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At least 199 records · Page 11

Synthesis and characterization of uranium trichloride in alkali-metal chloride media

Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride – potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride – potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Finally, chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Synthesis and characterization of uranium trichloride in alkali-metal chloride media

Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride – potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride – potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.

D Herrmann, Steven↗

Experimental determination of the electrochemical properties of bismuth chloride in eutectic LiCl–KCl and LiCl–KCl–CaCl 2 molten salts

Bismuth has been investigated as a potential liquid electrode for molten salt electrorefining, but the electrochemical behavior of Bi 3+ /Bi redox couple has received scant attention in LiCl–KCl and LiCl–KCl–CaCl 2 eutectics. Electroanalytical techniques were used to determine properties of Bi 3+ /Bi. Diffusivities of 6.71 and 8.93 × 10 –6 cm 2 s –1 and standard apparent potentials of 0.1166 and 0.1187 V versus Ag/AgCl (4.75 mol%) were recommended for BiCl 3 in LiCl–KCl and LiCl–KCl–CaCl 2 eutectics, respectively, at 680 ± 3.1 K. The ICP-MS analysis of BiCl 3 content limited the accuracy of the results. Furthermore, the reliability of the results improved with a more thorough drying procedure for CaCl 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

First-of-a-Kind Fuel-bearing Molten Chloride Irradiation Experiment

As a dozen MSR developers in the U.S. work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have never been irradiated. Licensing and operating these reactors requires an understanding of (1) the source term, radiation chemistry, and gas generation; (2) unanticipated irradiation-induced corrosion effects; and (3) the impact of burnup on thermophysical properties, all of which can be deduced through irradiation testing. Idaho National Laboratory (INL) is ideally suited to leverage its expertise in chloride salt chemistry, as well as its Neutron Radiography Reactor (NRAD) facility in the Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC), to conduct the world’s first fuel-bearing chloride salt irradiation and fill in the knowledge gaps pertaining to salt chemistry under irradiation. Molten-salt Research Temperature-controlled Irradiation (MRTI) consists of a salt-containing capsule that is internally heated inside of a secondary containment. The experiment must rely on resistive heating to melt the salt before irradiation (to avoid the impact of radiolysis) and once the reactor is turned on and fission reactions commence in the salt, the resistive power can then be reduced. Salt-immersed thermocouples coupled to a controller allow for the heater power to be adjusted as needed to meet experimental objectives. The bulk of the experimental results will be achieved through Post-Irradiation Examination (PIE). At which point a range of different measurements are anticipated to assess the salt/plenum/wall composition, the capsule corrosion rate, and the evolution of salt properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Measure the effect of molten halide salt exposure on creep rupture lifetime

Recent resurgence in the research and commercial interests in molten salt reactors (MSRs) as a viable advanced reactor concept to achieve the short- and long-term climate goals has resulted in ongoing efforts to demonstrate their commercial potential. These are relying on a combination of the extensive legacy knowledge from the molten salt reactor experiment (MSRE) and relatively recent data on materials compatibility of structural materials of interest such as 316H in molten salts environments. However, there is a critical lack of data on the mechanical behavior of alloys of interest for MSRS such as 316H, 617 and 709 in molten fluoride (FLiNaK or FLiBe) or chloride (NaCl-MgCl 2 ) salts. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy (Ni-15Cr-7Fe wt.%) in the molten fluoride NaF-ZrF4-UF4 (50-46-4 mol.%) salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Creep tests for 316H were conducted with fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts tat 650°C/150 MPa while alloys 709 and 617 were tested with FLiNaK at 700C/158 MPa and 750C.146 MPa respectively. Baseline tests were conducted in air to assess the impact of the molten salts on the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

In-situ anodic precipitation process for highly efficient separation of aluminum alloys

Electrorefining process has been widely used to separate and purify metals, but it is limited by deposition potential of the metal itself. Here we report in-situ anodic precipitation (IAP), a modified electrorefining process, to purify aluminium from contaminants that are more reactive. During IAP, the target metals that are more cathodic than aluminium are oxidized at the anode and forced to precipitate out in a low oxidation state. This strategy is fundamentally based on different solubilities of target metal chlorides in the NaAlCl4 molten salt rather than deposition potential of metals. The results suggest that IAP is able to efficiently and simply separate components of aluminum alloys with fast kinetics and high recovery yields, and it is also a valuable synthetic approach for metal chlorides in low oxidation states.

36 MATERIALS SCIENCE↗

Low temperature sodium battery comprising an electrochemically active molten inorganic catholyte

A high-performance electrochemically active sodium molten salt catholyte enables a dramatic reduction in molten sodium battery operating temperature from near 300° C. to less than 120° C. As an example, stable electrochemical cycling was demonstrated in a high voltage (3.65 V) sodium battery comprising a sodium iodide-gallium chloride (NaI—GaCl 3 ) molten salt catholyte for over 8 months at 110° C. The combination of high voltage, stable cycling behavior, and practical current densities supported by a molten catholyte enables a new generation of transformative high performance, low temperature molten sodium batteries.

