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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Characterization of Pu-238 Heat Source Granule Containment

The Milliwatt Radioisotopic Themoelectric Generator (RTG) provides power for permissive-action links. Essentially these are nuclear batteries that convert thermal energy to electrical energy using a doped silicon-germanium thermopile. The thermal energy is provided by a heat source made of 238 Pu, in the form of 238 PuO 2 granules. The granules are contained by 3 layers of encapsulation. A thin T-111 liner surrounds the 238 PuO 2 granules and protects the second layer (strength member) from exposure to the fuel granules. An outer layer of Hastalloy-C protects the T-111 from oxygen embrittlement. The T-111 strength member is considered the critical component in this 238 PuO 2 containment system. Any compromise in the strength member seen during destructive testing required by the RTG surveillance program is characterized. The T-111 strength member is characterized through Scanning Electron Microscopy (SEM), and Metallography. SEM is used in the Secondary Electron mode to reveal possible grain boundary deformation and/or cracking in the region of the strength member weld. Deformation and cracking uncovered by SEM are further characterized by Metallography. Metallography sections are mounted and polished, observed using optical microscopy, then documented in the form of microphotographs. SEM mat further be used to examine polished Metallography mounts to characterize elements using the SEM mode of Energy Dispersive X-ray spectroscopy (EDS).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Why SRS Matters - H Canyon

A video series presenting an overview of the Savannah River Site's (SRS) mission and operations. Each episode features a specific area/operation and how it contributes to help make the world safer. This episode features H Canyon's mission and operations. H Canyon and its adjoining H B Line facility represent the last full-scale radio chemical processing capabilities left in the United States.

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

Idaho National Laboratory Radioisotope Power Systems Nuclear Operations: Readiness Assessments Supporting a Nuclear-enabled NASA Mission

Idaho National Laboratory Radioisotope Power Systems Nuclear Operations: Preparations, Documentation, Readiness Assessments and Conduct of Operations Supporting a Nuclear-enabled NASA Mission Kelly L. Lively, Eric S. Clarke, Robert P. Gomez Idaho National Laboratory, Idaho Falls, ID 83415 208-533-7388; Kelly.Lively@inl.gov Abstract. The Radioisotope Power Systems (RPS) Program, located at Idaho National Laboratory (INL), is responsible for assembling, testing, and delivering plutonium oxide-fueled RPSs for use in powering missions in remote, harsh environments such as deep space. An informative presentation will be given discussing the preparations, documents, readiness assessments, human capital and conduct of operations involved in performing nuclear operations to support providing these systems to end users for the Department of Energy (DOE). DOE is the start-up authority for performing nuclear operations. Readiness for start-up is determined through independent assessment against established acceptance criteria to ensure activities can be performed safely and within a well defined nuclear safety envelope. There is also an RPS Program approval element requiring additional readiness review before nuclear operations can begin. The programmatic readiness review is performed to ascertain readiness to proceed with nuclear operations from the perspective of product quality. The assessments and reviews are conducted in series in the following order of progression: Management Self Assessment (MSA), Contractor Readiness Assessment (CRA), DOE Readiness Assessment (DOE RA) and lastly, the Programmatic Production Readiness Review (PRR). The CRA, DOE RA, and PRR requires an independent technical review team. Twelve assessment criteria are reviewed and a selection of the criteria is identified for review on a graded approach. Typically, the assessments/review criteria requires, at a minimum, review of operating instructions to ensure technical safety requirements are adequately identified, review of training records to ensure personnel are adequately trained to perform the specified work scope, personnel are interviewed to determine adequacy of level of knowledge for work scope, and a high-fidelity performance of the operation to ensure the operating instructions and conduct of operations are adequate to perform the work scope. As each assessment/review is conducted a formal report delineating any issues in the form of findings, observations, and noteworthy practices will be issued. Before start-up approval is obtained all issues must be resolved to the satisfaction of the individual teams. Start-up notification is formally communicated by memorandum from DOE. Programmatic approval is documented in a Segmented Readiness Review where vested Program representatives in the RPS community (to include DOE NE 75 and DOE ID representatives) ensure personnel, documentation, and materials are in place to perform the activity. The four assessments/reviews can be completed in a 16-week period of time, provided ample resources and a relatively small work scope is identified. RPS assembly and testing operations to support the Mars 2020 Mission, the next planned space mission using a nuclear power system, will require about of year of assessments/reviews before the nuclear operations are performed. Keywords: INL, Operations, Assessments, DOE

