Molten Salt Reactor Experiment Research Articles

Surface Reaction-Diffusion-Coupled Simulation of Ni–Fe–Cr Alloy under FLiNaK Molten Salt

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.

Extended development of the Monte Carlo code MCNP for effective delayed neutron fraction calculation in molten salt reactors

In this study the Monte Carlo code MCNP is extended to account for the fuel circulation effect in molten salt reactors. An algorithm is created to track delayed neutron precursors and the effective delayed neutron fraction is calculated using predicted fuel velocity fields from the computational fluid dynamics solver. Delayed neutron precursors are monitored throughout various sections of a closed fuel loop in a molten salt reactor. The Moten Salt Fast Reactor is chosen as the benchmark for the extended code, with the calculated multiplication factor and reactivity feedback coefficients agreeing well with referenced results. Dependence of the effective delayed neutron fraction on the fuel flow rate is studied. Validation of the extended code is performed using the Molten Salt Reactor Experiment, showing a good agreement between the calculated delayed neutron fraction loss and the experimental data. These results demonstrate that the extended MCNP code can be reliably used for delayed neutron fraction calculations in both fast and thermal spectrum molten salt reactors.

Implementation of a Drift Flux Model into SAM with Development of a Verification and Validation Test Suite for Modeling of Noncondensable Gas Mixtures

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

MSR Transient Simulation and MSRE Transient Benchmark with SAM and SPECTRA

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

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.

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.

Xenon behavior modeling for molten salt reactors by using multiple transport mechanisms

Molten Salt Reactor (MSR) is a liquid fueled reactor, in which the sparingly soluble fission gases including xenon continuously circulate with the fuel salt in the primary loop, and meanwhile are removed by helium bubbling. This unique operation feature introduces multiple mechanisms and in turn complicates the behavior of xenon, necessitating being accurately modeled. In this study, a migration model of 135Xe and 135mXe in the primary loop for MSR is established based on the equations that describe the nuclide concentration balance and mass transfer among the helium bubbles, fuel salt and graphite. The developed model is validated with the experimental data of Molten Salt Reactor Experiment (MSRE), and presents a better agreement compared with the previous xenon migration model. Sensitivity analysis of some key parameters including injected void fraction, bubble size, mass transfer coefficient, temperature and pressure, on the xenon poison under steady state are then analyzed. The results demonstrate that the injected void fraction, mass transfer coefficient and core operation pressure would impose a significant influence on the xenon poison, which should be paid particular attention in the core design and safety analysis.

Hydrogen and deuterium permeation in Hastelloy N

Hastelloy N was chosen as the fluoride salt-contacting structural material for the Molten Salt Reactor Experiment due to its excellent compatibility with the fuel salt FLiBe. FLiBe is currently investigated for several advanced fusion and fission reactor concepts where tritium generation in the FLiBe is anticipated. Knowledge of hydrogen transport properties through Hastelloy N is important to understand how tritium would permeate through this material and result in an unintentional release. In this study, the hydrogen and deuterium permeability, diffusivity, and solubility were measured from 500 to 700 °C at a primary-side pressure of 10 kPa in a well-characterized sample of Hastelloy N. The prepared polycrystalline Hastelloy N had C and O impurities present on the surface. These impurities were investigated using Auger Emission Spectroscopy and Ar depth profiling. The adventitious C was removed upon the first Ar sputter cycle whereas O persisted deeper into the sample. For permeation experiments, applied deuterium pressures ranged from 13 Pa to 100 kPa and deuterium transport remained in the diffusion-limited regime (J ∝ P0.5) throughout the pressure range examined. Two methods are employed to measure the effective hydrogen and deuterium diffusivity: rise and decline. The decline method produced improved statistical model fits for calculating the effective diffusion coefficient compared to the rise method. The resultant transport properties compared well to published values for other nickel alloys.

