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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

MicroRaman measurements for nuclear fuel reprocessing applications

Treatment and reuse of used nuclear fuel is a key component in closing the nuclear fuel cycle. Solvent extraction reprocessing methods that have been developed contain various steps tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper will focus on the development and application of a new MicroRaman probe for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid in solutions relevant to used nuclear fuel reprocessing. Previous research has successfully demonstrated the applicability on the macroscopic scale, using sample probes requiring larger solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 μL or less, which is a scale reduction of over 1,000 fold in sample size. Finally, this paper will summarize our current work in this area including: comparisons between the macroscopic and microscopic probes for detection limits, optimized channel focusing, and application in a flow cell with varying levels of HNO 3 , and UO 2 (NO 3 ) 2 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

C–H Functionalization of Polyolefins to Access Reprocessable Polyolefin Thermosets

Upcycling plastic waste into reprocessable materials with performance-advantaged properties would contribute to the development of a circular plastics economy. Here, we modify branched polyolefins and postconsumer polyethylene through a versatile C–H functionalization approach using thiosulfonates as a privileged radical group transfer functionality. Cross-linking the functionalized polyolefins with polytopic amines provided dynamically cross-linked polyolefin networks enabled by associative bond exchange of diketoenamine functionality. A combination of resonant soft X-ray scattering and grazing incidence X-ray scattering revealed hierarchical phase morphology in which diketoenamine-rich microdomains phase-separate within amorphous regions between polyolefin crystallites. The combination of dynamic covalent cross-links and microphase separation results in useful and improved mechanical properties, including a ~4.5-fold increase in toughness, a reduction in creep deformation at temperatures relevant to use, and high-temperature structural stability compared to the parent polyolefin. The dynamic nature of diketoenamine cross-links provides stress relaxation at elevated temperatures, which enabled iterative reprocessing of the dynamic covalent polymer network with little cycle-to-cycle property fade. Finally, the ability to convert polyolefin waste into a reprocessable thermoformable material with attractive thermomechanical properties provides additional optionality for upcycling to enable future circularity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

MELCOR Modeling of Accident Scenarios at a Facility for Aqueous Reprocessing of Spent Nuclear Fuel

The work presented in this report applies the MELCOR code to evaluate potential accidents in non-reactor nuclear facilities, focusing on Design Basis Accidents. Ten accident scenarios were modeled using NRC's best-estimate severe accident analysis code, MELCOR 2.2. The accident scenarios simulated a range of explosions and/or fires related to a nuclear fuel reprocessing facility. The objective was to evaluate the radionuclide source term to the environment following initiating explosion and/or fire events. The simulations were performed using a MELCOR model of the Barnwell Nuclear Fuel Plant, which was decommissioned before beginning reprocessing operations. Five of the accident scenarios were based on the Class 5 Design Basis Accidents from the Final Safety Analysis Report. Three of the remaining accident scenarios include sensitivity studies on smaller solvent fires. The final two accidents included an induced fire from an initial explosion. The radionuclide inventory was developed from ORIGEN calculations of spent PWR fuel with an initial enrichment of 4.5% U-235 by weight. The fuel aged for five years after a final 500-day irradiation cycle. The burn-up was conservatively increased to 60 GWd/MTU to bound current US operations. The results are characterized in terms of activity release to the environment and the building decontamination factor, which is related the leak path factor used in Department of Energy safety analyses. The MELCOR 2.2 results consider adverse consequences to the filters, ventilation system, and structures as a result of the explosions and fires. The calculations also include best-estimate models for aerosol transport, agglomeration, and deposition. The new calculations illustrate best-estimate approaches for predicting the source term from a reprocessing facility accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Requirements and Conceptual Design of Off-gas Systems for the Reprocessing of Metallic Fuels

