Search

This study presents a computational capability for fission product retention and release in two-phase, multi-species systems representing Molten Salt Reactors (MSR) with coupled thermal-hydraulics and fuel coolant chemical behaviours. This is demonstrated through four simulated cases centred on the proposed Molten Salt Fast Reactor (MSFR). This is achieved by two-way coupling the Computational Fluid Dynamics (CFD) code OpenFOAM and the Computational Thermodynamics (CT) code Thermochimica, using the Joint Research Centre Molten Salt Database (JRCMSD). Local chemical equilibrium is assumed, implying that chemical kinetics are predominantly governed by mass transport. Four simulations address normal operating conditions, exploring: (i) dilution of fission products injected within the molten salt coolant, (ii) molten salt coolant evaporation rate, (iii) release of radioactive gaseous species, (iv) shifts in the UF 4 /UF 3 ratio, and (v) comparison of vapour pressures of gaseous species. The influence of temperature-dependent viscosity on retaining fission products, compared to consistent values, is also discussed. The feasibility of integrating CFD with Thermochimica showed promising results, broadening insights into multiphysics systems and setting the stage for its application in more intricate scenarios. [Display omitted] • Coupling computational fluid dynamics and thermodynamics for MSR applications. • Local thermodynamic equilibrium using the MQMQA thermodynamic model. • Prediction of fission products retention and release. • Calculation of UF 4 /UF 3 ratio buffer to molten salt corrosion applications. [ABSTRACT FROM AUTHOR]

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for 233U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares 233U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the 233U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations. [ABSTRACT FROM AUTHOR]

1 1 Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (https://www.energy.gov/doe-public-access-plan). Safeguards verification measurements of nuclear material content in fresh fuel salt for liquid-fueled molten salt reactors (MSRs) are likely to be required as part of nuclear material accountancy for International Atomic Energy Agency safeguards. This paper presents a comprehensive review and evaluation of 18 potential candidate techniques to quantify total uranium and 235 U for input accountancy measurements for liquid-fueled MSRs. As part of an overall screening and down-selection effort to identify the most promising techniques for further development for an MSR feed monitoring system, this paper defines eight figures of merit (FOMs): reasonably achievable measurement uncertainty, measurement time required, capital cost, burden upon the facility operator, maintenance intensity, technological maturity, human capital requirements for operation, and whether the technique introduces a path for potential material removal. Each candidate technique is then evaluated across these FOMs to identify the techniques with the highest potential for future development for fresh fuel accountancy measurements in MSRs. Our findings indicate that no single technique or combination thereof currently has the requisite technological maturity for immediate implementation in nuclear material accountancy at a liquid-fueled MSR facility. While several promising techniques are identified, there is a critical lack of experimental data for most systems in the context of molten salt applications. • We propose criteria to screen feed accountancy techniques in molten salt reactors. • We evaluated 18 candidate techniques to identify the most promising for development. • Presently, no technique is sufficiently mature for MSR feed accountancy measurements. • Critical gaps in measurement data of actinide-bearing molten salts must be addressed. [ABSTRACT FROM AUTHOR]

• High-fidelity simulations for smMSRs with the complex shape plenum were performed. • The novel MSA technique is proposed and successfully accelerates the whole-core calculation for smMSRs. • The proposed MSA technique is easy to implement and suitable for special plenum regions. The planar method of characteristics (MOC) combined with the coarse-mesh finite difference (CMFD) method is highly effective for solving the neutron transport equation in pressurized water reactors (PWRs). However, applying this to small modular molten salt reactors (smMSRs) may lead to an inefficient CMFD acceleration due to a large void region outside the reactor vessel. A low order MOC-based synthetic acceleration (MSA) technique is then proposed, and addresses this issue by employing the same planar MOC method but with significantly relaxed computational parameters as the low-order solver. The MSA technique was validated with the C5G7-3D Rodded B case and the Molten Salt Reactor Experiment-like (MSRE-like) benchmark, where it achieved up to a speedup of 48 with a slight memory increase. The low order MOC-based synthetic acceleration scheme offers a promising approach for improving the computational efficiency of neutron transport simulations for smMSRs with a complex shape plenum. [ABSTRACT FROM AUTHOR]

Molecular dynamics simulations are a powerful tool for predicting the properties of molten salts at conditions sometimes challenging to achieve experimentally. Here, they are applied within the framework of the polarizable ion model (PIM) for the investigation of the NaCl-KCl-UCl 3 molten salt at two different compositions and for a temperature range between 1000 and 1400 K. This allowed us to understand the impact of sodium and potassium ions on the structure of the solvation shell of uranium ions, as well as predicting their density, viscosity, self-diffusion coefficients and heat capacity at different conditions. [ABSTRACT FROM AUTHOR]

Molten salt nuclear reactors are promising technologies for sustainable energy sources. In this context, understanding the physicochemical properties of molten salts is essential to the development, usage, and maintenance of reactors. Classical molecular simulations allow estimating the macroscopic properties of molten salts at a relatively low computational cost. In this work, we develop a new force field for plutonium-based molten salts. We validate our classical model by comparing molecular dynamics results with experimental measurements, obtaining a quantitative agreement. Our work demonstrates the reliability and transferability of classical force fields to represent the physicochemical properties of molten salts. [ABSTRACT FROM AUTHOR]

