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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]

Compatibility tests are carried out on Ni-based Hastelloy alloys and stainless steel under the influence of the molten chloride salt NaCl-CaCl2-CeCl3 to investigate the corrosion behavior of molten salts with structural materials. Weight change measurements before and after the reaction tests show that the Hastelloy alloys have better corrosion resistance than stainless steel. The effects of molten salt purification on the corrosion behavior in the Hastelloy alloys are explored. The composition of the reacting phases is identified from transmission electron microscopy analysis and the corrosion mode is homogeneous corrosion with Cr and Fe diffusion, followed by Mo and Ni elution. [ABSTRACT FROM AUTHOR]

Compatibility tests are carried out on Ni-based Hastelloy alloys and stainless steel under the influence of the molten chloride salt NaCl-CaCl2-CeCl3 to investigate the corrosion behavior of molten salts with structural materials. Weight change measurements before and after the reaction tests show that the Hastelloy alloys have better corrosion resistance than stainless steel. The effects of molten salt purification on the corrosion behavior in the Hastelloy alloys are explored. The composition of the reacting phases is identified from transmission electron microscopy analysis and the corrosion mode is homogeneous corrosion with Cr and Fe diffusion, followed by Mo and Ni elution. [ABSTRACT FROM AUTHOR]

Nuclear power's role as a reliable, baseload, low-carbon source and its importance in achieving clean energy goals are being increasingly recognized with growing urgency around decarbonization of the global energy systems. However, to deliver a long-term sustainable solution, it is essential to develop innovative nuclear technologies for improving the fuel utilization and reducing the nuclear waste disposal challenge. Zero Power Reactors (ZPR) are an essential initial step for developing new nuclear technologies because they allow for testing and refinement in a safe environment before large-scale deployment. This paper discusses the design of a ZPR experiments for the development of iMAGINE, a novel chloride-based molten salt reactor technology. The paper presents a detailed analysis of the neutronic design for the shutdown and control systems of an experimental ZPR based on the iMAGINE molten salt reactor technology. The study concludes that a split-core design with a lower corner reflector as an extension of the lower annular reflector offers the most robust ZPR configuration, offering optimum operational margins and maneuverability. This design ensures safety, regulatory compliance, and sufficient control and shutdown performance for the successful development of the iMAGINE technology. [ABSTRACT FROM AUTHOR]

Structural materials in fast reactors with harsh radiation environments due to high energy neutrons—compared to thermal reactors—potentially suffer from a higher degree of radiation damage. This radiation damage can change the thermophysical and mechanical properties of materials and, as a result, alter their performance and effective lifetime, in some cases leading to their disintegration. These phenomena can jeopardize the safety of fast reactors and thus need to be investigated. In this study, the effect of radiation damage on the vessels of molten salt fast reactors (MSFR) was evaluated based on two fundamental radiation damage parameters: displacement per atom (dpa) and primary knock-on atom (pka). Following the previous part of this article (Parts 1 and 2), an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) with stainless steel and nickel-based alloy material vessels, respectively, were considered as case studies. The SPECTER and SPECTRA-PKA codes and a PTRAC card of MCNPX, integrated with a module which has been developed in MATLAB, named PTRIM and SRIM-2013 (using binary collision approximation), were employed individually to calculate and compare dpa and PKA (this master module containing all three tools has been appended to the iMAGINE-3BIC package for future use during reactor operations). Additionally, SRIM-2013 was applied in a 3D simulation of a radiation damage map on a small sample of vessels based on the calculated PKA. Our results showed a higher degree of radiation damage in the iMAGINE vessel compared to the EVOL one, which could be expected due to the harder neutron flux spectrum of the iMAGINE core compared to EVOL. In addition, the nickel alloy vessel showed better radiation damage resistance against high energy neutrons compared to the stainless steel one, although more investigations are required on thermal neutrons and alloy corrosion mechanisms to determine the best material for use in MSFR vessels. [ABSTRACT FROM AUTHOR]

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]

