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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

In-situ Synthesis of PuCl 3 and Corresponding Raman and Density Functional Theory Vibrational Modes

The experimental and calculated Raman spectrum of PuCl 3 has been reported for the first time. PuCl 3 is a primary species found in plutonium metal refinement, specifically in pyrochemical salt processes including multicycle direct oxide reduction, metal chlorination, and electrorefining. As such, Raman signatures of PuCl 3 could serve as potential forensic indicators of material process history. A novel technique for synthesizing PuCl 3 from the in situ chlorination of plutonium metal with HCl was developed to establish these signatures. Cerium metal surrogates were utilized to ensure optimization of the plutonium experiments and to minimize personnel exposure, and all experiments were carried out in a Raman reaction chamber designed for air-tight, high vacuum environments. In situ Raman spectroscopy was employed in conjunction with density functional theory (DFT) to investigate the vibrational modes of PuCl 3 . Associated mixed oxy and hydroxyl phases are also reported. The combined Raman and DFT results have eliminated inconsistencies in Raman mode assignments for the MX 3 family of metal chlorides having P6 3 /m symmetry, and IR modes derived from the DFT calculations are additionally presented. The data observed in this study are of potential interest to nuclear forensic analyses, nuclear sample aging, nuclear energy, plutonium processing, and stockpile stewardship.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Gamma-ray spectra analyses of molten salts in spent nuclear fuels pyroprocessing facilities for mass measurement

To evaluate radioactive tracer dilution and gamma spectroscopy as a safeguards technique for monitoring the mass of liquid salt in molten salt systems, the gamma spectroscopy data of salt samples from the pyroprocessing facilities in Idaho National Laboratory were acquired and analyzed. The primary focus is the uncertainty analysis of the gamma radioactivity of 154 Eu isotope in the molten salts. The facilities include an electrorefiner for pyroprocessing spent oxide fuels (SOF-ER) and Mark-IV ER for processing the metallic fuels from Experimental Breeder Reactor-II (EBR-II). The relative gamma radioactivity uncertainty was consistently at 3% for 137 Cs for both Mark-IV ER and SOF-ER salts. However, for 154 Eu, it was 8% for SOF-ER salt and 7% for Mark-IV ER salt. Here, the main reason for the higher uncertainty for 154 Eu is believed to be the lower counting statistics in gamma spectroscopy. To improve the gamma radioactivity uncertainty of isotope 154 Eu, a longer data acquisition time (12 h instead of 4 h) and a higher count rate gamma detector for gamma spectroscopy were tested and it was found that the uncertainty for 154 Eu was significantly improved from 7% or 8–3%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Uncertainty improvement of 22 Na based radioactive tracer dilution for determining total mass of pyroprocessing molten salt systems by 154 Eu removal

To determine the total salt mass of the molten salt systems for pyroprocessing spent nuclear fuels, a 22 Na based radioactive tracer dilution was studied in Idaho National Laboratory in recent years. This 22 Na based RTD technique was deemed feasible, but due to the gamma energy peak of 22 Na coinciding with one of the energy peaks of 154 Eu radioisotope in the molten salt, the uncertainty of the 22 Na radioactivity in the 22 Na-spiked salt samples was quite high. To improve the uncertainty of the 22 Na based RTD technique, we proposed to chemically remove the 154 Eu of the salt samples by DGA resin for gamma spectroscopy. The effectiveness of removing 154 Eu on uncertainty improvement was evaluated. Furthermore, it was found that (1) the 154 Eu fission product effect on the uncertainty and detection limit can be effectively eliminated by chemically removing the 154 Eu during the salt sample preparation and (2) the uncertainty of 22 Na radioactivity in the salt samples for electrorefining was significantly improved from 13% to 2%, showing the potential of practical engineering application of 22 Na based RTD as a safeguards technique for molten salt systems for pyroprocessing spent nuclear fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Effects of composition and canister centerline cooling on microstructure, phase distribution, and chemical durability of dehalogenated iron phosphate waste forms

This paper discusses the effects of composition and canister centerline cooling (CCC) on the microstructures, phase distribution, and chemical durability of dehalogenated iron phosphate waste forms starting from chloride-based salt simulants including KCl (Simple-1), NdCl3 (Simple-2), SrCl2 (Simple-3), LaCl3 (Simple-4), LiCl (Simple-5), CsCl (Simple-6), as well as a complex simulant containing LiCl, KCl, NaCl, CsI, SrCl2, CeCl3, and NdCl3 called ERV2. The simulants represent salt constituents present in salt wastes from electrorefiners used for electrochemical reprocessing. These salts are reacted with ammonium dihydrogen phosphate (ADP) resulting in a NH4Cl (dechlorination) byproduct that is captured and phosphate product containing oxides of the salt cations. This product is further stabilized using Fe2O3 through vitrification at higher temperatures. The results presented herein describe the properties of CCC-treated materials following dechlorination and vitrification including phase identification and quantification, elemental distributions, and chemical durabilities. The chemical durability in the CCC-cooled samples are lower durability than quenched materials for the ERV2 samples. The phases formed upon CCC cooling are very complex compositionally and microstructurally.

