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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Potential application of curium neutron detection in the high reliability safeguards approach for remotely handled nuclear facilities

The study presented in this paper is part of a series of quantitative modeling efforts to characterize the neutron flux for the major subsystems in a pyroprocessing facility. An MCNPX model has established the electroreduction process, in which uranium oxide is converted into a metal product by electrochemical methods. This process is important because the electroreduction stage is the earliest point in the system where a diversion attempt could occur. This paper will present neutron flux for a commercial batch size. The intent of this study is to determine if the magnitude of the flux is sufficiently high as to affect facility design. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermodynamic Assessment of LiCl-KCl-PuCl{sub 3} Ternary System

Pyroprocessing in LiCl-KCl molten salt is being considered as an alternative to the PUREX for the used nuclear fuel treatment. As more and more spent fuel is reprocessed, the composition of molten salt electrolyte changes, which alters the thermodynamic properties of the system and subsequently requires the adjustment of operation conditions. In order to optimize the system design and carefully operate it, thermodynamic assessment of actinides and lanthanides in LiCl-KCl molten salt is needed. Even though several systems have been assessed recently, such as LiCl-KCl-CeCl{sub 3} and LiCl-KCl-UCl{sub 3}, as one of the major actinides in pyro-processing, PuCl{sub 3} in LiCl-KCl hasn't been calculated. In present work, LiCl- KCl-PuCl{sub 3} is being considered and studied. Sub-binary systems of LiCl-PuCl{sub 3} and LiCl-KCl system was calculated firstly by using two-sublattice model based on the CALPHAD (Calculation of Phase Diagram) theory. The result indicates both of LiCl-PuCl{sub 3} and LiCl-KCl show simple eutectic properties and no solid solution is found. This research will help to predict the properties in electro-refiner when the concentration of elements in LiCl- KCl molten salt electrolyte changes during pyro-processing. Currently, binary system of KCl-PuCl{sub 3} is also being assessed and after that, phase diagram of ternary system of LiCl-KCl-PuCl{sub 3} will be calculated to provide a whole sight of the properties of PuCl{sub 3} in LiCl-KCl eutectic salt. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Electrodeposition of Actinides on a Semiconductor Detector for Concentration Monitoring

Pyroprocessing is a promising technology for separation and recycle of actinides from spent nuclear fuel. It is being actively researched in the United States as well as the Republic of Korea for possible implementation into a closed nuclear fuel cycle. The features of the chemical technology behind pyro-processing, unfortunately, present difficulties for real time accounting of actinides. Development of optimal process monitoring technology and safeguards approaches are, thus, a high priority for completing the development of the overall process technology. Real time monitoring of actinide concentrations in the molten salt electro-refiners is particularly important to achieve because of the potential for multiple significant quantities of plutonium in the molten salt. To address this need, development of a semiconductor detector for real-time monitoring of actinides in molten chloride salt (LiCl-KCl) is being conducted. Towards this end, a novel alpha spectroscopy detection system is being developed using wide band-gap semiconductor material, such as GaN or SiC, with a Boron-Doped-Diamond (BDD) contact. The objective is to deposit a thin layer of actinides onto the BDD layer, which will be used to measure the alpha particle spectrum that will distinguish individual actinides and their concentration. In the first stage, the actinide electrodeposition onto a surrogate material for the detector has been undertaken. While this project is in its very early states, a couple of observations have been made. One is that the electrodeposition is a slow, diffusion-controlled process. Therefore the deposit should be as thin (≅1 m) as the alpha spectroscopy permits (the resolution gets inferior with lesser number of atoms). Also, applied potentials should not be too negative; otherwise the deposit does not adhere well to the surface. The application of this method to measuring actinides in molten salt appears to be feasible, and a considerable amount of further investigation is planned. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of a Smart Signal Detection Method for Cyclic Voltammetry via Artificial Neural Intelligence

