Liquid-fueled Molten Salt Reactor Research Articles

Development of fuel depletion code for molten salt reactor with very deep burnup

A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.

A Framework for Multi-Physics Modeling, Design Optimization and Uncertainty Quantification of Fast-Spectrum Liquid-Fueled Molten-Salt Reactors

The analysis of liquid-fueled molten-salt reactors (LFMSRs) during steady state, operational transients and accident scenarios requires addressing unique reactor multi-physics challenges with coupling between thermal hydraulics, neutronics, inventory control and species distribution phenomena. This work utilizes the General Nuclear Field Operation and Manipulation (GeN-Foam) code to perform coupled thermal-hydraulics and neutronics calculations of an LFMSR design. A framework is proposed as part of this study to perform modeling, design optimization, and uncertainty quantification. The framework aims to establish a protocol for the studies and analyses of LFMSR which can later be expanded to other advanced reactor concepts too. The Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) statistical analysis tool was successfully coupled with GeN-Foam to perform uncertainty quantification studies. The uncertainties were propagated through the input design parameters, and the output uncertainties were characterized using statistical analysis and Spearman rank correlation coefficients. Three analyses are performed (namely, scalar, functional, and three-dimensional analyses) to understand the impact of input uncertainty propagation on temperature and velocity predictions. Preliminary three-dimensional reactor analysis showed that the thermal expansion coefficient, heat transfer coefficient, and specific heat of the fuel salt are the crucial input parameters that influence the temperature and velocity predictions inside the LFMSR system.

Gamma-ray spectra of post-irradiated uranium salt for total mass accounting with sodium-22 tracer

In order to meet the safeguard requirements for Nuclear Material Control and Accountability (MC&A) in liquid-fueled molten salt reactors, fuel salt transportation, and fuel cycle processes, it is important to develop effective methods for measuring and monitoring the mass of molten fuel salt or coolant salt. This paper seeks to validate the radioactive tracer dilution (RTD) method used for determining the mass of irradiated uranium-bearing molten salt. Small scale testing was conducted for this method by irradiating 6.065 g of MgCl–KCl–UCl3 salt mixture with 1.21 mg of 235U at The Ohio State University, resulting in a neutron fluence of 4.0 × 1016 n/cm2. Gamma spectra were acquired over the course of a 42-day period starting in March 2023 to assess fission products and activity levels of the 22Na tracer that was added before irradiation. No additional interference peaks were observed to overlap with the 1274.54 keV energy peak from 22Na, aside from the known 154Eu peaks, which can be corrected based on prior studies. By placing a thin piece of lead between the source and detector, deadtime was effectively reduced while still allowing observation of the higher-energy peaks, most notably around the 154Eu peak at 1004.76 keV used for interference correction. Overall, the uranium salt irradiation techniques applied in this experiment demonstrate the potential of the RTD method to effectively measure small irradiated samples for large scale molten fuel salt applications.

Development and validation of machine learning-based transient identification models in a liquid-fueled molten salt reactor system

Safety is the most important aspect of nuclear power plants. Rapid identification and effective prevention of accidents in nuclear reactor system is a significant method to enhance the safety of the current fleet of reactors. Machine learning (ML) has been introduced in engineering applications of nuclear power plants and is becoming increasingly practical and powerful in recent years. Consequently, ML can also benefit rapid transient identification in nuclear power plants. The feasibility of ML-based identification models to identify transient events in liquid-fueled Molten Salt Reactor (MSR) is presented. Four transient identification models based on recurrent neural network (RNN), support vector machine (SVM), decision tree (DT) and k-nearest neighbor (KNN) were developed and validated. RELAP5-TMSR code was used to generate datasets including eleven operation conditions, and these datasets were used to train, optimize, and validate the identification models. Four metrics including accuracy, precision, recall and F1 score were utilized to evaluate all four identification models. Moreover, the robustness of the models under noise was tested. The four ML-based models were successfully applied to transient identification of liquid-fueled MSRs. The KNN-based model has the best performance and achieves high test scores under noise. In the future, these proposed intelligent identification models will have good potential and prospects in supporting the operation of nuclear power plants.