Currently, the properties of molten lithium, sodium and potassium fluoride eutectic mixtures with different additions are immensely important for the development of molten salt nuclear reactors. In the present work, the density of molten FLiNaK mixtures with additions of neodymium fluoride was studied by the Archimedean method. The neodymium fluoride addition increased the density of the 46.5 mol% LiF–11.5 mol % NaF–42.0 mol % KF (FLiNaK) and FLiNaK + 25 mol% NdF3 mixture from 2.00 g⋅cm−3 to 3.25 g⋅cm−3, respectively. Thermal diffusivity was measured by the laser flash method. It was found to decrease abruptly as the NdF3 concentration increased. Thermal conductivity of the FLiNaK–NdF3 system, which was calculated using thermal diffusivity, density and heat capacity values, was lower than that of molten FLiNaK at the same temperature. The composition with 25 mol % NdF3 (0.69 W⋅m−1⋅K−1) had a lower value of thermal conductivity than molten FLiNaK without additions (0.74 W⋅m−1⋅K−1) at the same temperature of at 973 K. It can be concluded that neodymium fluoride additions resulted in the density growth and decrease in the thermal diffusivity, heat capacity and thermal conductivity of molten FLiNaK. The change in the neodymium fluoride concentration can affect the technological process in nuclear reactor. [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]

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

To address the significant commercial interest in fusion energy, it will be necessary to accelerate the compatibility research associated with liquid breeders including Li, eutectic Pb–Li and LiF-BeF2 (FLiBe) molten salt. Particularly for FLiBe, compatibility understanding is limited especially for fusion relevant materials such as reduced activation ferritic-martensitic steels, SiC and V alloys. The historical knowledge associated with molten salt reactors (MSRs) and recent work to commercialize MSRs can benefit fusion research. Recent experimental and modeling work has improved understanding and this knowledge can be applied to fusion relevant materials. For liquid metals (LMs), the comparisons to Li and Pb–Li are less direct but nevertheless can help guide the pathway toward commercialization. For Pb–Li, Al-rich coatings have been shown to inhibit dissolution and potentially increase operating temperatures. For commercialization, the experience with sensors and on-line cleanup can help guide future developments. Thus, it is worth considering the potential for fission-related research with LMs and molten salts to help accelerate fusion research. [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]

Ionic liquid materials are viable candidates as a heat transfer fluid (HTF) in a wide range of applications, notably within concentrated solar power (CSP) technology and molten salt reactors (MSRs). For next-generation CSP and MSR technologies that strive for higher power generation efficiency, a HTF with wide liquid phase range and energy storage capabilities is crucial. Studies have shown that eutectic chloride salts exhibit thermal stability at high temperatures, high heat storage capacity, and are less expensive than nitrate and carbonate salts. However, the experimental data needed to fully evaluate the potential of eutectic chloride salts as a HTF contender are scarce and entail large uncertainties. Considering the high cost and potential hazards associated with the experimental methods used to determine the properties of ionic liquids, molecular modeling can be used as a viable alternative resource. In this study, the eutectic ternary chloride salt MgCl2-NaCl-KCl is modeled using ab-initio molecular dynamics simulations (AIMDs) in the liquid phase. Using the simulated data, the thermophysical and transport properties of eutectic chloride salt can be calculated: density, viscosity, heat capacity, diffusion coefficient, and ionic conductivity. For an initial model validation, experimental pair-distribution function data were obtained from X-ray total scattering techniques and compared to the theoretical pair-distribution function. Additionally, theoretical viscosity values are compared to experimental viscosity values for a similar system. The results provide a starting foundation for a MgCl2-NaCl-KCl model that can be extended to predict other fundamental properties. [ABSTRACT FROM AUTHOR]

Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors. [ABSTRACT FROM AUTHOR]

This paper reports on a new approach to chlorinating molten salts using vaporized NH4Cl. In separate experiments, chlorination of dissolved Li2O and submerged U metal was demonstrated. Ammonium chloride was thermally decomposed in a quartz glass vaporization chamber with an internal temperature ranging from 295 to 384 °C. These decomposition products were then bubbled into eutectic LiCl–KCl salt via argon carrier gas. The vapor appears to absorb as NH4+/Cl− into the molten LiCl–KCl based on the high dissolved concentration of acid compared literature reported solubility of HCl. In Li2O chlorination experiments, complete reaction was achieved starting with 1 wt% Li2O for reaction times of 24.8 h or less. In U metal chlorination experiments, dissolved U concentrations ranged from 1.75 to 4.17 wt% for reaction times ranging from 3.5 to 21.6 h. The rate limiting step is mass transfer and reaction between the molten salt and U metal pellet. Cyclic voltammetry peak heights were correlated with dissolved U concentration with overall scan results consistent with UCl3 being the dominant form of dissolved U. [ABSTRACT FROM AUTHOR]

