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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).

The 235,238UF4 was irradiated by photo-neutrons, distribution and behavior of the fission product 95Nb from irradiated 235,238UF4 in FLiBe salt were investigated by the measurement of its activity in the salt with the γ-ray spectroscopy. The experiments indicated that a part of 95Nb deposited on the surfaces of graphite and Hastelloy, as the moderator and the structural materials of molten salt reactor (MSR), respectively, and the majority of 95Nb maintained in molten salt. Addition of lithium metal made 95Nb in salt to be reduced and settled, leading to the decrease in its activity. Degree of the decrease was found to be correlated with niobium concentration. The experimental results supported the statement proposed early by ORNL, that 95Nb might be used as a redox indicator for MSR. Finally, the problem met with on-site monitoring for redox potential in MSR was pointed, and a possible protocol to resolve the problem was proposed. [ABSTRACT FROM AUTHOR]

The Molten Salt Reactor (MSR) is the most important system suggested by Generation IV for the future direction in the nuclear reactor field. For more development of the MSR reactor, the core system inside the tube is proposed by naturally circulating molten fuel salt. The nonlinear kinetic equations form a linearized function and are obtained in state-space form. Reactivity feedback and delayed neutrons are extremely important for reactor control. In this paper, a thermal-hydraulic system for the commercial computation dynamic model is proposed. Currently, there is no commercial software to simulate the natural circulation flow. The proposed method can be easily employed to detect faults and can provide a feasible overall system performance. [ABSTRACT FROM AUTHOR]

BackgroundIn molten salt reactors (MSRs), the liquid fuel salt circulates in the whole primary loop. Part of the delayed neutron precursors (DNP) decay outside the reactor core, but most of the reactor fission power releases directly in the fuel salt. Therefore, the neutronics kinetics (NK) and thermal-hydraulics (TH) are strongly coupled due to the liquid fuel in MSRs, hence traditional dynamic codes cannot be used for MSRs.PurposeThis study aims to develop and validate a three-dimensional time-spatial dynamic code named ThorCORE3D which can be used to carry out design and transient analysis of MSRs.MethodsNodal expand method was used to solve time-dependent neutron diffusion equations and a thermal-hydraulics solver for circulating liquid fuel with internal heat source was developed in ThorCORE3D. Then the code was validated using both steady state and transient experimental benchmark questions of Molten Salt Reactor Experiment (MSRE) of Oak Ridge National Laboratory (ORNL).ResultsThe results show that numerical results from ThorCORE3D are in good agreement with the experiment results and that of other codes. The unique phenomena in MSRs can be well modeled and simulated by the ThorCORE3D.ConclusionsThe ThorCORE3D dynamics code can be used reliably for steady state and transient analyses of liquid fuel MSRs.

Summary: As one of the representatives of fourth generation nuclear reactors, molten salt reactor technology has become the research focus worldwide. Higher passive safety requirements have been put forward for the nuclear reactor systems. In this paper, a conceptual design of a passive residual heat removal system for the molten salt reactor is presented and the mathematical and physical models are established. Based on FORTRAN programming language, the passive residual heat removal system analysis code is developed and validated. The code is used to analyze the safety of molten salt reactor under accident conditions, including unprotected reactivity introduction accident, unprotected overcooling, unprotected loss of flow, opening failure of the frozen valve and the startup accident of partial heat pipes. The results show that the passive residual heat removal system of molten salt reactor can discharge fuel salt decay heat effectively within 200 hours and ensure that the molten salt reactor to operate safely under all accident conditions. Above all, the system has good passive safety performance. The proposed passive residual heat removal system can satisfy the safety requirement of molten salt reactor. The feasibility of the conceptual design of the heat pipe cooled passive residual heat removal system is theoretically verified. [ABSTRACT FROM AUTHOR]

