Fast-spectrum Molten Salt Reactors Research Articles

Influence of transmutation products on the thermophysical properties of eutectic NaCl-UCl3 fuel salt in a fast-spectrum molten salt reactor

Influence of transmutation products on the thermophysical properties of eutectic NaCl-UCl3 fuel salt in a fast-spectrum molten salt reactor

CFD Simulations of the Molten Salt Fast Reactor

The molten salt fast reactor (MSFR) is a fast-spectrum molten salt reactor concept, where the fuel salt flows freely through a toroidal core (2 m high × 2 m in diameter). The flow through this core is turbulent (Reynolds number ≈ 5 × 10 5 ) and presents several recirculation areas. Several thermal-hydraulic studies of the steady-state MSFR flow have been performed, where the thermal power distribution was fixed and the Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) approaches were used. Different geometries were considered, from 1/16th of the core to the full core. The RANS simulations were performed first. The use of symmetry boundary conditions in these simulations had a very limited impact on the results while greatly speeding up the calculations. In the LES formulation, however, the symmetry boundary conditions had a large impact on the velocity and temperature fields, and so could not be used. While the RANS fields and the LES mean fields differed quite heavily, no particular hot spots appeared in the LES fields. Analysis of the LES temperature evolution in the center of the core revealed fluctuations of ± 20°C at a speed of 100°C·s−1.

On the need for multi-dimensional models for the safety analysis of (fast-spectrum) Molten Salt Reactors

This paper aims at characterizing the impact of adopting numerical models with different dimensionalities on the predicted behavior of fast-spectrum Molten Salt Reactors (MSRs). The study encompasses 1-D, 2-D, and 3-D representations of thermal-hydraulics and precursor transport/diffusion, along with spatial and point kinetics models for neutronics. We evaluate the accuracy of each model based on steady-state results and on the reactor response to 2 different transient initiators. The findings emphasize the significance of utilizing a 3-D representation with accurate thermal-hydraulics modeling, and with either spatial kinetics or carefully calibrated point kinetics incorporating a spatial description of precursors transport. 2-D and 1-D models can reproduce main trends and remain valuable tools for e.g. reactor design, control-oriented studies or uncertainty quantification. However, proper calibration of these models is needed and the user should be aware that alterations in flow patterns could jeopardize model calibration and hide first-order local effects.

Neutronics for fast spectrum, molten salt reactors and evolution over burnup

There are a wide variety of fast spectrum molten salt reactor (FSMSR) designs and corresponding fuel cycles that shown promise. These designs differ in features. Therefore, it becomes difficult to determine whether a set of parameters is applicable to all, one, or none of these designs. This paper presents a number of iterations of a FSMSR concept with a sodium chloride carrier salt. Different design characteristics and their effects on the neurotics parameters are shown. The results show that effective delayed neutron fraction, mean neutron generation time, neutron yield, and energy of neutrons causing a fission all follow trends based off the heavy metal concentration and leakage rate. This paper also gives preliminary fuel cycle calculations and the evolution of the neutronic parameters over burnup. A FSMSR initially fueled with low enriched uranium will naturally breed at a certain core volume, without a need for refueling, for decades of full power operation.

Adding MgCl2 to Molten NaCl-UCln (n=3, 4): Insights from First-Principles Molecular Dynamics.

Molten chlorides are proposed for fast-spectrum molten salt reactors. Molten MgCl2 with NaCl forms eutectic mixtures and is considered as a promising dilutant to dissolve fuel salts such as UCl3 and UCl4 . A previous study suggests the formation of a U-Cl network at the U : Na=1 : 1 binary salt. However, it is unclear how the structure of UCln (n=3, 4) in NaCl will change after adding MgCl2 in the salt. Here, we use first-principles molecular dynamics to investigate the molten ternary salts NaCl-MgCl2 -UCln (n=3, 4) at various concentrations of Mg2+ in NaCl-UCln with a fixed ratio of Na : U at 1 : 1. It is found that the addition of Mg2+ to NaCl-UCl3 leads to a higher coordination number (from 6.5 to 6.7) of Cl around U while the U-Cl network structure slightly decreases with the Mg concentration. Adding MgCl2 to NaCl-UCl4 , however, breaks down the U-Cl network more completely. We attribute the different behavior of adding Mg2+ into NaCl-UCl3 and NaCl-UCl4 to the difference between U(III) and U(IV) in attracting Cl- ions to form the first coordination shell. The present work reveals the impact of MgCl2 as a dilutant solvent for NaCl-UCln fuel salts, which will be helpful in further studies and understanding of the thermophysical and transport properties of ternary systems.