• The first proposal of a strategy for purifying waste salt through the adsorption of molecular sieve following electrolysis at high temperature is presented. • The high temperature adsorption of molecular sieve solves two main problems: (i) Low current efficiency; (ii) Molten salt can be recycled. • Electrochemical precipitation potential sequences and thermochemical parameters of Tm(III) and In(III) are provided. • On the basis of electrolytic refining, the removal rate of Tm(III) combined with molecular sieve greatly increased from 93.86 % to 99.5 %. • It provide important experimental ideas and data for spent fuel pyroprocessing and thorium-based molten salt reactor. In the process of pyroprocessing after spent fuel, the recycling of molten salt and electricity consumption are an important factor that restricts the cost of post-treatment. The implementation of high-temperature electrolysis and adsorption techniques can significantly enhance the extraction rate of Tm. This approach effectively addresses the issue of low current efficiency when ion concentration is low, resulting in reduced electricity expenses and facilitating the recycling of waste salt. In this paper, the electrochemical reduction and oxidation process of Tm(III) and In(III) were investigated on a tungsten (W) electrode at 773 K in the molten salts of LiCl-KCl. The findings suggest that the reaction process observed was a reversible, diffusion-controlled process. In-Tm alloys were synthesized using galvanostatic/potentiostatic electrolysis on W electrode in a LiCl-KCl-InCl 3 -TmCl 3 melt. The electrolysis process resulted in a Tm extraction rate of approximately 93.86 % from the In electrode. Subsequently, the removal of Tm (III) using 5A molecular sieve was investigated. Through the utilization of molecular sieves in high-temperature adsorption, the elimination rate can exceed 99.50 %, resulting in a significant enhancement in the purification of waste salt. The predominant adsorption mechanism was determined to be chemical (specifically ion exchange), as evidenced by the application of pseudo-first-order, pseudo-second-order, and intraparticle diffusion model of Weber-Morris. [ABSTRACT FROM AUTHOR]

The Molten Salt Reactor (MSR) represents a significant addition to India's clean energy initiative through nuclear energy. Noteworthy for its inherent safety features and the utilization of mixed fluoride salts (U, Th, Li) as molten-liquid fuel, this reactor stands out as a unique advancement. This paper showcases the application of Laser-Induced Breakdown Spectroscopy (LIBS) to ensure the chemical composition of the fuel. The micro-destructive nature of the analysis, minimal sample preparation requirements and the potential for remote analysis position LIBS as a viable alternative to Inductively Coupled Plasma Mass Spectrometry (ICP-MS). The study encompasses the quantification of UF 4 and ThF 4 , demonstrating both univariate and multivariate chemometrics calibration approaches. Various normalization methods on the raw spectral data are explored, and the results are meticulously compared. Upon evaluating different analytical methods, the optimal accuracy and precision achieved were in the range of approximately 2.6% and 3.9% for UF 4 , and about 0.4% and 0.5% for ThF 4 , respectively. This underscores the effectiveness of LIBS in ensuring the accurate determination of chemical composition in the context of MSR fuel. [Display omitted] • LIBS was used for quantitative characterization of Indian MSR fuel. • Different pre-processing & fitting models have been systematically compared. • SEC and SEP were used for different model quality comparison. • For UF 4 and ThF 4, precision of 3.9% and 0.5% was obtained respectively. [ABSTRACT FROM AUTHOR]

Spiral plate heat exchangers (SHE) have been proven in industrial applications involving viscous and fouling fluids. Their simplified maintenance requirements and high heat transfer capacity make them a great candidate for Molten Salt Reactor (MSR) applications. While traditional tube-and-shell heat exchangers also have proven industrial success, their complexity and significant maintenance requirements are disadvantageous in the corrosive environment of MSRs. Because the fuel salt of liquid fueled MSRs often passes through an external heat exchanger in a compact geometry with a potentially moderating secondary fluid, it is important to understand which conditions might lead to a criticality accident. Monte Carlo N-Particle® (MCNP®) code allows simulation of the criticality of the system and the neutron flux across the system. We evaluate the system's multiplication factor for several geometries and materials through parametric studies by varying SHE sizes, hot/cold channel gaps, and plate thickness for selected structural materials, fuel salts, and secondary loop heat transfer fluids. We characterize which configurations will likely lead to a critical system and identify some of the important SHE design considerations from a criticality perspective. Additionally, this work suggests the possibility of novel spiral MSR reactor core designs with an integrated heat exchanger under specific conditions. • SHE advantages in heat transfer capacity and maintenance could be applied to MSRs. • System multiplication factor is characterized for various SHE configurations. • MCNP simulations suggest several viable subcritical SHE configurations for MSRs. • Novel MSR core designs might attain an integrated heat exchanger with spiral designs. • SHE design considerations influencing the multiplication factor are discussed. [ABSTRACT FROM AUTHOR]

• The undocumented transient flow rates during the MSRE pump transient test are predicted. • The affinity law approximation is proved to be adequate for modeling the pump startup transient. • A novel approach is proposed to predict the coastdown flow rate based on the data obtained from the secondary pump. • The proposed model was validated by comparing the model predictions for the pump impeller speed with the measurements. Molten Salt Reactor Experiment (MSRE) is being extensively used for validating the computational tools for Molten Salt Reactors (MSRs). The pump transient tests were performed during the operation of MSRE to obtain the reactivity response to flow perturbation. The transient flow rate during this set of tests is required as input to predict the reactivity response. Since the primary loop of MSRE was not equipped with a flow rate meter, the transient flow rate, which is the crucial factor for the reactivity change in MSRE, is missing. Achieving an accurate simulation of the reactivity response to the MSRE pump transient tests necessitates a precise estimation of the transient flow rate. This paper endeavors to reconstruct the missing flow rate with the aim of offering a valuable input for simulating reactivity responses in the pump transient test series. In order to obtain an accurate transient flow rate, we employed a centrifugal pump transient model based on the affinity laws and solved the model for the MSRE secondary pump to validate the underlying assumptions of the model. In addition, we constructed the homologous pump head curves based on the water test data for the same purpose. The affinity law approximation is proved to be adequate in predicting the MSRE pump startup transient with the root-mean-square error (RMSE) in the normalized flow rate to be 0.03. On the other hand, for the MSRE pump coastdown transient, the preliminary results indicate that neither the affinity law approximation nor the homologous head curve is sufficient to provide acceptable predictions. To overcome this noticeable modeling deficit, we propose an innovative approach based on a straightforward data mining technique to regenerate the MSRE pump coastdown homologous relations using the measured data of the secondary pump. These relations are transferred to the primary pump and used to simultaneously solve for the impeller speed and the flow rate of the pump in the primary pump coastdown. With the new approach, the estimated RMSE in the normalized pump speed for the primary pump coastdown is reduced 0.0052. This excellent agreement validates the accurate calculations for the flow rate during the pump coastdown transient. We also performed an uncertainty analysis to quantify the confidence interval of the predicted quantities, which further justifies the robustness of the proposed approach. [ABSTRACT FROM AUTHOR]