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recycle of Am-241 obtained from long term stored plutonium for use in radioisotope power systems - 15035

Pu-238 has been used as a power source for spacecraft since the early days of space exploration. It has proven to be an effective source of power where the use of solar generated power is impractical. Its application for use in spacecraft power systems was developed by the USA and Russia who have spent many $ 100s of millions on its development and production. However, the original facilities for the production of Pu-238 are no longer in use and existing worldwide stockpiles have reduced to levels sufficient for only one or two further missions. Production of Pu-238 requires considerable facilities including a nuclear reactor and reprocessing plants that are very expensive to build and operate. The European Space Agency (ESA) has assessed the options for post-launch power generation in future European space missions and has made the decision to pursue the use of Am-241 as an alternative isotope to power future European Radioisotope Power Systems (RPS). What makes Am-241 attractive for the European market is its 100% isotopically pure production from the decay of Pu-241 in separated civil plutonium stockpiles. Historically, Europe has relied on collaborations with the USA or Russia to access these nuclear power sources. During 2009, the European Space Agency (ESA) funded a project to examine the cost and practicality of establishing a European source of material suitable for Radioisotope Power Systems (RPS). The study concluded that a small number of radioisotopes could be used in RPSs. On further review two leading candidates, Am-241 and Pu-238, were selected for further study. Whilst Pu-238 was recognized as the isotope of choice for use in RPSs, the costs of setting up a European based production facility was prohibitively expensive. Although Am-241 has a number of limitations compared to Pu-238, the much reduced production costs make Am-241 an attractive option for European based production of RPSs. Despite its lower power density of ∼ 0.11 Wth/g, Am-241 is viewed as a potential alternative to Pu-238 because of its availability within the nuclear fuel cycle. During the reprocessing of spent nuclear fuel plutonium is separated as a chemically pure PuO{sub 2} product, containing around 3-11% w/w of the Pu-241 isotope, and placed into storage. During storage the Pu-241 isotope beta decays to Am-241 with a half-life of 14.4 years. {sub 94}{sup 241}Pu → (β 14.4 y) {sub 95}{sup 241}Am. This process has resulted in many 100 kgs of americium growing into the civil plutonium stockpile in the UK; material that is of value in generating heat sources and is able to be recycled. The key benefit of Am-241 as a potential European RPS source isotope is the availability of feed material along with the facilities to process it. It is expected that aged civil PuO{sub 2} will contain 25-50 g Am-241 per kg PuO{sub 2} depending on the source. A conceptual flowsheet using facilities for separating americium from plutonium is expected to produce on average 8.4 kg Am/year. ESA are now following their published road-map of studies to develop flight ready RTGs and RHUs. This has included the development and testing of the flowsheet to be used for production of Am-241. The project has focused on underpinning the conceptual flowsheet, using aged plutonium in NNL's PuMA laboratory. This has underpinned the design of a plant to produce a steady supply of Am-241. The scope of the overall project is 'store to store' - the starting point is aged plutonium currently housed in existing Sellafield stores and the end point is the storage of separated plutonium and americium powders on the Sellafield site. The full scale process is planned to be housed within the NNL's Central Laboratory in an existing facility designed for plutonium active operations. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Opportunities to Reuse and Recycle Redundant Radioisotopes - 16023

Sixty plus years of nuclear research and operation in the UK has produced significant quantities of nuclear material. To those involved in decommissioning a site or facility, this material represents a liability that needs to be managed and eventually disposed of. However to those in the wider nuclear industry and supply chain, the radioisotopes represent a potential resource of valuable, and in some cases irreplaceable material for future work. There is therefore an opportunity to create a win-win by identifying material that could be recycled or reused, removing it from being a liability and creating value from what is considered a waste. The UK is currently focused on the clean up of nuclear sites and the safe disposal of legacy nuclear material. Whilst the bulk of legacy material on these sites is mixed wastes in various solid and liquid forms, some of it consists of material produced for research purposes in the form of radioisotopes or exotic fuels. This has resulted in a plethora of unirradiated and irradiated materials in store across the UK including: Depleted and natural Uranium, Enriched Uranium, Mixed U/Pu oxides, Uranium carbide and mixed U/Pu carbides, Thorium as metal and oxides, Research radioisotopes. The material exists in almost every conceivable state from pure metal and oxides to solutions. Some material is well characterised and free from extraneous debris, other material is poorly characterised and contaminated with chemicals, experimental residues and sweepings. In addition, to reduce the risk and potential hazard across the UK there has also been a period of consolidation of obsolete radioactive sources from a wide variety of industrial or medical facilities to secure storage within a licensed site. The waste hierarchy is an established principle for the management of wastes, radioactive or otherwise. Its principles are set out and it gives top priority to prevention in the first place. Where waste is produced, it gives priority to preparation for re-use, then recycling before consideration for disposal as a last resort. The term wastes may only be applicable to the owner of the material where they may view the material as unwanted, unusable, or a by-product. It does not necessarily follow that the material is a waste to all potential owners. Therefore it is important to consider the potential value of material before it is considered for disposal since there is a significant and growing market for radioisotopes across the world. This paper will outline some of the opportunities in the UK for the recycle of surplus radioisotopes and orphan material. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Collaborative Research Programme in Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories (DISTINCTIVE) - 16466