Application of Modelica/TRANSFORM to system modeling of the molten salt reactor experiment

The molten salt reactor (MSR), being a liquid-fueled reactor, exhibits a strong coupling between neutronics and thermal hydraulics originating from the simultaneous utilization of the fuel as the coolant. Analysis of the dynamic behavior of MSRs thus necessitates the use of coupled multiphysics tools. This work leverages the multiphysics capabilities of the Modelica-based, open-source component library Transient Simulation Framework of Reconfigurable Modules (TRANSFORM) and demonstrates its ability to simulate the behavior of an MSR with the focus on modeling the Molten Salt Reactor Experiment (MSRE). The model employs 1D thermal-hydraulic components with a coarse radial and axial discretization of the active core region approximating the design of the MSRE. The power generation within each node is determined by neutron point kinetics which are modified to account for the effects of fuel circulation. Using the developed model, neutronic and thermal-hydraulic behavior under steady-state conditions is characterized and compared to experimental data and alternative computational codes. Some discrepancies are found but most of them are explained by the modeling choices and uncertainties in input data. Performance of the model under transient conditions is then investigated by simulating experiments conducted on the MSRE; the results are again benchmarked against reference data. Overall, good agreement is attained.

A Consistent One-Dimensional Multigroup Diffusion Model for Molten Salt Reactor Neutronics Calculations

Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis.

Analysis of dynamic responses of circulating fuel reactors using in-house 3D space-time kinetics code ARCH

Molten Salt Reactors (MSRs) are characterized by power generation in circulating liquid fuel salt rather than in static fuel pins of traditional nuclear reactors like in water cooled reactors. The fuel flow in MSRs drift the delayed neutron precursors (DNP) and results in lower delayed neutron fraction. Thus, MSRs show added complexity in dynamic analysis due to inherently coupled neutronics and thermal hydraulics. With regard to the reactor safety analysis, accurate predictions of DNPs distribution and its effect on reactor transients turn out to be an interesting subject of research. Due to these unusual features of MSRs, the traditional safety analysis codes are not applicable in dynamic analysis of reactors with circulating fuel. Challenges in accurate modelling and analysis of dynamic responses of MSRs have been described in present work using new capabilities in 3D space-time kinetics code ARCH. The numerical results of the implemented methodology presented in this study are compared with data available from Molten Salt Reactor Experiment (MSRE), a reactor operated at ORNL.

Application of MELCOR for Simulating Molten Salt Reactor Accident Source Terms

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

NERTHUS thermal spectrum molten salt reactor neutronics and dynamic model

There are several Molten Salt Reactor (MSR) designs which are currently in development by various companies and institutions. However, many researchers, developers, and communicators in the nuclear industry are unfamiliar with the unique qualities and behavior of MSRs. To help remedy this issue, the NERTHUS combined neutronic and dynamic models are intended to be a tool for simulating generic MSR behavior under a variety of scenarios. The design of the models was inspired by that of ThorCon’s TMSR-500 and utilizes many features and the same topology as the Molten Salt Reactor Experiment (MSRE). NERTHUS’s neutronics model simulates a four year depletion cycle with refueling for any user-defined fuel salt, and includes the ability to perform feedback and control rod worth calculations. The point kinetics parameters at any stage of the burn-up simulation can be used to model various transients using the dynamic model. The NERTHUS dynamic model is a full y integrated model of a MSR power plant up to the balance of plant. It features two salt-to-salt heat exchangers, a once-through steam generator, it can track decay heat and fission product poisons, and it utilizes a modular ‘plug and play’ design. The dynamic model is lightweight and can simulate nominal and off-nominal transients relating to heat transfer through the system with simulation speeds of one minute simulated per few seconds actual. This paper details results from a FLiBe fuel salt [LiF-BeF2-UF4 (72-16-12 mole%)], and a “ThorCon” fuel salt [NaF-BeF2-ThF4-UF4 (76-12-9.5-2.5 mole%)]. The NERTHUS neutronics and dynamics models were made to be an educative tool to help researchers and students investigate and learn about MSR behavior. As such, it is freely available on GitHub.

