloys were systematically investigated. All the Ni-xW-6Cr alloys were irradiated with 7 MeV Xe26+ irons at room temperature with dose of 0.05, 0.1, 0.5, 1 and 2 dpa. The average size of defects decreases and the density of defects increases with the increase of W content in the Ni-xW-6Cr alloys, indicating that the increase of W content enhances the hindrance to the migration or accumulation of black dot defects and dislocation loops. The difference of average size and density of defects in the Ni-xW-6Cr alloys derives from the pinning effect on dislocation loops caused by W atoms. [ABSTRACT FROM AUTHOR]

Molten salt mixtures containing LiCl–KCl and NaCl–MgCl2 have been infused with UCl3 via reaction of U metal and FeCl2. The process starts with base salt (LiCl–KCl or NaCl–MgCl2) drying/purification using hydrochlorination via bubbling anhydrous HCl. An auto-titrator running in pH-stat mode was used to determine the point at which there is no net reaction with the salt. U metal is contained in a porous stainless steel basket as it is submerged in the molten salt. The byproduct Fe metal forms dendrites on the basket walls, allowing for simple separation from the molten salt. Based on analysis of salt samples using inductively coupled plasma mass spectroscopy, UCl3 yield of 90% was attained in both NaCl-MgCl2 at 550 °C and in LiCl–KCl at 450 °C. Open circuit potential difference between a W rod working electrode and a Ag/AgCl reference electrode indicated a UCl4 to UCl3 activity ratio of 1 × 10−5. The iron dendrites were comprised of linked cubic structures with length scale of about 50 µm. [ABSTRACT FROM AUTHOR]