Molten salts have been used as heat transfer fluids since the middle of the 20th century. More recently, molten chloride salts have been studied for use in concentrated solar power plants or molten salt reactors. However, none of the materials studied to date has been able to withstand this highly corrosive environment without controlling the salt's redox potential. The alumina-forming alloy was a promising option, as it has not yet been widely studied. To investigate this possibility, two iron-based alumina-forming alloys were corroded in NaCl-MgCl2 eutectic at 600 °C for 500 h after being pre-oxidised to grow a protective layer of α-alumina on each alloy. A salt purification protocol based on salt electrolysis was implemented to ensure comparable and reproducible results. During immersion, alumina was transformed into MgAl2O4, as shown by FIB-SEM observation. Inter and intragranular corrosion were observed, with the formation of MgAl2O4 in the corroded zones. The nature of the oxides was explained by the predominance diagram. Intragranular corrosion was 2 µm deep, and intergranular corrosion 10 µm deep. Alumina formed at the bottom of the intergranular corrosion zones. The depth of intergranular corrosion is consistent with O diffusion control at the grain boundary. [ABSTRACT FROM AUTHOR]

Next‐generation nuclear reactor technologies such as the molten salt reactor utilize alkali metal fluoride salts as both fuel and coolant. In the present study, the suitability of iron phosphate glass (IPG) as a vitrification matrix for alkali metal fluoride (NaF, CaF2) and simulated fission product loaded fluoride (NdF3, CeF3, SmF3) waste has been explored. The structural change in the metal fluoride–loaded IPG has been analyzed thoroughly using Raman and fourier transform infrared (FTIR) spectroscopy. Thermal analysis showed that the stability and glass forming ability of IPG improved upon loading the same with various mixed metal fluorides. Mössbauer data and X‐ray absorption spectroscopy at Fe K‐edge explored the minute changes in the local structure. The effect of radiation emanating from radioactive wastes in the fluoride‐loaded IPG has been scrutinized via 4.5 MeV proton beam irradiation. Our study firmly establishes the applicability of IPG as suitable vitrification matrix for radioactive metal fluoride–loaded nuclear wastes. [ABSTRACT FROM AUTHOR]

This paper investigated four different S-CO2 Brayton cycle layouts for nuclear energy conversion: simple recuperation cycle (SR), recompression cycle (RC), re-heating cycle (RH), and intercooling cycle (IC). We compared these S-CO2 Brayton cycle schemes for nuclear energy conversion using the G1+TOPSIS multi-index evaluation method. The evaluation results of different schemes based on their safety, thermodynamics, techno-economic and compactness are given. The results show that for Generation IV reactors with the same designed thermal power, the thermodynamic performance of the S-CO2 system is better for higher reactor exit temperature. Among the schemes, the gas-cooled fast reactor (GFR)+RC scheme has the highest thermal efficiency (47.4%) and exergy efficiency (56.48%). The GFR+IC scheme has the lowest specific cost (1861.3$/W) and the internal rate of return (24.8%). The re-heating cycle (RH) has worse indexes, but it requires the lowest initial investment cost. The intercooling cycle (IC) has the lowest levelized cost of electricity (0.0134 $/KW•h) coupling to GFR. Considering all indexes of four aspects, the reactor's performance ranking is MSR>LFR>SFR>GFR, and the S-CO2 system's performance ranking is RC>SR>IC>RH. For Generation IV nuclear energy conversion technologies, the molten salt reactor (MSR)+RC scheme should be given priority, while GFR+RH schemes should be carefully considered. [ABSTRACT FROM AUTHOR]

AbstractSoftware verification and validation constitute crucial phases in the development of simulation computer codes, particularly in the context of nuclear reactor safety analysis codes, where stringent safety requirements govern the development and deployment of nuclear technologies. This paper focuses on numerical verification study of the System Analysis Module (SAM) computer code, currently under development at Argonne National Laboratory. Specifically, we employed the Method of Manufactured Solutions (MMS) and proposed a verification technique tailored to the multiphysics simulation of molten salt reactors (MSRs). This research accomplished three main objectives. First, we have addressed key challenges associated with applying the MMS to MSR systems, arising from (1) the complex multiphysics coupling inherent in this problem and (2) the necessity to model the entire coolant loop for describing the drift of delayed neutron precursors outside the reactor core. The paper provides recommendations and guidelines to overcome these challenges, enabling the successful application of the MMS for simulating MSRs. Second, we have presented a comprehensive set of verification examples, serving as an exhaustive benchmark for code verification within the nuclear community. Third, we have established a robust verification of the SAM code’s capability to model MSR systems. [ABSTRACT FROM AUTHOR]