Three-Dimensional Surrogate Model Based on Back-Propagation Neural Network for Key Neutronics Parameters Prediction in Molten Salt Reactor

The simulation and analysis of neutronics parameters in Molten Salt Reactors (MSRs) is fundamental for the design of the reactor core. However, high-fidelity neutron transport calculations of the MSR are time-consuming and require significant computational resources. Artificial neural networks (ANNs) have been used in various industries, and in recent years are increasingly introduced in the nuclear industry. Back-Propagation neural network (BPNN) is one type of ANN. A surrogate model based on BP neural network is developed to quickly predict two key neutronics parameters in reactor core: the effective multiplication factor (keff) and the three-dimensional channel-by-channel neutron flux distribution. The dataset samples are generated by modeling and simulating different operation states of the Molten Salt Reactor Experiment (MSRE) using the Monte Carlo code. Hyper-parameters optimization is performed to obtain the optimal surrogate model. The numerical results on the test dataset show good agreement between the surrogate model and the Monte Carlo code. Additionally, the surrogate model significantly reduces computation time compared to the Monte Carlo code and greatly enhances efficiency. The feasibility and advantages of the proposed surrogate model is demonstrated, which has important significance for real-time prediction and design optimization of the reactor core.

Microstructural characterization of the CGB graphite grade from the molten salt reactor experiment

In the 1960s, the Molten Salt Reactor Experiment demonstrated the feasibility of molten salt reactors for civil applications. The Molten Salt Reactor Experiment used a nuclear graphite grade known as CGB in the core as the fast neutron moderator. For application in this purpose, CGB graphite underwent additional impregnation steps that are not typically used in conventional nuclear graphites. These additional impregnations significantly lowered the permeability and reduced the size of pore openings in CGB graphite, for the purpose of inhibiting molten salt ingression into the graphite. Although several publications report the mechanical properties and irradiation response of CGB graphite, little information has been published about the microstructure or sealant used for this graphite grade. Here the results of multiple advanced microscopy techniques are presented with the objective to investigate the unique microstructure and sealing technology of legacy material from the Molten Salt Reactor Experiment so it can be potentially readopted for modern reactors. The microstructural characterization shows that some pores were fully sealed by the impregnation whereas other regions contained foam-like materials that partially filled the void content. This manuscript also investigates and documents the nature and possible source of the sealant material. The pore sealing methods used for CGB are discussed and compared with other common technologies used to reduce the impregnation of molten salts into graphite components. Knowledge of CGB graphite's microstructure can enable development of new pore infiltration technologies to improve in operando behavior advancing the next generation of molten salt reactors.

Horizontal Split-Table Conceptual Design to Support Advanced Reactor Validation Needs

To improve the nuclear data testing and validation of advanced reactors, such as pebble-bed high-temperature gas-cooled reactors, molten salt reactors, and heat pipe microreactors, a conceptual design for a novel critical assembly using a horizontal split table (HST) was developed jointly by Oak Ridge National Laboratory and Lawrence Livermore National Laboratory. The mechanical design is led by Lawrence Livermore National Laboratory, whereas the neutronics considerations are led by Oak Ridge National Laboratory. The characteristics of the designed HST and the benefits of performing such an experiment to the community are included. As a proof of concept, a proposed critical experiment using tristructural isotropic fuel particles and a graphite moderator/reflector is described, mimicking a pebble-bed-type advanced reactor based on the HTR-10. A critical configuration corresponding to a footprint of about 4.5 m2 was determined with SCALE/KENO-VI to fit the planned dimensions of the HST. The similarity of the pebble-bed design and the HTR-10 reactor application was assessed using SCALE/TSUNAMI, and a similarity coefficient, ck, of 0.9982 was obtained, proving that the concept will be useful for cold-critical validation and for nuclear data validation and assimilation of pebble-bed-type advanced reactors. In the proposed design, the materials with the highest keff sensitivity are graphite and uranium, which demonstrates that particular care must be given to carbon-related cross-section data. A cross-section library study was performed to test the influence of the different recent releases of the ENDF/B cross-section library on the concept’s keff. The effect of mechanical uncertainties between the fixed and moving tables was also assessed by calculating the reactivity change caused by vertical and horizontal gaps, as well as angular and torsion offsets between the two sides of the HST concept. As a last analysis step, the performed nuclear data assimilation of the hypothetical experiment showed that uncertainties can be reduced by several hundred pcm. The same analysis process is currently being used to create a molten salt advanced reactor–type HST concept based on the Molten Salt Reactor Experiment.

