Salt Reactor Research Articles

Comparative Irradiated Dimensional Change Strain Analyses of Two Types of Graphite Components in a Thorium Molten Salt Reactor

Comparative Irradiated Dimensional Change Strain Analyses of Two Types of Graphite Components in a Thorium Molten Salt Reactor

Thermodynamic modeling of CsF with LiF-NaF-KF for molten fluoride-fueled reactors

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 (ΔmixH) 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.

Control and surveillance of redox potential for 233Pa dissolution in 2LiF-BeF2 molten salt.

233Pa, the precursor nuclide of 233U in the thorium-uranium conversion is prone to reductive deposition in 2LiF-BeF2 (66 : 34 mol%, FLiBe) molten salt. We explored the adjustment and control of the redox potential of the FLiBe melt to avoid the 233Pa reduction deposition. The experimental data indicated that the deposited 233Pa can be completely dissolved and reentered into the molten salt with the addition of oxidant NiF2, and the distribution and behaviour of uranium, thorium, neptunium, and most fission products did not have any significant change in the NiF2-oxidised FLiBe molten salt, showing the feasibility of this manner to make 233Pa exist stably in the melt. The effects of NiF2-addition on the behaviour of the fission products 95Nb and 131I in the FLiBe molten salt were also investigated. It was found that 131I could be used as a redox indicator to monitor the redox potential of the oxidation-enhanced FLiBe molten salt. All the information drawn from this study could provide significant support for the control and surveillance of the redox potential of the FLiBe molten salt in the upcoming thorium molten salt reactor (TMSR).

A new approach to predict pump transient phenomena in Molten salt reactor Experiment (MSRE) by missing data identification and regeneration

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.

Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor

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.

A theory and analysis of the impact of gas in the dynamical behavior of the molten salt research reactor leading to the computation of the "gas coefficients of reactivity"

Background Molten salt reactors, and other types of circulating, liquid fueled, nuclear reactors contain a certain amount of gas entrained in their liquid nuclear fuel. This gas induces an effect on the nuclear and dynamical behavior of the reactor as a whole. Gas voids respond to variation in temperature and pressure differently than liquids. When the gas voids in the reactor working fluid expand, the nuclear fuel is pushed from the core. Likewise, when the gas voids contract, more nuclear fuel enters into the core. Methods This paper examines the interplay of gas void fraction and reactivity in a molten salt reactor, and attempts to elucidate the dynamical response of the void fraction and the reactivity of the system to perturbation in system temperature, pressure, and gas quantity. A theory is presented that aims at describing the relationship between reactivity and gas behavior. This theory is then applied to the Molten Salt Research Reactor (MSRR) design, a facility currently under construction at Abilene Christian University campus. Results A result of this paper is the temperature and void fraction parameterized gas coefficients of reactivity for the Molten Salt Research Reactor. Conclusions The presence of voids accounts for 5-30% of the total temperature coefficient of reactivity, demonstrating their non-trivial contribution. Additionally, the study emphasizes the importance of considering gas content in MSR physics, especially in the context of pressure transients and system reactivity during pump trips. The initial system pressure, particularly in designs like the MSRR operating at sub-atmospheric pressures, is crucial due to its influence on reactivity changes during rapid pressure increases.

CAD and constructive solid geometry modeling of the Molten Salt Reactor Experiment with OpenMC

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

Material attractiveness of irradiated fuel salts from the Seaborg Compact Molten Salt Reactor

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.

Multi-objective capacity configuration optimization of a nuclear-renewable hybrid system

Integrating the energy storage and the base-load energy can be an efficient solution to cover the fluctuation of renewable energy. A nuclear-renewable hybrid energy system consisting of a small modular thorium molten salt reactor, solar photovoltaics, wind turbines, thermal energy storage and battery storage with two operation modes is proposed to meet the community load demand. The methodology combining multi-objective evolutionary algorithms with multi-criteria decision-making method is used to identify the optimal operation mode and gain the optimal configuration of the proposed nuclear-renewable hybrid energy system. In the first stage, the multi-objective evolutionary algorithm with better performance is utilized to gain Pareto solutions to minimize the levelized cost of energy while ensuring an acceptable level of deficiency of power supply probability. In the second stage, the multi-criteria decision-making method is adopted to select the best solution from the obtained Pareto frontiers. The optimization results demonstrate that the nondominated sorting genetic algorithm has the best performance compared with other algorithms and the operation mode with battery discharged first is more cost competitive. It is concluded from the results analysis that the selection of optimal mode is influenced by seasonal variation that contributes to large energy storage capacity requirement.

Electrochemical measurement of the corrosion rates of structural materials in molten NaCl–MgCl2 at 580℃

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–MgCl2 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. EuCl3-induced corrosion of Inconel 625 and stainless steel 316 (SS 316) in molten NaCl–MgCl2 was monitored by electrochemical detection of Ni2+ and Fe2+. The corrosion rates were linearly dependent on the EuCl3 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–MgCl2–EuCl3. 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.

Comparative Assessment of Molten Salt Reactor Neutronic Performance with Various U-233 Purity

Comparative Assessment of Molten Salt Reactor Neutronic Performance with Various U-233 Purity

Corrosion Behavior of Candidate Structural Materials for Molten Salt Reactors in Flowing NaCl-MgCl2

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.

Molecular dynamics simulations for the prediction of thermophysical properties of plutonium-based molten salts

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.

