Fluoride Salt Coolant Research Articles

Assessment of Reactor Physics Characteristics of Prismatic Fuel Concepts With a Hydrogen-Based Moderator for Use in a Fluoride Salt-Cooled Small Modular Reactor

Abstract Fluoride-salt-cooled high temperature reactor (FHR) effectively combines the solid fuel and moderator design of high-temperature gas-cooled reactor (HTGR) technology with the fluoride salt coolant (LiF-BeF2, FLiBe) of molten salt reactor (MSR) technology, enabling low-pressure (∼1 atm, 101.325 kPa), and high-temperature (∼700 °C) operations. The design and operational features of the FHR make it a potentially attractive option for a small modular reactor (SMR), provided that it can be modified and made physically small and operate at a low-enough power level (<350 MWth/<150 MWel). Most FHR-SMR designs use high-assay low enriched uranium (HALEU) fuel in the form of tri-structural isotropic (TRISO) fuel particles, combined with the use of a graphite moderator. However, there are alternative design concepts for an FHR-SMR that may offer superior performance characteristics, while utilizing an alternative fissile fuel supply option. In this exploratory study, lattice physics calculations were performed with Serpent to evaluate an alternative FHR-SMR prismatic fuel block design concept using coated annular fuel pellets instead of TRISO-particle fuel compacts, along with the use of hydrogen-based solid moderator rods made of 7LiH. In initial studies, it was found that fuel blocks with 120 moderator rods made of 7LiH tended to have large positive temperature reactivity coefficients (TRCs), which is undesirable for safety reasons. However, reducing the number of moderator rods to 90 or 54, while increasing the number of fuel rods and coolant holes led to low or negative temperature coefficients. For a prismatic fuel block design with 54-7LiH moderator rods, the isothermal temperature coefficient of reactivity (Isothermal TRC), with simultaneous changes in the fuel (F), graphite (G), hydrogen (H), and coolant (C) temperatures, ranges between −0.159 mk/K and −0.054 mk/K, depending on the operating temperature and fuel burnup. Such alternative FHR-SMR fuels could achieve a single-batch core life of ∼10 years with low enriched uranium (LEU) fuel, and ∼45 years with HALEU, in a 350-MWth reactor core.

A Novel Core Design with Movable Moderator for a Fluoride Salt–Cooled High-Temperature Reactor

The Fluoride salt–cooled High-temperature Reactor (FHR) is a Generation IV reactor concept that can operate under near atmospheric pressure circumstances and further enhance inherent safety. In this study, an FHR core design with 165 MW of thermal output [MW(thermal)] is proposed. The reactor core employs tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks as the basic fuel form, FLiBe [lithium-beryllium fluoride (2 7LiF-BeF2)] as the primary coolant, and a three-batch fuel cycle scheme. Sensitivity analyses on various parameters were performed to optimize the cycle length and neutronic parameters. The fuel cycle of this core design was evaluated in detail from four aspects: cycle length, power peaking factor (PPF), discharge burnup, and temperature coefficient. It was found that a larger fuel channel pitch would have a relatively harder neutron spectrum and yield a relatively longer cycle length, lower PPF, and better fuel temperature coefficient and moderator temperature coefficient (MTC). In addition, burnable poison (BP) (Er2O3) can effectively reduce PPF, hold down the multiplication factor, and more importantly it can improve the MTC. The preliminary design of control blades is also presented in this paper. Furthermore, on the basis of the proposed 165-MW(thermal) core, we propose a novel core design that incorporates “fuel inside radial moderator (FIRM)” assemblies, movable moderator, and movable BP. This new design can extend the fuel cycle length by approximately 45 days for an 18-month fuel cycle. In addition, improvements were also found in PPF, discharge burnup, and temperature coefficients.

