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In the framework of the European project SAMOSAFER, this numerical study focuses on some thermal aspects of the Emergency Draining Tank (EDT) located underneath the core of a Molten Salt Reactor. In case of an emergency, this tank passively receives the liquid fuel salt and is designed to ensure a subcritical state. An important requirement is that the fuel does not overheat to maintain the EDT Hastelloy container integrity. The present EDT is based upon a group of hexagonal cooling assemblies arranged in a hexagonal grid and cooled down thanks to conduction through the inert salt layer up to an air flow in charge of removing the heat. This numerical thermal study relies on a conjugated heat transfer analysis coupling a Finite Element solid thermal code (SYRTHES) and two instances of a Finite Volume CFD codes (Code_Saturne). Calculations on an initial design suggest that a simple center airpipe flow is likely to not sufficiently cool the device. Alternative solutions have been evaluated. Introduction of fins to enhance the heat transfer do not bring a noticeable improvement regarding maximum temperature reached. However, a solution in which the central pipe air flow is replaced by several cooling channels located closer to the fuel is investigated and suggests a better cooling.

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

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10−4); the splitting method could reach a precision of 10−2 level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors.

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.

BackgroundThe passive cooling system for the reactor cavity of the molten salt reactor (MSR) is an important guarantee to ensure the safe operation of the reactor, is one of the four engineered safety features for the MSR, and its structure design is an important part of the thermal hydraulic design.PurposeThis study aims to find out a suitable passive reactor cavity cooling system (RCCS) to meet the requirements of thermal shielding design for the lower reactor cavity of the MSR, and maximize the removal of reactor core decay heat under accident conditions.MethodsFirstly, based on the design parameters of a 153 MWt MSR, a 1/4 geometric model of the lower reactor cabin of this MSR was established. Then, ANSYS FLUENT 20.1 software was employed to conduct three-dimensional numerical simulation of the flow filed and temperature filed for the lower reactor cabin, the influence of thermal shielding of the lower cavity was analyzed by changing the structure and layout of the passive RCCS, the structure sizes of the passive RCCS with double channel, the thickness of thermal insulation cotton on the intermediate thermal shielding plate and the position of the air inlet pipe. Finally, a new and suitable structure of passive RCCS was proposed after step-by-step improvements for a 153 MWt MSR.ResultsThe simulation results show that the optimized new-style passive air-cooling system with a double channel in the lower reactor cavity is the best among the three structures. Changing the width of the RCCS has little effect on the thermal shielding results of the lower reactor cabin whilst increasing the thickness of thermal insulation cotton on the intermediate thermal shielding plate of the RCCS can significantly reduce the temperature of the inner surface of the concrete wall. The closer the inlet position of the air inlet pipe is to the top of the RCCS, the better the thermal shielding effect. Based on above results, a new-style passive air-cooling system with double channel in the lower cavity is designed to completely dsatisfy the requirements for shielding cooling for the lower reactor cavity of a 153 MWt MSR.ConclusionsThe results of this study provide an important reference for the further engineering optimization design of passive residual heat removal system in the hundred megawatt-scale molten salt reactor.

BackgroundMolten salt reactor is one of the six internationally recognized and recommended fourth generation reactors that can use liquid nuclear fuel. The production, transportation, and storage of nuclear fuel involve different processes from conventional solid-state nuclear fuel reactors.PurposeThis study aims to perform criticality safety analysis for nuclear fuel storage of molten salt reactor in compliant with requirements of nuclear fuel management and nuclear safety supervision.MethodsThe design parameters of MSRE (Molten Salt Reactor Experiment) were referred for criticality safety analysis, and the impact of different factors on nuclear fuel salt storage was analyzed. Calculation was completed by selecting storage modeling, critical parameter analysis, and Monte Carlo neutron transport software simulation for liquid fuel molten salt reactor nuclear fuel. The keff values of dry environment storage and water flooded environment storage under the design model were summarized, so did the changes in the total mass of fuel salt.Results & ConclusionsThe quality of subcritical safety control under different conditions are obtained and compared with corresponding raw material salts, intermediate products, and consideration of container walls. This study combines legal regulations and the process of nuclear material circulation for analysis and discussion, summarizes the critical safety characteristics of nuclear fuel salt, and for the first time proposes relevant supervision and evaluation points from the perspective of nuclear safety supervision.