An assessment has been conducted to determine how key regulations regarding volatile radionuclide emissions to the atmosphere may apply to the off-gas streams associated with electrochemical reprocessing. The scope of this assessment was based upon a generic electrochemical reprocessing scheme with a throughput rate of 200 MTIHM/y applied to metallic fuel discharged from a sodium fast reactor (SFR), but the findings are able to be translated to other advanced nuclear scenarios as merited. Air dispersion modeling was performed using the EPA CAP-88 model and evaluated the uncontrolled decontamination factors (DFs) that would be required to achieve regulatory compliance with the dose-based limits set forth by EPA regulation 40 CFR 190.10(a). These DFs were compared to those required by fuel cycle–based limits set forth by EPA regulation 40 CFR 190.10(b). Two theoretical sites with disparate climatological conditions were selected for air dispersion modeling (Idaho and Tennessee). The radionuclides modeled included 3 H, 85 Kr, 129 I, and selected alpha-emitting transuranic isotopes (referred to here as 239 Pu-TRU <1y ). It was found that the fuel cycle-based limits in 40 CFR 190.10(b) are most restrictive for 85 Kr and 239 Pu-TRU <1y , with DFs of 3 and 6.1E+09, respectively. The dose-based limit as derived from 40 CFR 190.10(a) could require mitigation of tritium in some scenarios, with an estimated DF of about 3 for the reference scenarios. The fuel cycle-based limit for 129 I resulted in a DF of about 240 for the reference scenario. The need for iodine mitigation based on dose to the public depended upon the physical form of iodine as either particulate or vapor-phase species. Emission of iodine from the facility as a vapor necessitated DFs of about 2 but emission as a particulate would require DFs >6,000 to meet thyroid dose-based limits. Effects of physical form on needed iodine mitigation are significant, but the understanding of speciation of iodine both during electrochemical reprocessing and after release to the atmosphere is limited. The electrochemical processing unit operations were evaluated to identify potential release points for the volatile radionuclides and to assess the potential for retention of the radionuclides within the process (thus decreasing the need for mitigation). Mitigation strategies for 3 H, 85 Kr, 129 I, and 239 Pu-TRU <1y were identified. In all cases, there are reasonably achievable pathways to regulatory compliance, although in some cases additional R&D is merited to verify the chemical speciation of these isotopes and to develop and demonstrate potential treatment technologies for this application. Whether or not additional off-gas controls (beyond common operations such as HEPA filtration and oxygen and moisture control) are needed for any of these regulated or volatile radionuclides depends on the (a) type of facility (NRC-regulated or DOE), (b) used fuel process rate, (c) used fuel burnup and composition, (d) speciation and retention of volatile radionuclides in the process and in the cell gas cleanup system, (e) site-specific parameters such as location, meteorology, stack height, and site boundaries, and (f) levels of conservatism and safety factors used in assessing compliance to air emissions regulations. Performance of this assessment revealed several areas where information is lacking or additional research is required in order to better determine if or what kinds of off-gas control might be needed. First, and most significantly, the understanding of the chemical speciation and physical form and partitioning of iodine during electrochemical processing operations is lacking and prevents the ability to accurately assess the potential iodine mitigation requirements. Future research in this area should be multifaceted and include thermodynamic modeling of iodine speciation in different process steps, experiments to quantify the kinetics of vapor-phase and melt-phase transitions, bench-scale experiments to determine the potential chemical and physical form of iodine emissions from the electrorefining process, and verification of iodine behavior with experiments utilizing operational facilities. Similarly, an improved understanding of iodine behavior in the environment after release from the facility stack will be required to refine dose estimations, as particulate and vapor-phase emissions result in significantly different doses to the MEI.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development of an MC&A toolbox for liquid-fueled molten salt reactors with online reprocessing (Final Report)