Spoerke, Erik D.↗

Measurement of partial vapor pressures of salt mixtures via combined horizontal transpiration and thermogravimetric analysis

A method combining thermogravimetric analysis (TGA) and horizontal transpiration with elemental analysis via inductively coupled plasma mass spectrometry or ion chromatography enabled calculation of partial pressures of individual salts in molten mixtures. TGA quantified total mass loss, while transpiration identified vapor-phase composition. Furthermore, two chloride (NaCl-MgCl 2 , NaCl-MgCl 2 + UCl 3 ) salts and one mixed halide (LiCl-LiF + Li 2 O) salt were analyzed at 750 °C and 550 °C, respectively. NaCl and MgCl 2 vapor pressures were 2.19–2.61 × 10 -4 atm and 2.47–2.48 × 10 -5 atm (dependent upon the identity of the invesitgated mixture); UCl 3 was 1.42 × 10 -7 atm. LiCl and LiF vapor pressures at 550 °C were 1.53 × 10 -6 and 6.32 × 10 -6 atm, respectively. Additionally, the TGA method was validated against values from the literature for unary LiCl and LiF.

36 MATERIALS SCIENCE↗

Molten Salt Electrochemical Reactors Outside of a Glovebox

Molten salt electrochemical reduction (MSER) offers an electrified method for the reduction, refining, and recycling of metals such as iron and metallurgical grade silicon at high efficiency and lower temperatures than traditional methods, which use temperatures >1000C and carbon-based reductants. However, salts appropriate for MSER are moisture and air sensitive, necessitating air-free methods of processing and testing which is often acomplished with a glovebox. Glovebox operations are cumbersome and difficult to troubleshoot reactor issues. We have developed a robust method of testing chloride and carbonate salts in molten salt reactors outside of a glovebox for silicon and carbon production and steel carburization. We highlight a practical guide for glovebox-free testing of molten salt electrochemical reactors to reduce the barriers to performing this type of work. Our aim is to enable other research groups to step into this field with simple and repeatable reactor configurations.

47 OTHER INSTRUMENTATION↗

Mechanistic transition of stainless steel 304L corrosion in MgCl 2 phase change material under thermal cycling

Magnesium chloride (MgCl 2 )-based molten salt phase change materials (PCMs) are promising candidates for thermal energy storage (TES) owing to their advantageous thermophysical properties, however, their high corrosivity in the molten state limits material compatibility. In this study, the corrosion behavior of 304?L stainless steel (SS304L) exposed to MgCl2 PCM was investigated as a function of thermal cycling (up to 9000 cycles) relevant to PCM-based TES systems. Microstructural changes and elemental redistribution at the alloy-salt interface under thermal cycling conditions were characterized using synchrotron-based X-ray fluorescence (XRF) and scanning electron microscopy coupled with energy dispersive X-ray spectroscopy (SEM-EDS). Results show that corrosion is governed by selective Cr-dissolution and elemental diffusion near the interface and grain boundaries. A clear transition in the dominant corrosion mechanism is observed with increasing thermal cycles. Surface corrosion is the dominant mechanism up to 1000 cycles, whereas intergranular corrosion becomes the primary degradation mode beyond 2000 cycles. These mechanistic changes are interpreted in terms of evolving interfacial chemistry, microstructure and the diffusion-driven redistribution of alloying elements.

36 MATERIALS SCIENCE↗

Progress Towards Synthesis of Uranium Chloride Fuel Salts Using Zinc Chloride

Reliable, scalable methods for producing high-purity actinide chloride salts are needed to support molten salt reactor fuel development and deployment. This report describes the continued development and demonstration of a bench-scale chlorination and purification apparatus using a zinc chloride-based method for synthesizing uranium chloride fuel salts. In this approach, uranium metal is chlorinated by ZnCl2 to produce LiCl-KCl-UCl3. Reaction with three aliquots of added uranium metal was used to generate a target uranium concentration of 30 wt %. While this concentration was chosen for initial testing of the apparatus and method, the final uranium concentration is not limited to 30 wt %. The zinc metal generated in the reaction forms an immiscible layer that was removed by volatilization at moderately high temperatures. Electrochemical measurements confirmed the removal of zinc and applied sensing methods indicated the uranium concentration to be approximately 25 wt %. These initial results demonstrate that the bench-scale chlorination apparatus is an effective platform for the synthesis and purification of uranium chloride salts using ZnCl2. This method shows promise for application to industry-relevant salt systems such as NaCl-UCl3. Further development is recommended to optimize reagent loading, zinc removal, and avoid possible U-Zn alloy formation.