The service of structural materials, such as steels and alloys, in molten salt and sodium-cooled reactors in nuclear power generation is associated with high temperature corrosion of these materials, which leads to degradation and failures of structural components. To prevent these issues and to extend the reactor components' service life, protective coatings on steels need to be employed. In the present work, thermal diffusion protective coatings based on borides and aluminides have been tested, for the first time, in molten Na-K-Mg-chlorides of eutectic composition and in molten sodium at 500 °C in high-purity Ar glovebox where the test conditions simulated molten materials environments in nuclear reactors. The structural and compositional examination of the proposed protective coatings revealed only their minimal changing and deterioration in the corrosive environments. Thus, no or minimal formation of the corrosive scale on the top of the coatings was observed with neither structure transformation, nor reduction of the protective layer after 30-days testing in the molten materials' environments. Furthermore, no penetration of the corrosive media to the steel substrates occurred. Due to promising test results, the selected protective coatings may be highly recommended for further long-term and higher temperature testing in modeling and actual molten salts and molten sodium conditions associated with the service in nuclear reactors. • Molten medium corrosion of boronized and aluminized coatings was studied. • Coatings exhibit robust structural integrity and corrosion resistance. • High temperature leads to minor cracks of boronized coatings. • Aluminized coating's ductility prevents formation of minor cracks. • Molten chloride salt corrosion resistance of these coatings outperforms bare steels. [ABSTRACT FROM AUTHOR]

• Reassessed the thermodynamic functions for CsF(g) and Cs 2 F 2 (g) to improve fit to experimentally measured vapor pressure and monomer–dimer distributions over condensed phase CsF. • Original phase equilibria determinations for the LiF-CsF, KF-CsF, LiF-NaF-CsF and LiF-NaF-KF-CsF systems. • First-time representation of the LiF-CsF system using non-default SNN coordination numbers to account for short range ordering in the melt. • First time representation of the LiF-NaF-KF-CsF system using the modified quasichemical model. • Successfully reproduced the experimentally measured vapor pressure of cesium-containing species over a dilute mixture of CsF with FLiNaK. Gibbs energy models were developed to describe the thermochemical behavior of CsF in molten FLiNaK (46.5LiF-11.5NaF-42KF mol%), a proposed molten salt reactor (MSR) fuel solvent and coolant, as cesium is of concern due to its high radiotoxicity and volatility. Initially, it was necessary to obtain a more accurate Gibbs energy function for CsF which required fitting parameters to reported vapor pressures over condensed phase CsF. The pseudo-binary systems CsF-LiF, CsF-NaF and CsF-KF were then evaluated utilizing phase equilibria and enthalpy of mixing (Δ mix H) values, together with original differential scanning calorimetry (DSC) measurements performed for the CsF-LiF and CsF-KF systems. The CsF-LiF-NaF, CsF-LiF-KF and CsF-NaF-KF pseudo-ternary system representations were obtained by interpolation of the constituent pseudo-binary systems, with DSC measurements performed for the CsF-LiF-NaF system to corroborate the calculated liquidus temperature. Ultimately, the pseudo-ternary systems were interpolated to obtain Gibbs energy models for the pseudo-quaternary CsF-LiF-NaF-KF system, supported by DSC measurements at low CsF compositions (1–10 mol%), yielding computed equilibria and cesium-containing vapor pressures that compare favorably with reported values. The Molten Salt Thermal Properties Database – Thermochemical (MSTDB-TC) was subsequently expanded to include these Gibbs energy models allowing description of the thermochemical behavior of the CsF-LiF-NaF-KF system. [ABSTRACT FROM AUTHOR]

Three sequential batch reactors (SBR) were operated to evaluate salt addition's impact on granulation, performance, and biopolymer production in aerobic granular sludge (AGS) systems. System R1 was fed without adding salt (control); system R2 operated with saline pulses, i.e., one cycle with salt (2.5 g NaCl/L) addition followed by another without salt; and R3 received continuous supplementation of 2.5 g NaCl/L. The results indicated that the reactors supplemented with salt presented higher concentrations of mixed liquor volatile suspended solids (MLVSS) and better settleability than R1, showing that osmotic pressure contributed to biomass growth, accelerated granulation, and improved physical characteristics. The faster granulation occurred in R2, thus proving the beneficial effects of intermittent salt addition through alternating pulses. Salt addition did not impair the simultaneous removal of carbon, nitrogen, and phosphorus. In fact, R2 showed better carbon removals. In conclusion, continuous or intermittent (pulsed) supplementation of 2.5 g NaCl/L did not lead to increased production of extracellular polymeric substances (EPS) and alginate-like exopolymers (ALE). This outcome could be attributed to the low saline concentration employed, a higher food-to-microorganism (F/M) ratio observed in R1, and possibly greater endogenous consumption of biopolymers in the famine period in R2 and R3 due to the greater solids retention time (SRT). Therefore, this study brings important results that contribute to a better understanding of the effect of salt in continuous dosing or in pulses as a selection pressure strategy to accelerate granulation, as well as the behavior of the AGS systems for saline effluents. [Display omitted] • Aerobic granulation and resource production were investigated in AGS systems. • Continuous and intermittent salt addition was evaluated. • Intermittent osmotic pressure favors the granulation process and granule stability. • Thauera was the most abundant genus found in the saline reactors. • Blastopirrellula, Phaeodactylibacter, and Rodhopirellula could also be found in the saline reactors. [ABSTRACT FROM AUTHOR]