Verification of moltres for multiphysics simulations of fast-spectrum molten salt reactors

Modeling strongly coupled neutronics and thermal–hydraulics in liquid-fueled MSRs requires robust and flexible multiphysics software for accurate simulations at reasonable computational costs. In this paper, we present Moltres and its neutronics and thermal–hydraulics modeling capabilities relevant to multiphysics reactor analysis. As a MOOSE-based application, Moltres provides various multiphysics coupling schemes and time-stepping methods, including fully coupled solves with implicit time-stepping. We verified Moltres’ MSR modeling capabilities against a multiphysics numerical benchmark developed for software dedicated to modeling fast-spectrum MSRs. The results show that Moltres performed comparably to participating software packages in the benchmark; the majority of the relevant quantities fell within one standard deviation of the benchmark average. Among the participating multiphysics tools in the benchmark, Moltres agrees closest to the multiphysics tool from the Delft University of Technology due to similarities in the numerical solution techniques and meshing schemes.

First-principles molecular dynamics simulations of UCln-MgCl2 (n = 3, 4) molten salts.

Molten chlorides are a preferred choice for fast-spectrum molten salt reactors. Molten MgCl2 forms eutectic mixtures with NaCl and is considered as a promising dilutant to dissolve fuel salts such as UCl3 and UCl4. However, the structure and chemical properties of UCln (n = 3, 4) in molten MgCl2 are not well understood. Here we use first-principles molecular dynamics to investigate the molten salt system UCln-MgCl2 (n = 3, 4) at various concentrations of U3+ and U4+. It is found that the coordination environment of Cl- around U3+, especially in the first coordination shell, varies only slightly with the uranium concentration and that both the 7-fold coordinate (UCl74-) and 6-fold coordinate (UCl63-) structures dominate at ∼40%, leading to an average coordination number of 6.6-6.7. A network or polymeric structure of U3+ cations sharing Cl- ions is extensively formed when the mole fraction of UCl3 is greater than 0.2. In contrast, the average coordination number of Cl- around U4+ is about 6.4 for a mole fraction of UCl4, x(UCl4), of 0.1 but decreases to 6.0 for x(UCl4) = 0.2 and then stays at about 6.0-6.2 with the uranium concentration. The 6-fold coordinate structure (UCl62-) is the most populous in UCl4-MgCl2, at about 60%. U-Cl network formation becomes dominant (>50%) only when x(UCl4) > 0.5. Unlike Na+, Mg2+ forms a network structure with Cl- ions and when x(UCl3) or x(UCl4) < 0.5, over 90% of Mg2+ ions are part of a network structure, implying the complex influences from Mg2+ on the coordination of Cl around U. The present work reveals the impact of MgCl2 as a solvent for UCln (n = 3, 4) on the U-Cl coordination and structure, and motivates further studies of their transport properties and the tertiary systems containing MgCl2-UCln.

Design and neutronic analysis of the intermediate heat exchanger of a fast-spectrum molten salt reactor

Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal power. The core power density was set to 100 MW m −3 with a core H/D [[EQUATION]] 1.0 amd four Intermediate Heat Exchanger (IHX). This leads to a power of 150 MW per IHX; this power is also comparable to the IHX proposed in the SAMOFAR framework. In this work, a preliminary design of a 150 MW helical-coil IHX for a chloride-fueled MSR is prepared and the fission rate, capture rate, and inelastic scatter rate are evaluated.