The UK government and many international experts have pointed out that nuclear energy has an important role to play in the transition towards a decarbonised energy system since it is the only freely manageable very low-carbon energy technology with 24/7 availability to complement renewables. Besides current investments in light water reactor technologies, we need innovation for improved fuel usage and reduced waste creation, like that offered by iMAGINE, for the required broad success of nuclear technologies. To allow for quick progress in innovative technologies like iMAGINE and their regulation, a timely investment into urgently needed experimental infrastructure and expertise development will be required to assure the availability of capacities and capabilities. The initial steps to start the development of such a new reactor physics experimental facility to investigate molten salt fast reactor technology are discussed, and a stepwise approach for the development of the experimental facility is described. The down selection for the choice for a diverse control and shutdown system is described through manipulating the reflector (control) and splitting the core (shutdown). The developed innovative core design of having the two core parts in two different rooms opens completely new opportunities and will allow for the manifestation of the request for separated operational and experimental crews, as nowadays requested by regulators into the built environment. The proposed physical separation of safety-relevant operational systems from the experimental room should on the one hand help to ease the access to the facility for visiting experimental specialists. On the other hand, the location of all safety-relevant systems in a now separated access-controlled area for the operational team will limit the risk of misuse through third party access. The planned experimental programme is described with the major steps as follows: core criticality experiments, followed by experiments to determine the neutron flux, neutron spectrum and power distribution as well as experiments to understand the effect of changes in reactivity and flux as a function of salt density, temperature and composition change. [ABSTRACT FROM AUTHOR]

Nuclear graphite plays a crucial role in thermal-spectrum thorium molten salt reactors (TMSRs) as both the neutron moderator and the construct for the coolant flowing channel. When subjected to irradiation and elevated temperatures, graphite components experience considerable deformation due to a combination of dimensional changes, thermal expansion, irradiation creep, elastic deformation, and changes in thermomechanical characteristics. The lifespan of the graphite component is a limiting factor in TMSR designs as it strongly correlates with the dimensional changes of the graphite. To evaluate the thermal and mechanical reactions of graphite component under TMSR core conditions, it is necessary to couple models of thermal-hydraulics, neutronics, and thermal-mechanics. This paper presents an enhanced methodology for analyzing the deformation of graphite components using the finite element method. Then, this method was applied to analyze a 10-year deformation history of a hexagonal prism assembly (HPA) and it was compared with the traditional hexagonal round channel assembly (RCA). The results demonstrate that the stress–strain field of both types of graphite components undergo significant variations with the increasing neutron fluence from irradiation. HPA graphite exhibits a slower deformation as compared to RCA graphite when subjected to identical operating conditions. In this case, HPA graphite has a lifespan of approximately 10 years, while RCA graphite lasts only 8.8 years. [ABSTRACT FROM AUTHOR]

As one of the six nuclear energy systems selected by the "Fourth Generation Nuclear Reactor International Forum", the Thorium Molten Salt Reactor (TMSR) has garnered significant attention due to its fascinating features, including excellent neutron economy, online fuel reprocessing, and reduced production of actinides. However, real-time online analysis of the nuclear elements remains a crucial technology throughout the fuel reprocessing. In this study, the femtosecond laser-induced breakdown spectroscopy (LIBS) was employed with both univariate and multivariate regression models for the quantitative analysis of Th and U elements in a graphite matrix. The concentrations of ThO 2 in six homemade samples were ranged from 4.998 to 40.012 wt % , while the concentrations of U 3 O 8 were ranged from 0.970 to 2.498 wt % . The mean relative error (MRE), root mean square error of prediction (RMSEP), root mean square error of calibration and coefficient of determination (R 2 ) were employed as indicators to evaluate the quantitative accuracy and stability of the LIBS analysis. For the univariate regress, the standard curve method was performed, utilizing a non-interferential emission line (Th II 286.99 nm) as analytical line for Th measurement. However, no suitable analysis line was found for U element. For the multivariate regression, the random forest (RF) method and partial least squares (PLS) regression method utilized three spectral regions (285.00–288.00, 330.00–370.00 and 390.00–425.00 nm) as analytical bands for ThO 2 , while two regions (380.00–389.00 and 514.00–517.00 nm) were employed for U analysis. The results suggest that the PLS regression exhibited the best performance, yielding RMSEP T h O 2 = 0.785 wt % , RMSEP U 3 O 8 = 0.135 wt % , MRE T h O 2 = 4.24 % , and MRE U 3 O 8 = 4.90 % . Therefore, the combination of LIBS with multivariate regression models exhibits significant potential as a robust in-situ analytical approach for the online fuel reprocessing in TMSR. [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]