BackgroundThe molten salt reactor (MSR) is one of the six advanced reactors identified by the Generation IV International Forum. The MSR exhibits unique characteristics, such as intrinsic safety, sustainable development, nuclear nonproliferation, natural resource protection, and economic efficiency. After a liquid-fuel molten salt reactor shuts down, residual decay heat in the reactor core is passively dissipated to the environment through a natural circulation loop of the molten salt. However, the decay heat from the molten salt in the main circuit can impact the thermal capacity of the overall system.PurposeThis study aims to determine the thermal characteristics of the passive system by establishing an analysis basis for a natural cycle model to examine the effects of natural circulation loop on the physical properties of the molten salt, the loop structure, and equipment resistance coefficient K.MethodsBased on the direct reactor auxiliary cooling system (DRACS) for MSR, the natural circulation model of the passive residual heat-removal loop of a liquid-fuel molten salt reactor was established using self-developed Python analysis program, and the temperature distribution of the molten salt in the loop was explored using the numerical model. Then, the effects of different physical properties, loop structures, and core and heat exchanger resistance coefficients K on the heat transfer and flow characteristics of the residual heat-removal system were analyzed. Finally, the effects of critical factors on the residual heat-removal capacity of the reactor core were analyzed using self-developed Python program for natural circulation calculation equations code (NCCC) and validated by using the natural cycle experimental results of the DRACS circuit in CIET1.0.ResultsThe findings indicate that the presence of decay heat from molten salt in a system loop decreases the natural circulation-driven heat transfer by the molten salt within the core. Based on the significance analysis results show that fuel salt density, specific heat capacity, and height difference between the hot and cold cores are parameters that significantly influence the natural circulation capacity. When these three parameter values are increased by 15% separately, the residual heat removal capacities increase by 26.02%, 15.00%, and 18.59%, respectively.ConclusionsResults of this study demonstrate that the molten salt properties, circuit structure, and equipment resistance coefficient all affect the natural circulation heat-removal capacity.

The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10−4); the splitting method could reach a precision of 10−2 level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors. [ABSTRACT FROM AUTHOR]

BackgroundThe extraction of 99Mo medical isotopes from off-gas of molten salt reactors (MSRs) has attracted significant attention. However, previous studies focused only on the estimation of 99Mo nuclide production in the reactor core, and limited studies have focused on the expansion of the entire reactor system.PurposeThis study aims to estimate 99Mo production in the off-gas of a molten salt reactor.MethodsIn this study, a 99Mo migration model with noble metal deposition and a flow effect in an MSR was established by using scientific computing software “Mathematica”, and a corresponding verification based on the deposition result of a molten salt reactor experiment (MSRE) was implemented. Subsequently, an analysis was conducted on the impact of changes in the yield, specific activity, and operational status of 99Mo crude product in MSRE off-gas on the stability of production.ResultsThe analysis results of Mo transport in the MSRE show that the production of 99Mo in the off-gas can reach 69.30 TBq with a specific activity of 0.39 PBq∙g-1 under normal operating conditions. In addition, an increase in the removal rate is conducive to the effect on the specific activity of 99Mo whilst the working conditions of bubble surface ratio bubble ratio and oxidation-reduction potential changes have little effect on the specific activity of the product.ConclusionsThe 99Mo in off-gas has high yield and specific activity, and the effects of the conventional operating conditions on the specific activity of 99Mo are beneficial, which indicates that this method can be a potential alternative for 99Mo production.