Knowledge about the microstructure of nuclear graphite is critical to the understanding of its irradiation behavior in the reactor. Using micro X‐ray diffraction (μXRD) two‐dimensional (2D) maps, the crystallite character of IG110 and reactor pebble graphite was characterized at the submillimeter range with a spatial resolution of about 5 μm. Various structures in the nuclear graphite were identified by comparing the X‐ray diffraction peak intensity, position, and full width at half maximum (FWHM) 2D maps. The two‐peak feature of the FWHM histograms seen in pebble graphite may be related to the raw coke and natural graphite used as raw materials. With these results, it can be concluded that the μXRD 2D map is an effective method to characterize the microstructure of nuclear graphite. [ABSTRACT FROM AUTHOR]

BackgroundNuclear reactors have become the main energy supply for future manned Mars exploration missions due to their high energy density, high power-mass ratio and small size. Heat pipe molten salt reactor is an innovative concept combining molten salt reactor and high temperature heat pipe technology.PurposeThis study aims to carry out preliminary neutronics design of a martian surface molten salt reactor (MS)2R core with 1 MWth and a lifetime longer than 5 a.MethodsFirst of all, the core model of (MS)2R was estabished according to the requirements. It mainly included active zone (including fuel salt and heat pipe), active zone wall, reflector, reactor vessel, etc. Then, the MCNP (Monte Carlo N particle Transport Code) and MOBAT were used to optimize the core size and reactivity control of (MS)2R.ResultsThe physical design parameters of the core are obtained: the height and diameter of the core are 90.94 cm and 88.94 cm, respectively, the uranium inventory is 146.08 kg, and the core mass is 2.09×103 kg. Reactivity control is achieved by a control drum system, in which the B4C (10B enrichment is 90%) with a thickness of 1.8 cm and a wrap angle of 120° is used as the neutron absorber.ConclusionsThe control drum arrangement can meet the critical safety requirements during the lifecycle of (MS)2R under full power (1 MWth) operation for five years, and the number of heat pipes can meet the heat transfer safety limit. This study provides a basic theoretical reference for the design of planet surface molten salt reactor.

Recent developments in high-magnetic-field fusion systems have created large incentives to develop flibe (Li2BeF4) salt fusion blankets that have four functions: (1) convert the high energy of fusion neutrons into heat for the power system, (2) convert lithium into tritium—the fusion fuel, (3) shield the magnets against radiation, and (4) cool the first wall that separates the plasma from the salt blanket. Flibe is the same coolant proposed for fluoride-salt-cooled high-temperature reactors that use clean flibe coolant and graphite-matrix coated-particle fuel. Flibe is also the coolant proposed for some molten salt reactors (MSRs) where the fuel is dissolved in the coolant. The multiple applications for flibe as a coolant create large incentives for cooperative fusion-fission programs for development of the underlying science, design tools, technology (pumps, instrumentation, salt purification, materials, tritium removal, etc.), and supply chains. Other high-temperature molten salts are being developed for alternative MSR systems and for advanced Gen-III concentrated solar power (CSP) systems. The overlapping characteristics of flibe salt with these other salt systems create significant incentives for cooperative fusion-fission-solar programs in multiple areas. We describe the fission and fusion flibe-cooled systems, what has created this synergism, what is different and the same between fission and fusion in terms of using flibe, and the common challenges. We review (1) the characteristics of flibe salts, (2) the status of the technology, (3) the options for tritium capture and control in the salt, heat exchangers, and secondary heat transfer loops, and (4) the coupling to power cycles with heat storage. The technology overlap between flibe systems and other high-temperature MSR and CSP salt systems is described. This defines where there are opportunities for cooperative programs across fission, fusion, and CSP salt programs. [ABSTRACT FROM AUTHOR]

