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DOE OSTI2023

SOFE2023 Presentation: "Tritium Safety and Sensitivity Analysis of Tritium Extraction eXperiment (TEX) PbLi Loop Using MELCOR-TMAP"

The efficient extraction of tritium from the breeder blanket (BB) in a fusion reactor is vital for maintaining the tritium economy for a sustainable and self-sufficient fusion power plant. The vacuum permeator (VP) design is one of several promising technology platforms for extracting tritium from the BB with the advantages of being capable of operating continuously at high temperatures with the aim of achieving high tritium extraction efficiencies. The Tritium Extraction eXperiment (TEX), under construction at the Safety and Tritium Applied Research (STAR) Facility at Idaho National Laboratory (INL), is designed to test the viability of VP technology by introducing atomic hydrogen (H, D, and T) from a source permeator (SP) into a forced convection lead lithium eutectic (PbLi) loop and determining the hydrogen extraction rates from the vanadium VP. Presented here is a computational analysis of the hydrogen inventory and permeation behavior in TEX using the hydrogen transport and accident analysis code MELCOR-TMAP developed at INL. The results of a sensitivity analysis based on MELCOR-TMAP simulations of hydrogen extraction efficiencies from the TEX test section using varying PbLi flow rates, hydrogen source rates from the SP, and material property uncertainties are also reported. Operation scenarios for TEX in consideration of the analyses based on results from MELCOR-TMAP are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Wetting Properties of Molten Salts

Wettability is a key piece of information in understanding the behavior of molten salts at material interfaces. Wettability is most easily characterized by contact angle which is measurable using the sessile drop method. A modified sessile drop method was used to measure the contact angle of the LiCl-KCl eutectic at varying temperatures on Nickel 200, Hastelloy C276, Glassy Carbon, and MACOR machinable ceramic. The contact angle was used with existing surface temperature data to find the work of adhesion at various temperatures. Finally, problems encountered during the experiment are listed and possible improvements are provided.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Power Electronics Materials and Bonded Interfaces - Reliability and Lifetime

High temperature operation of wide bandgap devices continue to be a challenge for the power electronics packages. Thermal performance and reliability are important factors that determine the viability of a bonded interface for operation at high temperatures. In this presentation, we present the technical approach and key results from the research on sintered silver, transient liquid phase alloy, and polymeric materials. A lifetime prediction model that incorporates the thermomechanical behavior of sintered silver at 200C was developed. The copper-aluminum transient alloy completed 350 thermal cycles from -40C to 200C and little increase in the defect level was observed. In addition to material research, we initiated a time-series analysis on the scanning acoustic microscope images of eutectic solder to explore statistical forecasting methods and machine learning techniques. Initial results that report the accuracy of a few different statistical models are presented.

ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2013

A physics study for negative void reactivity in compact supercritical CO{sub 2}-cooled fast reactor

A compact S-CO{sub 2}-cooled fast reactor which has negative Coolant Void Reactivity (CVR) has been investigated. A negative CVR is important for the gas cooled fast reactor as an inherent safety mechanism to prevent the sudden positive reactivity insertion when the loss of coolant accident happens. An alternative solution to reduce the CVR is investigated in this study by using O-17 instead of O-16 in UO{sub 2} fuel. By using O-17 in the fuel, it is found that the CVR can even be negative. Impacts of the radial reflector on the CVR are also evaluated for the small SCO{sub 2} cooled fast reactor in this study. We have considered a pure lead (Pb) reflector and a lead magnesium eutectic (LME) reflector as alternative radial reflectors of the S-CO 2-cooled fast reactor. It has been shown that, with the LME radial reflector, the CVR can be negative, while the pure lead reflector provides a slightly positive CVR. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Methods of forming a metal coated article

A method of forming a metal coated article, comprises forming a metal halide in a molten salt plating bath at a first temperature, wherein forming the metal halide in the molten salt further comprises forming at least one functional metal halide electrolyte; and forming at least two auxiliary metal halide electrolytes at eutectic conditions; increasing the first temperature to a second temperature; forming a plated metal coating from the at least one functional metal halide electrolyte onto a thermally conductive substrate; and introducing at least one of deuterium and tritium into the plated metal coating.