• Catalytic effect follows the order of KF > NaF > LiF. • Basic plane of the salt-infiltrated graphite is seriously damaged. • Adsorption energy of O 2 on graphite surface increases in the presence of alkali atom. Air ingress accident scenarios may result in the oxidation of graphite matrix materials and cause serious safety problems in molten salt reactor. However, the oxidation behavior of the molten-salt-infiltrated graphite remains elusive. In this study, the effect of alkali fluoride salts (LiF, NaF, KF and ternary salt FLiNaK) on the oxidation behavior of graphite powder under air atmosphere was investigated. Thermogravimetric analysis and oxidation tests manifest that the initial oxidation temperature of the graphite is reduced from 720 °C to 510–610 °C, because of the presence of the alkali fluoride salts. The catalytic oxidation effect follows the order of KF > NaF > LiF. Careful characterizations reveal that the pristine graphite is mainly oxidized at the graphite edges, while salt-infiltrated graphite is mainly damaged at the graphitic basal planes. Density functional theory calculations suggest that the doping of alkali metal atoms does not change the physical adsorption feature of the oxygen molecule on the graphite plane surface, but increases its adsorption energy to facilitate the graphite oxidation reactions. [Display omitted] [ABSTRACT FROM AUTHOR]

Recent studies related to the thermal neutron scattering (TNS) data in reactors were generally focused on fuels and moderators, while the TNS data for structural materials has received little attention. The TNS effect of the GH3535 alloy, a nickel-based alloy widely used in molten salt reactor (MSR), is studied in this work. The partial phonon densities of states (PDOS) for the main metals (Ni, Mo, Cr and Fe) in the alloy, which are required for the TNS data generation, are calculated using the density functional perturbation theory (DFPT) based on a special quasi-random structure (SQS) of the alloy. The TNS cross sections for the main metals in the alloy are generated using the modified NJOY code. To preliminarily validate the TNS data, the calculated scattering cross section, combined with the evaluated absorption cross section of the alloy, is compared with the experimental total cross section, which shows a good agreement. Moreover, an integral neutronics experiment of neutron leakage spectrum is designed to further validate the TNS data. The neutron leakage spectrum is simulated by the Monte Carlo method, which can be compared with the future experimental result. Based on the experiment geometry modeling, the TNS effects of the GH3535 alloy are analyzed at multiple incident neutron energies, sample thicknesses and temperatures, respectively. It is found that the TNS effect of the GH3535 alloy on the neutron leakage spectrum cannot be ignored, and the scattering cross section of the alloy with high precision should be included in designs of MSR. • The partial phonon densities of states for main metals in the GH3535 alloy are calculated. • The thermal neutron scattering (TNS) data for the alloy is generated. • The impacts of the alloy's TNS effects on neutron leakage spectra are analyzed. [ABSTRACT FROM AUTHOR]

Combining with the appealing technologies of the liquid-fuel molten salt reactor and the passive heat transfer of heat pipes, the micro Molten Salt Reactor (micro-MSR) has prospective applications in space, the ocean and other off-grid remote areas. Due to the distinctive features of these applications and liquid fuel in a micro-MSR, special attention should be paid to the safety issues of the micro-MSR. The study aims to evaluate the safety features of the micro-MSR core under various accident scenarios and to further develop effective control schemes. The core reactivities under cool-down conditions are initially assessed based on the micro-MSR core optimized in the early work. The results indicate that the micro-MSR is slightly subcritical or supercritical due mainly to a shrunken salt volume. Then, an analysis of the core reactivities under various submersion and flooding accidents is performed. It is found that the micro-MSR would be highly supercritical under the accident scenarios where the core is submerged in wet sand, seawater or fresh water and meanwhile, the heat pipes are saturated with water. The core reactivity would grow by up to 28563 pcm in the worst-case scenario where the core is surrounded by wet sand and the heat pipes are saturated with seawater. Two control schemes, including the safety-rod and the fuel-out schemes, are proposed to redesign the core and prevent inadvertent criticality accidents before startup. Both control schemes can ensure an adequate shutdown margin under all accident scenarios by introducing more neutron poisons within the core or lowering the initial fuel loading. The two control schemes are also compared in terms of core size, initial fuel distribution, battery's capacity and startup sequence. The fuel-out scheme is considered superior to the safety-rod scheme owing to its smaller size, lower weighted mass and safer startup sequence. [Display omitted] • The core reactivities under accident scenarios are evaluated for micro-MSR. • The safety-rod and the fuel-out control schemes are proposed and designed for micro-MSR. • The fuel-out scheme is superior owing to its compact, low-weighted and safe features. [ABSTRACT FROM AUTHOR]

[Display omitted] • EuCl 3 -based method for corrosion rate of materials in NaCl–MgCl 2 molten salt proposed. • Current for reduction of corrosion products, Ni2+, Fe2+, measured as function of time. • Relationship between EuCl 3 concentration and corrosion rate determined. • Corrosion rates of Inconel 625 and SS 316 at 580 ℃ estimated. • This is a facile solution for in-situ corrosion monitoring in molten salts. Molten chlorides find varied applications in engineering disciplines, notably in the nuclear energy sector. However, the utilization of molten chlorides in molten salt reactors (MSRs) requires investigation of the corrosion behavior of the reactor's structural materials in such environments. In this study, we proposed a facile method for determining the corrosion rates of structural materials immersed in a high temperature NaCl–MgCl 2 molten salt, based on electrochemical measurements by cyclic voltammetry. The method involves measuring the electrochemical signals of the soluble corrosion products as a function of time. EuCl 3 -induced corrosion of Inconel 625 and stainless steel 316 (SS 316) in molten NaCl–MgCl 2 was monitored by electrochemical detection of Ni2+ and Fe2+. The corrosion rates were linearly dependent on the EuCl 3 concentration, and first-order rate constants of 1.03 × 10-4 and 3.20 × 10-5 s−1 at 580 ℃ were obtained for Inconel 625 and for SS 316, respectively. Arrhenius plot analysis of the temperature dependence of the rate constant provided an activation energy of 43.3 kJ mol−1 for Inconel 625 corrosion in molten NaCl–MgCl 2 –EuCl 3. The rate constant became temperature-invariant above 620 ℃ indicating the mass-transfer limitation of corrosion at higher melt temperatures. The proposed approach enables accurate and real-time evaluation of the corrosion rate in the molten salt, and accordingly, offers a straightforward solution for in-situ corrosion monitoring in MSR systems. [ABSTRACT FROM AUTHOR]