Improvements in the decay heat model of the RELAP5 for liquid-fueled molten-salt reactor

The liquid-fueled Molten-Salt Reactor (MSR) is one of the Generation-IV nuclear power systems. One of the significant characteristics of the liquid-fueled MSR is that the decay heat is released both in the reactor core and external primary loop due to the circulation of fuel salt in the whole primary loop. However, the existing zero-dimension decay heat model of the RELAP5 cannot describe the spatial distribution of the decay heat in the primary loop of a liquid-fueled MSR. In order to accurately model and simulate the special decay heat phenomenon of a liquid-fueled MSR, the RELAP5 code is improved with a one-dimensional decay heat model. The proposed model is based on the coupling of multi-node depletion and nuclides transport calculation, and a 2nd-order splitting method is adopted for efficient numerical solution. Several scenarios are modeled and simulated for two different power-level liquid-fueled MSR systems with modified RELAP5. The numerical results show that the one-dimensional decay heat model can accurately describe the time evolution and spatial distribution of the decay heat in the primary loop of the liquid-fueled MSR, which indicates that the improved RELAP5 with proposed model is more suitable for liquid-fueled MSRs.

A Comprehensive Approach to Molten Salt Reactor Source Term Generation and Shielding Analyses

The research presented herein outlines a comprehensive process for characterizing the major radiological source terms necessary for radiation protection and licensing activities that one would expect in a liquid-fueled molten salt reactor. This process leverages organic simulation tools in the SCALE modeling and simulation code suite to provide an “off-the-shelf” solution for shielding assessments of this reactor type. Ultimately, this source development process is applied to a representative molten salt reactor system to assess the impact of ex-core source terms on shielding in varying operating conditions. The results of the analysis determined that while the prompt core source is the major dose contributor outside the radiological shielding, specific ex-core features, such as the primary salt loop components and configuration, can have an appreciable dose impact, and thus must be accounted for.

Improvement of a deterministic fuel management code using artificial neural network for liquid-fueled molten salt reactor

The fuel management code of liquid-fueled molten salt reactor (MSR) relies on the coupling of the neutronics code with depletion code. The deterministic method can improve the efficiency of neutronics calculation. However, updating the microscopic cross-sections in depletion calculation is still time-consuming using Monte Carlo code. A surrogate model based on artificial neural network (ANN) is developed to update the cross-sections online. Then a new fuel management code named ThorNEMFM-ANN was developed by integrating the surrogate model into the deterministic fuel management code. The surrogate model and ThorNEMFM-ANN code were validated with the molten salt breeder reactor (MSBR) and molten salt fast reactor (MSFR) benchmarks. Furthermore, the ThorNEMFM-ANN was applied to the fuel cycle analysis of a molten chloride salt fast reactor (MCSFR) for the demonstration of the applicability. The numerical results indicate that the surrogate model is feasible and effective in the fuel cycle simulation of liquid-fueled MSR.

Machine learning approaches to equilibrium burnup analysis for Molten Salt Reactor

During the operation of a liquid-fueled molten salt reactor (MSR), the deep depletion to an equilibrium state can be attained with online refueling and reprocessing techniques. However, the equilibrium burnup simulation of an MSR using current depletion codes requires considerable computational cost and time, and even fails when the equilibrium conditions are unsatisfied. To efficiently evaluate the specific characteristics of the equilibrium burnup in MSRs, a fast-filtering model for equilibrium state and a fast prediction model for equilibrium neutronic properties are developed based on the machine learning (ML) technique. A database of 4860 cases including the neutronic information of equilibrium and non-equilibrium of a fuel assembly with variable geometric parameters in an MSR is first generated, and then the neutronic properties of 232Th-233U, 232Th-EU (enriched uranium) and 232Th-Pu fuel cycles are evaluated. Based on the MSR database, fifteen different ML classification algorithms are implemented to establish the fast-filtering model for equilibrium burnup state. Moreover, a model for the fast prediction of equilibrium neutronic parameters of neutron flux (Φ) and breeding ratio (BR) is developed with fifteen ML regression algorithms. Considering the various performance metrics for measuring the predicted performances of ML models in the classification and regression, the LightGBM (LGBM) model is the most favorable for filtering the burnup state and predicting the neutronic parameters in MSRs. With the test data, the LGBM model can obtain the accuracy of about 0.98 for classification and the relative error of about 1.10% for regression.