A critical barrier to the deployment of MSRs is the absence of a well-defined nuclear material control and accounting (MC&A) approach, a vital prerequisite to meet NRC licensing requirements as well as facilitating future international exports. Liquid-fueled MSR variants (especially those incorporating online refueling or reprocessing) present a unique set of challenges to traditional nuclear material control and accountancy. Unlike solid-fueled light-water reactors, traditional item-counting methods cannot be applied as an accountancy strategy. Rather, MC&A approaches to MSR variants (including both uranium and thorium fueled designs) are more analogous to bulk material handling facilities (e.g., enrichment and reprocessing); yet further complicating matters is the fact that fuel medium is also highly radioactive. Meanwhile, the space of MSRs covers a broad range of design parameters, including thermal and fast spectra designs; operation in actinide breeder or burner modes, choice of actinide fuel used (e.g., 235 U, 232 Th / 233 U, denatured 233 U, etc.), pool or loop-type configuration, and even different salt chemistry. Each of these design choices introduce significant challenges to MC&A approaches within MSR facilities. We propose to bridge this gap for liquid-fueled MSRs by developing a modular, component-based test framework for evaluating viable process monitoring and MC&A techniques specifically suited to liquid-fueled MSR system variants employing online refueling or reprocessing. This test platform will consist of a toolbox of independent process modules representing discrete physical units (such as the reactor core, off-gas processing, decay tanks, and actinide separation units), each with its own self-contained physics responsive to the input mass flow, along with appropriate measurement models that can be coupled to key flow points. These dynamic physical signatures thus afford the ability to test the viability and efficacy of potential accountancy techniques under the full range of reactor operating conditions. As process modules are connected via mass flows, the result is a reconfigurable, generic MSR mass flow model capable of serving as an MC&A test platform for a broad spectrum of possible MSR configurations. The resulting MSR MC&A toolbox will enable robust assessment of accountancy strategies for this unique facility type, including analysis of physical feedbacks arising both from depletion of the fuel over time as well as from potential off-normal events, including those introduced by equipment failures (e.g., a pump failure) as well as by malicious action (i.e., attempts to divert material). The proposed toolbox both addresses a critical needs area for the MPACT analysis toolkit while leveraging existing MPACT-sponsored tools, especially with respect to simulation of measurement and accountancy techniques for advanced fuel cycle facilities. Beyond enabling new analysis capabilities for MSR systems, the design of this toolbox will be to such to maximize compatibility with existing MPACT tools, such to enhance existing facility MC&A analysis capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Application of Hybrid K-Edge Densitometry in Reprocessing Facilities

During reprocessing operations, accurately determining the elemental concentrations of uranium and plutonium is critical for nuclear security. Reprocessing facilities processes hundreds of tons of nuclear material annually, requiring accurate measurements to ensure effective nuclear material control and accountability. Hybrid K-edge densitometry (HKED) system combines K-edge absorption densitometry with x-ray fluorescence to measure actinide elemental concentrations with a low uncertainty. This system uses high-purity germanium gamma-ray detectors and advanced signal processing equipment to detect x-ray accurately. This document provides guidance on how to achieve effective performance from an HKED system for measuring uranium and plutonium concentrations in reprocessing facilities. It includes best practices for the setup and operation of an HKED system, while highlighting factors influencing uncertainties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Comparison of Industrialized Late 20th Century Flowsheets for Reprocessing Used Nuclear Fuel

This study identified and compared three flowsheets for reprocessing used nuclear fuel (UNF) industrialized in plants in the United Kingdom (the Thermal Oxide Reprocessing Plant), France (UP2-800/UP3) and Japan (the Rokkasho Reprocessing Plant). The study also identified the major implications for a plant in the United States if it were initiated. All flowsheets employed the established Plutonium Uranium Reduction Extraction (PUREX) solvent extraction technology to separate uranium and plutonium from UNF dissolved in nitric acid. However, differences in the approaches to managing iodine-129, tritium and technetium were identified in the flowsheets. A US plant would also need to separate krypton-85 as well as iodine-129 and tritium for immobilization and disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Regulatory Considerations for Domestic Reprocessing Facility Physical Security

U.S. advanced non-light-water reactor vendors may pursue collocated on-site reprocessing activities. Therefore, these facilities are likely to possess formula quantities, or Category I quantities, of special nuclear material (SNM) during normal operations. The U.S. Nuclear Regulatory Commission (U.S. NRC) has yet to formally establish a regulatory framework for commercial reprocessing. While Category I requirements would explicitly not apply in this circumstance under current regulatory requirements, regulatory certainty does not exist. A novel framework should be developed to ensure public health and safety while also risk-informing the physical security requirements. This report reviews the relevant background of related rulemaking activities and proposes risk-informed physical protection requirements to satisfy these objectives. Insights from NRC security-related rulemaking activities provide a substantial technical basis to approach potential establishment of physical security requirements for reprocessing facilities. If a licensee can provide justification that the material satisfies a sufficient self-protecting radiation dose threshold, the material may not be subject to theft or diversion requirements and only potential sabotage requirements would apply. Furthermore, if the material can be justified to be moderately dilute, a set of risk-informed requirements could provide adequate protection of public health and safety. A revised performance objective for prevention of theft of moderately dilute Category I SNM may be detection to allow prompt recovery by a local law enforcement agency. However, a significant caveat to the proposed categorization scheme is the unknown integration of radiological sabotage with requirements for the protection against theft. Future licensees should consult with the NRC regarding treatment of this regulatory topic. Additionally, the self-protecting radiation dose threshold (either the existing or a proposed future threshold) would need to be considered. An integrated approach may apply graded potential requirements for protection against the design basis threat of radiological sabotage currently applicable to commercial nuclear power plants and Category I SNM facilities defined within 10 CFR 73.1(a).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1993