Tripathy, Prabhat K.↗
DOE OSTI2013

Development of pyro-processing technology at CRIEPI for carving out the future of nuclear fuel cycle

Pyro-processing has been attracting increasing attention as a promising candidate as an advanced nuclear fuel cycle technology. It provides economic advantage as well as reduction in proliferation risk and burden of long live radioactive waste, especially when it is combined with advanced fuels such as metallic or nitride fuel which gives excellent burning efficiency of minor actinides (MA). CRIEPI has been developing pyro-processing technology since late eighties with both domestic and international collaborations. In the early stage, electrochemical and thermodynamic properties in LiCl-KCl eutectic melt, and fundamental feasibility of core technology like electrorefining were chiefly investigated. Currently, stress in the process chemistry development is also placed on supporting technologies, such as treatment of anode residue and high temperature distillation for cathode product from electrorefining, and so on. Waste treatment process development, such as studies on adsorption behavior of various FP elements into zeolite and conditions for the fabrication of glass-bonded sodalite waste form, are steadily improved as well. In parallel, dedicated pyro-processing equipment such as zeolite column for treatment of spent electro-refiner salt is currently in progress. Recently, an integrated engineering-scale fuel cycle tests were performed funded by Japanese government (MEXT) as an important step before proceeding to large scale hot demonstration of pyro-processing. Oxide fuels can be readily introduced into the pyro-processing by reducing them to metals by adoption of electrochemical reduction technique. Making use of this advantage, the pyro-processing is currently under preliminary evaluation for its applicability to the treatment of the corium, mainly consisting of (U,Zr)O{sub 2}, formed in different composition during the accident of the Fukushima Daiichi nuclear power plant. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling. In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Study on the electrochemical extraction of rare earth elements from FLINAK

Electrochemical behaviors of rare earth elements, such as NdF{sub 3}, GdF{sub 3}, SmF{sub 3}, YF{sub 3}, and EuF{sub 3}, were investigated in a LiF-NaF-KF (46.5-11.5-42.0 mol %, FLINAK, m. p. 454 Celsius degrees) solvent. The results indicated that it is possible to extract Nd, Gd and Y directly by electrochemical deposition since the reductions of those cations to metal are located in the electrochemical window of the FLINAK eutectic, while the reductions of Sm and Eu metal are out of the range of the medium. Subsequently electro-deposition of Nd was carried out with two kinds of cathodic materials, namely, an inert cathode, Pt, and a reactive electrode, Cu. The collected products were characterized by various techniques revealing that a Nd-rich product was obtained. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2013

Decommissioning and PIE of the MEGAPIE spallation target

A key experiment in the Accelerated Driven Systems roadmap, the MEGAwatt PIlot Experiment (MEGAPIE) (1 MW) was initiated in 1999 in order to design and build a liquid lead-bismuth spallation target, then to operate it into the Swiss spallation neutron facility SINQ at Paul Scherrer Institute. The target has been designed, manufactured, and tested during integral tests, before irradiation carried out end of 2006. During irradiation, neutron and thermo hydraulic measurements were performed allowing deep interpretation of the experiment and validation of the models used during design phase. The decommissioning, Post Irradiation Examinations and waste management phases were defined properly. The phases dedicated to cutting, sampling, cleaning, waste management, samples preparation and shipping to various laboratories were performed by PSI teams: all these phases constitute a huge work, which allows now to perform post-irradiation examination (PIE) of structural material, irradiated in relevant conditions. Preliminary results are presented in the paper, they concern chemical characterization. The following radio-nuclides have been identified by γ-spectrometry: {sup 60}Co, {sup 101}Rh, {sup 102}Rh, {sup 108m}Ag, {sup 110m}Ag, {sup 133}Ba, {sup 172}Hf/Lu, {sup 173}Lu, {sup 194}Hg/Au, {sup 195}Au, {sup 207}Bi. For some of these nuclides the activities can be easily evaluated from γ-spectrometry results ({sup 207}Bi, {sup 194}Hg/Au), while other nuclides can only be determined after chemical separations ({sup 108m}Ag, {sup 110m}Ag, {sup 195}Au, {sup 129}I, {sup 36}Cl and α-emitting {sup 208-210}Po). The concentration of {sup 129}I is lower than expected. The chemical analysis already performed on spallation and corrosion products in the lead-bismuth eutectic (LBE) are very relevant for further applications of LBE as a spallation media and more generally as a coolant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Very high temperature measurements: Applications to nuclear reactor safety tests; Mesures des tres hautes temperatures: Applications a des essais de surete des reacteurs nucleaires