Energy markets are changing because of (1) the addition of nondispatchable wind and solar electric generating capacity and (2) the goal of a low-carbon energy system. The large-scale addition of wind and solar photovoltaics results in low wholesale electricity prices at times of high wind and solar output and high prices at times of low wind and solar input. The goal of a low-carbon energy system requires a replacement energy production system with assured peak energy production capacity. To minimize costs, capital-intensive nuclear reactors should operate at base load. To maximize revenue (minimize sales at times of low prices and maximize sales at times of high prices), the power cycle should provide variable heat and electricity. This requires the power cycle to (1) include heat storage that enables peak heat and electricity output that may be several times base-load reactor output and (2) provide assured peak power production. Assured peak power production requires the capability to efficiently burn low-carbon fuels such as hydrogen and biofuels. Alternatively, nuclear systems with base-load reactors can be built to produce peak electricity and storable hydrogen for industry, biofuels, and other markets. All power reactors with appropriate system designs can meet these requirements. The lowest-cost technologies for heat storage, assured peak power production, and hydrogen production require high-temperature heat. This economically favors salt-cooled reactors with the average temperature of delivered heat of about 650°C versus heat delivered at lower average temperatures from other reactors such as light water reactors: 280°C, sodium-cooled reactors: 500°C, and high-temperature helium-cooled reactors: 550°C. Salt-cooled reactors include (1) Fluoride-salt-cooled High-temperature Reactors (FHRs) with solid fuel and clean salt, (2) Molten Salt Reactors (MSRs) with fuel dissolved in the salt, and (3) fusion reactors with salt blankets. Future energy markets, nuclear systems (heat storage, assured peak energy production capacity, and hydrogen production) designed for such markets and the power cycle technologies that economically favor salt reactors are described. [ABSTRACT FROM AUTHOR]

The fluoride-salt-cooled high-temperature reactor and some proposed fusion reactors use clean fluoride salts as reactor coolants that have melting points above 450°C and generate tritium. Tritium diffuses through most hot metals, thus methods to capture tritium and prevent its release to the environment are required. Molten salt reactors (MSRs) use fluoride or chloride salts with high melting points where the fuel is dissolved in the coolant. MSR systems produce volatile fission products (Xe, Kr, etc.) and some produce significant tritium. We examine the use of heat exchangers with multiple heat pipes for salt-cooled fission and fusion systems that serve four functions: (1) transfer heat from primary coolant to power cycle, secondary loop, or environment; (2) provide the safety function of a secondary loop by isolating the reactor salt coolant from the high-pressure power cycle; (3) stop heat transfer if the reactor coolant approaches its freezing point to prevent blockage of the primary loop; and (4) block tritium escape to the environment with recovery of the tritium. Each of these capabilities in some form has been demonstrated in a heat pipe system, but not all the functions have been demonstrated in a single system because there has been no need for all of these capabilities in a single system. We review the status of heat pipe technology and the limits of heat pipe technology as the starting points for decisions on the development of such heat pipe systems. [ABSTRACT FROM AUTHOR]

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In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

Nuclear reactor systems that use fuel dissolved in a liquid have the potential for enhanced safety characteristics, improved fuel-cycle outcomes, and more efficient isotope-production configurations. In these reactor systems, the fueled liquid may simultaneously undergo irradiation, physical and chemical removal processes, and fueling. The modeling and simulation of this transmutation and decay with material additions and removals is an ongoing research area. An accurate simulation tool is critical to the reactor and fuel-cycle design, reactor deployment, and source-term characterization for these advanced reactor systems. The work described herein involved implementing, testing, and applying the capability to perform reactor physics simulations within the Oak Ridge National Laboratory-developed SCALE suite for nuclear systems analyses and design, leveraging much of its pedigree in quality-assurance and reactor-analysis capabilities. The functionalities to simulate irradiation with material feeds and removals had been added in ORIGEN, and the TRITON reactor physics sequence was extended to calculate the total removed material and track external nonirradiated mixtures to estimate separate processing or waste streams. Results from these capabilities align with analytical expectations obtained from ORIGEN for simplified test cases and with expectations for a molten salt reactor application. This implementation, available with the SCALE 6.3 release, provides for a more efficient and accurate material accountability methodology, allowing for the characterization, design, and analysis of the complete isotopic material inventory of advanced liquid-fueled systems for a variety of applications. • Implementation of the capability to simulate a liquid-fueled system in the SCALE6.3/TRITON. • TRITON can accept feed materials and track and decay removed materials. • The features are demonstrated in a Molten Salt Reactor Experiment simulation. [ABSTRACT FROM AUTHOR]

