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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Multiplicity of Th(IV) and U(VI) HEH[EHP] Chelates at Low Temperatures from Concentrated Nitric Acid Extractions

Organophosphorus extractants have been widely investigated for lanthanide recovery from ore and for application in the reprocessing of spent nuclear fuel, such as in Advanced TALSPEAK schemes. Determining the speciation of the extracted metal complex in the organic phase remains a significant challenge. A better understanding of the variability of HEH[EHP]–actinide complexes and the speciation of chelates for tetra- and hexavalent actinides can improve the predictability of actinide phase transfer in such biphasic systems. Here, the extraction of Th(IV) and U(VI) from nitric acid media using HEH[EHP] in heptane is examined. The distribution ratio as a function of nitric acid concentration was quantified using UV–vis spectroscopy, and then the speciation of HEH[EHP]–metal complexes in the organic phase was investigated using Fourier transform infrared (FTIR) spectroscopy and low-temperature 31 P nuclear magnetic resonance (NMR) spectroscopy. In addition to perturbation of the vibrational modes proximal to the phosphonic moiety in HEH[EHP] in the FTIR spectra, the appearance of a nitrate signal was found in the organic phase following extraction from the highest acidity conditions for U(VI). The 31 P NMR spectra of the organic phase at a low temperature (-70 °C) exhibited a surprising number (n) of resonances (n ≥ 7 for Th(IV) and n ≥ 11 for U(VI)), with the distribution between these resonances changing with the initial concentration of nitric acid in the aqueous phase. These results indicate that the compositions of the inner and outer spheres of the extracted actinides in the organic phase are more diverse than initially thought.

31P NMR↗
DOE OSTI2023

Structural Investigation of Technetium Dibutylphosphate Species Using X-ray Absorption Fine Structure Spectroscopy

The speciation of Tc after the extraction of Tc(IV) from H 2 O and 1 M HNO 3 by dibutylphosphoric acid (HDBP) in dodecane has been studied by X-ray absorption fine structure (XAFS) spectroscopy. Results show the formation of dimeric species with Tc 2 O 2 and Tc 2 O units, and the formulas [Tc 2 O 2 (DBP·HDBP) 4 ] (1) and [Tc 2 O(NO 3 ) 2 (DBP) 2 (DBP·HDBP) 2 ] (2) were, respectively, proposed for the species extracted from H 2 O and 1 M HNO 3 . The interatomic Tc–Tc distances found in the Tc 2 O 2 and Tc 2 O units [2.55(3) and 3.57(4) Å, respectively] are similar to the ones found in Tc(IV) dinuclear species. It is likely that the speciation of Tc(IV) in dodecane is due to the extraction of a species with a Tc 2 O unit for (2) and to the redissolution of a Tc(IV)-DBP solid for (1). The XAFS results for (1) and (2) were compared to that obtained for the extraction of Tc(IV) with TBP/HDBP/dodecane from 0.5 M HNO 3 , (3) which highlight the formation of Tc mononuclear nitrate species {i.e. [Tc(NO 3 ) 3 (DBP)] or [Tc(NO 3 ) 2 (DBP·HDBP)]}. These results confirm the importance of the preparation and speciation of the Tc(IV) aqueous solutions prior to extraction and how much this influences and drives the final Tc speciation in organic extraction. Here, these studies outline the complexity of Tc separation chemistry and provide insights into the behavior of Tc during the reprocessing of used nuclear fuel.

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

Influence of Nd(II) and Nd(III) Ions on the Speciation and Kinetic Dynamics of Radiolytic Transients in Molten LiCl-KCl Eutectic Salt Mixtures