A foreseen feature of the Molten Salt Fast Reactor is the adoption of a bubbling system for the removal of gaseous and metallic fission products (FPs). This mechanism injects helium bubbles into the core to remove FPs from the salt through floating and mass transfer mechanisms for metallic and gaseous FPs, respectively. The present work is aimed at analyzing this helium bubbling system, focusing on gaseous FPs. We investigate both operational and safety-related features in order to get information useful for the design and to assess the convenience of its adoption. Accordingly, our investigations split into two strands: (1) analyzing the characteristics of the bubbling system itself and (2) assessing the safety features of the reactor in its presence. In order to perform the above analyses, we add the capability to simulate production, transport, and mass transfer of an arbitrary number of gaseous FPs to a preexisting multiphysics solver, built with the OpenFOAM suite. In terms of operational characterization, our analyses quantify the removal efficiency through a characteristic removal time and estimate the poisoning effect of gaseous FPs. In addition, we evaluate the activity and decay heat of the removed gas, which is an aspect crucial for the design of the off-gas unit, and the effect of the bubbling system on the power versus the fuel mass flow rate curve, which is a possible control mechanism. Among our safety-related studies, we first evaluate the void coefficient, determining upper bounds on the helium flow rate in order to avoid prompt supercriticality in case of prompt loss of helium injection. The latter accidental scenario is also analyzed considering the thermal-hydraulic dynamics of the system. We also discuss another accident: complete loss of helium removal. [ABSTRACT FROM AUTHOR]

Advanced reactors are characterized by passive safety systems to be activated in case of an accident. Molten salt reactors are equipped with a passive emergency draining system. In case of an accident, the reactor shuts itself down, the freeze plug melts due to decay heat, and fuel salt is discharged into the draining tank. The behavior of the freeze plug during liquid-to-solid transition and reactor operation, as well as in the event of an accident was investigated taking two important safety criteria into consideration i.e. the plug shall not melt during normal operation (unwilling opening) and shall passively melt in case of an accident. To evaluate the effectiveness of the freeze plug, computational fluid dynamics method was used. The results showed that draining pipe blocking time is 900 s and 70% of the interior of the freeze plug solidifies in 2700 s. Furthermore, it was concluded that the freeze plug did not cause any unwilling opening. In addition, the freeze plug opening time was calculated as 400 s. Since this value is less than the limiting time of 1100 s which depends on material strength temperature, the freeze plug can be utilized as a passive safety system component. [ABSTRACT FROM AUTHOR]

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution. [ABSTRACT FROM AUTHOR]

The increasing installment of solar and wind renewable energy systems create a volatile energy demand to be met by electricity providers. A nuclear hybrid energy system is a nuclear reactor with energy storage that integrates into the grid with renewable energy sources. The Natrium design by TerraPower and GE Hitachi is a sodium fast reactor with molten salt energy storage. The Natrium design operates at steady state of 345 MWe and can boost up to 500 MWe for 5.5 hours. This study uses Dymola and the Modelica language to model the Natrium‐based nuclear‐renewable hybrid energy system. The dynamic system model is tested using hourly historical data from the state of Texas 2021 to show how renewables affect the electricity demand and how energy storage affects the Natrium system response to the demand. According to the results, while the available storage will allow the Natrium design to boost electricity production when the demand and electricity price is high making it more economically viable, the current molten salt storage is slightly undersized for the ERCOT market. [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]

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

This Special Topic "Radiation Effects in Materials" is a broad collection of scientific studies addressing the interaction of radiation with condensed matter, a long-standing issue that is relevant for a range of key energy technologies. With the discovery of nuclear fission and its use for energy production, the study and understanding of the effects of radiation on materials got a strong engineering dimension. As noted above, the principle of scintillators is based on the interaction between radiation and matter, and the increasing uses of radiation in our society stimulate the search for better performing materials. [Extracted from the article]