BackgroundMolten salt reactors (MSRs) feature high temperature, low pressure, high chemical stability, and nuclear non-proliferation, which make them promising for a wide range of applications. However, owing to the fluid nature of the fuel, changes in the core structure of MSRs are bound to affect the fuel distribution and loading, consequently affecting the physical parameters of the core. Currently, the research on the impact of structural changes of the MSR core, composed of dispersed graphite components, on reactivity is insufficient.PurposeThis study aims to explores the influence of core structure changes on reactivity.MethodsThe core design model of new solid hexagonal graphite module MSR was taken as a reference, the small geometric changes, such as deformation and displacement of the core components, caused by factors including thermal expansion, fluid erosion, core vibration, and graphite irradiation, were analyzed using MCNP code. Additionally, the impact of these structural changes in the core on reactivity was investigated in details.ResultsThe results indicate that the structural changes in the core caused by thermal expansion introduce negative reactivity. The reactivity introduced by fluid erosion and core vibration, causing graphite component displacement, fluctuates. However, the overall trend shows that a shift of the graphite components towards the center of the core introduces positive reactivity, whereas shift towards the outer periphery of the core introduces negative reactivity. Graphite irradiation-induced deformation initially decreases reactivity and then increases it. At the end of the core's lifespan, the reactivity is still lower than at the beginning of the lifespan. This reactivity change can be compensated for by online feed or control rod movement, which has a limited impact on the operation of the MSR. However, the critical issue of control should be considered when this batch of fuel is reinserted into a new core.ConclusionsResults of this study suggest the necessity of constraining the graphite components within a certain range to ensure the safe operation of the reactor, providing an important reference for the design, operation, and maintenance of MSRs.

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D2-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D2-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D2-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D2-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

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. [ABSTRACT FROM AUTHOR]

This paper presents a molten salt reactor (MSR) design parameter sensitivity study using a nodal dynamic modelling methodology with explicitly modified point kinetics equation and Mann’s model for heat transfer. Six parameters that can impact MSR safety are evaluated. A MATLAB-Simulink model inspired by Thorcon’s 550MWth MSR is used for parameter evaluations. A safety envelope was formed to encapsulate power, maximum and minimum temperature, and temperature-induced reactivity feedback. The parameters are perturbed by ±30%. The parameters were then ranked by their subsequent impact on the considered safety envelope, which ranks acceptable parameter uncertainty. The model is openly available on GitHub.

This paper describes the methodology used to create a fuel data library comprising safeguards-relevant quantities that may be useful for verification of spent nuclear fuel (SNF) produced by simulating a concept Molten Salt Reactor (MSR). The Monte-Carlo particle transport code, Serpent2 and the calculation code SOURCES 4C were used to compile this fuel data library. The data library is based on the Compact Molten Salt Reactor (CMSR) concept being developed by Seaborg Technologies (based in Copenhagen, Denmark). The library includes data such as nuclide mass densities for a total of 1398 nuclides (in g/cm3), as well as total decay heat production (denoted by suffix the ‘TOT_DH’) in Watts, total gamma photon emission rates (denoted by the suffix ‘TOT_GS’) in photos per second, and the total activity (denoted by suffix ‘TOT_A’) in Becquerel. Lastly, the data also includes total neutron emission rates from 1) spontaneous fission (denoted by ‘SF’ and reported in neutrons per second per cm3), and 2) (ɑ, n) reactions (denoted by ‘AN’ and reported in neutrons per second per cm3) for the fuel salt. These quantities are reported for a range of burnup-initial enrichment-cooling time (or collectively known as, BIC) parameters. The resulting fuel data library is an extension of a previously published data library for the same reactor concept but with one significant change. The current library is based on a more realistic model of the CMSR involving movement of gaseous and volatile fission products (GFP and VFP) from the core via an Off-Gas System (OGS). The dataset is made available for public use in a compressed binary format as an HDF5 (or Hierarchical Data Format) file that can be parsed using data analysis tools such as Pandas.