In the pyrochemical reprocessing of used nuclear fuel, understanding the effects of ionizing radiation on the chemical behavior of fission products in molten chloride eutectic salt is crucial. Here, this study investigates the impact of radiation on the chemical and kinetic speciation of Nd ions in molten LiCl-KCl eutectic, which are challenging to separate due to their multivalent oxidation states and comproportionation reactions. Electron pulse radiolysis techniques were used to determine transient absorption spectra, chemical kinetics, and activation parameters for the reaction of Nd ions with the primary products of molten LiCl-KCl eutectic radiolysis: the solvated electron (e S - ) and dichlorine radical anion (Cl 2 •- ). Initially, Nd(III) reacted rapidly with e S - , forming Nd(II) with a second-order rate coefficient of k = (4.54 ± 0.07) × 10 10 M -1 s -1 at 673 K. The resulting Nd(II) ions then reacted slowly with Cl 2 •- , regenerating Nd(III) with an estimated rate of k = (1.72 ± 0.04) × 10 10 M -1 s -1 at 673 K. Additionally, the comproportionation reaction of Nd(III) and Nd(s) to form Nd(II) was monitored chemically and electrochemically. This work suggests that the comproportionation reaction of Nd(III) with Nd(s) to form Nd(II) is unstable in molten LiCl-KCl eutectic.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Adsorption of Radioactive Iodine Using Nanocarbon on ETS-10 as Adsorbent

Here, laboratory-synthesized nanocarbon pelletized with titanosilicate (ETS-10) as a support matrix has been investigated for the capture of radioactive iodine present as methyl iodide (CH 3 I) in the off-gas streams produced during aqueous reprocessing of used nuclear fuel. The mass fraction of carbon in the sorbent matrix was 0.10. The effects of residence time and CH 3 I concentration were investigated using a continuous flow column setup to quantify the adsorption and desorption capacities of adsorbent under dynamic conditions from an air stream containing CH 3 I present at concentrations representative of those expected in the off-gas streams. Air with CH 3 I gas as a source in the column resulted in quantifiable CH 3 I adsorption with 0.98 mg/g of adsorption capacity. Laboratory-made nanocarbons had a larger adsorption capacity than those of the other carbons reported in the literature. Additionally, the adsorption capacity of nanocarbon on ETS-10 is compared to that of nanocarbon coated on cordierite in previous studies.

ETS-10↗
DOE OSTI2021

Single-component-at-a-time variation study for glass-ceramic waste forms

Here, a 51-sample composition variation study was performed on glass-ceramic waste forms for a raffinate waste stream from aqueous reprocessing of used nuclear fuel containing high fractions of Mo, alkalis, alkaline earths (AEs), and rare earths (REs). The study was designed with a single-component-at-a-time variation approach off a centroid composition. The components that were varied included Al, B, Ca, Li, Mo, Na, REs, Si, Zr, and Others (containing minor components). Data analysis included crystallization curves, microstructure, and phase compositions. A number of components (i.e., Li 2 O, B 2 O 3 , REO x , MoO 3 , Na 2 O, and ZrO 2 ) significantly impacted the concentration and chemistry of phases, especially the primary phases of oxyapatite [i.e., Ca 2 RE 8 (SiO 4 ) 6 O 2 ] and powellite (i.e., AEMoO 4 ), precipitated in the slow-cool heat-treated waste forms; minor phases included cerianite [i.e., Ce x Zr (1-x) O 2 ], Ba-molybdate [i.e., Ba(Gd 0.67 Mo 0.33 O 3 ], noble metals, pollucite (i.e., CsAlSiO 4 ), and RE-borosilicate (i.e., RE 3 BSi 2 O 10 ).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Effect of NO x and Water Variations on Iodine Loading of AgZ

Reprocessing of used nuclear fuel will result in the release of several volatile radionuclides that need to be removed from facility off-gas streams before their release to the environment. Iodine, one of these radionuclides, is expected to be primarily released in the dissolver off-gas stream as iodine (I 2 ) or methyl iodide (CH 3 I). Sorbents that target I 2 and CH 3 I removal from the DOG stream will need to adsorb iodine under elevated temperatures and in the presence of water vapor and nitrogen oxides (NO x ). Previous studies examining the adsorption of CH 3 I and I 2 in the presence of NO x gases by silver-exchanged mordenite (AgZ), a zeolite mineral considered for use in this application, have resulted in slightly inconsistent conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Comparison of Extended and Accelerated VOG Tests