The UK government remains committed to nuclear power as an important part of the energy generation capacity over the next 30-50 years and possibly beyond, with aspirations for both new-build power stations and lifetime extension of the existing fleet. A central component of public acceptance for any new-build programme is the need to demonstrate our ability to safely manage and dispose of high- and intermediate-level wastes from legacy nuclear operations. Clear milestones associated with the needs of waste management and decommissioning have been provided by the UK government, and most rely on further research and technical developments being delivered over the next 10-20 years. The DISTINCTIVE (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) consortium is carrying out research that addresses the broad area of nuclear waste and decommissioning, bringing together key industry partners and leading academic researchers from 10 of the UK's leading research intensive universities. The research being carried out is multi-disciplinary and covers both fundamental and applied topics, with research projects clustered in four major themes: AGR, Magnox and exotic spent fuels; PuO{sub 2} and fuel residues; legacy ponds and silo wastes; and structural integrity. The consortium builds upon and consolidates the work of previous Research Councils' UK Energy Programme-funded projects in the same field although, importantly, this new project draws in researchers from a larger group of universities and increases the multi-disciplinary nature of the group. All participating research groups have a strong track-record of supporting the nuclear industry through basic and applied research as well as consultancy activities. Key aims of the consortium are to extend and develop a relevant academic skill base in the UK associated with the needs of the nuclear waste management industry, to train the next generation of UK researchers, equipping them with skills and experiences relevant to nuclear waste management and decommissioning issues, and to carry out cutting-edge research that has the potential to provide new and innovative pathways to better management of both legacy and future nuclear wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

An Industrial Perspective of Research Within the DISTINCTIVE Program - 18045

UK Engineering and Physical Sciences Research Council (EPSRC) and three industrial partners, the Nuclear Decommissioning Authority (NDA), the National Nuclear Laboratory (NNL) and Sellafield Ltd (SL) are supporting the DISTINCTIVE (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) consortium of eleven universities to carry out research in four theme areas of importance to the UK nuclear industry: Advanced Gas-cooled Reactor (AGR), Magnox and exotic spent fuels; PuO{sub 2} and fuel residues; legacy ponds and silo wastes; and structural integrity. This paper describes the main industry challenges in each of these theme areas and some of industry led activities to address these challenges, and provides examples of where the DISTINCTIVE research output is assisting the nuclear industry in meeting these challenges. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

The DISTINCTIVE University Consortium: An Overview - 18102

The UK government remains committed to nuclear energy having an important role to play in delivering a secure, low carbon and affordable energy future. Successful delivery of their strategy recognises the need for underpinning research, in addition to the provision of a skilled workforce and international collaboration as key enablers. One crucial area is the decommissioning and clean-up of all existing civil nuclear licensed sites and the implementation of geological disposal of highly active wastes, with the Nuclear Decommissioning Authority responsible for the delivery of policy aims with respect to legacy wastes. A central component of public acceptance of this strategy is a demonstrated ability to safely manage and dispose of waste, with acceptance linked to having scientifically robust routes for its safe clean-up and disposal. Similarly, the UK Engineering and Physical Sciences Research Council (EPSRC) strategy is to maintain investment in nuclear fission, with the development of approaches for implementing safe, cost-effective clean-up, decommissioning and waste disposal of existing and future nuclear sites and facilities a strategic focus. The EPSRC-supported DISTINCTIVE (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) consortium is carrying out research that addresses the broad area of nuclear waste and decommissioning, bringing together academic researchers from 11 of the UK's leading research intensive universities and key industry partners. When allied to the strong links of the partners with the industrial and international R and D community, the project provides an excellent route for building a greater capacity for R and D within the UK, underpinning the strategic needs noted. The consortium also addresses the skills agenda in training a significant number of researchers with high level skills of direct relevance to nuclear waste and decommissioning. This paper describes the aims and objectives of the work being undertaken, and the technical work underway within each of the research theme areas: AGR, Magnox and exotic spent fuels; PuO{sub 2} and fuel residues; legacy ponds and silo wastes; and structural integrity. Projects using radioactive materials are also considered. As the project enters its final year, consideration is given to the impacts of the project for both academic and non-academic beneficiaries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Geochemical Simulations of Plutonium Redox Speciation at Increasing Ionic Strengths - 18567

