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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Synthetic diversity in the preparation of metallic uranium

Uranium metal is associated with several aspects of nuclear technology; it is used as fuel for research and power reactors, targets for medical isotope productions, explosive for nuclear weapons and precursors in synthetic chemistry. The study of uranium metal at the laboratory scale presents the opportunity to evaluate metallic nuclear fuels, develop new methods for metallic spent fuel reprocessing and advance the science relevant to nuclear forensics and medical isotope production. Since its first isolation in 1841, from the reaction of uranium chloride and potassium metal, uranium metal has been prepared by solid-state reactions and in solution by electrochemical, chemical and radiochemical methods. The present review summarizes the methods outlined above and describes the chemistry associated with each preparation.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Demonstration of Electroreduction Technology to Convert GNF Uranium Oxide Powder to Metal

Global Nuclear Fuel – Americas LLC (GNF) has partnered with Argonne National Laboratory to demonstrate electroreduction of uranium oxides produced by deconversion of UF 6 to uranium metal through the Gateway for Accelerated Innovation in Nuclear (GAIN) program under the U.S. Department of Energy to accelerate the domestic production of metallic advanced reactor fuels. Electroreduction of uranium oxide was first demonstrated and patented by Argonne in the early 2000s as a technology to convert used oxide nuclear fuel from light water reactors to metal for further fuel reprocessing. More recently, electroreduction has been proposed as a front-end technology for metallization of uranium oxides produced by deconversion of UF 6 and for scrap recovery of oxide materials. During the electroreduction process, UO 2 powder is contained in a stainless-steel mesh basket with a cathode lead located in the center of the UO 2 bed. The basket is immersed in lithium chloride molten salt electrolyte containing 1 wt% lithium oxide along with a platinum anode and a nickel/nickel oxide (Ni/NiO) reference electrode. Current is applied to the cell between the cathode and anode to reduce the UO 2 to metallic uranium via a solid-state reduction reaction. Oxide ions released from the UO 2 during reduction are transported through the salt to the anode where oxygen gas is evolved. Once reduction is complete, the basket containing the metallicized uranium is removed from the salt and can be processed to remove the salt and consolidate the uranium into an ingot for use in metallic fuel fabrication. This project was performed to provide evidence of the electroreduction technology readiness level for metallization of UO 2 powder, identify and retire technical risks for industrialization of electroreduction, and accelerate the path to commercialization for metallic fast reactor fuel production. To that end, five electroreduction tests were performed with UO 2 provided by GNF and the resulting product was analyzed for the extent of conversion to metal and for impurity contents of the metal product to verify that electroreduction does not introduce impurities that would prevent use of the product in metallic fuel fabrication.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Supercritical Fluid Extraction and Separation of Uranium from Other Actinides

This paper investigates the feasibility of separating uranium from other actinides by using supercritical fluid carbon dioxide (sc-CO2) as a solvent modified with tri-n-butylphosphate (TBP) for the development of an extraction and counter current stripping technique, which would be a more efficient and environmentally benign technology for used nuclear fuel reprocessing compared to traditional solvent extraction. Several actinides (U(VI), Np(VI), Pu(IV), and Am(III)) were extracted in sc-CO2 modified with TBP over a range of nitric acid concentrations and then the actinides were exposed to reducing and complexing agents to suppress their extractability. According to this study, the separation of uranium from plutonium in sc-CO2 modified with TBP was successful at nitric acid concentrations of less than 3 M in the presence of acetohydroxamic acid or oxalic acid, and the separation of uranium from neptunium was successful at nitric acid concentrations of less than 1 M in the presence of acetohydroxamic acid, oxalic acid, or sodium nitrite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Quantify the Extent of Physisorption on Silver Based Sorbents under VOG Conditions