Environmentally-assisted cracking is a well-known and extensively studied phenomenon in light water reactor environments but remains relatively unexplored in proposed molten salt reactor systems where the mechanisms at play are expected to be quite different. Here, slow strain rate testing has been performed on 316L stainless steel tensile specimens during simultaneous exposure to a molten LiF-KF-NaF eutectic salt environment at 600 °C with a strain rate between 1E-5 to 1E-7 s-1. Sample ductility and ultimate tensile strength were both observed to decrease for slower strain rates/longer salt exposures. Post-mortem characterization of fractured specimens revealed Fe deposits and Cr- and O-rich precipitates decorating surface grain boundaries and crack surfaces, suggesting a cracking mechanism driven by embrittlement of the crack tip due to oxide formation. Unlike previous studies, intergranular Cr corrosion into the bulk was not observed, and distinct differences in surface morphology between electropolished and non-electropolished sample surfaces are reported. Trace metal salt chemistry measurements showed accumulation of 316L alloying elements over the course of the tests. It appears likely that the surface oxide layer induced by electropolishing, in addition to oxygen and moisture impurities in the salt, led to the formation of LiCrO 2 which had a net passivating effect that prevented significant bulk corrosion, though some interplay between mechanical and environmental stressors is still observed. [ABSTRACT FROM AUTHOR]

[Display omitted] • Improved experimental results and thermodynamic models were obtained for the NaI-CsI and KI-CsI pseudo-binary systems using DSC measurements. • New reciprocal salts measurements of the NaF-CsI system. • Experimental sensitivity analyses for sample mass and heating rates, determined that 50 mg and 5 °C/min are reasonable values. • Contribution to the Joint Research Center Molten Salt Database (JRCMSD). The present study describes the thermodynamic assessment of three pseudo-binary systems relevant to CsI solubility in molten iodide salts: KI-CsI, NaI-CsI, and NaF-CsI. The motivation for this study was to corroborate a single previously reported data set of the NaI-CsI system, resolve inconsistencies reported by two different data-sets of the KI-CsI system, and generate new experimental data on the NaF-CsI system. Equilibrium data for all systems were obtained using Differential Scanning Calorimetry. Thermodynamic treatments of the three pseudo-binary systems were revised using the CALPHAD method with the thermodynamic software FactSage and Thermochimica. Both experimental and computational investigations provide increased confidence in the thermochemical behaviour of CsI in Molten Salt Reactor nuclear systems. [ABSTRACT FROM AUTHOR]

Onion-like carbon spheres (OLC) were doped into the traditional graphite raw material system to decrease the pore diameter to ∼400 nm, with potential for use in molten salt reactors. The obtained graphite (OLCG) prepared from OLC was subjected to 7 MeV Xe26+ irradiation in this work, and a systematic investigation was conducted to verify its irradiation behavior. The surface morphology and microstructural evolution in graphite samples under different irradiation conditions were studied with different characterization methods. The surface roughness of OLCG increased and the hill appeared at lower irradiation doses, which was similar to that of IG-110, but the filler particles of OLCG gradually become spherical at 2.5 dpa. The analysis by X-ray diffraction reveals that OLCG's graphitization degree is approaching but slightly worse compared to IG-110. Raman studies indicate that the irradiation causes an increase in defect density and in-plane disorder both in graphite. The OLC filler exhibits a greater sensitivity to radiation, leading to OLCG being slightly sensitive to radiation as compared to IG-110. Neutron irradiation of the resulting graphite should be continued as OLCG showed good irradiation resistance at low irradiation doses and had a much smaller pore diameter. • Preparation of small pore graphite by doping with OLC. • First study on irradiation behaviors of graphite surface changes with OLC fillers. • Onion-like carbon spheres fillers became spherical after irradiation. • Surface changes are attributed to irradiation-induced microstructural evolution. [ABSTRACT FROM AUTHOR]

The optimal shape of a fusion-based transmutation reactor with a molten salt coolant was determined by plasma physics, technology, and neutronic requirements. System parameters such as neutron multiplication, power density, shielding, and tritium breeding, were calculated in a self-consistent manner by coupling neutron transport analysis with conventional tokamak systems analysis. The plasma physics and engineering levels were similar to those used in the International Thermonuclear Experimental Reactor. The influence of aspect ratio of the tokamak and fusion power on the radial build, and the transmutation properties associated with two molten salt options, FLiBe and FliNaBe, were investigated. Being compared with a transmutation reactor with a small aspect ratio, a transmutation reactor with large aspect ratio was smaller in size and had a larger maximum fusion power. This type of reactor also revealed increased tritium-breeding capability and a smaller initial transuranic (TRU) inventory with a slightly lower burn-up rate. The burn-up rate for molten salt using either FLiBe or FLiNaBe was similar, but the initial TRU inventory and the tritium-breeding capability were smaller with FLiNaBe compared with FLiBe. [ABSTRACT FROM AUTHOR]