The electrorefiner (ER) is the heart of pyroprocessing technology which is a high-temperature method for separating uranium from Used Nuclear Fuel (UNF). It is important to improve this technology with respect to nuclear materials detection and accountability. Artificial Neural Intelligence (ANI) is a novel data analysis and simulation method that can be applied to electrochemical data sets. This computational code, which has been performed using the commercial software MATLAB, can be trained to generalize adequate electrical current and potential simulated data sets for the unseen data with a high accuracy of prediction. For this purpose, a massive collection of cyclic voltammetry (CV) data sets by Hoover (2014), for 0.5, 1, 2.5, and 5 wt% of zirconium chloride in LiCl-KCl molten salt with different scan rates at 773 K has been considered. The computer is trained via ANI to predict the unseen data after providing suitable hidden layers and validation numbers. In addition, this work can trace the CV plot for a blind condition by interpolating between two simulated data set. The different hidden layers with various neurons (from 5 to 30) at several validation numbers (from 5 to 30) has been studied and the average percent error between experimental and theoretical data for 0.5 wt% with 200 and 450 mV/s has been calculated. Preliminary results demonstrate that if the number of hidden layers increases from one to three, the average error falls down from 44% to around 8%. The best condition which gives a minimum average percent error has been discovered and the simulated CV graph has been compared with the experimental data. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Molten Salt Sampling Techniques and Analytical Approaches

Recent global interest in pyroprocessing and molten salt reactors has brought salt sampling methods and techniques back to the forefront of nuclear safeguards concerns. Issues with uranium supplies have also encouraged various countries to pursue advanced nuclear fuel cycles. Tracking nuclear material in molten salt has proven to be a challenge and updating molten salt sampling will greatly help in this endeavor. Molten salt is problematic to sample due to salt stratification, lack of homogeneity, solids, and difficulty with hot cell adaptations. Various salt sampling techniques have been used since before the 1960s including surface, spoon/spatula, and bar solidification. Since then, new types of sampling techniques have been developed to improve sampling results. These include rod/dip, pipet, suction, filtered sampling along with devices such as the Valve Core Sampler and the Multi-Level Sampler. These different approaches are being analyzed and improved upon along with developing requirements for an improved salt sampling device. Work continues to develop salt samplers that are more robust, easier to segment, collect at a specific depth, can work with filters, and can collect fines. Sampling parameters are also being narrowed in terms of stirring, settling time, filtration, depth, etc. In the future, we hope to address deficiencies for process control and nuclear material accountancy control by determining the best way to collect samples that minimizes contaminants and is representative. A compilation of salt sampling approaches, analyses techniques, and an evaluation of findings will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Electrochemical Characterization of Molten Salt Chemistry During Atmospheric Ingressions

This report describes the salt chemistry and sedimentation studies that were completed in FY23 as part of the Pyrochemical Fuel Cycles – ANL project. The primary objective of these activities was to employ electrochemical methods in order to quantify and gain insights into the reaction mechanisms associated with the interaction of O2 and moisture impurities with the molten chloride salts used for the pyrochemical processing of nuclear materials. If not monitored and controlled, the ingression of these atmospheric impurities can lead to changes in the salt redox conditions and initiate the formation of oxide particles, potentially resulting in operational and safeguards challenges. By elucidating the mechanisms underlying the formation of oxide particles, we can effectively monitor the conditions of the salt and design systems to mitigate the influence of atmospheric O2 and H2O. To achieve these goals, we designed a test system that allowed for precisely controlled gas ingressions into the molten salt vessel. This system included on-line monitoring provided by electrochemical probes that enable near real-time measurements of the salt conditions throughout the course of the experiment. Using this system, we systematically varied experimental conditions, including O2/Ar flow rates and O2 concentrations, to comprehensively understand their impact on solid particle formation. We also used a particle size and shape analyzer to characterize the particles that were generated. Tests in FY23 concentrated on the formation of CeO2 particles, but we also prepared a separate apparatus targeting UO2 particles for use in FY24. We additionally conducted extensive modeling activities in support of this work. This included multiphysics simulations of the gas ingressions along with the development of a machine learning workflow to enable accelerated molecular dynamics modeling of the molten salt chemistry of LiCl-KCl-UCl3. The combination of experimental and modeling tools has allowed us to begin to get a more complete understanding of the reactions that occur when atmospheric ingressions occur in pyroprocessing systems.