In salt repositories like the Waste Isolation Pilot Plant (WIPP), reducing conditions are expected, supporting two aqueous species of plutonium: Pu(III) and Pu(IV). Of these states, Pu(III) is more soluble than Pu(IV). Because Pu behavior is uncertain in WIPP brines under reducing conditions, performance assessment calculations assume a 50/50 partitioning between the two, justified through calculations using dilute solutions. Recent experimental evidence suggests an increasing importance for Pu(III). To reexamine this assumption, we modeled Pu speciation in NaCl and dilute WIPP brine solutions equilibrated with a Pu solid, including PuO{sub 2} and Pu(OH){sub 3}. To maintain reducing conditions, sulfate and sulfide were added. Models were performed using PHREEQC and the Specific Ion Theory (SIT) database on dilutions ranging from 0.1 M to 3 M. Models were compared to calculations using the Pitzer database with PuO{sub 2} and with analogs Am(III) and Th(IV), all up to 11 M. Results indicate that under reducing conditions, Pu(III) concentrations increase with ionic strength and can be up to 80 fold higher than Pu(IV). While concentrations varied due to differences in solubility for each solid, Pu(IV) concentrations remained nearly constant in all solutions. Analogs showed similar trends, though Am(III) is 3 times more soluble than Pu(III) while Th(IV) is 6 times more soluble than Pu(IV). Pu(IV) concentrations using Pitzer calculations were similar to those with SIT. Our study indicates that assuming equivalent partitioning of Pu species may under-represent concentrations during a release. They also show that in PA calculations there is a need to consider redox state at high ionic strength rather than adopting an assumption of equal partitioning between the two charges. (authors)

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

Oxygen potential measurement and point defect chemistry of UO{sub 2}

Many studies on the oxygen potential of UO{sub 2} have been carried out so far. However, the oxygen potential data for UO{sub 2} near the stoichiometric composition in the high temperature region (1673-1873 K) are limited. Recently, Kato's group had investigated oxygen potential of (U, Pu)O{sub 2} as functions of Pu content, oxygen-to-metal (O/M) ratio and temperature. They evaluated the data based on defect chemistry. The UO{sub 2} data are fundamental to evaluate oxygen potential in UO{sub 2}-PuO{sub 2} solid solution. In this work, the oxygen potential data of UO{sub 2+x} were extended to the high temperature range of 1673-1873 K by gas equilibrium method. The measured data were analyzed based on a defect chemistry model. Kroeger-Vink's diagram were obtained in UO{sub 2+x}. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Custom Equipment Development for Processing of Surplus Plutonium

The Strategic Laboratory Assessment (SLA), a collaborative team of SRNL and ORNL personnel, has been established to advance the objectives of the Surplus Plutonium Disposition (SPD) Project, by identifying and developing technologies to accelerate disposition, reduce life cycle costs, minimize worker radiation exposure, improve worker safety, and minimize Surplus Plutonium Disposition Program risks. [1] The SLA team has identified can cutting and plutonium (Pu) oxide size reduction as two glovebox processes where technology enhancements would be valuable. The DOESTD-3013 package currently in use for Pu downblending requires cutting two nested cans before the inner convenience can that holds the Pu oxide may be accessed for further processing. A rotary tubing-style cutter is used for opening the 3013 packages within the glovebox. Collet changeouts are required between cutting of the outer and inner cans. The SLA team is currently developing and testing an adjustable-clamp can cutter design that eliminates collet changeouts and allows cutting of the outer and inner can at the same time, resulting in significant reduction of radiological dose and process time, as well as improved ergonomics. To meet the Pu oxide particle size requirement, size reduction of Pu oxide agglomerations must be performed within the process gloveboxes. The SLA team has identified jaw crushing technology as an alternative to the currently employed rotary mill. Jaw crusher advantages include reduced dust within the glovebox, increased batch sizes, and easier integration with other glovebox processes due to the flow-through nature of jaw crushing. Commercially manufactured jaw crushers are either too large and/or too heavy for implementation in the SPD gloveboxes, so the SLA team is developing and testing a custom jaw crusher to meet the needs of the SPD Project.