Numerical Investigations of Molten Salt Pump Journal Bearings Under Hydrodynamic Lubrication Conditions for FHRs

Fluoride salt-cooled, High-temperature Reactors (FHRs), featuring particle fuel, graphite moderator, and molten fluoride salt coolant, are used for electricity generation and process heat applications. The primary loop of an FHR is a closed loop that operates slightly above the atmospheric pressure with the fluoride salt temperature over 600°C. Reliable high-temperature molten salt pumps are critical to the successful deployment of FHRs. To stabilize rotating shafts and reduce the associated friction coefficients, well-designed bearings are required for molten salt pumps. Therefore, it is necessary to investigate the detailed hydrodynamic performance of bearings under high-temperature molten salt conditions. In this study, a computational fluid dynamics software package, i.e., STAR-CCM+, was used to predict the performance of fluoride salt–lubricated bearings. The numerical models were verified and validated respectively based on an analytical solution derived from the Reynolds equation and experimental data published in the literature. Good agreement was observed between the simulation results and the analytical solution and experimental data with a maximum relative discrepancy of less than 5%. The validated numerical model was then employed to predict the pressure distributions, applied static loads, and power losses of high-temperature fluoride salt–lubricated bearings with various Sommerfeld numbers. In addition, a parametric analysis was performed to investigate the influence of the axial and helical grooves of bearings on applied static load and power loss. It is found that under the same salt lubrication conditions, the bearings with helical grooves and axial grooves respectively yield 20% off and 14% off power loss compared with the bearing without grooves.

Tritium Content and Chemical Form in Nuclear Graphite from Molten Fluoride Salt Irradiations

Advanced reactor applications that use a molten fluoride salt coolant and graphite moderator are under consideration as next-generation energy technologies. For molten salts with lithium or beryllium, such as flibe (2LiF-BeF2), the production of tritium from neutron irradiation is a significant technical challenge. Understanding the expected quantities and mechanisms for tritium retention in graphite is important for designing tritium management strategies in these advanced reactors. In this work, the tritium content of IG-110U graphite from a 2013 in-core flibe irradiation experiment was measured by leaching in water and thermal desorption. Five total samples were tested, with an average measured tritium content per salt-contacting surface area of 3.83 ± 0.25 Ci/m2. The tritium measured from the thermal desorption experiments was primarily in a water-insoluble form. Compared to the overall tritium generation during the irradiation, the total amount of retention in graphite predicted by the desorption measurements is significant.

Scaling Methodology for Integral Effects Tests in Support of Fluoride Salt–Cooled High-Temperature Reactor Technology

The capability to validate integral transient response models is of critical importance for licensing new reactor designs. The Kairos Power testing program has developed a methodology to design scaled experiments that predict the thermal fluid behavior of the Kairos Power Fluoride Salt–Cooled High-Temperature Reactor (KP-FHR). Such experiments will be used as part of the assessment base of evaluation models supporting KP-FHR safety analysis. The Hierarchical Two-Tiered Scaling (H2TS) methodology was selected for Kairos Power scaling efforts that will be applied to integral effects tests (IETs) for system-level testing. In this paper, the scaling methodology is presented for thermal fluid IETs that will model the KP-FHR primary heat transport system under normal operations and transient conditions involving transition to natural circulation. This paper provides a basis for using surrogate fluids for testing that requires the thermal fluid performance of the KP-FHR primary coolant lithium fluoride–beryllium fluoride [2LiF/BeF2 (Flibe)] to be replicated. Kairos Power intends to use heat transfer oil as a surrogate fluid for Flibe in thermal fluid IETs. This paper demonstrates that this class of surrogate fluids is an acceptable substitute for Flibe salt for some types of scaled IETs and that the principal thermal fluid properties can be properly scaled with minor distortions over the range of conditions expected for both normal and off-normal operating conditions of the KP-FHR.

A stylized 3D Advanced High Temperature Reactor (AHTR) benchmark problem

A set of Stylized 3D Advanced High Temperature Reactor (AHTR) benchmark problems in full core and single fuel assembly configurations is developed in this paper. The configurations include the lower support plate, the bottom reflector, the fuel zone (AHTR assemblies), the top reflector, and the upper support plate. The benchmark problems retain the multiple heterogeneities and other important neutronics features such as the detailed geometric and material distributions of the Tristructural-Isotropic (TRISO) fuel and burnable poison particles, the fluoride salt coolant, the graphite moderator, and the reflectors. Monte Carlo results are presented for the benchmark problem in both the uncontrolled and controlled single assembly configurations. These benchmark problems can be used for evaluating the performance of neutronics codes.