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.

The impact of nuclear data uncertainty on identifying plutonium diversion in liquid-fueled Molten Salt Reactors

Molten salt-fueled reactors (MSRs) have gained popularity recently, but there are challenges in accounting for nuclear material due to the inability to count discrete items (e.g., fuel assemblies). The focus of this work is to calculate what changes would occur due to diversion of fissile material, and to evaluate the effect of nuclear data uncertainties on these quantities. The impact of other sources of uncertainty (e.g., measurement error) is out of the scope. A thermal-spectrum MSR was modeled using Serpent 2 and SCALE 6.3.b12 (beta), including plutonium diversion scenarios. The feed and removal capabilities were recently added into the depletion schemes of these codes. Protracted and abrupt diversion scenarios were developed for the removal of 1 and 10 significant quantities (SQs) of plutonium. The SCALE module Sampler was used to perform nuclear data uncertainty propagation along the depletion interval from neutron cross-sections, fission product yield and decay data. Results showed that along with plutonium species, other actinides and fission products presented significant changes between the reference scenario (i.e., no diversion) and the 10 SQ plutonium protracted diversion scenario. Considering their propagated nuclear data uncertainty from Sampler simulations, some actinides such Am and Cm showed changes higher than their nuclear data uncertainty, which goes from 2.23% for 241Am up to 4.88% for 242mAm. Some fission products also presented notable changes, Sr and Y isotopes are examples with positive changes, while Cd, Eu and Sm isotopes showed more prominent changes on the negative side, with uncertainties in the range of 1.89% for 149Sm to 3.31% for 113Cd. The analysis extended to nuclides regularly moved to other modeled inventories, showing that 105,106Ru can be good candidates given their low uncertainty that stands within the 1% range. Many other isotopes with considerable changes presented high uncertainty values and methods to improve their nuclear data uncertainty estimation will be sought in future work.

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.

Simulation of the Axial-Flow Centrifugal Bubble Separator for Liquid-Fueled Molten Salt Reactors Using Eulerian Two-Fluid Model

The axial-flow centrifugal bubble separator designed for the gaseous fission product removal system in liquid-fueled molten salt reactors is simulated using the Eulerian two-fluid model coupled with the Adaptive Multiple Size Group method to account for the significant coalescence and breakup in the bubble separator. The behavior of the gas core in the bubble separator is mimicked by the symmetric interfacial area concentration model. The separator efficiency, local velocity, and pressure profiles at various conditions are compared with experimental data. Good agreement is found between the experiment and the simulation for the separator efficiency. With the coalescence and breakup being accounted for, the effect of the inlet void fraction on the separator efficiency is correctly captured. For the local pressure and velocity profiles, the agreement is only quantitative due to the simplifications on the geometry and potential limitations of the current computational fluid dynamics models. As good agreement is found for the separator efficiency, the sensitivity study is performed for various operational and design parameters with further simplified two-dimensional axisymmetric simulation.