MULTIPHYSICS MODELING OF PRECURSORS IN MOLTEN SALT FAST REACTORS USING PROTEUS AND Nek5000

The goal of this work was to calculate the impact of the delayed neutron precursor drift in fast spectrum Molten Salt Reactors (MSRs) using coupled solutions from the neutronics code PROTEUS and the computational fluid dynamics code Nek5000. Specifically, using a multiphysics approach to solve the effective delayed neutron fraction (βeff) or delayed neutron precursor distribution for reactors with flowing fuel salts would provide valuable information for transient simulations and safety assessments. Given the multiple options for the flux solution and geometric resolution/fidelity in PROTEUS, two approaches were developed and applied to various test cases: PROTEUS-NODAL/Nek5000 and PROTEUS-SN/Nek5000. For the former, the precursors are tracked in the built-in precursor drift model in PROTEUS-NODAL, whereas in the latter, Nek5000 directly tracks the precursors. Both approaches were used to solve a single test channel problem and showed excellent agreement in the calculated βeff. Separately, a 3D hourglass-shaped core was modeled using the PROTEUS-SN/Nek5000 approach. This problem was designed to demonstrate the capability of the discrete ordinates (SN) solver and Nek5000 to model complex core designs with axially varying geometries and the ability for Nek5000 to track the precursors and calculate the resulting βeff. In addition, the Nek5000 calculations revealed the presence of recirculation zones in the hourglass design, which could lead to significant temperatures in the fuel salt and surrounding materials. These first coupled solutions show why these approaches may be necessary for not only predicting the precursor drift effect in fast MSRs but also for reactor design and performance assessments.

Test task for transmutation of 241Am in molten salt reactors

Molten Salt Reactor (MSR) is one of the six innovative systems, which has been selected by the international Generation IV forum as a candidate for the next-generation reactors. In this paper, the transmutation of241Am, the main contributor to the waste toxicity in the nuclear fuel cycle, in both thermal- and fast-spectrum molten salt reactors is compared. The Am-cylinder is irradiated in the center of the thermal (SD-TMSR) and fast (SMSFR) spectrum MSRs, respectively, and during a long period of time. The total neutron fluxes in SD-TMSR and SMSFR are 4.1×1014 and 1.8×1015 n.cm-2.s-1, respectively. The overall change in actinides and fission products (FPs) mass during the irradiation has been investigated using direct Serpent2 calculations. Under fast spectrum irradiation, transmutation leads to produce short-lived FPs (by fission) rather than the accumulation of heavier actinides. The ratio between the maximum and initial released energy from all isotopes in the Am-cylinder has been calculated in this paper.

Results from a multi-physics numerical benchmark for codes dedicated to molten salt fast reactors

Verification and validation of multi-physics codes dedicated to fast-spectrum molten salt reactors (MSR) is a very challenging task. Existing benchmarks are meant for single-physics codes, while experimental data for validation are absent. This is concerning, given the importance numerical simulations have in the development of fast MSR designs. Here, we propose the use of a coupled numerical benchmark specifically designed to assess the physics-coupling capabilities of the aforementioned codes. The benchmark focuses on the specific characteristics of fast MSRs and features a step-by-step approach, where physical phenomena are gradually coupled to easily identify sources of error. We collect and compare the results obtained during the benchmarking campaign of four multi-physics tools developed within the SAMOFAR project. Results show excellent agreement for all the steps of the benchmark. The benchmark generality and the broad spectrum of results provided constitute a useful tool for the testing and development of similar multi-physics codes.