Variable renewable energy availability has increased the volatility in energy prices in most markets. Nuclear power plants, with a large ratio of capital to variable costs, have historically operated as base load energy suppliers but the need for more flexible operation is increasing. We simulate the techno-economic performance of a 950 MW t nuclear power plant, based on the Westinghouse lead-cooled fast reactor, coupled with molten salt thermal storage as a method for flexible energy dispatch. We use the System Advisor Model to model the nuclear reactor thermal power input and power cycle operating modes. We combine this robust engineering model with a mixed-integer linear program model for optimized dispatch scheduling. We then simulate the coupled nuclear and thermal storage system under different market scenarios with varying price volatility. We find that the coupled plant outperforms the base plant under markets where energy price peaks fluctuate by a factor of two or more about the mean price. We show that a calculated power purchase agreement price for the plant improves by up to 10% when operating under California energy market conditions. Sensitivity analysis on the thermal storage cost shows that the optimal design remains unchanged even when doubling costs. [ABSTRACT FROM AUTHOR]

An efficiency indicator is proposed for the transmutation of Np, Am, and Cm that characterizes the relation between the burnup of these elements, the consumption of plutonium, and the production of high-level waste (except for fission products). A comparative analysis of transmutation in a molten-salt burner reactor with the salts LiF–NaF–KF and LiF–BeF2 that high transmutation efficiency, going asymptotically to 100%, depending on the volume of the core, obtains with the first salt and concentration of actinide fluorides in the range 10–20 mol.%, which experimental data show to be lower than their solubility. For the second salt LiF–BeF2, the highest transmutation efficiency ~40% obtains with core volume >15 m3. Such a reactor consumes 1.5 times more plutonium than Np, Am, and Cm. [ABSTRACT FROM AUTHOR]

A numerical framework for modeling depletion and mass transport in liquid-fueled molten salt reactions is presented based on exponential time differencing. The solution method involves using the finite volume method to transform the system of partial differential equations (PDEs) into a much larger system of ordinary differential equations. The key part of this method involves solving for the exponential of a matrix. We explore six different algorithms to compute the exponential in a series of progression problems that explore physical transport phenomena in molten salt reactors. This framework shows good results for solving linear parabolic PDEs with each of the six matrix exponential algorithms. For large problems, the series solvers such as Padé and Taylor have large run times, which can be mitigated by using the Krylov subspace. [ABSTRACT FROM AUTHOR]

The continued development of molten salt reactors and concentrated solar power plants requires highly efficient and stable instruments that can efficiently monitor the chemical conditions of the molten salt during long‐term operation in both the fuel and coolant/heat transfer fluid loops. Sapphire (Al2O3) fibers have shown tremendous potential due to inherent radiation resistance and a broader operational range of temperature. In this work, computational thermodynamic modeling (CALPHAD) using the ThermoCalc software in conjunction with the SGTE (Scientific Group Thermodata Europe) Molten Salts (SALT1) and Pure Substances (Pure5) databases is applied to understand the compatibility of Al2O3 fibers with NaCl‐MgCl2 eutectic molten salt in the temperature range of 1500–2500 K. The thermodynamic calculations show that sapphire fibers are not expected to be stable over the long‐term when exposed to molten chloride salts at these temperatures. To improve the stability of these diagnostic fibers in molten salt environments, various pure metallic elements were evaluated as potential cladding materials for Al2O3 fibers. Based on the thermodynamic analysis, molybdenum (Mo) and nickel (Ni) could be effective cladding materials to enhance the stability of Al2O3 in NaCl‐MgCl2 chloride salt molten bath in the desired temperature range. Thus, the presence of Mo and Ni cladding can provide a protective coating against the corrosive molten salts, thus improving the stability of Al2O3. Additionally, it is also shown that Al2O3 remains stable up to 2400 K in the presence of preexisting Al2MgO4 and Al2NiO4 in the eutectic molten chloride bath environment. [ABSTRACT FROM AUTHOR]

Molten chloride salt fast reactors (MCFRs) will require UCl3 dissolved in molten salt mixtures as fuel for nuclear fission. For infusing the salt with UCl3, bubbling HCl into NaCl-CaCl2 in contact with U metal was investigated. The reaction was run up to 9 h and yielded U concentration up to 0.652 wt.%. Open circuit potential between a W electrode and Ag/AgCl reference electrode yielded a potential consistent with uranium existing as U(III) in the salt. This demonstrates that HCl can be a very effective chlorinating agent to infuse MCFR fuel with UCl3 starting from U metal. [ABSTRACT FROM AUTHOR]

Nuclear fission technologies have the potential to play a significant role in the energy mix of a net-zero and sustainable society. However, to achieve the sustainability goal two significant challenges remain: efficient and sustainable fuel usage and the minimization of long-term nuclear waste. Civil nuclear molten salt systems and technologies offer the opportunity to address both, delivering future reactors at scale for efficient and effective power production and nuclear waste burnup. Potentially, both objectives could be fulfilled in one reactor system, which could significantly improve sustainability indices. The key to this innovation is demand driven development of a significantly reduced fuel cycle with enhanced proliferation resistance which offers further potential for improvement. To achieve these goals, a transformative approach for salt clean-up during molten salt reactor operation is proposed, by concentrating on the detection and removal of key neutron poisoning elements which prevent the reactor from long-term operation. To enable this highly innovative development work, a novel analysis of the evolving elementary fuel composition, their concentrations, and their criticality influence is now provided in this work. This, combined with consideration of the oxidation states of each of these elements then provides the basis for the selection of these key poisons and the development of advanced separation processes and process monitoring. This work also discusses the importance of the effective integration of physics and chemistry when systems modelling in achieving these system development goals. [ABSTRACT FROM AUTHOR]