This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Control blade degradation test under temperature gradient in steam atmosphere

The core of a boiling water reactor (BWR) consists of an array of fuel assemblies with cross-shaped control blades located between these assemblies. The typical control blade consists of a control blade sheath made of 316L-type S.S., a center pole made of 316L-type S.S., and control rods. In order to understand the degradation process of the control blade channel, control blade degradation tests were performed under various atmospheres with a temperature increase condition and in the presence of a temperature gradient. A test device was built. Four carbon heaters are set on both sides at the upper and the middle parts of the test device. The bottom ceramic part of the test device is cooled by water. Heating in the upper region and cooling at the bottom of the test piece give rise to the temperature gradient in the test piece. Temperatures of the test piece are measured by five W-Re type thermocouples placed close to the test piece. Steam carried by the argon gas is supplied from the bottom of the device and flows to the top inside a square carbon muffle in which the test piece is set. Three windows for visual observations are installed, with one located above the furnace and the others located at the back of the furnace. In the steam atmosphere with a steam flow rate of 0.0125 g/s/rod, the control blade, channel box, and fuel rods were degraded, especially at the upper part of the test piece. This result was similar to that observed in the argon atmosphere test. However, the observed degradation was rather different at a higher steam flow rate of 0.0417 g/s/rod. At the upper part of the test piece, only the control blade was degraded preferentially and did not react with the channel box wall. In contrast, the eutectic reaction of the stainless steel/boron carbide melt and zircaloy occurred at the lower part by coexisting with the relocated melt from the top part. These observations suggested the existence of a threshold condition for the control rod degradation between 0.0125 and 0.0417 g/s/rod which is significantly affected by the thickness of the oxide layer on zircaloy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermo mechanical analysis of fully ceramic microencapsulated fuel during in-pile operation

A proposed fuel for improved performance in LWRs (Light Water Reactors) is FCM (fully encapsulated micro-ceramic) fuel concept that involves TRISO (Tri-structural-Isotropic) fuel particles embedded in a NITE (Nano Infiltrated Eutectic) silicon carbide matrix. TRISO fuel particles contain a spherical fuel kernel that will likely range from about 500 to in excess of 800 μm in diameter. The kernel and buffer layer are then typically coated with 3 isotropic layers consisting of a dense inner pyrolytic carbon (IPyC), a silicon carbide (SiC) layer, and an outer layer of dense, pyrolytic carbon (OPyC). These layers are typically around 40 μm thick, but can range from 20 to 60 μm. The TRISO particle packing fraction in the NITE-SiC matrix is expected to be 40-45 vol.%. The release of radioactivity into the coolant is dependent on the integrity of the silicon carbide layer of the TRISO particles and the NITE-SiC matrix. BISON, a code under development by Idaho National Laboratory, has been built on the Multi-physics Object Oriented Simulation Environment (MOOSE). MOOSE is a massively parallel, finite element computational system that uses a Jacobian-free, Newton-Krylov (JFNK) method to solve coupled systems and non-linear partial differential equations. A single particle comparison to fuel performance codes such as PARFUME was conducted and demonstrated that BISON is able to simulate the required physics. FCM pellets embedded with homogenous particles were simulated. It was found that the interior matrix temperature was predicted to be considerably lower than for an equivalent UO{sub 2} pellet. The swelling dependence on temperature created a compressive region around the perimeter of the pellet. However, at fluences above 6.10{sup 25} neutrons/m{sup 2}, the swelling of the pyrolytic carbon might cause significant tensile stresses and possible cracks near the surface of the pellet. This suggests that the pyrolytic carbon swelling may be an important factor in the lifetime of this fuel type.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Decontamination of Irradiated Nuclear Graphite Using High Temperature Molten Salt - 17447