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas streams of a processing plant, including ¹²⁹I. The dissolver off-gas and more dilute vessel off-gas streams (VOG) must both be targeted when mitigating ¹²⁹I environmental release because each contains some amount of iodine. Iodine-129 will likely be found as elemental iodine (I₂) and methyl iodide (CH₃I) in these off-gas streams. Reduced silver-exchanged mordenite (AgZ or Ag⁰Z) has been investigated as a potential sorbent for iodine abatement and is studied here under prototypic VOG conditions. Because of the relatively low iodine abundance and high flow rates of the VOG, total iodine concentrations are expected to be in the parts per billion range. Thus, VOG experiments need to run for extended durations at low concentrations to reach sorbent saturation. Because the sorbent will be exposed to oxidizing gas streams for extended periods of time, aging effects and sorbent degradation need to be considered when designing a sorbent-based abatement system. Three sets of experiments were performed with the aim of determining how the adsorption of iodine by AgZ is affected by differing test durations and gas compositions. The first tested the capacity of sorbent aged for 8 months under a humid air stream. The second tested differences in sorbent capacity and mass transfer zone (MTZ) length during high-concentration (1200 ppb v ) CH₃I loading in a humid nitrogen gas (N₂) stream and a humid air gas stream over 28 days. The third tested sorbent capacity and MTZ length during low concentration (< 200 ppb v ) CH₃I and I₂ loading in a humid air stream over 9 months. The results of these tests suggest that AgZ capacity drops by ~50% after 1 month of aging, by ~60% after 2 months of aging, and then does not continue to decrease significantly at up to 8 months of aging. One-month aging tests that compared N₂ and air as the gas stream diluent resulted in similar maximum loading capacities and overall loading curves, indicating that the effects of aging cannot be mitigated by avoiding air as the balance gas. The 9-month extended VOG tests did not result in clear sorbent saturation at the inlet, but it can be inferred using an assumption of a maximum capacity of 45 mg I/g AgZ. Applying this assumption, then the MTZ of CH₃I is 12.8 cm, and the MTZ of I₂ is 12.4 cm. These results are similar to previous tests. Comparisons to tests completed at Idaho National Laboratory suggest that the CH₃I MTZ may be dependent on concentration. Future work should re-evaluate the design of the VOG iodine capture system with this updated data and should seek to understand fundamental characteristics of CH₃I adsorption by AgZ, such as the effect of concentration, gas velocity, and gas composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Actinide N-Donor Thermodynamics: Expanding the f-element Covalency Dialogue. Final report

Resolving the chemistry and physics of f-electrons is one of the grand challenges of science for energy technology. A central component of this challenge is describing covalency in actinide-ligand interactions. Crystallography, x-ray spectroscopy and computational techniques have provided reasonable, and in some cases remarkable, support for covalency in actinide-ligand interactions. Courtesy the quality of the results obtained, a significant portion of the covalency dialogue has been forwarded through solid-state and computational studies of the more stable actinides (Th, U, Pu, Np). However; recent reports have questioned whether covalency in actinide interactions increases or decreases across the series. Examination of less-stable trans-actinides (Am, Cm, Bk, Cf and Es) by the aforementioned approaches can be rapidly limited due to material availability, radiological hazards and experimental data to validate computational models. Furthermore, limited thermodynamic data exists to confirm covalency in actinide solution phase interactions that are highly relevant to the remediation and reprocessing of used nuclear fuel. The best-defined interactions of actinides with soft donors in aqueous solutions involve (poly)aminopolycarboxylate (APC) ligands. The chelate effect and binding affinity with the acetic acid APCs subgroups encourages amine interactions with the actinide metal center. In the absence of these factors, amine interactions with actinides are too weak to overcome the protective hydration shell of the dissolved ion. The ability for APCs to force actinide interactions with soft nitrogen donors (as defined by Pearson’s Hard Soft Acid Base theory) encourages the application of these ligands in a variety of processes for actinide recovery from nearly chemically identical lanthanides. The ability to functionalize the amine center of the ligand in a variety of capacities (adding additional amine groups, exchanging the conventional acetate group for an acetate group, etcetera), allows for the APC ligand to serve as a thermodynamic probe for actinide-nitrogen interactions. Perhaps the single most significant breakthrough during the previous funding cycle was the observation that covalency for the transplutonium part of the actinide series can be increasingly influenced with energy degeneracy driven covalency as the actinides become heavier. This was observed most predominantly with dipicolinic acid, but extensions of f-orbital degeneracy were found to affect aliphatic aminopolycarboxylate-actinide complexes through einsteinium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Analysis of Organoiodide Adsorption Mechanisms