Molecular dynamics simulations of structural and transport properties of molten NaCl-UCl3 using the polarizable-ion model

Molten chlorides are a good choice for fast-spectrum molten salt reactors, but many of their thermophysical and transport properties are not available. In this work we use classical molecular dynamics simulations based on the polarizable-ion model to investigate the structural and transport properties of NaCl-UCl3 at various U3+ mole fractions. Molar volume, coordination structure and number, network structure, heat capacity, diffusivity, and ion conductivity have been simulated. We find strong variation of the second coordination shells (U-U and Cl-Cl) with U3+ mole fraction. This is consistent with the increasing network formation or oligomerization of UCl63− and UCl74− units via shared Cl− ions. The movement of Cl− is strongly coupled with U3+ as the U-Cl network structures become dominant in the mixture when the mole fraction of UCl3 is >0.25. Evaluation of a heat-transfer figure-of-merit suggests that 0.20 to 0.30 is an optimal mole fraction of UCl3 to achieve both lower operating temperatures and better heat transfer.

Modeling and simulation functional needs for molten salt reactor licensing

The modeling and simulation functional needs to license and design molten salt reactor concepts were assessed and an evaluation methodology was developed using a licensing-focused approach considering the modeling and simulation components that contribute to a molten salt reactor safety analysis report. These functional needs were prioritized by identifying the relationships between functional needs and using a quantitative weighting method to identify the most critical components to build a licensing case. This novel weighting method provided a list of functional needs prioritized relative to key advanced reactor licensing outcomes. In addition, functional need inputs were ranked to reveal those that feed the most modeling and simulation components. The findings from this prioritization effort were used to inform the modeling and simulation goals of the US Department of Energy Office of Nuclear Energy (DOE-NE) Molten Salt Reactor campaign.Modeling and simulation needs were grouped into three main categories: (1)source term definition, material accountancy, and radiological consequences, (2)safety analysis, and (3)steady-state neutronics and thermal hydraulics. Five near-term goals were identified among these focus areas that add value: (1) deploying a systems and safeguards analysis computational tool, (2) defining source terms for molten salt reactor analysis and interfacing with the US Nuclear Regulatory Commission on accident progression analysis, (3) developing point designs for liquid-fueled thermal-spectrum and fast-spectrum molten salt reactors, (4) building safety analysis specifications and reference solutions for Molten Salt Reactor Experiment transients, and (5)interfacing with the thermophysical and thermodynamic salt chemistry databases.

Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code

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.

Multiphysics analysis of the MSFR helium bubbling system: A comparison between neutron diffusion, SP3 neutron transport and Monte Carlo approaches

The Molten Salt Fast Reactor is a fast-spectrum molten salt reactor under development in the framework of the European H2020 SAMOFAR Project (http://samofar.eu/). Among the design peculiarities, this circulating fuel reactor features a helium bubbling system aimed at removing on-line gaseous fission products, and metallic particles as well. From a modelling point of view, the presence of helium bubbles in the core needs to be assessed both from a neutronics and thermal-hydraulics point of view. In this paper, the attention is paid to the first aspect, analysing the void reactivity effect induced by the bubbles by means of a Monte Carlo, an SP3 neutron transport and a neutron diffusion approach. Since the distribution of helium bubbles is not uniform in the core but strongly depends on the fuel salt flow conditions, as well as on the location of their injection and extraction, a coupling scheme between Serpent 2 and OpenFOAM codes is adopted for the Monte Carlo analysis. In this way, the Monte Carlo code Serpent 2 is provided with a real bubble distribution calculated by means of a two-phase Euler-Euler solver implemented in OpenFOAM. The outcome puts in evidence the difference arising when a uniform and a real distribution of bubbles is considered in a Monte Carlo calculation. These results are also compared with an SP3 neutron transport and a neutron diffusion solver implemented in OpenFOAM, highlighting the difference among the three neutronics approaches herein adopted.

First-Principles Molecular Dynamics Simulations of UCln–NaCl (n = 3, 4) Molten Salts

Molten chloride salts are one of the two main choices to dissolve nuclear fuels for fast-spectrum molten-salt reactors, as a generation IV reactor design. Both U3+ and U4+ can stably exist in molte...