Irradiated graphite is one of the most significant, large volume waste streams in the UK. After shut down of gas cooled reactors there will be ∼96,000 tonnes of nuclear graphite arising from pressure vessels, sealed or unsealed stacks, temporary surface storage and in silos, which may account for up to 30% by ILW volume of any future UK geological disposal facility. High temperature molten salt treatment (HMST) could be considered as a partitioning process of the activation and fission products from irradiated graphite. The main objective of the research is to optimize a specific graphite treatment technology compatible with older, current and future reactors to provide a safe and effective process to decontaminate graphite and reduce the waste inventory. In order to reach that purpose principal radionuclides contained in irradiated samples from Magnox reactors were investigated by germanium (Ge) gamma spectrometry. The next key tasks were to employ treatment at 450 deg. C in LiCl-KCl eutectic which included the following procedures: electrochemical cleaning of the salt, initial cyclic voltammetry (CV) of graphite followed by several steps of chrono-potentiometry (CP) with different current applied and CV was taken after each step. Once the treatment was established, experiments investigating the electro-refinement of the resultant salt mixtures were conducted as well as the comparison of gamma spectroscopy results of graphite before and after the treatment providing the result of total activity reduction of around 60%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Molten Salt Oxidation of Ion Exchange and Oils in Carbonate Salts and Evaporator Residue - 16531

Molten Salt Oxidation is a well-studied and robust technology. Recently, renewed interest in the technology has been seen due to limitations on the use of incineration for hazardous and radioactive waste streams. The nuclear research center at Rez, near Prague, has been investigating a bench scale molten salt oxidation system to provide insight for the design of a pilot scale system to condition spent ion exchange resin and scintillator oils. The use of both an eutectic K/Na/Li carbonate salt and a simulated residue salt for liquid waste obtained from evaporated primary coolant of nuclear reactors was examined. In these studies the usable capacity of the salts were examined. Materials doped with surrogate radioisotopes were processed and the capture efficiency was measured. Corrosion of the reactor vessel is of significant engineering concern. Corrosion coupon tests were performed for Inconel HX and stainless steel. These tests provided corrosion rates to assist in design of pilot scale systems. While the bench scale reactor's detention time was too short to provide complete combustion, the values of emissions were tracked and showed an increase in the combustion efficiency at temperatures greater than 750 deg. C. A two part reactor has been designed and work is planned to continue to optimize the system. Additionally, further work has examined the final conditioning of salt which has been saturated with radionuclides after processing organic waste streams A conceptual pilot plant has been designed to solidify radionuclide saturated salt into a form acceptable for repositories according to EU directives. Saturated salt is milled to a powder and immobilized in a matrix with a high percentage of waste loading. Both geo-polymer and polysiloxane matrices have been methods for low temperature immobilization. Conditioning in this manner decreases operating and capital costs when compared to current conditioning technologies and procedures utilized for high volume reduction factors of final waste forms. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Methods of recovering an elemental rare earth metal, and methods of forming a rare earth metal

A method of recovering an elemental rare earth metal comprises placing a rare earth-containing material comprising a rare earth metal in a reaction solution comprising a reducing agent and a non-aqueous solvent comprising an ionic liquid or a eutectic mixture, reducing the rare earth metal with the reducing agent to form a metallic rare earth metal and cations of the reducing agent, transferring the cations of the reducing agent from the reaction solution to an electrochemical cell through an ion exchange membrane, and reducing the cations of the reducing agent in the electrochemical cell. Related methods of forming an elemental rare earth metal, and related systems are disclosed.

Case, Mary E.↗
DOE OSTI2016

Development and Optimization of Normal Pulse Voltammetry for Real Time Analysis of Electrorefiner Salt with High Concentrations of MgCl{sub 2} and UCl{sub 3}