Guo, Jicheng↗
DOE OSTI2024

Nuclear waste reduction: Exploring new pathways one step at a time

In my home country of Venezuela, nuclear energy is not a topic that attracts much attention. The government briefly oversaw some nuclear energy programs during the 1950s, but currently there are no active nuclear power facilities in the country. In fact, the Venezuelan government signed and ratified the treaty of the prohibition of nuclear weapons in 2021, which states that Venezuela has never owned, possessed or controlled nuclear weapons or programs of any kind. When I moved to the United States, however, nuclear energy became an extremely relevant topic. In the 1940s, the U.S. government established and oversaw the Manhattan Project to build atomic bombs for use in World War II. After the war, the government encouraged scientists to use this information on nuclear reactions to develop nuclear energy for peaceful civilian purposes instead.1 During these early days of nuclear research, there were no formal regulatory standards for nuclear waste management. Policies usually were self-regulated and often created based on existing policies of disposal for non-nuclear waste.2 As a result, there were instances of nuclear waste leaching into the environment and affecting local communities. So, much research has been conducted since then to characterize and store nuclear waste safely and securely.3 I first became interested in nuclear energy during my undergraduate studies when I worked on a project involving ligand synthesis to help extract actinides from nuclear waste. I then studied electrochemistry in molten salt systems for nuclear energy applications during my Ph.D. As I approached graduation, I started looking into national laboratories that have programs involving nuclear energy and waste management. At Idaho National Laboratory (INL), the focus is more on applied processes and how nuclear energy can be innovated to realize next-generation reactor design and technologies. This focus led me to apply for a Seaborg distinguished postdoctoral position at INL, for which I was chosen based on my proposal of a way to improve nuclear waste recycling. To understand my proposal, we must familiarize ourselves with the makeup of nuclear waste. After uranium dioxide is used as nuclear fuel in a reactor, the fuel matrix is then characterized by various fission products, including rare earth elements, alkali and alkaline earths, and actinides. Some of these fission products can potentially be recovered through pyroprocessing, 4 which involves the electrochemical dissolution of the used nuclear fuel in a molten chloride salt mixture at high temperatures. Though some of the fission products can be easily recovered—for example, uranium is reduced onto an inert cathode by applied potentials—numerous other fission products such as rare earth elements are difficult to recover due to their multivalent oxidation states and side reactions.5 To improve the recovery efficiency of rare earth elements specifically, I proposed investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form (Figure 1). The kinetic pathways and the chemical reactions of these elements, which will be elucidated through spectro-electrochemistry at high temperatures, will give insights on how the recovery efficiency can be improved. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Bubbler Design Updates for Nuclear Safeguards Applications

The accurate monitoring of molten salt compositions is crucial for nuclear safeguards, particularly in the context of mixtures used in molten salt reactors as well as pyroprocessing. However, for nuclear material accountancy in actinide bearing molten salts, the solution volume is also needed to determine the total mass of material in the salt. This study explores the effectiveness of two bubbler configurations, triple and single, for density and level measurements within these molten salts for salt volume determinations. The triple bubbler system utilizes three gas dip-tubes strategically positioned at different heights within the molten salt. With this design, the molten salt level, density, and surface tension can be determined simultaneously. The single bubbler configuration, on the other hand, relies on a dynamic single dip-tube to determine the density and level. Both configurations will be tested in aqueous solutions as well as in LiCl-KCl and NaCl-MgCl2 molten salts between 450°C and 750°C. Dip-tube diameters and bubbler inferences will also be explored. A new bubbler is being designed based on results. The findings of this study will provide valuable insights into the optimal bubbler design and configuration for ensuring representative sampling of molten salt compositions in nuclear safeguards applications. This will contribute to enhanced accountability and transparency in the monitoring of actinide bearing molten salts, promoting nuclear safeguards objectives.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Undercooling minimization in ultrasound coupled DTA measurements of molten salts