• The impact of graphite moderator density on the neutronic parameters of a Molten Salt Reactor (MSR) was analysed. • Higher graphite density is associated with significantly weaker temperature coefficient of reactivity (TCR) and slightly higher conversion ratio (CR). • Fuel economy in MSR with low graphite density can be severely compromised if the U-233 production cost is high. • Despite being weaker in TCR, high graphite density is still recommended for the application in graphite-moderated MSR due to its superiority in other neutronic parameters and fuel economy. Typical thermal molten salt reactor (MSR) employs graphite as its neutron moderator, due to its chemical compatibility and low neutron absorption. It is not without issues, however, since graphite provides positive temperature feedback when the MSR is fueled with U-233. Reducing the graphite density can possibly ameliorate this issue, but at the cost of other neutronic parameters, such as fertile conversion into fissile nuclide, and fuel economy, such as fissile makeup cost. This study evaluates the influence of graphite moderator density on the neutronic performance and fuel economy of an MSR. Passive Compact Molten Salt Reactor (PCMSR) design was used as the MSR model. To evaluate the neutronic parameters, MCNP6.2 radiation transport code with ENDF/B-VII.0 neutron cross section library were used. The evaluated neutronic parameters were temperature coefficient of reactivity (TCR), kinetic parameters, and conversion ratio (CR). Meanwhile, the fuel economy evaluation comprised the initial fuel inventory cost and fissile makeup cost. From the calculation results, lowering graphite density remarkably improves TCR, but also decreased CR in a notable amount. While effective delayed neutron fraction (β eff) was not particularly affected, mean neutron generation time (Λ) was reduced in lower graphite density. Considering the fuel economy, initial fuel cost is not significantly affected, but the fissile makeup cost can be sensitive of graphite density. [ABSTRACT FROM AUTHOR]

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. • Leveraging Modelica/TRANSFORM to simulate MSR behavior • Coupling of 1D thermal hydraulics and distributed model of modified point kinetics • Steady-state and transient benchmarking of MSRE model. [ABSTRACT FROM AUTHOR]

• A transport model for 131I in MSRE was developed by combining the noble metal migration mechanism with a helium bubble distribution model. • Yield evaluation of a production method based on 131I extraction from a molten salt reactor. • Stability and economics of the 131I production method based on molten salt reactor off-gas extraction were evaluated. Isotope extraction methods based on molten salt reactor off-gas treatments can provide a new solution for the shortage of medical isotopes. Current research work involves only static estimation, which does not consider the influence of the variation of parameters brought about by the actual reactor operation and may lead to inaccurate estimation of the yield and cooling time. In this work, a model of species migration is developed, and a corresponding program based on Mathematica is constructed to estimate the above problem. Results reveal 2.50 TBq of 131I was delivered to the off-gas system after 50 days of routine operation. This accounts for only 0.041 % of the total 131I in the reactor, which is sufficient to support 4,000 doses of hyperthyroidism treatment. Compared to the extraction based on the main loop salt, 133I remains the main radioactive impurity and product would meet the medical limit for radioactivity impurity after 8 days of cooling. Changes in fuel composition due to normal reactor operation have a small impact on isotope production. The power and flow rate variations due to peak load operation and the corresponding changes in helium bubble surface area due to temperature–pressure fluctuations have essentially no effect on the cooling time of the product but partially affect the production in the off-gas system. Bubble removal efficiency has a large perturbation on the results, but this needs to be investigated in the future. The stability of the cooling time indicates that nuclide production can follow an established scheme, suggesting that a molten salt reactor tail gas-based production scheme is stable and a promising production method. [ABSTRACT FROM AUTHOR]