Electro-refiners can accumulate large holdup of fissile actinides from used nuclear fuel including plutonium and will need to be well safeguarded in a commercial facility. Idaho National Laboratory has operated the Fuel Conditioning Facility, an engineering scale demonstration facility, with two electro-refiners for two decades using mass tracking models supplemented with destructive analysis for the requirements of domestic safeguards. This approach is known to be insufficient of meet international safeguards timeliness requirements. Since there are non-nuclear weapons states that are interested in commercial electro-refiners, there is a need to develop a real time concentration measurement technique. In addition to being useful for safeguards, such a real time measurement would be functional for maintaining optimal process control. Normal pulse voltammetry (NPV) is a powerful electrochemical analysis method that has been identified to have great potential for measuring concentrations of various ions in a given salt mixture. While cyclic voltammetry (CV), chronoamperometry (CA), and chronopotentiometry (CP) result in progressive growth of the working electrode surface area, NPV potentially can circumvent this problem via periodically stripping deposited metals from the electrode (WE) surface. With each successive pulse, the potential is dropped to a more negative level. The corresponding current is a summation of partial currents from each of the reducing species. The ultimate objective of this work is to develop correlations between limiting currents measured via NPV with known concentrations. This is potentially a more accurate method than CV and other methods, because it should keep the WE surface area from growing significantly. Previous work on using NPV for actinide concentration determination in eutectic LiCl-KCl has been reported by M. Iizuka et al. However, only low uranium concentrations were studied in that paper. The maximum concentration of uranium tested was 2.48 wt%. Limiting currents from NPV were found to be proportional to their concentrations up to about 1.7 wt%. They were unable to maintain double plateaus (U and Pu) at uranium concentrations above 1.69 w%. This issue was resolved during our high concentration study. This result represents a significant breakthrough for optimizing voltammetry for analysis of actinides in molten salt.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Development of SiC-based Cladding for Accident Tolerant Fuels

Fuels and core structures in current light water reactors (LWR's) are vulnerable to catastrophic failure in severe accidents. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Alternative materials to zirconium alloys with reduced oxidation kinetics are sought to enable enhanced accident-tolerant fuels and cores. Among the candidate alternatives, silicon carbide (SiC) - based materials, in particular continuous SiC fiber-reinforced SiC matrix ceramic composites (SiC/SiC composites), are considered to offer outstanding safety benefits. The primary objective of this work is to provide with updated overviews of the concepts, development status, and future prospect for the SiC-based accident-tolerant fuel (ATF) claddings. A systematic technology evaluation program plan was developed and updated to review the technologies needed, to identify and analyze critical gaps, and to lay out a road-map toward establishing key technologies. The technology review and the gap analysis in this exercise were based on an early study and community workshops. Many of the technical gaps identified are related with the three key feasibility issues: hydrothermal corrosion, mechanical failure, and fuel compatibility. Additional performance issues including accident-tolerance features and fission product retention during the normal operation have also been identified. The program plan was designed to systematically address the key gap issues and is formulated in a work breakdown structure. Simultaneously, the plan is being established to set a technical program for advancing the technological readiness levels of concept non-specific essential technologies in three top-level categories of Design and Failure, Environmental Effects, and Off-normal Behavior. The execution of the program plan is in progress across the U.S. Department of Energy laboratories, industries, universities, and international collaborators. The critical issues for SiC composite-based fuel cladding concepts were found to be related to the performance in the normal operation. In particular, radiation-assisted hydrothermal corrosion and the gaseous fission product (FP) retention are identified to be the most critical issues. The primary ORNL concept pursues the use of dual-purpose coating on the SiC composite cladding wall to mitigate the corrosion and the gas permeation issues at the same time, whereas the alternative approaches such as the dense monolithic SiC protective layer combined with the water chemistry control may be feasible. Surface metallization through variations of physical vapor deposition and electroplating are being demonstrated. The FP gas retention is an issue associated with the cracking for the ceramic matrix composite cladding. To establish the probabilistic failure behavior, the relevant test methods were developed in ASTM Committee C28 on Advanced Ceramics. Toward the end-plugging technology development, various methods for joining SiC to SiC had been developed and tested for mechanical properties, irradiation effects, and environmental effects. Multiple methods including the pressureless transient eutectic SiC joining and CVD-based SiC joining were identified particularly promising. In addition, a multi-physics, three-dimensional finite element-based modeling is in progress toward an integrated simulation of the SiC composite-cladded fuel evolutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experimental Capability for Preparation and Purification of Molten Salts to Support Advanced Reactor Concepts, Nuclear Hybrid Energy Systems and Dynamic Grid Energy Storage Concepts

High-temperature multi-fluid, multi-loop test facility to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications is being developed at Idaho National Laboratory (INL). In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. The main objectives to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers at prototypical operating conditions, flow, and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for light water reactors at prototypical conditions. Fluoride salt coolants are eutectic binary or tertiary mixtures of fluoride salts with melting points in the range of 320 to 500 deg. C. Fluoride high temperature reactors have reactor outlet temperatures of 600 deg. C or higher for high-efficiency power generation or process heat applications. To address research needs related to salt coolants experimental capability has been added to study thermodynamic properties such as melting point temperature, heat capacity, and thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