Dulovic, Stephanie↗

Molten Chloride Reactor Experiment

MCRE will be the first fast spectrum, molten chloride reactor in operation. Small critical experiment to provide data for future reactors. - Collaboration between Southern Company Services, TerraPower, and Idaho National Laboratory (INL), as well as other partners. - Expected to be built and operated at the LOTUS testbed (ZPPR). - Authorized by the U.S. Department of Energy. - Technology development of salt synthesis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Measure Effects of Molten Halide Salt on Creep Rupture Lifetime for 316H, 709 and 617

Ongoing commercialization of different molten salt reactor concepts can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy in molten fluoride salt. This task evaluated the creep behavior of candidate structural materials 316H, 709 and 617 for advanced nuclear reactors under molten halide salt environments at 650, 700 and 750°C respectively. Creep tests were conducted in air, Ar, fluoride (FLiNaK) and chloride (NaCl-MgCl 2 ) salts. The molten salt environment caused a 50% reduction in creep rupture lifetime for 316H in FLiNaK and an 80% reduction in creep rupture lifetime in NaCl-MgCl 2 compared to air exposures. The molten fluoride salt environment resulted in almost a 90% and 50% reduction in creep rupture time for 709 and 617 irrespective of the cover gas environment (air and Ar). Significant differences in the compositional changes between the salt- and air-facing surfaces were observed in the alloy subsurface which might have potentially lead to formation of phases detrimental to the creep behavior.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Quality Ranking of Unary Chloride Salt Property Data Included in MSTDB-TP

Molten salt reactor developers rely on thermal property data to design, license and operate the reactors. The Molten Salt Thermal Database-Thermophysical Properties (MSTDB-TP) was established under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program and is managed by Oak Ridge National Laboratory to serve as a single source of thermophysical property values measured for a wide variety of molten salt systems for use by researchers, molten salt reactor developers, and regulators. These properties include density, viscosity and thermal diffusivity and conductivity. Published measurements of molten salt properties are lacking for many salts of interest and the data that are available are often inconsistent. This creates a challenge for MSR developers when determining which property values to use when designing their reactors. It is the purpose of this work to apply a consistent ranking system to all data entries that indicates the quality of property values listed in the database. These rankings will be the technical basis for down-selections by the database developers and alert users about the quality of the available property values. MSTDB-TP collects all available property data and indicates preferred data sets or correlations. However, all available data sets are included in the database. Quality assessments and rankings are being applied to data in MSTDB-TP to provide an indication of the quality of each data set independent of consistency with other data. Previous reports detailed the ranking system that was followed and assessments of unary fluoride data sets. Documentation of the quality of data in MSTDB-TP was continued by reviewing and assessing all available sources of density, viscosity and thermal diffusivity or conductivity values for unary chloride salts in MSTDB-TP V3.0 using the same criteria.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Multielectrode Array Sensors to Enable Long-Duration Corrosion Monitoring and Control of Concentrating Solar Power Systems

Molten mixtures of MgCl2-KCl-NaCl are promising heat transfer fluid (HTF) candidates for the next generation of concentrating solar power systems (Gen3 CSP), because of their low cost, high decomposition temperature, and favorable heat transfer characteristics. However, if the chemistry of the salts is not properly controlled, molten chlorides can cause corrosion of high-temperature metal alloys, such as Inconel (R), in CSP systems. Here we demonstrate a corrosion monitoring system that can be used to monitor the health of the structural alloys over long durations. The monitoring system includes a multielectrode array voltammetry sensor that measures the redox potential of molten MgCl2-KCl-NaCl mixture as well as the presence of corrosion products and impurities. The sensors have been successfully operated in static molten salt environments for long durations, and are capable of detecting corrosion products, such as Cr2+ and Fe2+, quantitatively at low concentrations (i.e., approximately 100 ppm) with less than 4% error.

Concentrated solar power↗

Recent Improvements in Pronghorn for Advanced Reactor Modeling

Pronghorn is a thermal-hydraulics computational tool developed using the Idaho National Laboratory's Multiphysics Object-Oriented Simulation Environment (MOOSE). It is designed to support Computational Fluid Dynamics (CFD) modeling, ranging from subchannel and porous media analysis to Reynolds Averaged Navier-Stokes (RANS) turbulence modeling. As an integral part of the MOOSE-based suite of tools, Pronghorn seamlessly couples with other MOOSE-based applications to simulate a variety of physical phenomena. This article highlights recent significant enhancements to Pronghorn's CFD modeling capabilities and demonstrates their application to advanced nuclear reactor designs. The recent improvements in Pronghorn primarily focus on modifications to its turbulence modeling capabilities, near-wall corrections and numerical schemes. In terms of turbulence modeling, the two-equation $k-\epsilon$ and $k-\omega$ SST models have been implemented and validated with both equilibrium and non-equilibrium wall treatments. Additionally, corrections for wall roughness, and curvature, and wall-channeling in pebble beds have been introduced in the near-wall modeling. These developments enable more accurate simulations of advanced nuclear reactors. Two case studies are presented in this work: a pool-type Molten Chloride Reactor and a salt-cooled Pebble-Bed High Temperature Reactor. In both cases, the previous models in Pronghorn are compared with the new implementations, demonstrating the improved accuracy achieved with the updated models.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