US regulations govern release limits for volatile radionuclides contained in the gaseous effluents of used nuclear fuel reprocessing facilities. Of the four volatile radionuclides that are most restricted by the release limits ( 3 H, 14 C, 85 Kr, and 129 I), 129 I will require the greatest degree of abatement. Within a reprocessing facility, the vessel off-gas stream will likely contain 1%–5% of the total iodine at parts per billion (ppb) concentrations. Studies conducted by the authors have now investigated both I 2 and methyl iodide (CH 3 I) adsorption by two silver-bearing sorbents, hydrogen-reduced silver-exchanged mordenite (Ag 0 Z) and silver functionalized aerogel (AgAerogel) at a range of concentrations up to 1,000 ppb. Throughout this suite of tests, and particularly in the case of CH3I adsorption tests, there was a significant discrepancy in the amount of iodine recovered on sorbent beds compared to the amount of iodine that was believed to be delivered to the test system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Off-Gas Adsorption Model Capabilities and Recommendations

Off-gas treatment is required to reduce emissions from aqueous fuel reprocessing. Evaluating the products of innovative gas adsorption research requires increased computational simulation capability to more effectively transition from fundamental research to operational design. Early modeling efforts produced the Off-Gas SeParation and REcoverY (OSPREY) model that, while efficient in terms of computation time, was of limited value for complex systems. However, the computational and programming lessons learned in development of the initial model were used to develop Discontinuous Galerkin OSPREY (DGOSPREY), a more effective model. Initial comparisons between OSPREY and DGOSPREY show that, while OSPREY does reasonably well to capture the initial breakthrough time, it displays far too much numerical dispersion to accurately capture the real shape of the breakthrough curves. DGOSPREY is a much better tool as it utilizes a more stable set of numerical methods. In addition, DGOSPREY has shown the capability to capture complex, multispecies adsorption behavior, while OSPREY currently only works for a single adsorbing species. This capability makes DGOSPREY ultimately a more practical tool for real world simulations involving many different gas species. While DGOSPREY has initially performed very well, there is still need for improvement. The current state of DGOSPREY does not include any micro-scale adsorption kinetics and therefore assumes instantaneous adsorption. This is a major source of error in predicting water vapor breakthrough because the kinetics of that adsorption mechanism is particularly slow. However, this deficiency can be remedied by building kinetic kernels into DGOSPREY. Another source of error in DGOSPREY stems from data gaps in single species, such as Kr and Xe, isotherms. Since isotherm data for each gas is currently available at a single temperature, the model is unable to predict adsorption at temperatures outside of the set of data currently available. Thus, in order to improve the predictive capabilities of the model, there is a need for more single-species adsorption isotherms at different temperatures, in addition to extending the model to include adsorption kinetics. This report provides background information about the modeling process and a path forward for further model improvement in terms of accuracy and user interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Copper and bismuth-based sorbent characterization in simulated iodine off-gas streams

The effective capture of volatile radioiodine, a fission product present in used nuclear fuel (UNF), is of paramount importance for development of used fuel reprocessing schemes to prevent release of radioiodine during unit operations and to meet regulatory standards for air emissions. A well-studied method for iodine capture in off-gas streams is the use of silver-functionalized zeolite phases (AgZ), which exploit chemisorption of I to Ag. Advances into other Ag-functionalized materials, including aerogels and metal organic frameworks (MOFs), are underway [1]. Additional metals with the capability to chemisorb iodine, including Cu, Bi, and Sn, [1] are being evaluated as alternatives to Ag for potential applicability to iodine management in off-gas systems. The design of novel functionalized sorbents with Cu and Bi, including composites with metal particles embedded in PAN substrates [2] and composites with metal sulfides embedded in PAN [3], is an ongoing area of study for improved iodine capture. Previously reported work on novel PAN-based metal sorbents has provided the synthesis, characterization, and iodine capture efficiency of this new class of sorbent. Specifically, the metal sulfide PAN composites are found to be easy to produce and reproduce, as well as having a high iodine loading potential under static conditions [3]. Due to the favorable testing previously performed with metal sulfide PAN composites, further testing into the performance of these composites under gas streams containing I2(g) in combination with NO2(g) and H2O(g) is needed. Humid streams of NO2(g) may arise from dissolver off-gas streams, when used fuel is dissolved in HNO3(aq) [4]. NO2(g) has been found to reduce AgZ sorption capacity for I because of oxidation of Ag, the chemisorbing agent, to Ag2O [5]. It follows that performance evaluation of novel sorbents under highly oxidizing conditions such as NO2(g) streams is critical. Therefore, the objective of this study is to determine the effect of flowing NO2(g) and H2O(g) streams on iodine sorption capacity and sorbent performance. This work utilizes custom-built gas handling capabilities for sorbent exposure along with solid-state characterization techniques to assess the physical and chemical properties of sorbents before and after exposure.