The purification of salt is indispensable for mitigating the corrosion of structural materials for molten salt reactors or other molten salt applications. This study develops a salt-purification system to synthesize and purify NaF-KF-UF4 salt (FUNaK) using Ar purging and hydrofluorination of impurities. Chronoamperometry is also used to remove metallic impurities in the hydrofluorinated FUNaK. This purified FUNaK is then used for a static corrosion test of stainless steel 316H (SS316H) to study the effectiveness of salt purification in mitigating its corrosion. For comparison, results from a previous study about the corrosion of SS316H by thermally purified FUNaK are used. FUNaK with UF3 is also synthesized for a static corrosion test with the same condition to investigate the impact of UF3 on corrosion. The results show that the corrosion of SS316H is significantly reduced by using the chemically purified FUNaK compared to thermally purified FUNaK. [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 CADbased particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer. [ABSTRACT FROM AUTHOR]

Thermochemical processes used to produce clean hydrogen (H 2) on a large scale by utilizing renewable energy and nuclear power sources are considered one of the most environmentally friendly strategies. In this study, the neutronic performance and hydrogen production potential of the PACER fusion blanket have been investigated during the operation period. Firstly, neutronic calculations have been performed to produce an electrical energy of 1000 MW from fusion explosions of 8.36 TJ to be repeated every 40 min with Monte Carlo Calculation Method (MCNP). Minor actinide flouride (MA) fuel out of the nuclear waste of LWRs with volume fractions of 2% have been mixed homogenously in the Flibe coolant. Tritium breeding ratio (TBR) at startup is calculated as 1.27. TBR and energy multiplication factor (M) decrease gradually. TBR >1.05 can be kept over 7 years for a self-sustained reactor. The M due to higher fissionable fuel content in the molten salt computes as 3.3 at initial. The burn up exceeds ∼800 GWd/tM at the end of operation period. Secondly, the potential of hydrogen production for four step iron chloride (Fe–Cl) and three step magnesium chloride (Mg–Cl) thermochemical cycles of the hydrogen unit PACER fusion blanket. The ratio of thermal power (1 − ψ), total thermal power (P hpf) and rate of mass flow of hydrogen ( m ˙ H 2 ) depends on M for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) thermochemical cycle have been presented. The ratio of thermal power (1 − ψ) for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) have been obtained as 0.6494, 0.131 and 0.095 whereas, total thermal power (P hpf) have been computed as 5298.8, 3608.43 and 3488.925 MW, respectively, at start up. The produced of hydrogen amount for Fe–Cl, Mg–Cl (option I) and Mg–Cl (option II) have been calculated as ∼ 534.65; ∼1236.82 and ∼806.03 kg/year, respectively, at startup. • Burning value of the PACER computed as ∼800 GWd/tM. • Hydrogen production obtained as 12.82 (Fe–Cl) and 39.22 (Mg–Cl) kg/s. • Energy efficiency calculated as 45% (Fe–Cl) and 68% (Mg–Cl). [ABSTRACT FROM AUTHOR]

Abstracts. A high-temperature molten salt natural convection loop was designed and manufactured for chloride-based salts. NaCl-MgCl2 salt was prepared and injected into the loop. Corrosion coupons composed of SS304, SS316L, and high-Ni alloys were also prepared as candidate structural materials for molten chloride salt reactors. After a corrosion experiment was conducted for 500 h, the salt was drained from the loop to the drain tank; inductively coupled plasma optical emission spectroscopy was used to investigate metal dissolution from the materials into the salt. The corroded materials were analyzed using scanning electron microscopy and energy-dispersive X-ray spectroscopy to calculate the corrosion rate. Materials exposed to NaCl-MgCl2 molten salt showed varying levels of corrosion resistance. The high-Ni alloy demonstrated the highest resistance, followed by SS316L and SS304. Furthermore, corrosion products were observed to migrate along the molten salt through natural convection, eventually depositing onto the surface of the high-Ni alloy in the cold leg of the loop. [ABSTRACT FROM AUTHOR]