Liquid Fuel Molten Salt Reactors for Thorium Utilization

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride based or chloride based, as either a coolant with a solid fuel (such as fluoride salt–cooled high-temperature reactors) or as a combined coolant and fuel with the fuel dissolved in a carrier salt. For liquid-fueled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as for introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with the online removal of parasitic absorbers enable the design of a thermal-spectrum breeder reactor. However, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at Oak Ridge National Laboratory in the 1950s and 1960s: the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR) with multiple configurations that could breed additional fissile material or maintain self-sustaining operation and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation resistance. MSRs have been selected as one of the Generation IV systems, and development activity has been seen in fast-spectrum MSRs, waste-burning MSRs, and MSRs fueled with low-enriched uranium as well as in more traditional thorium fuel cycle–based MSRs. This paper provides a historical background of MSR R&D efforts, surveys and summarizes many of the recent developments, and provides analysis comparing thorium-based MSRs by way of example.

GeN-Foam: a novel OpenFOAM® based multi-physics solver for 2D/3D transient analysis of nuclear reactors

The FAST group at the Paul Scherrer Institut has been developing a code system for reactor analysis for many years. For transient analysis, this code system is currently based on a state-of-the-art coupled TRACE-PARCS routine. This work presents an attempt to supplement the FAST code system with a novel solver characterized by tight coupling between the different equations, parallel computing capabilities, adaptive time-stepping and more accurate treatment of some of the phenomena involved in a reactor transient. The new solver is based on OpenFOAM®, an open-source C++ library for the solution of partial differential equations using finite-volume discretization. It couples together a multi-scale fine/coarse mesh sub-solver for thermal-hydraulics, a multi-group diffusion sub-solver for neutronics, a displacement-based sub-solver for thermal-mechanics and a finite-difference model for the temperature field in the fuel. It is targeted toward the analysis of pin-based reactors (e.g., liquid metal fast reactors or light water reactors) or homogeneous reactors (e.g., fast-spectrum molten salt reactors). This paper presents each “single-physics” sub-solver and the overall coupling strategy, using the sodium-cooled fast reactor as a test case, and essential code verification tests are described.

The molten salt reactor (MSR) in generation IV: Overview and perspectives

Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled High-temperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel.Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.

Liquid Salt Applications and Molten Salt Reactors

In the last 5 years, there has been a rapid growth in interest in the use of high-temperature (700 to 1000°C) molten and liquid fluoride salts as coolants and for other functions in nuclear systems. This interest is a consequence of new applications for high-temperature heat and the development of new reactor concepts. These salts have melting points between 350 and 500oC; thus, they are of use only in high-temperature systems. Nitrate salts with a peak operating temperature of ~600°C are the highest-temperature commercial liquid coolant available today; thus, the development of higher-temperature fluoride salts as coolants opens new nuclear and non-nuclear applications. These salts are being considered for intermediate heat transport loops between all types of hightemperature reactors (helium and salt cooled) and hydrogen production systems, oil refineries, and shale oil processing facilities. Historically, steam cycles with temperature limits of ~550°C have been the only efficient method to convert heat to electricity. This limitation produced few incentives to develop high-temperature reactors for electricity production. However, recent advances in Brayton gas-turbine technology now make it possible to convert higher-temperature heat efficiently into electricity and thus have created the enabling technology for more efficient cost-effective high-temperature reactors. The near-term advanced high-temperature reactor (AHTR) uses a graphite-matrix coated-particle fuel and a liquid salt coolant. There is the longer-term potential of a liquid-salt-cooled fast reactor (LSFR) that uses metal-clad fuel and a liquid salt coolant. The molten salt reactor (MSR), with the fuel dissolved in the molten salt coolant, is receiving increased attention because of (1) the advancing salt-coolant technology and Brayton cycles that improve the economics, (2) advances in salt chemistry that enable the development of fast-spectrum MSRs with the safety advantages of large negative void coefficients, and (3) the interest in actinide burning where MSRs avoid the need to fabricate fuel of highly active actinides. Last, there is a developing interest in liquid-wall fusion machines with much higher power densities than solid-wall fusion machines.