High temperature interaction of LiF-NiF2 system is characterized with the ternary compound, Li2NiF4(s). The compound has been synthesized by solid state route. Li2NiF4(s) has been characterized with XRD technique and its micro-homogeneity and surface morphology have been tested with EDS and SEM technique, respectively. In order to study the thermodynamic stability, the Gibbs energy of formation, ΔfG°m(T), of Li2NiF4(s) has been measured with solid electrolyte galvanic cell technique and its measured ΔfG°m(T) can be expressed as ΔfG°m(T) [Li2NiF4(s)] / k.J mol−1 = (-1902.2 ± 1.1) + (0.3491 ± 0.0016) × (T/K). Using the experimental thermodynamic data, ternary phase diagram and fluorine potential diagram of Li-Ni-F2 system have been calculated in order to study the stability domain and coexisting phases of Li2NiF4(s) at reactor operating temperature. The stability of Li2NiF4(s) has also been investigated in the presence of oxygen impurity by calculating chemical potential diagram of Li-Ni-F-O system and ternary phase diagram of LiF-NiF2-O2 system at the same temperature. In the study, interaction of major structural components, Ni, Cr, and Fe, with fuel salt components (ThF4 and UF4) of MSR has also been investigated at different temperatures. [ABSTRACT FROM AUTHOR]

The corner defects in the casting slab greatly influence the product quality. These defects may extend during the heating and rolling process and even result in the discarding of the rolled plate as scrap. A corner cleaning equipment based on the scarfing machine is proposed to eliminate the defects in slab corners for producing high-quality steel. Unlike the flat surface scarfing process, the flame jet and the shape of the molten pool have an essential impact on the effectiveness of the flame cleaning in the corner of the casting slab. A three-dimensional fully coupled model for the flame cleaning nozzle is developed to simulate the flow pattern of the flame jet, Oxygen concentration distribution, and temperature field in the corner of the slab. The simulated flame jet flow field and temperature results agreed well with the factory trial results. Additionally, a three-dimensional thermal model for simulation of the molten pool formed by flame cleaning in the corner of the casting slab has also been developed. For the sake of simplicity, the 2D elliptic and 3D Gauss heat source models are used to simulate the flame heating on the upper and right surfaces of the slab corner and the reaction heating between oxygen and heated iron along the corner, respectively. The simulation results show that the length is 58.1 mm and 57.9 mm on both sides and the corner melting depth is 29.9 mm. The error is 7.04%. The numerical simulation results showed good agreement with the factory trial results, indicating that the proposed models of the flame jet and the heat sources analysis are feasible to study the flame cleaning process of the slab corner, it provides the scientific theoretical basis for the design and practical application of corner scarfing machine. [ABSTRACT FROM AUTHOR]

Nuclear energy represents an important option for generating largely clean CO2-free electricity. Zirconium is a fission product in the nuclear reaction that needs to be extracted from irradiated fuels used in Gen-IV molten salt reactors. The present investigation addresses the electrochemical reduction of solution soluble Zr4+ (soln) to surface confined Zro (s−c) at a molybdenum electrode in a LiF—CaF2 eutectic molten salt at 840 °C using DC cyclic, square-wave and AC voltammetry. Cyclic voltammograms simulated by the reaction scheme: Zr4+ (soln) + 4e− → Zro (soln); Zro (soln) ↔ Zr*(s−c); Zr*(s−c) → Zr4+*(s−c) + 4e−; Zr4 +*(s−c) → Zr4+ (soln); Zr*(s−c) + Zr4+ (soln) ↔ 2Zr2+ (soln); Zr2+ (soln) ↔ Zr2+*(s−c) and Zr2+*(s−c) → Zr4+*(s−c) + 2e− provided excellent agreement with experimental data over the scan rate range of 50 to 500 mV s−1. The interpretation of the simulation is that the reduction of Zr4+ (soln) to Zro (metal) takes place via a transiently soluble Zro (soln) in an overall 4-electron essentially reversible diffusion-controlled process having a reversible formal potential (Eo f) of −1.22 V (vs Pt). A minor oxidation process observed at −0.455 V (vs Pt) on the reversing the potential scan direction is simulated via the reaction step Zr(s−c) + Zr4+ (soln) ↔ 2Zr2+ (s−c) followed by Zr2+ (s−c) → Zr4+ (sol) + 2e−. The sharply rising initial component where reduction of Zr4+ (sol) commences, contains evidence of a nucleation-growth mechanism associated with the electrocrystallisation of zirconium metal. This initial rapid growth of current is not fully accommodated in the simulations, but all features found beyond the peak potential are supported by the theory. A comparison with theory based on a direct reduction of Zr4+ (soln) to the metallic state having unit activity is provided. It is proposed that an analogous mechanism applies at a Ni electrode, except that a Ni-Zr alloy formation occurs instead of Zr metal. [ABSTRACT FROM AUTHOR]

The aim of this paper is to present results of fast neutron behavior analysis within the chloride salts environment using simulations based on Monte Carlo method (MCNP 6.2). Three non-fueled salts (NaCl, KCl, MgCl2) and two salts containing fissile material (UCl3, ThCl4) were studied. Results of this theoretical study will be used for design of an experimental assembly, which will serve to achieve goals of the international research project ADAR (Accelerator Driven Advanced Reactor). One of the project objectives is to investigate chloride salts as potential coolant and a dissolved fuel carrier of advanced nuclear reactor cooled by molten salts and driven by an accelerator.