Monte Carlo modeling and simulations of AHTR fuel assembly to support V&V of FHR core physics methods

The Fluoride Salt Cooled High Temperature Reactor (FHR) is a liquid salt-cooled reactor utilizing TRISO-based fuel. Its low operating pressure and large thermal margin offer attractive safety. However, its complex core design, in particular the double heterogeneity introduced by TRISO particles, as well as the presence of several moderating materials, require careful verification and validation (V&V) of the core physics modeling and simulation (M&S) methods. Due to their ability to accurately simulate complex geometries and utilize continuous-energy nuclear data, Monte Carlo methods are used in this study to evaluate core physics phenomena and provide reference solutions for verification of lower fidelity methods. This work is focused on the FHR design with hexagonal fuel elements consisting of fuel plates (“planks”), i.e., the Advanced High-Temperature Reactor (AHTR) developed by ORNL. The pebble-bed FHR design, PB-FHR, is not directly considered in this work, although the results of this study are relevant for PB-FHR as well. The SERPENT Monte Carlo code is selected and utilized to perform core physics analyses of an FHR fuel assembly. The issues investigated include detailed fission distribution, quantification of certain approximations, study of neutron spectra in different material regions, and impact of graphite density, temperature, and scattering matrix data. This study contributes to V&V of modeling and simulation methods for FHR.

The feasibility research of thorium breeding using fluoride salt as a fast reactor coolant

Breeder reactors are considered as a unique tool for fully exploiting natural resources. Fast breeder reactors based on thorium fuel can enhance inherent safety. Fluoride salt has good performance as a coolant in high-temperature nuclear systems. However, there is some doubt about the fuel breeding ability using fluoride salt coolant for fast spectrum due to its moderating ability. The aim of this study was to choose a proper fluoride salt mixture for Liquid-salt-cooled Solid-fuel Fast Reactor (LSFR) based on thorium-uranium fuel and give parametric studies to provide a design window for flexible self-sustaining core design. Infinite assembly model was used to analyze the salt selection from five candidate fluorides for fast spectrum as coolant. Combining neutron balance analysis with linear least squares fitting method based on 0D model, parametric studies at the neutron balance equation unique solution with burn-up for several parameters such as fuel volume fraction, removing fission gases process, total neutron losses and power density were presented in this paper. It was found that BeF2-NaF was a promising coolant in the five candidate fluoride salt mixture. This study proved that the design of a self-sustaining core for fluoride-salt-cooled fast breeder based on thorium fuel is achievable. A design window was found in the definition of a self-sustaining core for various fuel volume fractions and neutron loss fractions. Core design and fuel management strategy will be given in future.

Use of Nitrogen Trifluoride To Purify Molten Salt Reactor Coolant and Heat Transfer Fluoride Salts

The molten-salt-cooled nuclear reactor is one of the Generation IV reactor types. One of the challenges of implementing this reactor is purifying and maintaining the purity of the various molten fluoride salts that will be used as coolants. The method used for Oak Ridge National Laboratory’s molten salt experimental test reactor was to treat the coolant with a mixture of H2 and HF at 600 °C. In this Article, we evaluate thermal NF3 treatment for purifying molten fluoride salt coolant candidates based on nitrogen trifluoride’s (1) past use to purify fluoride salts; (2) other industrial uses; (3) commercial availability; (4) operational, chemical, and health hazards; (5) environmental effects and environmental risk management methods; (6) corrosive properties; (7) thermodynamic potential to eliminate impurities that could arise due to exposure to water and oxygen. Our evaluation indicates that nitrogen trifluoride is a viable and safer alternative to the previous method.