copper, bismuth, iodine capture, iodine sorbent, p↗
DOE OSTI2022

Gamma Radiation-Induced Degradation of Acetohydroxamic Acid (AHA) in Aqueous Nitrate and Nitric Acid Solutions Evaluated by Multiscale Modelling

Acetohydroxamic acid (AHA) has been proposed for inclusion in advanced, single-cycle, used nuclear fuel reprocessing solvent systems for the reduction and complexation of plutonium and neptunium ions. For this application, a detailed description of the degradation of AHA in dilute aqueous nitric acid is required. To this end, we present a comprehensive, multiscale computer model for the coupled radiolytic and hydrolytic degradation of AHA in aqueous sodium nitrate and nitric acid solutions. Rate coefficients for the reactions of AHA and HA with the oxidizing nitrate radical were measured for the first time using pulsed electron radiolysis and used as inputs in the kinetic model. The computer model results are validated by comparison to experimental data from steady state gamma irradiations, for which the agreement is excellent. Here, the presented model accurately predicts the yields of the major degradation products of AHA: acetic acid, hydroxylamine, nitrous oxide, and molecular hydrogen.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Bismuth-Functionalized Silica Aerogels for Iodine Capture

The U.S. Department of Energy is looking into alternative sorbents for the removal of radioiodine from off-gas streams in a used nuclear fuel reprocessing plant. One class of sorbents considered are bismuth-functionalized silica aerogels which offer an efficient capture of iodine from off-gas streams and are environmentally friendly and lower cost when compared to silver-functionalized sorbents. Two types of bismuth-functionalized silica aerogels were successfully manufactured using a hierarchical synthesis approach:1) The Bi 3+ -aerogel with bismuth oxide nitrate hydroxide hydrate particles and 2) Bi-aerogel with bismuth metal and bismuth sulfide nanoparticles. Both sorbents exhibited high sorption capacity for iodine. However, presence of particles of different composition, size, and distribution on aerogel support resulted in different iodine loadings. The Bi 3+ -aerogel exhibited sorption capacity of 289 mg/g. In contrast, Bi-aerogel exhibited more that 20% higher sorption capacity. The heat-treatment of Bi 3+ -aerogel at 225? under hydrogen atmosphere increased iodine loading capacity to 351 mg/g.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Ion-Exchanged SAPO-34 membranes for Krypton-Xenon Separation: Control of Permeation Properties and Fabrication of Hollow Fiber Membranes

Separation of radioisotope 85 Kr from 136 Xe is of importance in used nuclear fuel reprocessing. Membrane separation based on zeolite molecular sieves such as chabazite SAPO- 34 is an attractive alternative to energy-intensive cryogenic distillation. We report the synthesis of SAPO-34 membranes with considerably enhanced performance, via thickness reduction based upon control of a steam-assisted vapor-solid conversion technique followed by ion exchange with alkali metal cations. The reduction of membrane thickness leads to a large increase in Kr permeance from 7.5 gas permeation units (GPU) to 26.3 GPU with ideal Kr/Xe selectivities > 20 at 298 K. Cation-exchanged membranes show large (>50%) increases in selectivity at ambient or slight sub-ambient conditions. The adsorption, diffusion, and permeation characteristics of ionexchanged SAPO-34 materials and membranes are investigated in detail, with potassium exchanged SAPO-34 membranes showing particularly attractive performance. Lastly, we then demonstrate the fabrication of selective SAPO-34 membranes on α-alumina hollow fibers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Vibroacoustic Process Monitoring of 5-Stage Centrifugal Contactors