Abstracts. A high-temperature molten salt natural convection loop was designed and manufactured for chloride-based salts. NaCl-MgCl2 salt was prepared and injected into the loop. Corrosion coupons composed of SS304, SS316L, and high-Ni alloys were also prepared as candidate structural materials for molten chloride salt reactors. After a corrosion experiment was conducted for 500 h, the salt was drained from the loop to the drain tank; inductively coupled plasma optical emission spectroscopy was used to investigate metal dissolution from the materials into the salt. The corroded materials were analyzed using scanning electron microscopy and energy-dispersive X-ray spectroscopy to calculate the corrosion rate. Materials exposed to NaCl-MgCl2 molten salt showed varying levels of corrosion resistance. The high-Ni alloy demonstrated the highest resistance, followed by SS316L and SS304. Furthermore, corrosion products were observed to migrate along the molten salt through natural convection, eventually depositing onto the surface of the high-Ni alloy in the cold leg of the loop. [ABSTRACT FROM AUTHOR]

This paper introduces a novel concept of a TRISO-fueled salt-cooled reactor (TFSCR). The core contains circulating pipes filled with molten salt carrying TRISO particles. The reactor achieves online refueling by slowly circulating the molten salt through the pipeline. The reactor utilizes the same molten salt as the coolant and graphite as the moderator. The reactor design has the characteristics of safety, economy, and nonproliferation. TRISO particles exhibit greater resistance to neutron irradiation, corrosion, oxidation, and high temperatures compared to conventional fuels. A molten salt-cooled reactor can also operate at higher temperatures, consequently enhancing power generation efficiency. Furthermore, lower operating pressures can mitigate the risk of significant damages and loss of coolant caused by accidents, thereby enhancing reactor safety. This paper presents the basic nuclear design of TFSCR under the assumptions concerning the average fuel power density, the volumetric core power density, and the core temperature. At the same time, the feasibility of online refueling and long-life operation was evaluated by fuel burnup calculation. [ABSTRACT FROM AUTHOR]

Molten salt systems have become of growing interest in the energy industry due to a wide range of applications (concentrated solar power systems, energy storage, Generation IV fission reactors, and high magnetic field fusion reactors). Because of the high temperature that characterizes such materials, radiative heat transfer (RHT) may become a nonnegligible heat transfer mechanism in molten salt components. In this paper, an investigation of FLiBe RHT has been conducted, with a focus on Affordable, Robust, Compact (ARC)–class fusion reactors, a preconceptual design proposed by Commonwealth Fusion Systems and the Plasma Science and Fusion Center. This class of reactors largely employs FLiBe molten salt due to its thermal and neutronic properties. The reactor is characterized by high temperatures, and its 0.5-m-thick liquid immersion blanket is a component where RHT contribution to the temperature distribution is yet to be evaluated. Therefore, this study is the first work that quantifies the contribution of RHT in ARC-class reactor FLiBe systems. FLiBe optical property spectral-banding assessment is carried out, and the impact of RHT in FLiBe systems is assessed in operational ARC-class scenarios through computational fluid dynamics models by taking advantage of COMSOL® Multiphysics. Heat transfer, thermal-dependent properties, and buoyancy effects are considered in a comparison between scenarios with and without RHT modeling. The flow field in the tank is unaffected by RHT effects, even when considering buoyancy effects. The external layer of the vacuum vessel shows an average decrease in the temperature of 5.4 K and an average decrease in temperature on the surface in contact with the FLiBe tank of 8.1 K. Results indicate that for ARC-class reactors, RHT phenomena are negligible (<1% increase in heat transfer) in operational conditions. [ABSTRACT FROM AUTHOR]

Off-gas systems are implemented in molten salt reactor designs to control the release of gaseous fission products. Two-phase flow in molten salt must be studied to understand how the system will behave in comparison to traditional working fluids like water. Flow visualization experiments and particle image velocimetry measurements were performed for three argon bubble sizes injected into a co-current stream of molten salt in a natural circulation loop facility. Similar bubble sizes were injected in experiments with water to compare the bubble shape, trajectory, and wake flow behavior of the fluids. The bubble region of interest was used to calculate the equivalent diameter and terminal velocity. Results for water showed a wobbling bubble surface and less stable bubble trajectory due to lower surface tension and viscosity compared with molten salt. Particle image velocimetry results demonstrated the increased viscosity of salt dampens turbulent fluctuations for the smaller bubble size. For a cap bubble, turbulent fluctuations were larger and longer lasting than in results for the wake flow of an argon cap bubble in water. [ABSTRACT FROM AUTHOR]