• Narrow annular channel heater is proposed to heat molten salt. • Rectangular wire coil is used to enhance the heat transfer. • The mechanism of enhancement is explained by the synergy field principle. To heat the simulator core of a solid-fuel thorium molten salt reactor, the narrow annular channel heater with rectangular wire coil separating from the tube wall is proposed. Three-dimensional numerical models are developed, and the simulated result shows that negative axial, radial and tangential velocities in the enhanced narrow annular channel heater are generated compared to the smooth narrow annular channel heater. As the wire width increases and helical pitch decreases, the Nusselt number and friction factor increase. The maximum Nusselt number for enhanced narrow annular channel heater is 5.6 times that for smooth narrow annular channel heater. The overall performance of enhanced narrow annular channel heater is evaluated at a constant pumping power, the value of which is in the range of 1.21–1.85. The best enhanced narrow annular channel heater for the simulator of solid-fuel thorium molten salt reactor is determined based on the overall performance. The maximum outer wall and fluid temperatures of enhanced narrow annular channel heater for the simulator of solid-fuel thorium molten salt reactor are 937.94 K and 931.01 K, which are within the safety limit values of 1143.15 K and 977.15 K, respectively. This design will provide a new method to heat molten salt. [ABSTRACT FROM AUTHOR]

In the Molten Salt Fast Reactor (MSFR), fuel is dissolved in a liquid salt, causing the formation and decay of fission products, such as noble metals, to be inside the salt. Noble metals have very low solubility and do not dissolve in the salt. They are known to agglomorate into particles which, in turn, diffuse, drift or sediment to interfaces such as walls, free surfaces and bubbles, where they deposit. Leveraging this in the MSFR, online or offline helium bubbling is foreseen to act as a means to extract noble metal particles from the salt. This can potentially limit strong deposition of particles onto walls (i.e., plating), preventing dangerous decay heat hotspots in the heat exchangers or other primary circuit components. In this paper, we investigate the efficiency of particle removal by helium bubbles in the MSFR using Computational Fluid Dynamics (CFD). Theoretical estimates of equilibrium particle size distributions are used as initial condition, based on previous work. CFD calculations are performed in a simplified MSFR geometry. Bubble expansion, turbulent bubble coalescence and turbulent break-up are accounted for using the poly-disperse Log-normal Method of Moments (LogMoM) model, embedded in the two-fluid framework in OpenFOAM. The CFD analyses presented in this paper, in conjunction with the developed theory of noble metal particle formation, growth and capture, will help in the understanding and optimization of the fuel cycle in the MSFR or any other molten salt reactor design. The precise knowledge of how noble metal particle populations in the salt can be controlled, and how plating of noble metal particles onto primary circuit reactor components can be reduced, contributes to safe molten salt reactor operation. • Noble metal particle behavior and removal by bubbles is studied in molten salt reactors. • CFD simulation is performed of noble metal particle flotation. • Particle formation, coagulation and removal by bubbles is considered. • Noble metal cycle times are calculated. [ABSTRACT FROM AUTHOR]

Most molten salt reactor designs present a liquid fuel that, mixed with the coolant composed of molten salts, circulates through the reactor. This circulation introduces important phenomena in the study of reactor physics, thus requiring changes in the existing physical models applied in conventional reactors with solid fuel. One of the main influences of the fuel flow in the reactor is in the delayed neutron precursors' concentration, which partially decay outside the core, not contributing to the fission chain reaction. In this paper, we investigate the physical models used for this type of reactor, especially the neutron point kinetics, in order to develop an inverse kinetics model, which has practical applications such as core reactivity monitoring. We develop a semi-analytical solution for the proposed model, which is used to solve two transients: a linear variation between power levels, and the determination of criticality regions during a decrease in flow velocity. • Circulation of a fluid fuel influences delayed neutrons precursors (DNP) distribution. • DNP heavily affect the system's neutron kinetics. • Inverse Point Kinetics model developed for determining reactivity. • Numerical and semi-analytical solutions for obtained model were proposed. [ABSTRACT FROM AUTHOR]