Vibroacoustic monitoring was investigated to aid in informing process operators with parameters regarding aqueous separation techniques. These techniques have a wide range of applications including nuclear fuel reprocessing. A small-scale solvent extraction system was set up with five magnetic drive contactors and two piston pumps to collect realistic operation. Vibration sensors were placed on three of the contactors and both pumps, and a microphone was placed near the system. Contactor rotational speed and pump flow rate were varied with known values to compare against measured vibroacoustic signatures. Spectral analysis allowed for the determination of contactor rotational speed and the ratio of organic to aqueous flow rate, an important parameter for ensuring optimal extraction. Vibration metrics such as kurtosis and crest factor provided additional details regarding individual rotational speeds and have potential in examination of faults for predictive maintenance. Overall, results indicate the capability of vibroacoustic monitoring to improve operator awareness and process outcomes.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sludge Processing Options for early HLW Treatment at Hanford

The U.S. Department of Energy’s (DOE) Hanford Site has 177 underground storage tanks that contain wastes from past nuclear fuel reprocessing and waste-management operations. Over 20% of this waste is in the form of an insoluble sludge that will require solids concentration and washing prior to vitrification for long-term disposal. An assessment of potential flowsheet operations to support feed preparation activities prior to high level waste (HLW) vitrification has been conducted to better evaluate pretreatment processing options. Settling studies assessing the baseline approach of a settle-decant method were explored as well as a crossflow filtration system to be used alternatively for concentrating and washing HLW sludge. Significant variations in behavior of settling rates and sludge characteristics give reason to evaluate alternative pretreatment options for the HLW. Non-radioactive sludge containing iron oxide, boehmite, and gibbsite were evaluated via gravity settling and crossflow filtration to determine the behavior of these compounds in various tank waste matrices. Understanding the predictive capabilities of HLW solids settling as well as sludge concentration via crossflow filtration can help provide technical guidance during flowsheet planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Complete Iodine Loading of NO-Aged Ag 0 -Functionalized Silica Aerogel (Milestone Report)

In off-gas treatment systems within a nuclear fuel reprocessing plant, capture materials will be exposed to a gas stream for extended periods during their lifetime. This gas stream may be at elevated temperature and could contain water, NO x gas, or a variety of other constituents. For this reason, it is important to understand the effects of long-term exposure, or aging, on proposed capture materials. One material under consideration for iodine sequestration is silver-functionalized silica aerogel (Ag 0 -aerogel). The aim of this study was to determine the effect of extended exposure at 150°C to an air stream containing NO on the iodine capture capacity of Ag 0 -aerogel. Ag 0 -aerogel was provided by the Pacific Northwest National Laboratory (PNNL), which manufactures the material at a lab scale. Prior to aging, the material has an iodine loading capacity of approximately 290 mg I/g Ag 0 -aerogel. Previous studies have aged the material in a dry air stream or in a moist air stream for up to 6 months. Both tests resulted in a 22% loss in iodine capacity. Aging the material in a static 2% NO 2 environment for up to 2 months results in a 15% loss of iodine capacity.3 In this study, exposure of Ag 0 -aerogel to 1% NO at 150°C for 2 months produced a loss of 43% in iodine loading capacity. This is largest loss observed for aerogel aging studies to date. The performance of Ag 0 -aerogel in this study was compared to the performance of reduced silver mordenite (Ag 0 Z) in similar studies. Ag 0 Z is a zeolite mineral considered to be the current standard technology for iodine removal from off-gas streams of a potential US used fuel processing plant. In an aging study exposing Ag 0 Z to 1% NO for 2 months, an iodine capacity loss of over 80% was observed. This corresponds to a silver utilization of 13.5% for 2 month NO-aged Ag 0 Z, compared to 57% silver utilization for 2 month NO-aged aerogel. While iodine loading capacity and silver utilization are critical parameters in evaluating these materials, other properties must also be considered when selecting the appropriate material (e.g., relative material densities and potential waste form production technology). The resistance of Ag 0 -aerogel to NO is promising, and investigations of this material for use in iodine capture should continue to be pursued.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Direct Electrical Detection of Iodine Gas by a Novel Metal–Organic-Framework-Based Sensor