If the United States were to engage in the reprocessing of used nuclear fuel, radioactive iodine must be removed from multiple plant off-gas streams to comply with governing regulations. One of these streams is the vessel off-gas (VOG), which arises from the separations process and is expected to contain iodine in primarily organic iodine forms and at parts-per-billion concentrations. The relative lack of knowledge surrounding organic iodine removal from the VOG prompted the US Department of Energy’s Office of Nuclear Energy to initiate experimental efforts targeted at understanding organic iodide removal from prototypic VOG streams. At Oak Ridge National Laboratory (ORNL), this effort has focused on developing a comprehensive understanding of iodine removal from VOG streams by using silver-based sorbents. An important aspect of this testing is developing an understanding of how the sorption of methyl iodide (CH 3 I), the most studied organic iodide species to date, compares with the sorption of other volatile organic iodide species potentially present in the VOG. The work presented here reflects initial testing that will continue to be developed, and final results will be incorporated into an end-of-year report that details a multiyear testing campaign designed to understand iodine mitigation from VOG streams. The goal of these experiments was to determine whether similar reaction pathways govern both CH 3 I and iodobutane (C 4 H 9 I) sorption onto silver mordenite (AgZ), a common silver-based iodine sorbent. More specifically, the authors hypothesized that the sorption of C 4 H 9 I by AgZ will result in the formation of butanol (C 4 H9OH). Effluent monitoring of CH 3 I sorption testing has confirmed methanol production, but analogous monitoring of effluents from C 4 H 9 I sorption studies has not been performed. A series of tests was completed to test this hypothesis with the aim of detecting C 4 H 9 OH downstream of the AgZ bed. Test conditions varied the bed depth, gas stream humidity, and bed temperature post sorption. The effluent gas stream downstream of the AgZ bed was sampled by using gas-tight syringes, and these samples were analyzed by a gas chromatograph coupled to a mass spectrometer. Thin bed C 4 H 9 I tests conducted at -65 and 0°C dew points did not result in C 4 H 9 OH detection in the effluent. This suggests that the sorption mechanism and subsequent reactions could be different from those observed for CH 3 I. The sorption testing continues; additional results will be included in final end-of-year report and will increase the fundamental understanding of organic iodide sorption by AgZ from VOG streams.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Iodine Capture Studies of Silver Mordenite and Novel Alternative Metal Sorbents

Radioiodine is one of the radionuclides of concern when considering reprocessing of used nuclear fuel (UNF). It is expected to be released primarily in the dissolver off-gas (DOG) stream. Sorbents that target I 2 removal from a UNF processing off gas must perform efficiently under elevated temperatures in the presence of water vapor and nitrogen oxide gasses (NOx). Many studies have been conducted that examine the adsorption of iodine species in the presence of NOx gases through reduced silver-exchanged mordenite (Ag 0 Z), a zeolite mineral. This study characterizes the effects of iodine, water vapor, and nitrogen dioxide on the adsorption of I 2 by Ag 0 Z at lower temperatures suitable for comparison with novel Cu- and Bi- bearing sorbents which offer a non-RCRA alternative. When exposed to ‘harsh’ conditions such as I 2 , H 2 O, and NO 2 vapors carried by air, Cu 2 S-polyacrylonitrile (PAN) is more efficient than Ag 0 Z and all other novel sorbents, providing with 11% I capture efficiency. However, under ‘ideal’ conditions in which I 2 is carried by dry air), Ag 0 Z is more efficient than Cu 2 S-PAN and other novel sorbents, providing 74% I capture efficiency. To investigate this result and the overall performance of all sorbents tested, several characterization techniques were employed, including thermal gravimetric analysis, scanning electron microscopy – electron dispersion X-Ray spectroscopy, powder X-ray diffraction, and X-ray photoelectron spectroscopy. These techniques are discussed within this report. This document also includes a design and test plan update from Idaho National Laboratory as the next step in testing the high performing sorbents from thin-bed tests carried out by Oak Ridge National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Aluminophosphate Waste Forms for Immobilizing Cations from Electrochemical Salt Wastes