The International Reactor Physics Benchmark Experiment's MSRE benchmark provides zero-power critical validation data that has been used to assess the accuracy and consistency of SCALE, MCNP, and Serpent for design and licensing of a modern MSR. The codes were used to model the benchmark and a wide range of variations in the geometry, nuclear data library, code versions, and problem specifications. Most of these cases demonstrated excellent consistency, within two standard deviations of the stochastic error, for the reactivity and associated reactivity worth of the variation. However, the codes over-predicted the initial criticality of the MSRE by 3%, which is larger than the provided experimental uncertainty. It was shown that care must be taken to ensure that the bound scattering treatment is consistent with comparing codes. It was also shown that for the MSRE, the use of the ENDF/B-VII.0 library introduced a significant increase in the reactivity (> 200 pcm). Code predictions for two coefficients of reactivity (fuel and isothermal) were validated with data from the MSRE and were well within the experimental uncertainty, thus providing confidence in each code to provide accurate data for safety analysis calculations. These results provide quantifiable estimates of the computational variation one could expect due to the choice of any one of these codes, or a particular nuclear data library, for both initial criticality and reactivity coefficients. • Ensure all participants have a common understanding of detailed reactor modeling. • Ensure we have a demonstrated process for comparing models to ensure consistency in user input (geometry and materials). • Identify limitations of the three codes for modeling criticality and coefficients of reactivity that are necessary inputs for safety analyses. • Quantify the accuracy of the three codes, including their nuclear data, using the MSRE as validation data. • Quantify the variability in predictions of criticality and reactivity coefficients due to the choice of code and data library. [ABSTRACT FROM AUTHOR]

• First experimental results on the phase equilibria and phase stabilities in the ThF4-PuF3 system. • New thermodynamic model for the ThF 4 -PuF 3 system. • Re -assessment of the ThF 4 -PuF 3 and LiF-ThF 4 -PuF 3 phase diagrams. • Discussion on advantages and limitations of using proxy compounds to derive thermodynamic models from experimental data. • Data contribute for extension and improvement of the JRCMSD thermodynamic database on molten halide systems. Phase equilibria in the ThF 4 -PuF 3 system were measured by differential scanning calorimetry. Samples were encapsulated to prevent leakage during the measurements and to prevent fluoride contamination of the instrument. Formation of intermediate compounds and solid solutions was investigated by X-ray diffraction. A thermodynamic model based on the CALPHAD approach was developed for the measured system. Good fit between measurements and phase equilibria calculated using the model was achieved. [ABSTRACT FROM AUTHOR]

• Effective two-equation model for heat transport in molten salt reactors is derived. • Effective heat transfer coefficients for the reactor scale are presented. • The theory predicts thermal equilibrium (non-equilibrium) for small (large) gas volume fractions. • Numerically calculated effective coefficients are presented along simple correlations. Modeling the heat transfer process in a molten salt nuclear reactor is challenging due to the non-uniform distribution of fuel material and cooling fluid. Developing theoretical tools to calculate effective parameters and temperature distribution is essential for ensuring the safety of reactor design. In this study, we utilized the theory of volume averaging to upscale the local model for heat transport in a fluid fuel reactor, known as the Molten Salt Fast Reactor (MSFR). This upscaled model, which is valid for the bulk of the reactor core, combines the heat transport physics that occurs at the microscale through mathematically derived effective coefficients. These coefficients can be obtained by solving related boundary-value problems. The standalone core of the MSFR with inert gas bubbles was studied and the local thermal equilibrium (non-equilibrium) condition between the fluid and bubbles was found to be achievable for small (large) bubbles volume fractions. We present the numerically calculated values of the effective coefficients in terms of the volume fraction of bubbles under operating conditions and provide correlations to facilitate their use in computational codes. [ABSTRACT FROM AUTHOR]