Basis for Fluoride Salt–Cooled High-Temperature Reactors with Nuclear Air-Brayton Combined Cycles and Firebrick Resistance-Heated Energy Storage

The fluoride salt–cooled high-temperature reactor (FHR) with a nuclear air-Brayton combined cycle (NACC) and firebrick resistance-heated energy storage (FIRES) is a new reactor and plant concept. The development of a new reactor is a large undertaking that requires a strong commercial case and a strong basis for government support to enable commercialization. The goals are (1) increase plant revenue by 50% to 100% relative to base-load nuclear plants with capital costs similar to light water reactors, (2) enable a zero-carbon nuclear renewable electricity grid, (3) no potential for major fuel failure and thus no potential for major radionuclide off-site releases in a beyond-design-basis accident, and (4) offer a pathway to more advanced energy systems. The approaches to achieve those goals are described herein.The FHR uses liquid-salt coolants originally developed for molten salt reactors (MSRs) where the fuel is dissolved in the coolant. However, in the FHR the fuel is not dissolved in the coolant. Instead, the FHR uses graphite-based high-temperature gas-cooled reactor (HTGR) coated-particle fuel. This combination enables delivering heat to the power cycle between 600°C and 700°C that, in turn, enables the FHR to couple to NACC. Using an air-Brayton power cycle enables the FHR to operate as a base-load reactor and produce added electricity in a peaking mode with the addition of auxiliary heat (natural gas, stored heat, or hydrogen). The auxiliary heat-to-electricity production is a thermodynamic topping cycle with efficiencies of 66%. Because an FHR with NACC is more efficient in converting natural gas into peak electricity than a stand-alone natural gas plant (60%), it can economically compete with a natural gas plant for peak electricity production because it uses less fuel. NACC can also incorporate FIRES heat storage that enables buying electricity at low prices to later sell electricity at high prices. For every 100 MW(electric) of base-load capacity, the station output can vary from minus several hundred megawatts to +242 MW(electric) because of the capabilities of the NACC with FIRES. The FHR can be built in different sizes.The use of high-temperature liquid-salt cooling and coated-particle fuel enables near-term reactor designs where large-scale fuel failure cannot occur, and thus, large-scale off-site releases of radionuclides cannot occur. Under extreme accident conditions decay heat can be passively conducted to the environment at temperatures below fuel failure temperatures and thus avoid the potential for large-scale radionuclide releases.The FHR is the gateway technology to several advanced salt-cooled reactor systems that have additional capabilities. The MSR can be designed to enable breeding and actinide waste burning. New high-magnetic-field fusion systems may require liquid-salt cooling. The FHR would provide the required salt and power cycle technology for these advanced reactors. There are significant development challenges. The United States has a competitive advantage in developing the FHR because it leads in gas turbine technology, high-temperature materials, and HTGR fuels.

Thermal-Hydraulic Analyses of Transportable Fluoride Salt–Cooled High-Temperature Reactor with CFD Modeling

The technology for the 20-MW(thermal) Transportable Fluoride Salt–Cooled High-Temperature Reactor (TFHR) is proposed by Massachusetts Institute of Technology for off-grid applications such as Antarctic bases and remote mining sites. The preliminary thermal-hydraulic analyses and improvements based on a 1/12th full-core model were performed using three-dimensional computational fluid dynamics (CFD). A benchmark study was conducted by comparing the CFD results against empirical correlations and experimental data obtained by Cooke, Silverman, and Grele. In the 1/12th full-core analysis, three practical considerations that may challenge the TFHR temperature limits are evaluated as bounding analysis. These include (1) helium gap between fuel compact and graphite block, (2) thermal conductivity degradations of graphite matrix due to neutron irradiation, and (3) full-core scale power distribution obtained from neutronic calculations. These design considerations lead to insufficient margin between the normal operating condition and the predefined thermal limits. In this context, additional design features are implemented to improve the thermal-hydraulic safety of the TFHR. First, bypass flow in the interstitial gaps between the active core and the reflector is found capable of reducing the temperature peaks at the core periphery. Second, improvements of the flow distribution from the central downcomer to individual coolant channels enable a higher mass flow rate to the regions with compromised cooling access. Overall, thermal-hydraulic performance was significantly improved with a fuel temperature margin from 10 to 150 K and a coolant temperature margin from 16 to 160 K, as well as the more uniform temperature distribution across the reactor core. Furthermore, thermal-hydraulic safety can be maintained at a 20% overpower operating condition [i.e., 24 MW(thermal)]. Overall, this study provides an engineering basis for the TFHR thermal-hydraulic design to improve its safety margin.