This report provides experimental details and results of evaluating aluminophosphate waste forms for treating and immobilizing the salt cations from salt wastes generated during electrochemical reprocessing of used nuclear fuel. In the waste form process for these materials, chloride salt streams are reacted with NH 4 H 2 PO 4 , the chlorine is removed from the salts and driven off as NH 4 Cl (a solid condensate that can be captured), and then the product can be vitrified in conjunction with glass-forming chemicals (e.g., Fe 2 O 3 , Al 2 O 3 ) to create a high-durability waste form. This study was initiated with some literature review on aluminophosphates containing high alkali oxide content and some of this information is summarized in this report. Following literature review, three new samples were synthesized where two contained Fe 2 O 3 +Al 2 O 3 (i.e., samples G3 and G5 ) and one was only Al 2 O 3 (Fe 2 O 3 -free) (i.e., sample G6 ). In addition to these samples, G1 was also made, which is the baseline reference waste form referred to as DPF5-336 (made without Al 2 O 3 ). Samples G1, G3, G5, and G6 had phases of Li 3 Fe 2 (PO 4 ) 3 (likely), monazite (below XRD detection limits), AlPO 4 , and AlPO 4 , respectively. Characterizations on these materials included optical images, scanning electron microscopy, energy dispersive X-ray spectroscopy, and X-ray diffraction. These samples were shipped to Argonne National Laboratory for chemical durability testing. Depending on how the samples perform in these tests, an additional phase of aluminophosphate formulations could be designed and tested. This report completes the milestone M4FT-23PN030104041 with details provided in Appendix B.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Radiological Degradation of Tetraoctyl Diglycolamide in n-Alkyl Solvents: Influence of Solvent Ionization Potential

Reprocessing of used nuclear fuel reduces the requirements for long-term storage repositories. Cost-efficient processes depend on knowledge of radiation effects, but gaps occur in understanding how radiation results in the degradation of fuel separation molecules. This project investigates the influence of different n-alkyl solvents on the radiolytic fragmentation behavior of N,N,N',N'-Tetraoctyl-3-oxapentane-1,5-diamide (TODGA), a promising ligand for extraction of the lanthanides and minor actinides from dissolved nuclear fuel. Findings from this work suggest a linkage between the ionization potential of the solvent and the degradation products observed in the resulting mass spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Crystalline Aluminosilicates for Radionuclide Remediation

Crystalline aluminosilicates have a wide range of applications such as media for the capture and/or immobilization of radionuclides. Several different types of metal-exchanged zeolites have been successfully demonstrated as effective sorbents for gaseous, aqueous, or solid-state capture and immobilization of radionuclides including iodine [e.g., I2(g), CH3I(g)], xenon, krypton, radium (e.g., 226Ra and 228Ra), selenium (e.g., 79Se6+), technetium (e.g., 99TcO4-), strontium (e.g., 90Sr2+), cesium (e.g., 135Cs+, 137Cs+), mixed alkali halide streams (i.e., iodide and chloride salts), and actinides (e.g., U6+, Th4+). These zeolites include sodalite (SOD), mordenite (MOR or Z-type), faujasite (FAU), Linde type-A (LTA), cancrinite (CAN), gismondine (GIS), and heulandite (or clinoptilolite) (HEU). The structural building blocks of these seven materials vary from cages (e.g., a-cages, ß-cages, e-cages) to channels linking various cage types and these are mostly comprised of aluminosilicates. This chapter will delve into the crystallographic differences between these framework materials and provide a brief overview of literature data where these seven crystalline aluminosilicates have been utilized for the capture and/or immobilization of radionuclides. Metal-exchanged zeolites (e.g., MOR, FAU, LTA) have often been the source of study for gaseous capture of iodine where the more promising compositions include Ag-exchanged sorbents; these types of materials can be made in a variety of robust forms including berl saddles and extrudates. Sodalite can be used as an immobilization strategy for halide waste streams (e.g., Cl, I), including those from electrochemical reprocessing of used nuclear fuel, as well as TcO4–. Cancrinite has been demonstrated to enhance sorption capacity of sediments for several problematic ionic radionuclides including Cs+, SeO42–, and TcO4–. Actinides such as U6+ and Th4+ have been successfully captured from solutions using MOR, FAU, LTA, and HEU-type aluminosilicates. Finally, mixtures of aluminosilicates have been included in a set of zeolite materials that show promise as sorbents for 226Ra and 228Ra removal from mine waters.