High-fidelity detection of iodine species is of utmost importance to the safety of the population in cases of nuclear accidents or advanced nuclear fuel reprocessing. In this paper, we describe the success at using impedance spectroscopy to directly detect the real-time adsorption of I 2 by a metal–organic framework zeolitic imidazolate framework (ZIF)-8-based sensor. Methanolic suspensions of ZIF-8 were dropcast onto platinum interdigitated electrodes, dried, and exposed to gaseous I 2 at 25, 40, or 70 °C. Using an unoptimized sensor geometry, I 2 was readily detected at 25 °C in air within 720 s of exposure. The specific response is attributed to the chemical selectivity of the ZIF-8 toward I 2 . Furthermore, equivalent circuit modeling of the impedance data indicates a >10 5 × decrease in ZIF-8 resistance when 116 wt % I 2 is adsorbed by ZIF-8 at 70 °C in air. This irreversible decrease in resistance is accompanied by an irreversible loss in the long-range crystallinity, as evidenced by X-ray diffraction and infrared spectroscopy. Air, argon, methanol, and water were found to produce minimal changes in ZIF-8 impedance. Finally, this report demonstrates how selective I 2 adsorption by ZIF-8 can be leveraged to create a highly selective sensor using >10 5 × changes in impedance response to enable the direct electrical detection of environmentally relevant gaseous toxins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Initial Evaluation of CaSDB MOF Xe and Kr Capacities

Idaho National Laboratory (INL) and Pacific Northwest National Laboratory (PNNL) have been conducting parallel research, as part of the off-Gas Sigma Team, producing and testing sorbents for the selective capture of xenon and krypton from surrogate nuclear fuel reprocessing off-gas streams. Sorbents promising on a laboratory scale are tested further at INL in a system designed to emulate anticipated plant process conditions. This report summarizes INL’s initial testing of CaSDB MOF developed and produced at PNNL. Previous testing of the same base CaSDB MOF material, conducted at INL in 2016, indicated good saturation capacity for xenon, but a low xenon capacity at initial breakthrough. Additionally, the material displayed a propensity to break down into fine powder during use and handling. The current test series looks at the same base material, but with refinements to production and activation techniques. A series of ambient temperature tests were conducted to determine capacities of CaSDB MOF for krypton and xenon at initial breakthrough as well as at saturation. Adsorption tests were run in triplicate at superficial velocities of 21 and 53 cm/min, along with a single test at 373 cm/min. Saturation capacity across all tests was 30 mmol/kg +/- 2, with capacity at initial breakthrough of 8 mmol/kg +/- 1. It should be noted that while the usable portion of the material has a much tighter size distribution than that previously tested, propensity to physically break down into very fine powder with use and handling continues to be problematic. An improved engineered form is recommended prior to further testing

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

NM in HLW Borosilicate Glass: Focus on the High Performances of the IHMM Operated at La Hague Plant - 19019

Ruthenium, palladium and rhodium, which belong to Platinum-Group Metals (PGM), are part of fission products resulting from nuclear fuel reprocessing. All High Level Waste (HLW) vitrification facilities have experienced operational issues that can be related to the presence of Noble Metals (NM) in the melter. These metals have low solubility and modify glass properties such as viscosity or electrical and thermal conductivity. These dense metals also tend to settle to the bottom of the melter. Whatever the vitrification process operated, the specific properties and behavior of NM have a propensity to impact the production capacity by decreasing the glass throughput and increasing downtimes. Orano and the CEA have studied the behavior of NM in borosilicate glass for many years. It has led to several technological improvements and process operation optimizations on the Induction Heated Metallic Melter (IHMM) currently operated at La Hague plant. These enhancements, implemented since active commissioning, have headed to a high efficiency in terms glass production and NM incorporation (more than 21000 canisters produced with up to 3 % of Noble Metals in the glass). This paper sets out the continuous improvement program implemented on the La Hague vitrification lines that has led to outstanding records of operation in terms of production capacity and NM incorporation. The benefits of the close collaboration between R and D, engineering, and operation teams is particularly emphasized. (authors)