A reprocessing model for the ultraviolet and optical light from 4U 1820-30

We show that the recently discovered optical and ultraviolet light from the X-ray burst source 4U 1820-30 in the globular cluster NGC 6624 is due to reprocessing of the X-rays in the outer regions of an optically thick, geometrically thin accretion disk. We suggest that observation of orbital modulation of the reprocessed light, due to the variable contribution made as the heated face of the companion turns toward and away from the observer, would provide constraints on the inclination of the binary orbit, and we suggest that detection in the reprocessed flux of the 'red noise' already observed in the X-rays would provide useful constraints on the geometry and physics of the accretion disk.

Arons, Jonathan↗
NASA NTRS2003

Comparing SeaWiFS Reprocessing Versions (R3 vs. R4)

Satellite observations of global ocean chlorophyll from SeaWiFS were recently reprocessed to incorporate calibration and algorithm improvements. Here, comparisons are made between the newly reprocessed SeaWiFS Level-3 chlorophyll product and the previous version using in situ measurements. The results show that the newly reprocessed SeaWiFS data matches up better with the surface measurements than the previous version did. Globally, the slope of the match-ups improves to 0.85 from 0.78 in log-log scale. A significant trend that contributed to this improvement was the overall decrease in SeaWiFS chlorophyll levels less than 1.0 mg m(exp -3). Regional analyses reveal that the matchups improve in every oceanic basin, except the Antarctic. However, SeaWiFS continues to exhibit poor correspondence with in situ data in the North Atlantic where the match-ups have a slope of 0.54. Also, an examination of monthly images for May 1999 revealed that the number and magnitude of high-value chlorophyll pixels had increased in the high-latitude open ocean of the South Pacific.

Casey, Nancy W.↗
NASA NTRS2000

Intensive HST, RXTE, and ASCA Monitoring of NGC 3516: Evidence against Thermal Reprocessing

During 1998 April 1316, the bright, strongly variable Seyfert 1 galaxy NGC 3516 was monitored almost continuously with HST for 10.3 hr at ultraviolet wavelengths and 2.8 days at optical wavelengths, and simultaneous RXTE and ASCA monitoring covered the same period. The X-ray fluxes were strongly variable with the soft (0.5-2 keV) X-rays showing stronger variations (approx. 65% peak to peak) than the hard (2-10 keV) X-rays (approx. 50% peak to peak). The optical continuum showed much smaller but still highly significant variations: a slow approx. 2.5% rise followed by a faster approx. 3.5% decline. The short ultraviolet observation did not show significant variability. The soft and hard X-ray light curves were strongly correlated, with no evidence for a significant interband lag. Likewise, the optical continuum bands (3590 and 5510 A) were also strongly correlated, with no measurable lag, to 3(sigma) limits of approx. less than 0.15 day. However, the optical and X-ray light curves showed very different behavior, and no significant correlation or simple relationship could be found. These results appear difficult to reconcile with previous reports of correlations between X-ray and optical variations and of measurable lags within the optical band for some other Seyfert 1 galaxies. These results also present serious problems for "reprocessing" models in which the X-ray source heats a stratified accretion disk, which then reemits in the optical/ultraviolet : the synchronous variations within the optical would suggest that the emitting region is approx. less than 0.3 It-day across, while the lack of correlation between X-ray and optical variations would indicate, in the context of this model, that any reprocessing region must be approx. greater than 1 It-day in size. It may be possible to resolve this conflict by invoking anisotropic emission or special geometry, but the most natural explanation appears to be that the bulk of the optical luminosity is generated by some mechanism other than reprocessing.