High-fidelity neutronics simulation of channel-type molten salt reactors

As one of the generation IV reactors, molten salt reactors (MSRs) have attracted much attention in recent years. Generally, molten salt acting as both fuel and coolant circulates through graphite channels and external primary loop in a liquid-fueled MSR with a channel-type core. To accurately simulate the neutronics of channel-type MSRs, a GPU paralleled high-fidelity neutron transport code ThorMOC has been improved by implementing a 3D rasterized coarse mesh finite difference (RCMFD) method and a quasi-2D delayed neutron precursor (DNP) drift model. A Molten Salt Reactor Experiment (MSRE) model and a small Thorium-based MSR (sTMSR) model with stationary and flowing fuel are used for the verification of the improved ThorMOC. The reference results of the two stationary reactor models are obtained from Monte Carlo code OpenMC, which indicates that ThorMOC performs a high accuracy for stationary MSR simulations. The reactivity loss caused by fuel flowing in the steady state MSRE is calculated by ThorMOC and agrees well with the measured value from the MSRE. Moreover, the influences of geometry simplifications of fuel channels and downcomer on the reactivity loss are evaluated by comparing their results with those of the high-fidelity model. The reactivity loss in the sTMSR core is computed, and the effects of fuel velocity in core channels and residence time outside the core on the reactivity loss are also analyzed. The whole-core transport calculations of the 3D MSRE and sTMSR core models with 8 energy groups cost several hundred seconds.

An Analysis of Transport Effects in Steady-State Simulation of the Molten Salt Reactor Experiment

We use the deterministic neutron transport code QuasiMolto to simulate steady-state operation of the Molten Salt Reactor Experiment (MSRE). Comparisons are made to similar results from the MOST benchmark, the MOOSE-based code Moltres, and the design calculations for the MSRE. In the course of these comparisons, we calculate a value of 0.1799 for the graphite-to-fuel power density ratio, which differs significantly from that seen in other works. We also find uniform graphite heating inadequate to reproduce the characteristic graphite temperature distribution of the MSRE. Leveraging the multilevel projective methodology of QuasiMolto, the influence of transport effects on the modeled problem is found to produce average and maximum group flux variations of 2% to 5% and 30%, respectively, with a 12% variation in the reactivity loss due to delayed neutron precursor drift.

Steady-state radiochemical transport model of the molten salt reactor experiment

Steady-state radiochemical transport model of the molten salt reactor experiment

Review of Redox Potential Control Options for Molten Salt Reactors

Redox potential control is a fundamental requirement to minimize corrosion of metals in neutron irradiated molten salts, including fission reactor fuel, fission reactor coolant, and fusion reactor blanket. In this presentation, the focus will be placed upon metallic redox buffers—including beryllium, zirconium, and lithium. The study of beryllium for this function spans several decades and includes use in the Molten Salt Reactor Experiment (MSRE) in addition to proposed use in a fusion blanket. A fascinating aspect of this subject is the phase behavior of the metallic redox buffer. Evidence has been published that some of the metals—including beryllium and lithium actually dissolve into their host salts in the zero oxidation state. This provides the capability for a fast, homogenous reaction sans mass transfer limits. But redox buffering has also been reported from the reaction of molten salt with zirconium metal rods. It is proposed that these reactions can be categorized as either short range (homogenous) or long range (electrochemical cell) electronically mediated reactions. Data on metal solubility, electrochemical measurements, and corrosion control will be presented from the MSRE to very recent work in support of the Versatile Test Reactor. Open questions will be posed with the intention of encouraging the audience to propose new theories and experiments to test those theories.