zeolites, aluminosilicates, radionuclide remediati↗
DOE OSTI2024

Acoustic Monitoring of Pyroprocessing for Safeguards

As pyroprocessing continues to be an attractive option for the reprocessing of spent nuclear fuel worldwide, safeguards technologies are needed to address the monitoring capabilities that can help state level authorities, or the International Atomic Energy Agency (IAEA) maintain continuity of knowledge of the plant operations. Idaho National Laboratory (INL) is studying the possibility of using acoustic monitoring as a means to monitor a pyroprocessing facility for safeguards purposes. This paper discusses the experimental design and some preliminary results of tests conducted at the Fuel Conditioning Facility, a pyrochemical capable facility, at INL.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2026

Non-Equilibrium Actinide Radiation Chemistry and the Nuclear Fuel Cycle

Invited John and Naomi Fackler Lectureship in Chemistry and English seminar at Valparaiso University, IN, USA. Actinides are inherently unstable elements that frequently coexist with other radioisotopes, generating intense ionizing radiation fields that drive the formation of non-equilibrium oxidation states. These transient species exert a profound mechanistic influence on the radiation response of actinide-containing systems due to their unique redox chemistry. Despite their importance, they remain poorly understood, yet such insight is essential for advancing actinide science and accurately predicting radiation-driven behavior. Actinide separations—critical for nuclear energy technologies, strategic deterrence, space exploration, and nuclear medicine—depend on precise control of actinide oxidation states to recover targeted elements from complex matrices such as used nuclear fuel. However, during these processes, actinides, their coordination complexes, and the separation media are all exposed to intense, multicomponent (alpha, beta, gamma, etc.) radiation fields that can alter process efficiency, selectivity, and chemical stability. Understanding, controlling, and mitigating radiation-induced reactions is therefore key to innovating and optimizing next-generation separation technologies. This seminar will provide an overview of the nuclear fuel cycle and non-equilibrium actinide radiation chemistry in the context of recovering actinides from used nuclear fuel, with a particular emphasis on direct-dissolution–based reprocessing strategies. We will explore time-resolved electron pulse radiolysis and gamma dose accumulation studies to elucidate the molecular-level roles of radiation-driven, non-equilibrium actinide species in process performance and in the radiolytic stability of organic ligands used for actinide recovery. These insights offer new pathways for designing advanced separation methods and next-generation solvent systems, with broad implications for the future of the nuclear fuel cycle.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
NASA NTRS1975

Reconstituted asbestos matrix for fuel cells

Method is described for reprocessing commercially available asbestos matrix stock to yield greater porosity and bubble pressure (due to increased surface tension), improved homogeneity, and greater uniformity.

Mcbryar, H.↗
DOE OSTI2020

Progress and Technology Assessment of Melt-Cast Ceramic Waste Forms

The Savannah River National Laboratory (SRNL) conducted research from FY2011 – FY2020 to develop ceramic waste forms to treat High-Level Waste (HLW) resulting from reprocessing of commercial Used Nuclear Fuel (UNF) as part of the Department of Energy (DOE) office of Nuclear Energy (NE) Fuel Cycle Technologies (FCT) research program. The objective of the research was to develop a reference ceramic waste form and process technology needed to immobilize the combined HLW raffinate stream, including undissolved solids and soluble technetium, and potentially the TRU waste stream resulting from reprocessing processes. This report compiles the relevant literature produced under, and pertinent to, the program objectives and provides a perspective on the technology development as well as prospects for future.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