Breed and Burn (B&B) fuel cycle in molten salt reactors (MSRs) qualifies this reactor type as one of the best candidates to be developed for the Gen-IV R&D program. This feature can be approached by employing a closed fuel cycle and application of a molten salt reactor as a spent nuclear fuel burner; the features promise sustainable and clean energy in the future. In this study, a complete package has been developed to calculate core inventory, fuel burnup, and salt clean-up systems of molten salt reactors during their lifetime. To achieve this, the iMAGINE-3BIC package ("iMAGINE 3D-Reg Burnup & Inventory Calculator package") has been developed in MATLAB R2023a by employing a CINDER90 module of MCNPX 2.7 for burnup-calculation and multi-linear regression method (MLR). The package can estimate the core inventory (concentration of 25 actinides and 245 non-actinides elements) and the burnup of the reactor core during MSR lifetime (up to 100 years) while optimizing the computational resources (time, CPU and RAM), and it can even be hassle-freely executed on standalone PCs in an appropriate time due to its generous database. In addition, the salt clean-up module of the iMAGINE-3BIC package can be employed to evaluate the effects of the salt clean-up system on the above parameters over the MSRs' lifetime. Finally, the iMAGINE-3BIC package has been applied to an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) to evaluate and compare the performance of chloride/fluoride-based salt fuel MSRs from the point of burnup, core inventory, and salt clean-up systems. The results confirm that while a chloride-based salt fuel system has some advantages in less dependency on the salt clean-up system and fewer poisoning elements inventory, the fluoride-based system can achieve higher burnup during the reactor lifetime. The outcome of this study, along with the first part of this article, provides evidence to support the neutronic decision matrix as well as the pros and cons of employing chloride- or fluoride-based fuel systems in MSR cores. [ABSTRACT FROM AUTHOR]

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective. [ABSTRACT FROM AUTHOR]

The electrodeposition process of carbon in molten lithium carbonate electrolysis and the associated gas-liquid flow hydrodynamics characteristics are for the first time investigated using computational fluid dynamics (CFD). The high-temperature (750 °C) process is challenging for conducting measurements, making CFD a valuable tool for providing insights into the novel coaxial-type cell design. The CFD simulation addresses the electric field distribution, oxygen gas evolution, and electrodeposition of carbon. The effect of gas bubble sizes (1, 0.8, and 0.6 mm) on the electrolysis process was examined at different electrical current densities (0.15 ± 0.01 A c m – 2 ). The CFD results reveal that gas holdup increases by decreasing the bubble size and that the bubble size significantly impacts the current density distribution by affecting the two-phase flow dynamics. The study highlights the potential of CFD for optimization of molten salt processes, while also suggesting the need for further improvements in the model. [Display omitted] • CFD simulation of bubble behavior and carbon deposition in molten salt electrolysis. • Computational modeling of two-phase flow in molten salt electrolysis processes. • Understanding the impact of gas bubbles on current density and carbon deposition. [ABSTRACT FROM AUTHOR]

This paper describes the methodology used to create a fuel data library comprising safeguards-relevant quantities that may be useful for verification of spent nuclear fuel (SNF) produced by simulating a concept Molten Salt Reactor (MSR). The Monte-Carlo particle transport code, Serpent2 and the calculation code SOURCES 4C were used to compile this fuel data library. The data library is based on the Compact Molten Salt Reactor (CMSR) concept being developed by Seaborg Technologies (based in Copenhagen, Denmark). The library includes data such as nuclide mass densities for a total of 1398 nuclides (in g/cm3), as well as total decay heat production (denoted by suffix the ‘TOT_DH’) in Watts, total gamma photon emission rates (denoted by the suffix ‘TOT_GS’) in photos per second, and the total activity (denoted by suffix ‘TOT_A’) in Becquerel. Lastly, the data also includes total neutron emission rates from 1) spontaneous fission (denoted by ‘SF’ and reported in neutrons per second per cm3), and 2) (ɑ, n) reactions (denoted by ‘AN’ and reported in neutrons per second per cm3) for the fuel salt. These quantities are reported for a range of burnup-initial enrichment-cooling time (or collectively known as, BIC) parameters. The resulting fuel data library is an extension of a previously published data library for the same reactor concept but with one significant change. The current library is based on a more realistic model of the CMSR involving movement of gaseous and volatile fission products (GFP and VFP) from the core via an Off-Gas System (OGS). The dataset is made available for public use in a compressed binary format as an HDF5 (or Hierarchical Data Format) file that can be parsed using data analysis tools such as Pandas.