Transmutation and Breeding Performance Analysis of Molten Chloride Salt Fast Reactor Using a Fuel Management Code with Nodal Expansion Method

The transmutation of transuranic (TRU) elements produced by pressurized water reactors (PWRs) can effectively reduce their radioactive hazards. The molten chloride salt fast reactor (MCSFR) is a type of liquid-fueled molten salt reactor (MSR) using fuel in the form of molten chloride salts. The MCSFR utilizing a fast neutron spectrum and high actinide fraction is considered to be a potential reactor type for TRU transmutation. An online refueling and reprocessing scenario is the unique feature of liquid-fueled MSRs. On account of this characteristic, a new fuel management code named ThorNEMFM with a nodal expansion method (NEM) was developed and validated with the molten salt breeder reactor (MSBR) and the molten salt fast reactor (MSFR) benchmarks. Then, the transmutation and breeding performances of the MCSFR were simulated and analyzed with the ThorNEMFM code. The MCSFR adopts TRU elements as initial fissile loads and online feeding fissile materials. The results show that the transmutation ratio of TRU elements in the MCSFR can reach 50%, and the breeding ratio can reach 1.359. Moreover, the MCSFR has low radiotoxicity due to lower buildup of fission products (FPs).

Informing Performance Metrics of Advanced I&C Systems for Liquid Fueled Fast Molten Salt Reactors

This work establishes a generic multiphysics tool for liquid-fueled molten salt reactors (LFMSRs) to select key installation locations and specify the expected operating temperature range for the development of advanced instrumentation and control systems, particularly distributed temperature sensors using fiber optics. A commercial computation fluid dynamics package (STAR-CCM+) is used to formulate a neutronics and thermal-hydraulic coupled solver, showing good agreement with a recent benchmark problem developed for evaluating the coupling methodology of neutronics and thermal hydraulics. The multiphysics model is then applied to the reference molten chloride salt fast reactor (MCFR) design under development by TerraPower based on publicly available information. The available two-dimensional axisymmetric model for the reactor core is used for coupling calculations, and system component details are leveraged using the lumped method to complete the energy balance. The dynamic responses of the MCFR model are investigated during operational transients, such as unprotected loss-of-flow and uniform perturbation scenarios. Maximum temperature and local temperature distributions are characterized during unprotected loss of flow and unprotected loss of heat sink events. The thermal responses of the fuel salt and core components are analyzed from induced perturbation of the system parameters, such as the flow rate and the heat sink capacity. The results motivate the use of continuous monitoring of the temperature variation in real time along the reflector region with the use of fiber optics to validate the multiphysics code to support a reactor’s licensing basis, as well as to support the structural longevity and improve safety in LFMSRs.

Flow field effect of delayed neutron precursors in liquid-fueled molten salt reactors

Flow field effect of delayed neutron precursors in liquid-fueled molten salt reactors

Analysis of load-following operation characteristics of liquid fuel molten salt reactor

As the IEA reports indicated that nuclear power owns a growing market share, which means that nuclear power plants (NPPs) must be able to respond quickly to grid fluctuations. A good load-following capability is the basic requirement for the flexibility and safety of reactor operation. In order to study the load-following capacity of the molten salt reactor, this study took the liquid fuel molten salt reactor (MSR) as the research object. The system model and power control model were established with RELAP5/MOD4.0 program, which has been implanted with relevant physical properties of molten salt and calculation formula. The response characteristics of the control system under step load change, linear load change and typical daily load-following operation were simulated and analyzed. The results show that MSR has a good load-following capacity under the designed control logic, the core power can respond to the load variation quickly, the power overshoot and temperature overshoot are small, and the operating parameters of the reactor are under a reasonable operating range. The simulation analysis provides the reference for subsequent safety analysis and power control of molten salt reactor.

Improvement of the delayed neutron precursor transport model in RELAP5 for liquid-fueled molten salt reactor