iron phosphate glass, iron phosphate waste forms, ↗
DOE OSTI2019

Liquid cadmium cathode performance model based on the equilibrium behaviors of U and Pu in molten LiCl–KCl/Cd system at 500°C

The liquid cadmium cathode is a pyroprocessing technology used for the recovery of group actinide metals from the molten chloride salts that support uranium electrorefining. This article presents a model for the performance of a liquid cadmium cathode based on considerations of salt/alloy equilibrium thermodynamics, Cd-U-Pu phase diagram behaviors, and mass balance constraints. In conclusion, the model demonstrates how the salt/cadmium system responds as incremental additions of uranium are made to the liquid cadmium cathode.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Dehalogenation of electrochemical processing salt simulants with ammonium phosphates and immobilization of salt cations in an iron phosphate glass waste form

This study demonstrates the conversion of simulated pyroprocessing salt wastes [i.e., ER(SF) and ERV2] into chemically durable iron phosphate glass waste forms through reactions with ammonium hydrogen phosphate precursors [i.e., NH4H2PO4 and (NH4)2HPO4] and added Fe2O3. The reaction of the salt simulants with the phosphate precursors resulted in dehalogenation of the salt through the production of NH4Cl and NH4I (along with water), of which the NH4Cl can be used to produce UCl3 oxidant for use in the electrorefiner during pyroprocessing. These iron phosphate glasses contain a notably higher waste loading than the baseline glass-bonded sodalite ceramic waste form, which requires about 70 vol% sodalite to accommodate Cl from the salt, because they are dehalogenated prior to vitrification. The high waste loadings coupled with higher bulk densities (~3×103 kg/m3) than glass-bonded sodalite (~2.3×103 kg/m3) mean that these glasses have a higher storage volume, i.e., the volume required to immobilize a given mass of waste. The iron phosphate glass waste forms were characterized using a variety of techniques including physical property measurements and chemical durability testing. Following initial testing, a proof-of-concept demonstration was conducted to show that the NH4Cl generated from this process could be used to produce UCl3 using depleted uranium dendrites as a feedstock.

phosphate glass, pyroprocessing, waste forms, salt↗
DOE OSTI2022

Potential of electrolytic processes for recovery of molybdenum from molten salts for 99 Mo production

Molten salt reactor (MSR) technologies are receiving significant interest for commercialization because of their potential safety features and efficient energy production. In addition to producing reliable and clean energy, high-value radioisotopes could be harvested from MSRs as an additional revenue stream. Although significant quantities of 99 Mo will be produced in MSRs through fission, how accessible 99 Mo will be for direct recovery from a fuel salt is unclear. Electrolytic processes in fused salts have been used extensively for processing molten salts and have the potential to enable the extraction of 99 Mo directly from the fuel salt. This paper reviews the behavior of 99 Mo in the molten salt reactor experiment and summarizes theoretical aspects of the electrolytic process in high-temperature fused salt and the published experimental results on Mo extraction through electrolytic processes. Here, the general principle of fused-salt electrolytic processes is described, by electrodeposition of Mo with the following aspects: electrowinning, electrorefining, electroplating, and electroextraction with several experimental considerations. Finally, important considerations for recovery of 99 Mo from MSR relevant molten salts via electrodeposition are discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Near-infrared spectra and molar absorption coefficients of trivalent lanthanides dissolved in molten LiCl–KCl eutectic