Edelson, Rick↗
DOE OSTI2016

Extending Spent Fuel Storage until Transport for Reprocessing or Disposal

Spent fuel (SF) must be stored until an end point such as reprocessing or geologic disposal is imple-mented. Selection and implementation of an end point for SF depends upon future funding, legisla-tion, licensing and other factors that cannot be predicted with certainty. Past presumptions related to the availability of an end point have often been wrong and resulted in missed opportunities for properly informing spent fuel management policies and strategies. For example, dry cask storage systems were originally conceived to free up needed space in reactor spent fuel pools and also to provide SFS of up to 20 years until reprocessing and/or deep geological disposal became available. Hundreds of dry cask storage systems are now employed throughout the world and will be relied upon well beyond the originally envisioned design life. Given present and projected rates for the use of nuclear power coupled with projections for SF repro-cessing and disposal capacities, one concludes that SF storage will be prolonged, potentially for several decades. The US Nuclear Regulatory Commission has recently considered 300 years of storage to be appropriate for the characterization and prediction of ageing effects and ageing management issues associated with extending SF storage and subsequent transport. This paper encourages addressing the uncertainty associated with the duration of SF storage by de-sign – rather than by default. It suggests ways that this uncertainty may be considered in design, li-censing, policy, and strategy decisions and proposes a framework for safely extending spent fuel storage until SF can be transported for reprocessing or disposal – regardless of how long that may be. The paper however is not intended to either encourage or facilitate needlessly extending spent fuel storage durations. Its intent is to ensure a design and safety basis with sufficient margin to accommodate the full range of potential future scenarios. Although the focus is primarily on storage of SF from commercial operation, the principles described are equally applicable to SF from research and production reactors as well as high-level radioactive waste.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Radiolytic evaluation of acetohydroxamic acid (AHA) under biphasic (n-dodecane and TBP/DEHBA/DEHiBA) used nuclear fuel reprocessing conditions

Acetohydroxamic acid (AHA) has been proposed as a substitute for uranium(IV) and hydrazine as a plutonium complexant and neptunium reductant for simplified, single-cycle used nuclear fuel reprocessing flowsheets. However, the chemical behavior of AHA in an intense multi-component radiation field is poorly understood, especially under representative biphasic reprocessing solvent system conditions. In response to this critical knowledge gap, this study has investigated the gamma radiolytic integrity of AHA in aqueous nitric solutions in contact with an organic phase, comprising current and future reprocessing ligands (TBP, DEHBA, and DEHiBA) dissolved in n-dodecane diluent. Our data show negligible effect of the organic phase on the radiolytic behavior of AHA compared to complementary single-phase experiments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Actinide Radiation Chemistry and Used Nuclear Fuel Reprocessing

Although the actinides boast many unique physical and chemical properties, their inherent susceptibility to radioactive decay—and subsequent consequences of radiation-induced chemistry—are what make them truly interesting and challenging elements to understand. From a closed nuclear fuel cycle perspective, the ability to predict and control the effects of actinide-driven radiolysis is critical for the design, development, and deployment of advanced used nuclear fuel reprocessing strategies and technologies. The absorption of ionizing radiation from actinide decay leads to the formation of a variety of transient and steady-state radicals, ions, and molecular radiolysis products that can lead to significant changes in a reprocessing solvent system’s physical and chemical properties, which ultimately limits that process’ efficiency and longevity. Presented here is an overview of recent advances in actinide radiation chemistry as it applied to used nuclear fuel reprocessing.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Advances in Online Monitoring for Measurement within Nuclear Fuel Reprocessing Streams

Abstract – Significant effort has been focused on the back end of the nuclear fuel cycle including treatment and reuse of irradiated fuel. Solvent extraction reprocessing methods that have been developed often contain various steps, with each one tailored to the separation of specific radionuclides, which are highly dependent upon solution properties. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. Our group has been investigating the use of optical spectroscopy for the on-line monitoring of actinides, lanthanides, and acid strength within fuel reprocessing streams. This paper describes the application of the electronic and vibrational spectroscopic techniques for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and UV-vis spectroscopic techniques for online real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(VI/IV)/Np(VI/V/IV)/Nd(III) in solutions relevant to used nuclear fuel reprocessing. These techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved used fuel solutions. Currently this has been demonstrated on the macroscopic scale, using sample probes requiring large solution volumes. In an effort to minimize waste and reduce dose to personnel, we have modified this technique to allow measurement at the microfluidic scale using a Raman microprobe. Under the current sampling environment, Raman samples typically require upwards of 10 mL and larger. Using the new sampling system, we can sample volumes at 10 µL or less, which is a scale reduction of over 1,000 fold in sample size. This will significantly reduce the requirements for handling hazardous materials, and greatly reduce the costs associated with worker protection from exposure to radioactive and hazardous materials, and costs due to waste disposal.