Design Summary of the Mark-I Pebble-Bed, Fluoride Salt–Cooled, High-Temperature Reactor Commercial Power Plant

The University of California, Berkeley (UCB), has developed a preconceptual design for a commercial pebble-bed (PB), fluoride salt–cooled, high-temperature reactor (FHR) (PB-FHR). The baseline design for this Mark-I PB-FHR (Mk1) plant is a 236-MW(thermal) reactor. The Mk1 uses a fluoride salt coolant with solid, coated-particle pebble fuel. The Mk1 design differs from earlier FHR designs because it uses a nuclear air-Brayton combined cycle designed to produce 100 MW(electric) of base-load electricity using a modified General Electric 7FB gas turbine. For peak electricity generation, the Mk1 has the ability to boost power output up to 242 MW(electric) using natural gas co-firing. The Mk1 uses direct heating of the power conversion fluid (air) with the primary coolant salt rather than using an intermediate coolant loop. By combining results from computational neutronics, thermal hydraulics, and pebble dynamics, UCB has developed a detailed design of the annular core and other key functional features. Both an active normal shutdown cooling system and a passive, natural-circulation-driven emergency decay heat removal system are included. Computational models of the FHR—validated using experimental data from the literature and from scaled thermal-hydraulic facilities—have led to a set of design criteria and system requirements for the Mk1 to operate safely and reliably. Three-dimensional, computer-aided-design models derived from the Mk1 design criteria are presented.

Hydrogen Diffusion and Trapping in Fluoride Salt Melts and Graphite Electrodes

This paper presents an approach currently under development at University of Wisconsin Madison to study tritium transport in high temperature fluoride salt systems and absorption in graphite. This work has applicability to tritium management in Fluoride Salt Cooled High Temperature Nuclear Reactors (FHRs). The applicability of the Double Potential Step Method (DSPM) and Electrochemical Impendence Spectroscopy (EIS) originally developed for the study of hydrogen transport in metals from water solutions is discussed for the salt-graphite system. Mechanistic models for hydrogen transport and trapping in graphite are reviewed. The experimental setup and electrochemical cell used with 2LiF-BeF2 (flibe) salt at 500-700°C is described. The design and preliminary characterization data for a Ni/Ni(II) flibe thermodynamic reference electrode are presented.

Thermal-hydraulic analysis of a fluoride-salt-cooled pebble-bed reactor with CFD methodology

The Fluoride Salt Cooled High Temperature Reactor (FHR) is an innovative concept reactor that inherits the technical foundation and advantages of the six optional generation-IV reactors and pressurized water reactors, which is mainly in process in both China and the United States. In this paper, the porous and realistic modeling approaches are adopted to analyze the thermal hydraulic characteristics of a FHR core and a unit segment of pebbles in the core respectively. The distributions of temperature and pressure of the fluoride salt, as well as the reflector temperature profile, are obtained using the porous model. The detailed local flow and heat transfer are investigated by the realistic modeling method for the locations which may have the maximum coolant temperature based on the results of the porous model. The profiles of temperature, velocity, pressure and Nusselt number (Nu) of the coolant on the surface of the pebble are also obtained and analyzed. Numerical results showed that the flow field between the fuel pebbles is complex including secondary flow and back-flow phenomenon, which are hard to measure by experiments. This work can provide useful information for the experimental and mechanism research of FHRs.