One user-friendly laboratory investigation method for tellurium-induced intergranular cracking has been preliminarily developed in this study. The Te exposure experiments were carried out using FLiNaK salts with NiF 2 and Cr 3 Te 4 at 700 °C in GH3535 alloy tubes. Te exists in the form of intergranular segregation without the formation of surface tellurides, and Te-induced intergranular cracking can occur under unstrained conditions, which agrees with the observations in molten salt reactors. The acceleration effect of NiF 2 can be ascribed to its reaction with Cr 3 Te 4 to form Te metal. • NiF2 addition promotes the molten salt corrosion of GH3535 alloy. • NiF2 addition increases oxidation potential to promote intergranular cracking. • Acceleration effect of NiF 2 results from its reaction with Cr 3 Te 4 to form Te metal. • Intergranular cracking formed in alloys exposed to FLiNaK salts with NiF 2 and Cr 3 Te 4. [ABSTRACT FROM AUTHOR]

• A cross-section online updating surrogate model based on artificial neural network is proposed and developed. • The deterministic fuel management code was improved by integrating with the surrogate model. • The surrogate model and ThorNEMFM-ANN code were validated by the MSBR and MSFR benchmarks. • The applicability of the code is further validated with the fuel cycle simulation of the MCSFR. 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. [ABSTRACT FROM AUTHOR]

This work presents a data-driven Reduced-Order Model (ROM) for parametric convective heat transfer problems in porous media. The intrusive Proper Orthogonal Decomposition aided Reduced-Basis (POD-RB) technique is employed to reduce the porous medium formulation of the incompressible Reynolds-Averaged Navier–Stokes (RANS) equations coupled with heat transfer. Instead of resolving the exact flow configuration with high fidelity, the porous medium formulation solves a homogenized flow in which the fluid-structure interactions are captured via volumetric flow resistances with nonlinear, semi-empirical friction correlations. A supremizer approach is implemented for the stabilization of the reduced fluid dynamics equations. The reduced nonlinear flow resistances are treated using the Discrete Empirical Interpolation Method (DEIM), while the turbulent eddy viscosity and diffusivity are approximated by adopting a Radial Basis Function (RBF) interpolation-based approach. The proposed method is tested using a 2D numerical model of the Molten Salt Fast Reactor (MSFR), which involves the simulation of both clean and porous medium regions in the same domain. For the steady-state example, five model parameters are considered to be uncertain: the magnitude of the pumping force, the external coolant temperature, the heat transfer coefficient, the thermal expansion coefficient, and the Prandtl number. For transient scenarios, on the other hand, the coastdown-time of the pump is the only uncertain parameter. The results indicate that the POD-RB-ROMs are suitable for the reduction of similar problems. The relative L2 errors are below 3.34% for every field of interest for all cases analyzed, while the speedup factors vary between 54 (transient) and 40,000 (steady-state).

BackgroundIn the reactivity initiated accidents under low power operating conditions of molten salt reactor (MSR), different reactivity insertion rates will trigger different emergency shutdown signals. At the same time, the initial conditions such as the initial reactor power and the temperature coefficients of reactivity affect the accident process and cause differences in accident consequences.PurposeThe study aims to conduct a sensitivity analysis of the impact of the reactivity insertion rate, the initial reactor power, and the reactivity temperature coefficient on transient consequences.MethodsFirst of all, 7 reactivity insertion rate conditions were selected and simulated through RELAP5-TMSR.Then, 25 combinations of the initial reactor power and the temperature coefficients of reactivity were assumed as initial conditions. Finally, the effects of these three parameters on the consequences of the accident were discussed separately by using local sensitivity analysis method.ResultsThe insertion rate that causes a concurrent trigger of the high outlet temperature and the high-power shutdown signal leads to the most unfavorable consequence. The difference between the peak temperatures of the fuel salt and structural materials and their respective initial values under the worst reactivity insertion rate condition is negatively correlated with initial power. However, the temperature difference of each parameter caused by different initial power does not exceed 3 ℃. The difference between the peak temperatures decrease first and then increases with the increase of the temperature coefficients of reactivity, but the maximum difference does not exceed 0.5 ℃.ConclusionsUnder low power operating conditions of MSR, the consequences of reactivity introduced events are highly sensitive to the reactivity insertion rate and low sensitivity to the initial power and temperature coefficients of reactivity.