Molten salts are ionic liquids that are used for the electrolytic pyroprocessing of metals and as heat transfer fluids in very high temperature processes. Recently, halide-based molten salt reactors (MSR) have gained momentum for high density and environmentally responsible electricity generation. The function of these reactors and their fuel cycle depend on a knowledge of the halide salt’s thermodynamic properties [1]. Therefore, high accuracy phase equilibria of MSR relevant base halide and actinide containing salts are needed. Modern thermal analysis of the phase transitions of halide salts is usually done during heating at relatively high scan rates with commercial devices. Normally such measurements are adequate for pure or pseudo-binary salts. However, as the number of components in the mixture increases, accurately resolving the liquidus becomes increasingly difficult. Reversing the scanning mode greatly increases the sensitivity of phase transition measurements but can decrease accuracy due to undercooling [2] — a common occurrence in molten halide systems [3]. This work presents the development of a differential thermal analysis (DTA) cell intended for use in radiological gloveboxes. Non-contact ultrasonic agitation of the halide salt is implemented to limit kinetic limitations on crystallization during cooling to minimize undercooling [4]. Measurements on halide salts are also presented to elucidate the effect of non-contact mixing on undercooling. Reference [1] S. Boyd and C. Taylor, “3 - Chemical fundamentals and applications of molten salts,” in Molten Salt Reactors and Thorium Energy, T. J. Dolan, Ed., Woodhead Publishing, 2017, pp. 29–91. doi: 10.1016/B978-0-08-101126-3.00003-8. [2] K. Nitsch, A. Cihlár, and M. Rodová, “Molten state and supercooling of lead halides,” J. Cryst. Growth, vol. 264, no. 1, pp. 492–498, Mar. 2004, doi: 10.1016/j.jcrysgro.2004.01.011. [3] L. Rycerz, “Practical remarks concerning phase diagrams determination on the basis of differential scanning calorimetry measurements,” J. Therm. Anal. Calorim., vol. 113, no. 1, pp. 231–238, Jul. 2013, doi: 10.1007/s10973-013-3097-0. [4] Md. H. Zahir, S. A. Mohamed, R. Saidur, and F. A. Al-Sulaiman, “Supercooling of phase-change materials and the techniques used to mitigate the phenomenon,” Appl. Energy, vol. 240, pp. 793–817, Apr. 2019, doi: 10.1016/j.apenergy.2019.02.045.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

Riley, Brian J.↗
DOE OSTI2019

Dechlorination of molten chloride waste salt from electrorefining via ion-exchange using pelletized ultra-stable H-Y zeolite in a fluidized particle reactor

Dechlorination of eutectic LiCl–KCl based electrorefner (ER) salt is reported via ion-exchange reaction with protonated ultrastable Y-type (USHY) zeolite bound into mechanically fuidized 45–250 μm diameter particles. Evidence of exchange of cations from the salt (Li + , K + , and fission product cations) into the zeolite lattice replacing H + ions was found based on a change in unit cell size, ICP-MS, XRD and TEM–EDS in addition to detection of HCl of gas. Ion exchange reaction was carried out at 625 and 650 °C, temperatures above the melting point of eutectic LiCl–KCl. Experiments were carried out to optimize zeolite drying temperature, estimate maximum ion-exchange capacity, and determine the thermal stability of USHY zeolite. The results indicate over 90% dechlorination can be achieved without zeolite structure collapse at 625 °C. Finally, this provides a promising route to stabilizing waste from radioactive chloride salts into dechlorinated waste forms for permanent geologic disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Time dependent chlorination of CeO 2 , La 2 O 3 and Nd 2 O 3 by ZrCl 4 dissolved in eutectic LiCl–KCl

Partial chlorination of CeO 2 , La 2 O 3 , and Nd 2 O 3 was achieved using ZrCl 4 dissolved in molten LiCl-KCl at 450°C over a period of 4-5 hr. ZrCl 4 was pre-volatilized into an argon stream to maintain a constant vapor pressure for equilibration with the salt. This approach was effective at maintaining the ZrCl 4 concentration at the solubility limit of 1.3 mol% in the molten salt throughout the duration of two of the chlorination experiments and within about 60% of the solubility limit in the third experiment. Based on the analysis of salt samples, fractional conversion after four hours of reaction for CeO 2 , and five hours for La 2 O 3 , and Nd 2 O 3 were calculated to be 0.97, 0.28, and 0.65 respectively. Assuming a shrinking core diffusion-limited model, the diffusivity of ZrCl 4 through the hypothesized ZrO 2 layer was calculated for the chlorination of CeO 2 , La 2 O 3 , and Nd 2 O 3 were 1.76 x 10 -10 cm 2 /sec, 7.85 x 10 -12 cm 2 /sec, and 7.42 x 10 -12 cm 2 /sec, respectively. Given the extremely low values calculated for diffusivity, it was concluded that the process is chemical reaction rate limited and that a complete reaction of each rare earth oxide could be achieved. Furthermore, a chemically controlled shrinking core model was shown to also fit the experimental data closely.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

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↗
DOE OSTI2022

Fundamental Issues Identified for Thermodynamic Description of Molten Salt Systems