MAX phases are a large family of damage‐tolerant structural materials for the nuclear energy industry. The anti‐corrosive property to the tritium fluoride (TF) produced by the neutron irradiation of FLiBe is an issue for the MAX phases applied to the molten salt reactors (MSRs). Herein, the Ti2SC phase was used for the first time to evaluate the corrosion resistance to the concentrated hydrofluoric acid as the simulated service condition. The comprehensive characterizations including weight‐loss curves, X‐ray diffraction, field emission scanning electron microscopy, and energy dispersive spectroscopy were employed to analyze the corrosion behavior and mechanism of Ti2SC in comparison with Ti3AlC2 and Ti3SiC2. The good intrinsic anti‐corrosive property of Ti2SC was originated from the strong Ti‐S bond. Considering the acceptable preliminary performance at low displacement per atom (dpa), the Ti2SC probably has a potential to be used as the structural material in the MSRs. [ABSTRACT FROM AUTHOR]

We report the successful infusion of molten salt mixtures containing LiCl–KCl, NaCl, and NaCl-CaCl2 with UCl3 via reaction of U metal with iron chlorides (FeCl2 and FeCl3). Reaction in LiCl–KCl and NaCl-CaCl2 resulted in a yield of 93% and 96.7% using FeCl2 at 500 and 600 °C, respectively. Reaction to form NaCl-UCl3 at 850 °C had a yield of 70.6%. Volatilization of the oxidant may explain low yields. Reaction with the more volatile FeCl3 in NaCl-CaCl2 at 600 °C resulted in 80.7% yield. Open circuit potential measurements were made and yielded values consistent with very high selectivity for UCl3 rather than UCl4. [ABSTRACT FROM AUTHOR]

The long-lived minor actinides (MA); americium, neptunium, and curium are main contributors to the long-term radiotoxicity of used fuel. Thus, the transmutation of these MAs is considered as an alternative to direct burial. Until now, no unambiguous internationally recognized quantitative criterion for the effectiveness of MA transmutation has been developed, although this would be highly desirable. The absolute and relative decrease in the total mass of MA is completely inadequate, since they ignore the accumulation of higher radiotoxic MA from the transmuted nuclide. In this paper, we propose a new criterion for the efficiency of MA transmutation in nuclear reactors and demonstrate its efficiency when comparing two molten salt reactors; Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and Small Molten Salt Fast Reactor (SMSFR). In addition, the proposed criterion takes into account the mass of all useful MA, short-lived MA, and short-lived fission products (FPs). We present a new approach to loading MA in SD-TMSR and SMSFR. The total change in the mass of actinides and FPs during irradiation was calculated using SERPENT-2 Monte Carlo code. The results show that the transmutation efficiency of 241Am (a major candidate for transmutation) in SD-TMSR is much higher than in SMSFR. After 1500 days of irradiation, the transmutation efficiency reaches 82.6% for SD-TMSR, but for SMSFR it reaches 52.5%. [ABSTRACT FROM AUTHOR]

The problems of utilization of waste from re-enrichment of regenerated uranium from 232U isotope in a reactor with circulating liquid fuel based on melts of metal fluoride salts are discussed. The effect of isotopic composition of the purification cascade selection fraction on the spectrum and reactivity coefficients in a liquid salt nuclear reactor depending on the content of 235U isotope is studied. It is shown that compositions with a high content of 235U isotope, on one hand, are more promising, since less minor actinides are produced owing to the low content of 238U isotope, and on the other hand, more favorable reactivity effects are provided with moderate enrichment of 235U isotope. [ABSTRACT FROM AUTHOR]

The manuscript describes the development of a novel pathway to produce hydrogen based on the thermal decomposition of methane with carbon black as the catalyst. The thermal reaction occurs at about 1000°C in a movable bed of carbon black particles. The thermal energy is provided by molten salt MgCl2-KCl flowing from a hot thermal energy storage tank to the cold thermal energy storage tank through the reactor, and from the cold tank to the hot tank through a higher concentration solar receiver. The reaction occurs at the surface of the particles that are recycled in a loop until they reach the desired size. This pathway has a minimum energy requirement that is much less than white and green hydrogen from the splitting of the water molecule. Even if based on hydrocarbon fuel, it has no direct production of CO2, but rather it produces carbon black particles of commercial interest. The pathway has the potential to deliver at a cost less than any other pathway carbon dioxide-free hydrogen. The most critical aspect of the proposed pathway is the design and development of a high-efficiency reactor for the proposed temperature range capable of continuous operation. [ABSTRACT FROM AUTHOR]

Nuclear technologies have strong potential and a unique role to play in delivering reliable low carbon energy to enable a net-zero society for future generations. However, to assure the sustainability required for its long-term success, nuclear will need to deliver innovative solutions as proposed in iMAGINE. One of the most attractive features, but also a key challenge for the envisaged highly integrated nuclear energy system iMAGINE, is the need for a demand driven salt clean-up system based on the principles of reverse reprocessing. The work described provides an insight into the dynamic interplay between a potential salt clean-up system and reactor operation in a plutonium-started core in a dynamic approach. The results presented will help to optimise the parameters for the salt clean-up process as well as to understand the differences which appear between a core started with enriched uranium and plutonium as the fissile material. The integrated model is used to investigate the effects of the initial fissile material on core size, achievable burnup, and long-term operation. Different approaches are tested to achieve a higher burnup in the significantly smaller Pu-driven core. The effects of different clean-up system throughputs on the concentration of fission products in the reactor salt and its consequences are discussed for general molten salt reactor design. Finally, an investigation into how a plutonium loaded core could be used to provide fuel for future reactors through fuel salt splitting is presented, with the outcome that one Pu-started reactor of the same size as a uranium-started core could deliver fuel for 1.5 new cores due to enhanced breeding. The results provide an essential understanding for the progress of iMAGINE as well as the basis for inter-disciplinary work required for optimising iMAGINE.

This editorial discusses the technological advancements in Gen4 nuclear energy systems and small reactors. These advanced systems, such as sodium-cooled fast reactors, lead-cooled fast reactors, and gas-cooled fast reactors, offer superior sustainability, safety, reliability, and economics compared to current-generation systems. The article highlights specific research on the safety analysis of a lead-based reactor, the operational characteristics of a micro gas-cooled reactor, the flow stability of supercritical carbon dioxide in a Brayton cycle, and the fault diagnosis of rotating machinery using deep learning models. The authors hope that these articles will serve as a valuable reference for researchers in the field of Gen4 nuclear energy systems and small reactors. [Extracted from the article]