Thermal hydraulics analysis of the Advanced High Temperature Reactor

The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

Spent Nuclear Fuel and Graphite Management for Salt-Cooled Reactors: Storage, Safeguards, and Repository Disposal

The fluoride salt–cooled high-temperature reactor (FHR) is a new reactor type that combines the graphite-matrix coated-particle fuel and graphite moderator from high-temperature gas-cooled reactors (HTGRs) with a clean liquid fluoride salt coolant. No FHR has yet been built. The proposed fuel cycle is a once-through fuel cycle—essentially identical to that of HTGRs. There is the option of adopting closed fuel cycles. Relative to light water reactor (LWR) spent nuclear fuel (SNF), all graphite-matrix coated-particle SNFs share the common characteristics of superior proliferation resistance and long-term performance as a waste form in a geological repository. The allowable HTGR and FHR SNF storage temperatures are much higher than allowable LWR SNF storage temperatures. These SNF characteristics are (a) a consequence of the high-temperature fuel form with a graphite matrix and SiC coating of the fuel microspheres and (b) to a first-order approximation independent of the reactor type in which the fuel is used.There are differences. The FHR reactor core power density is four to ten times higher than in an HTGR, so the short-term decay heat of the SNF per unit volume upon discharge is four to ten times higher. The volume of FHR SNF is one-half to one-third that of an HTGR per unit energy output because (a) the salt provides some neutron moderation thus reducing the carbon-to-uranium ratio of the fuel and (b) the economic optimization with higher power densities increases the fuel loading. The FHR SNF volume is about four times that of a LWR per unit of electricity. The coolant generates significant tritium that is partly absorbed by the graphite and can be partly desorbed at higher temperatures. Last, any residual solid salt coolant with the SNF at low temperatures can undergo radiolysis with the potential generation of fluorine gas. The presence of the salt coolant on the SNF and graphite moderator will require treatment, removal of residual coolant salt, or demonstration that the small quantities of radiolysis products of frozen salt do not impact long-term performance of storage or disposal facilities.

ICONE23-2171 Design Considerations for Insulation System for Fluoride Salt Cooled High Temperature Reactor under Beyond Design Basis Accident Conditions

ICONE23-2171 Design Considerations for Insulation System for Fluoride Salt Cooled High Temperature Reactor under Beyond Design Basis Accident Conditions

Analysis of the Limiting Safety System Settings of a Fluoride Salt–Cooled High-Temperature Test Reactor

The fluoride salt–cooled high-temperature reactor (FHR) is an advanced reactor concept, which uses high-temperature TRISO fuel with a low-pressure liquid salt coolant. The design of a fluoride salt–cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress in both China and the United States. An FHTR based on a pebble bed core design with coolant temperature 600·C to 700·C is being planned for construction by the Chinese Academy of Sciences” Thorium Molten Salt Reactor Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal-hydraulic licensing analysis of an FHTR using SINAP”s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting safety system settings (LSSSs) considering uncertainties for forced convection and natural convection are obtained. The LSSS power and coolant outlet temperature, respectively, are 24.83 MW and 720·C for forced convection and 1.19 MW and 720·C for natural convection. The maximum temperature for the structural materials of 730·C is the most limiting constraint of the FHTR design.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Corrosion of oxide dispersion strengthened iron–chromium steels and tantalum in fluoride salt coolant: An in situ compatibility study for fusion and fusion–fission hybrid reactor concepts

Primary candidate classes of materials for future nuclear power plants, whether they be fission, fusion or hybrids, include oxide dispersion strengthened (ODS) ferritic steels which rely on a dispersion of nano-oxide particles in the matrix for both mechanical strength and swelling resistance, or tantalum alloys which have an inherent neutron-induced swelling resistance and high temperature strength. For high temperature operation, eutectic molten lithium containing fluoride salts are attractive because of their breeding capability as well as their relatively high thermal capacity, which allow for a higher average operating temperature that increases power production. In this paper we test the compatibility of Flinak (LiF–NaF–KF) salts on ODS steels, comparing the performance of current generation ODS steels developed at Kyoto University with the commercial alloy MA956. Pure tantalum was also tested for comparative purposes. In situ data was obtained for temperatures ranging from 600 to 900 °C using a custom-built high temperature electrochemical impedance spectroscopy cell. Results for ODS steels show that steel/coolant interfacial resistance increases from 600 to 800 °C due to an aluminum enriched layer forming at the surface, however an increase in temperature to 900 °C causes this layer to break up and aggressive attack to occur. Performance of current generation ODS steels surpassed that of the MA956 ODS steel, with an in situ impedance behavior similar or better than that of pure tantalum.