Thermodynamic modeling of the molten salt systems is critical in the design of advanced nuclear power plants, materials recycling, and many other important clean energy applications. As one of the most capable computational thermodynamic approaches, the CALPHAD (Calculation of Phase Diagrams) method has been widely used for multicomponent thermodynamic database development and further support more physics-based modeling. Although considerable CALPHAD modeling efforts for molten salt systems have been made, we found that no evaluation has been performed regarding the unary thermodynamic models for the molten salt systems despite the fact that more than one pure substance database is available for the CALPHAD modeling. Therefore, the thermodynamic descriptions of the pure salts should be critically evaluated to ensure a high-fidelity molten salt thermodynamic database. In this work, we comprehensively analyzed some selected molten salts using different available thermodynamic descriptions from two databases as an example. One is the SSUB database released by the SGTE (Scientific Group Thermodata Europe), and the other is the FactPS database from the FactSage. Finally, the significant discrepancies observed in the thermodynamic description between the existing CALPHAD pure substance databases and experiments call for an urgent effort to improve.

42 ENGINEERING↗
DOE OSTI2019

A new inventory tracking method for Mark-V electrorefiner

Here, we discuss the current practice of assessing the Mark-V electrorefiner inventory and a proposed new way of tracking this inventory. The presented approach only involves the total mass measurement of the transferred materials and the consideration of metallic uranium phase in the inventory while the current practice involves the estimation/calculation of the salt phase mass in the transferred material mass and the calculation of the holdup. Without estimating the involved salt inventory in the incoming and outgoing material streams, the proposed method yields the salt inventory agreeing with the current mass tracking practice. The estimated holdup in the current mass tracking practice is entirely explained with the introduced metallic uranium phase in the proposed model. The current practice of closing out the special nuclear material inventory is revisited with the insight gained from the proposed model. Suggestions are given to achieve the bounded uncertainty of the inventory estimates with the emerging in-situ measurement technique on the molten salt density.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Uranium exchange kinetics in a molten LiCl-KCl/Cd system at 500 °C

A dissolution experiment is performed in order to examine the kinetics of uranium isotope equilibration in a molten salt and cadmium system. The salt contains a solute concentration of UCl 3 in LiCl-KCl eutectic and the cadmium contained dissolved uranium at 500 °C. A small piece of high enriched uranium is introduced to the cadmium, which perturbs the equilibrium with respect to the uranium enrichment of the salt and cadmium phases. The series of salt and cadmium samples trace how the enrichment of the two phases reach equilibrium (in the absence of a chemical driving force) over the course of approximately 24 h. Here, a model linking dissolution and isotope exchange kinetics is developed to track the evolution of uranium isotope composition. With the assumption of unaccounted salt mass incurred from the dated sample reporting practice, the model explains the trend of the measured uranium isotope composition well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Analysis of undissolved anode materials of Mark-IV electrorefiner

The Mark-IV electrorefiner contains a molten-salt/molten-cadmium system for electrorefining uranium. The anode basket is loaded with chopped sodium-bonded metallic driver fuel. For the present investigation, the bulk of the fuel was U-10Zr binary alloy. The undissolved anode materials after electrorefining are sampled for chemical analysis. The estimation of the undissolved anode materials based on individual chemical analysis results are challenging due to inherently high sampling errors associated with heterogeneous undissolved material compositions. Investigation of chemical analysis data as a whole reveals noticeable trends in the compositions of the primary phases and materials. Based upon this observation, we propose a methodology for accounting the primary phases relying on the measurement of the total anode mass only. Furthermore, the methodology is based on an empirical model explaining the mass of the undissolved anode materials constructed from historical anode material analysis data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

A Parametric Study of Operating Carbon Anodes in the Oxide Reduction Process

A parametric study of the electrolytic reduction of uranium dioxide in molten lithium chloride – lithium oxide using carbon anodes was conducted to determine the operational parameter values necessary for high yields. Operational parameters evaluated in this study include anode and cathode current densities, anodic polarization, UO 2 batch size, and method of electrochemical control. Seven oxide reduction experiments were conducted with between 25 g and 100 g UO 2 . The current density on the cathode was the critical parameter, which indicates the reduction process is kinetically controlled by cathodic reactions. High cell currents were achieved without the application of high anodic potentials by using a large anode surface area. As a result, this approach facilitates efficient reduction at high throughput without production of chlorine or other corrosive gasses.

36 MATERIALS SCIENCE↗