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

FY-2016 Methyl Iodide Higher NOx Adsorption Test Report

Deep-bed methyl iodide adsorption testing has continued in Fiscal Year 2016 under the Department of Energy (DOE) Fuel Cycle Technology (FCT) Program Offgas Sigma Team to further research and advance the technical maturity of solid sorbents for capturing iodine-129 in off-gas streams during used nuclear fuel reprocessing. Adsorption testing with higher levels of NO (approximately 3,300 ppm) and NO2 (up to about 10,000 ppm) indicate that high efficiency iodine capture by silver aerogel remains possible. Maximum iodine decontamination factors (DFs, or the ratio of iodine flowrate in the sorbent bed inlet gas compared to the iodine flowrate in the outlet gas) exceeded 3,000 until bed breakthrough rapidly decreased the DF levels to as low as about 2, when the adsorption capability was near depletion. After breakthrough, nearly all of the uncaptured iodine that remains in the bed outlet gas stream is no longer in the form of the original methyl iodide. The methyl iodide molecules are cleaved in the sorbent bed, even after iodine adsorption is no longer efficient, so that uncaptured iodine is in the form of iodine species soluble in caustic scrubber solutions, and detected and reported here as diatomic I2. The mass transfer zone depths were estimated at 8 inches, somewhat deeper than the 2-5 inch range estimated for both silver aerogels and silver zeolites in prior deep-bed tests, which had lower NOx levels. The maximum iodine adsorption capacity and silver utilization for these higher NOx tests, at about 5-15% of the original sorbent mass, and about 12-35% of the total silver, respectively, were lower than for trends from prior silver aerogel and silver zeolite tests with lower NOx levels. Additional deep-bed testing and analyses are recommended to expand the database for organic iodide adsorption and increase the technical maturity if iodine adsorption processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Conceptual Model of Iodine Behavior in the Subsurface at the Hanford Site

Isotopes of iodine were generated during plutonium production within the nine production reactors at the U.S. Department of Energy Hanford Site. The short half-life 131I that was released from the fuel into the atmosphere during the dissolution process (when the fuel was dissolved) in the Hanford Site 200 Area is no longer present at concentrations of concern in the environment. The long half-life 129I generated at the Hanford Site during reactor operations was (1) stored in single-shell and double-shell tanks, (2) discharged to liquid disposal sites (e.g., cribs and trenches), (3) released to the atmosphere during fuel reprocessing operations, or (4) captured by off-gas absorbent devices (silver reactors) at chemical separations plants (PUREX, B-Plant, T-Plant, and REDOX). Releases of 129I to the subsurface have resulted in several large, though dilute, plumes in the groundwater. There is also 129I remaining in the vadose zone beneath disposal or leak locations. The fate and transport of 129I in the environment and potential remediation technologies are currently being studied as part of environmental remediation activities at the Hanford Site. A conceptual model describing the nature and extent of subsurface contamination, factors that control plume behavior, and factors relevant to potential remediation processes is needed to support environmental remedy decisions. Because 129I is an uncommon contaminant, relevant remediation experience and scientific literature are limited. In addition, its behavior in subsurface is different from that of other more common and important contaminants (e.g., U, Cr and Tc) in terms of sorption (adsorption and precipitation), and aqueous phase species transformation via redox reactions. Thus, the conceptual model also needs to both describe known contaminant and biogeochemical process information and identify aspects about which additional information is needed to effectively support remedy decisions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of a bench-scale dissolution concept for the direct extraction of nuclear fuel

Current used nuclear fuel reprocessing efforts utilize a hydrometallurgical approach in which the UNF is dissolved in hot nitric acid followed by solvent extraction into an organic solvent to harvest target nuclides. Previous studies have shown that the dissolution and loading process could be combined into a single organic dissolution/extraction step, producing loaded organic in a single step process. This single step process also includes the advantage of selectively targeting key nuclides in the dissolution while leaving undesirable constituents as part of the undissolved solids. This process is referred to hereafter as direct extraction. Ongoing research from multiple national labs has proven the effectiveness of this technique at research scale. Therefore, potential methods to implement direct extraction at both bench and industrial scale have been developed. The key features of potential dissolver system designs were identified via the team at Pacific Northwest National Laboratory and several designs based on industrial counterparts were assessed for feasibility. This report summarizes the advantages and disadvantages of multiple methods, concluding with a path forward to create multiple unique dissolver designs. The first design will be a single stage recirculating eductor mixer. The second design recommendation is a stator rotor static mixing flow loop design. Each dissolver could be utilized separately, simultaneously, or in series to answer questions surrounding reaction kinetics including residence time, provide a proof of concept for targeted extractions of specific nuclides, and inform needs for industrial scale implementation

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