Highly selective removal of thorium from acidic solutions achieved on phosphonic acid functionalized core–shell magnetic nanocomposite: Understanding adsorption and binding mechanisms

Thorium has received widespread attention as potential nuclear fuel alternative for the next-generation thorium-based molten salt reactors. With the growth of nuclear energy, it is needful to exploit magnetic nanomaterials to remove thorium from aqueous solutions, owing to their superparamagnetism and large specific surface energy. Nevertheless, most of magnetic nanomaterials often suffered from low thorium selectivity and poor structural ability in acidic solutions. Herein, a novel core–shell magnetic nanocomposite Fe3O4@SiO2/P (AA-MMA-VPA) possessing phosphonic acid was developed by a controlled radical polymerization of vinylphosphonic acid (VPA) in the presence of Fe3O4 encapsulated by SiO2 (Fe3O4@SiO2). Due to the presence of the high content of phosphonic acid groups and the absence of any universal groups, Fe3O4@SiO2/P (AA-MMA-VPA) displayed exceptional selectivity for thorium in the presence of 11 co-existing cations. The separation factor STh/M value for Fe3O4@SiO2/P(AA-MMA-VPA) concerning all other competing cations surpassed 74.4, and the largest value for STh/M even amounted to 343.5. Further, Fe3O4@SiO2/P (AA-MMA-VPA) possessed a maximum adsorption capacity (qmax) of 485.4 mg g−1 at pH of 3.0, higher than that of other magnetic sorbents. By reason of its core–shell structure, the sorbent possessed an outstanding structure stability even after immersed into HNO3 solution at pH of 0.5 for 72 h. The sorbent could as well be magnetically recouped, and recycled at least six times without evident diminution in sorption amount. The sterling removal performance of the sorbent was assigned to the interaction of thorium with phosphonic acid, corroborated by FT-IR, XPS and DFT method. Thus, this study affords a new viewpoint for exploiting magnetic nanomaterials with sterling acidic resistance for the strongly selective entrapment of thorium from acidic media.

Study on the cantilever ratio optimization of high-temperature molten salt pump for molten salt reactor based on structural integrity

The high-temperature molten salt pump is the core equipment in the small modular molten salt reactor with media temperatures up to 700 °C. The cantilever ratio of the molten salt pump is usually large. Excessively large cantilever ratios cause increased deformations and rotational amplitudes at the impeller, thus affecting the operational stability of the main pump; small cantilever ratios cause heavy temperature gradients, thus affecting the structural integrity evaluation. This paper used numerical simulation methods to calculate and analyze the temperature field, stress, and structural integrity, optimized the pump shaft cantilever length of the original scheme based on structural integrity using the dichotomy method, and analyzed the rotor dynamics of the optimization results. The results of this study show that the thermal expansion load caused by the temperature difference has a significant mechanical effect on the structure; the first-order critical speed of the rotor system of the optimized schemes has been improved, and the amplitude of the unbalanced response has been significantly reduced, which not only improves the operational stability of the rotor, also contributes to the compact design of the main pump of a small modular molten salt reactor.

Heterogeneous Structure, Mechanisms of Counterion Exchange, and the Spacer Salt Effect in Complex Molten Salt Mixtures Including LaCl3.

Complex molten chloride salt mixtures of uranium, magnesium, and sodium are top candidates for promising nuclear energy technologies to produce electricity based on molten salt reactors. From a local structural perspective, LaCl3 is similar to UCl3 and hence a good proxy to study these complex salt mixtures. As fission products, lanthanide salts and their mixtures are also very important in their own right. This article describes from an experimental and theory perspective how very different the structural roles of MgCl2 and NaCl are in mixtures with LaCl3. We find that, whereas MgCl2 becomes an integral part of multivalent ionic networks, NaCl separates them. In a recent article (J. Am. Chem. Soc. 2022, 144, 21751-21762) we have called the disruptive behavior of NaCl "the spacer salt effect". Because of the heterogeneous nature of these salt mixtures, there are multiple structural motifs in the melt, each with its particular free energetics. Our work identifies and quantifies these; it also elucidates the mechanisms through which Cl- ions exchange between Mg2+-rich and La3+-rich environments.

A theoretical model of gas column breakup under a swirling flow field based on the jet theory

Maintaining the stability of the gas column/liquid film in a swirling flow field in thorium molten salt reactor (TMSR) and pressurized water reactor (PWR). The linear stability analysis in jet theory can be used for this problem. This method focuses on the maximum growth rate of the perturbation at the gas–liquid interface and determines the stability of the gas column based on it. This approach is usually applied in the cases under non-swirling flow field. In this paper, we extend this method to the study of stability problem under a swirling flow field. We classify the stability problem within different reactors into two categories. The detailed calculation method of the maximum growth rate is given separately for the two cases. Based on the calculated maximum growth rate, we experimentally verified the stability method. Considering the differences between non-swirling flow field and swirling flow field, we improve the original method and give a new prediction correlation. The prediction errors of new correlation are within ±20 %.

Flow visualization experiments of argon injection in a molten salt natural circulation loop

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.

Plutonium utilization in a small modular molten-salt reactor based on a batch fuel reprocessing scheme

Plutonium utilization in a small modular molten-salt reactor based on a batch fuel reprocessing scheme

Development of Two-Step Method for Fast Chloride MSR Neutronics

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