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450°C to 700°C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

BackgroundThe control rod guide tube (CRGT) is needed to facilitate the movement of the control rod in the molten salt reactor (MSR). Graphite material is resistant to neutron radiation, high temperature and molten salt corrosion, and has small neutron absorption cross section, hence it is an ideal CRGT material except for the comparatively low strength and toughness. It is believed that reinforcing the graphite CRGT with carbon fiber fabric (CFF) is a promising way to promote its application in molten salt reactor.PurposeThis study aims to optimize the preparation parameters of the CFF reinforcement for CRGT.MethodsFirst of all, the CFF was pre-impregnated with precursor solution in a vacuum environment, and wound on the graphite tube with a certain tension. Secondly, samples with different densification cycles and winding layers were prepared through multiple pressure impregnation, curing and carbonization cycles of the Precursor-Infiltration-Pyrolysis (PIP) process. Then, the mechanical properties of carbon fiber cloth reinforced graphite tube samples and the reference samples were tested. Finally, the microstructure and failure mode of the damaged samples were analyzed for further optimizing the preparation process to improve the material performance.Results & ConclusionsThe results show that the CFF winding and PIP method enhances the strength and toughness of the graphite tube, and the preparation parameters have been optimized that make the reinforced graphite CRGT with carbon fiber fabric (CFF) possible to be applied to MSR.

BackgroundMolten salt reactor (MSR), a candidate of the Generation IV reactors, has the advantages of inherent safety and good neutron economy. The circulation flow of fuel, on-line refueling and reprocessing are the three features of liquid fuel in MSR. Due to the features of liquid fuel, the fuel management regulations for conventional reactors that use solid fuels are not applicable to liquid-fueled MSR.PurposeThis study aims to develope a new fuel management code named ThorNEMFM to analyze liquid fuel management of MSR.MethodsFirst of all, the self-developed deterministic code ThorCORE3D was coupled with the cross section processing module and a burnup calculation module. Then a fuel management analysis program ThorNEMFM was developed for liquid fuel molten salt reactor to realize on-line feeding and fission product processing. Finally, the ThorNEMFM was verified on molten salt reactor experiment (MSRE), molten salt breeder reactor (MSBR) and molten salt fast reactor (MSFR).ResultsThe results of benchmarks for MSRE, MSBR and MSFR indicate that ThorNEMFM has a high accuracy with benchmarks in comparison with the reference.ConclusionsThorNEMFM is suitable for calculation and analysis of fuel management of MSR.

BackgroundMolten salt reactor (MSR) is an ideal advanced reactor with good neutron economy, inherent safety and online refuelling.PurposeThis study aims to evaluate the transient behavior of reactive insertion transients of MSR.MethodsBased on the RELAP5/FLUENT coupling program and the establishment of the primary circuit model, the graphite-moderated channel type molten salt reactor was studied, and transient analysis was conducted when the reactor core was inserted with 0.000 1, 0.000 2, and 0.000 5 reactivities.ResultsThe results show that under the concerned reactivity insertion, the maximum temperature of the reactor core and the outlet temperature of the salt are both lower than the safety allowable limit due to the negative temperature reactivity of the reactor core.ConclusionsThis study verifies the good response ability of MSR to reactivity insertion events.

• Various deployment scenarios of a future FSMSR fleet are evaluated with ORION. • Efficiency of transitioning from an LWR fleet to an FSMSR fleet is analyzed. • FSMSRs can be easily deployed due to key attributes discussed in this paper. Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. A solid-fueled reactor only discharges a fraction of a core every year. [ABSTRACT FROM AUTHOR]