process monitoring, U, nitric acid, microfluidic, ↗
DOE OSTI2024

Elucidating the Radiation-Induced Redox Chemistry of Plutonium Under Used Nuclear Fuel Reprocessing Conditions

Plutonium plays a critical role in the development of sustainable nuclear fuel cycles, and yet, our fundamental understanding of this element’s inherent radiation-induced redox chemistry and associated impacts on nuclear fuel cycle technologies is limited. Unanticipated changes in oxidation state distribution can influence the speciation and transport of plutonium in a given process. Control of these parameters is especially important for used nuclear fuel reprocessing technologies, wherein the separation and recovery of plutonium is typically achieved by the selective formation, maintenance, and complexation of specific oxidation states. Furthermore, plutonium’s inherent radiation-induced redox chemistry has the capacity to influence the radiolytic behavior of its complexes, the longevity of which are critical in the design of efficient and cost-effective advanced reprocessing technologies. These radiation-induced processes are unavoidable under fuel cycle conditions owing to the inherency of ionizing radiation fields to the decay of plutonium’s isotopes and to the various other radioisotopes generated by nuclear fission and neutron-capture process and the subsequent radioactive decay of their products. As such, mechanistically understanding the response of plutonium’s multiple oxidation states to multi-component ionizing radiation fields is essential for predicting the behavior of this critical element under used nuclear fuel reprocessing conditions. Here, through a combination of time-resolved (electron pulse) and steady-state (alpha and gamma) irradiation experiments complemented by quantitative, multiscale modeling calculations, we present advances in our understanding of radiation-induced plutonium redox chemistry!

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Radiation Effects in Next Generation Used Nuclear Fuel Reprocessing Strategies

With the global community committed to significantly expanding nuclear energy capacity, the development of efficient used nuclear fuel (UNF) management strategies has become more critical than ever. These strategies are vital to fostering the widespread adoption of closed fuel cycles, which are essential for sustainable nuclear energy production and security. Achieving this ambitious goal necessitates a comprehensive understanding of radiation effects on next-generation technologies, as radiolysis can often limit the longevity and performance of these systems. This seminar will provide an overview of next-generation UNF reprocessing strategies, highlighting the latest advancements and innovative approaches in the field. Particular attention will be given to two key areas of recent research: 1. Radiation robustness and performance of advanced sulfur chloride-based chlorination technologies. We will explore the efficacy of sulfur chloride-based chlorination processes in the presence of surrogate cladding materials, specifically aluminum. These processes have shown promise in the dissolution, decontamination, and recovery of cladding materials for reuse. Detailed findings on how the composition and performance of these sulfur chloride solvents respond to radiation exposure will be discussed. 2. Impacts of metal ion complexation and direct dissolution conditions on monoamide-based reprocessing strategies. We will delve into the time-resolved and dose accumulation effects of irradiation on the direct dissolution of voloxidized uranium and rhenium using N,N-di-(2-ethylhexyl) butyramide (DEHBA) or N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) in pre-equilibrated n-dodecane solvent. The implications of these interactions on dissolution efficiency, radiolytic stability, and overall process performance will be examined. These studies aim to underscore the importance of understanding radiation effects in the development of next-generation UNF reprocessing technologies and the global transition towards more sustainable and efficient nuclear energy systems.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2017

Promising Technology for Reprocessing Complex Uranium Ores from the Elkon Deposit

The results of research on reprocessing of complex uranium ores from the Elkon deposit with recovery of uranium, gold, silver, and vanadium by methods of radiometric sorting, combining of autoclave and atmospheric leaching of ore followed by sorption and extraction recovery of valuable components, are presented. The capacity was 90 mg/g in sorption of uranium on strongly basic anionite and the uranium concentration ≤0.003 g/dm{sup 3} in sorption filtrates. The choice of eluent is determined by the formulation of the problem of obtaining the finished uranium product of the requisite form and quality. A fundamentally new composition of the extracting agent is proposed for direct extraction of uranium from nitrate-containing solutions. An iron-vanadium concentrate with vanadium and iron content 2.58 and 22.8%, respectively, suitable for subsequent reprocessing, was obtained. The anionite PuroGold S-992 is recommended for recovery of gold and silver. The gold and silver concentrations in the commercial desorbate were 270 and 380 mg/dm{sup 3}, respectively; a gold-containing concentrate was obtained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