Determining the concentration of the dissolved lanthanide species in LiCl–KCl eutectic salt is important to the development of pyrochemical reprocessing of used nuclear fuel. In this process, lanthanide fission products are found dissolved in the electrorefiner electrolyte in their trivalent oxidation state. The presence of dissolved trivalent lanthanides increases the liquidus temperature of the electrolyte mixture and can lead to the formation of insoluble oxide or oxychloride phases and must therefore be continuously monitored and controlled during the operation. Absorbance spectroscopy is a promising method for continuous measurement of the concentration of lanthanides and other elements dissolved in the electrolyte. The absorption of light by elements is linearly proportional to the concentration of the element for relatively dilute solutions according to the Beer-Lambert law. Although measurement of the absorption of ultraviolet and visible range light by lanthanides in LiCl–KCl eutectic molten salt have been explored previously, near infrared (NIR) absorption spectroscopy has received far less attention. It may, however, provide a better analytical signal when insoluble phases are present due to less Raleigh scattering compared to shorter wavelength radiation. Additionally, it may allow for concentration determination for certain elements using NIR absorption features where UV and visible range features are overlapping with features from other species. In this study, we report the UV–Vis–NIR spectra of the trivalent lanthanide chlorides of neodymium, samarium, and dysprosium in LiCl–KCl eutectic. Molar absorption coefficients are reported for analytically useful absorption maxima, with a focus on the molar absorption coefficients for NIR absorption maxima which have not been reported previously. Additionally, we observe a NIR-range absorption band of Nd3+ which was previously predicted but never experimentally observed. Here, we compare the calculated crystal field levels to the newly observed absorbance band and find them to be in good agreement with previous predictions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Organic Acid-Assisted Thermal Dehalogenation of Halide Salt Nuclear Wastes: From Waste Salts to Borosilicate Glass

Only a handful of high-halide salt waste forms have been demonstrated for vitrification-based immobilization strategies for halide-salt nuclear waste streams (e.g., pyroprocessing wastes, molten salt reactor wastes) and they all have low waste loading potential and most have low chemical durabilities for high-alkali streams. An alternative approach to direct salt immobilization is salt partitioning prior to waste form fabrication and one option for partitioning is halide removal (called dehalogenation). Removing the halogen fraction through dehalogenation can significantly reduce the waste volume required for disposal in the primary waste form. Furthermore, when dehalogenation is performed using organic acids, the dehalogenation reagent can decompose during high-temperature vitrification, reducing waste loading limitations in the waste form. In the current work, different organic acids (i.e., oxalic, formic, acetic, oxamic, and citric) were evaluated for dehalogenation efficiency of a simple chloride salt simulant (7.19% LaCl 3 , 53.77% LiCl, and 39.04% KCl, by mole) and a more complex chloride salt simulant called ERV3 (electrorefiner version 3) at 150 °C–300 °C and using H + /Cl – molar ratios of 1:1, 2:1, and 3:1. Additionally, a borosilicate glass waste form called TARS (or the average of refined specifications) was formulated, produced, and characterized for dehalogenated ERV3.

Amorphous materials↗
DOE OSTI2025

Unraveling Impurity-Dependent Morphological and Chemical Evolution of Ni–20Cr Alloy in Eutectic LiCl–KCl Molten Salt

Understanding the interfacial evolution of alloys in molten salt with different amounts of water (H 2 O) and oxygen (O 2 ) impurities is significant for applications in many fields, including concentrated solar power, molten salt reactors, and applications in pyrochemical reprocessing and electrorefining. Additionally, the impurity-driven corrosion mechanisms that lead to various morphological and chemical evolution characteristics at the interfaces of structural alloys and molten salts are not fully understood. In the present work, the three-dimensional (3D) morphological evolution of Ni-20Cr microwires in LiCl-KCl was studied at 500 °C under different moisture and oxygen conditions using in situ synchrotron transmission X-ray microscopy (TXM) and scanning transmission electron microscopy (STEM) techniques. No significant morphological changes were observed in Ni-20Cr microwires under vacuum conditions. However, the wires exhibited distinct morphological evolutions when exposed to molten salt containing H 2 O alone, as well as when both H 2 O and O 2 were present. Furthermore, Cr 2 O 3 precipitates were observed in the molten salt during corrosion with only H 2 O present, while Cr 6+ species were identified in the salt when O 2 was added. Further, these findings are crucial for understanding the corrosion mechanisms of molten salt with different amounts of H 2 O and O 2 contamination, providing insights for developing corrosion mitigation methods and improving the stability of containment alloys in molten salt applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Conceptual Design of a Pilot-Scale Pyroprocessing Facility

Argonne National Laboratory and Merrick & Company introduced the conceptual design of a pilot-scale (100 T/year) pyroprocessing facility for the treatment of used fuel generated by commercial light water reactors and subsequent treatment of waste streams generated during the process. The primary purpose of this study was to perform sufficient engineering for the pilot facility conceptual design so that credible capital and operating cost estimates could be developed. Initial safety, safeguards, and security assessments were also completed to provide a detailed evaluation in these areas that can significantly affect both capital and operating costs. Electrorefining-based pyroprocessing resulted in a compact hot-cell facility with few process equipment systems. The process equipment and support systems were estimated to cost $93 million and the facility $305 million for a project total cost of $398 million. The annual operating cost was estimated at $53 million/year Scaling up to a commercial-scale (400 T/year) was also assessed and the capital cost was estimated at $911 million with annual operating cost of $90 million/year.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Small-scale production of bespoke accelerated aging plutonium alloy