The Liquid-Fueled Molten Salt Reactor (LF-MSR) with high-temperature molten salt is one of the Generation IV nuclear power systems. One of the significant differences between the LF-MSR and the Pressurized Water Reactor (PWR) is that the delayed neutron precursors (DNP) circulate in the primary loop of the LF-MSR and leads to different neutron kinetics between LF-MSR and PWR. Consequently, a system code for PWR is not applicable to LF-MSR, and the safety analysis code RELAP5-TMSR is modified by adding a one-dimensional DNP transport model to improve its applicability to LF-MSR. A second-order Godunov method is adopted in this 1-D DNP model. Then the improved code is tested with experimental benchmarks from Molten Salt Reactor Experiment (MSRE) designed and operated by Oak Ridge National Laboratory (ORNL) and the results show good consistency. Moreover, the modified RELAP5-TMSR code is applied to analyzing transient characteristics of the single-fluid Molten Salt Breeding Reactor (MSBR) so that the applicability of the modified RELAP5-TMSR code is further verified. It is concluded that the one-dimensional DNP transport model can describe effectively the movement of the DNP and the improved RELAP5-TMSR code is suitable for modeling and simulation of reactor transient responses in LF-MSR.

Dynamic mass accountancy modeling of a molten salt reactor using equilibrium thermodynamics

A mechanistic-based mass accountancy model in the context of liquid-fueled molten salt reactors was implemented in the dynamic systems modeling software library TRANSFORM by way of coupling with the equilibrium thermodynamics code Thermochimica. Liquid-fueled molten salt reactors present new challenges for mass accountancy because of the dissolution of fuel and evolved fission products, which may be soluble in the salt, off-gas, or precipitate. Two cases of mass loss from the molten salt were addressed: off-gassing and precipitation. The software implementation was tested through a series of increasingly complex demonstration problems, culminating in a model of the primary fuel and primary coolant loops of the molten salt demonstration reactor. Analysis shows that negligible mass was lost from the salt under normal operating conditions, but an overheating event caused by partial loss of fuel loop cooling resulted in release of measurable amounts of uranium (among other elements) via off-gassing. The tools developed here are primarily aimed at capability development but are readily available for use in further modeling of molten salt reactor concepts. These tools have not yet been validated, and future experimental work to perform this validation is recommended.

Effect of process parameters on the recovery of thorium tetrafluoride prepared by hydrofluorination of thorium oxide, and their optimization

Liquid fueled molten salt reactors (MSRs) have seen renewed interest because of their inherent safety features, higher thermal efficiency and potential for efficient thorium utilisation for power generation. Thorium fluoride is one of the salts used in liquid fueled MSRs employing Th–U cycle. In the present study, ThF4 was prepared by hydro-fluorination of ThO2 using anhydrous HF gas. Process parameters viz. bed depth, hydrofluorination time and hydrofluorination temperature, were optimized for the preparation of ThF4 in a static bed reactor setup. The products were characterized with X-Ray diffraction and experimental conditions for complete conversion to ThF4 were established which also corroborated with the yield values. Hydrofluorination of ThO2 at 450 °C for half an hour at a bed depth of 6 mm gave the best result, with a yield of about 99.36% ThF4. No unconverted oxide or any other impurity was observed. Rietveld refinement was performed on the XRD data of this ThF4, and Chi2 value of 3.54 indicated good agreement between observed and calculated profiles.

Rasterized coarse mesh finite difference acceleration on method of characteristics for a reactor core with a generalized boundary

The method of characteristics (MOC) is widely employed in neutron transport solvers to obtain accurate solution of a direct transport calculation for a whole reactor core model. Coarse mesh finite difference (CMFD) method is often used to speedup the convergence of MOC solution especially for a large geometry. In this work, a rasterized CMFD method is proposed to improve the convergence rate of the high-order neutron transport calculation for reactor cores with generalized boundaries. The rasterized CMFD method is implemented in a 2D neutron transport MOC code ThorMOC which has been developed to perform the high-fidelity criticality calculation for channel type molten salt reactors (MSRs). The rasterized CMFD method is validated by two benchmarks, a C5G7 hexagonal benchmark and a LFMSR (Liquid-Fueled Molten Salt Reactor) benchmark. The computational result obtained from ThorMOC accelerated by the rasterized CMFD method shows a good agreement with that from the Monte Carlo code OpenMC for the two benchmarks. The rasterized CMFD method can reduce the MOC stand-alone solution time of the LFMSR benchmark by more than 100 times.