Here, this study demonstrates the 100 g scale manufacture of a plutonium alloy that ages at an accelerated rate. The resulting alloy ages six times faster than typical weapons-grade plutonium due to the addition of 238 Pu. As a major innovation, the process involved using a partial direct oxide reduction technique. This method was achieved by developing a new, complex geometry stirrer using additive manufacturing to reduce the 238Pu oxide and efficiently incorporate it into weapons-grade plutonium metal. The material was then purified by molten salt extraction and electrorefining before being alloyed with gallium. The alloy was then cold-rolled and annealed in a homogenization heat treatment. The resulting disk was characterized by metallography and differential scanning calorimetry, and the impurity content was determined using analytical chemistry techniques. The results show that a homogeneous delta phase plutonium alloy was achieved with expected microstructure and minimal impurities. This study was also successful in changing the plutonium isotopic composition by incorporating additional 238 Pu to accelerate the effects of radiation damage. This enables researchers to study long-term aging phenomena in a reduced time frame, thus avoiding the need for large-scale material production and circumventing the limitations of using naturally aged, archived plutonium.

Aging theory↗
DOE OSTI2023

Review—Fundamental Uranium Electrochemistry and Spectroscopy in Molten Salt Systems

Uranium is a key element used for nuclear energy production. Some advanced reactor designs, specifically molten salt reactors, will continue to use uranium as the fissile material for energy production. These new technologies require an intimate understanding of uranium chemistry during and after energy production. This review covers contemporary research on the coordination chemistry and behavior of uranium with the coolant and pyroprocessing salts as proposed for use in future reactor designs. Discussed topics include the nature of U redox reactions involving the reduction of U(III) to U metal and oxidation of U(III) to U(IV). These systems have been interrogated using cyclic voltammetry, chronopotentiometry, and optical and X-ray absorption spectroscopies. Insights obtained into the electrode potentials, the uranium species, and their diffusion coefficients in alkali halide melts from decades of research are summarized selectively. Further, perspectives are provided on the importance of unifying studies for comparison across multiple institutions. The application of synchrotron radiation research and multimodal approaches involving two (or more) probes, such as the widespread combination of UV–visible spectroscopy and electroanalysis known as spectroelectrochemistry, can provide new knowledge about the main process of uranium electrorefining—diffusion, as will be demonstrated in this review through the lack of comparable results.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Electrochemical Investigation of Moisture Byproducts in Molten Calcium Chloride

Residual water in molten CaCl 2 reacts to form different byproducts, such as HCl, which can impact the corrosivity of the salt and efficiency of electrochemical operations, such as electrolytic oxide reduction and electrorefining. The ability to detect and quantify these byproducts electrochemically can provide feedback on the efficacy of vacuum drying and other purification methods, as well as the impact of these byproducts on process operations. An electrochemical signal’s association with the production of H 2 is verified and characterized using cyclic voltammetry (CV) and residual gas analysis. CV estimated a 2-electron exchange process associated with H 2 production. CV detected trace quantities of an oxidized species containing hydrogen in the salt on the order of 10 ppm. Different salt handling methods were compared for their impact on the hydrogen electrochemical signal. It was found that 30 min of exposure of CaCl 2 in a beaker to low-humidity air (<20%) had minimal impact on the H 2 production signal.

Electrochemistry↗
DOE OSTI2024

Electrolysis in Chloride Molten Salts for Sustainable Critical Metals Production and Recovery

Critical materials, such as rare-earth metals, are essential to numerous applications, including clean energy; however, the present industrial practices for producing rare-earth metals involve environmentally damaging and thus unsustainable chemical and electrochemical processes. An alternative moderate-temperature chloride-based molten salt electrolysis process can address these issues, providing energy efficient and sustainable metal production. While it is being developed presently for rare-earth electrowinning, one can easily envision its broader application to rare-earth electrorefining and the electrolytic production of high-volume metals like Fe and Al. Presently, these high-volume metals industries account for nearly 10% of global greenhouse gas emissions. Furthermore, the chloride MSE process presents a huge opportunity for truly achieving sustainability if it is developed further for producing Fe, Al, Ti, Mg, and other commodity metals.

36 MATERIALS SCIENCE↗