Krementz, Daniel [Savannah River National Laborato↗
DOE OSTI2017

Modulation of Tetravalent Actinide Oxalate Morphology through Organic Compounds

Spent nuclear fuel of pressurized water reactors mainly consists of uranium, plutonium, minor actinides and fission products. The long term radiotoxicity of wastes is mainly due to plutonium and minor actinides such as neptunium, americium and curium. So, in order both to reduce the radiotoxicity of the ultimate wastes and to save uranium resources, researches for current and new generations of nuclear plants focus on the recycling of plutonium and minor actinides, to fabricate new mixed oxides nuclear fuel pellets (MOX) or transmutation targets. In this aim, production of actinide oxides is necessary. Currently, according to the Purex process, the spent fuel is dissolved in nitric acid and then valuable actinides (uranium and plutonium) are separated and purified from fission products and minor actinides in the Areva plant 'La Hague'. The resulting plutonium nitric solution is converted into a solid phase by oxalic conversion: the oxalic precipitation is followed by the calcination of the obtained oxalate into plutonium oxide. Considering either the MOX fuel or the transmutation targets fabrication from actinide oxides, the control of the particle structure and microstructure is a key factor related to the ability of oxide powder to sinter and shape as pellets. In the oxalic conversion process, the morphology of the precipitated oxalate is strongly influenced by the precipitation conditions and is kept during the calcination from oxalate into oxide. In this context, the aim of this work is to control the morphology of the final actinide oxide by changing the oxalic precipitation conditions. A lot of possibilities can be considered to modulate the morphology of actinide oxalates. Numerous parameters such as temperature, pressure, the presence of a magnetic field, the operating process (seeding, emulsions)... can modify the precipitated particles shape. Additives are also able to play a major role at the precipitation stage by means of different mechanisms involving either complexation, surface covering or structure insertion. This study focuses on this way to modify the morphology of actinides oxalates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Radiation damage and annealing in plutonium tetrafluoride

Plutonium tetrafluoride that was separated prior to 1966 at the Hanford Site in Washington State was analyzed at the Pacific Northwest National Laboratory (PNNL) in 2015 and 2016. The plutonium tetrafluoride, as received, was an off-normal color and considering the age of the plutonium, there were questions about the condition of the material. These questions had to be answered in order to determine the suitability of the material for future use or long-term storage. Therefore, Thermogravimetric/Differential Thermal Analysis and X-ray Diffraction evaluations were conducted to determine the plutonium’s crystal structure, oxide content, and moisture content; these analyses reported that the plutonium was predominately amorphous and tetrafluoride, with an oxide content near ten percent. Freshly fluorinated plutonium tetrafluoride is known to be monoclinic. During the initial Thermogravimetric/Differential Thermal analyses, it was discovered that an exothermic event occurred within the material near 414°C. X-ray Diffraction analyses were conducted on the annealed tetrafluoride. The X-ray Diffraction analyses indicated that some degree of recrystallization occurred in conjunction with the 414°C event. The following commentary describes the series of Thermogravimetric/Differential Thermal and X-ray Diffraction analyses that were conducted as part of this investigation at PNNL, in collaboration with the University of Utah Nuclear Engineering Program.

McCoy, Kaylyn↗
DOE OSTI2013

Implications of Plutonium isotopic separation on closed fuel cycles and repository design

Advances in laser enrichment may enable relatively low-cost plutonium isotopic separation. This would have large impacts on LWR closed fuel cycles and waste management. If Pu-240 is removed before recycling plutonium as mixed oxide (MOX) fuel, it would dramatically reduce the buildup of higher plutonium isotopes, Americium, and Curium. Pu-240 is a fertile material and thus can be replaced by U-238. Eliminating the higher plutonium isotopes in MOX fuel increases the Doppler feedback, simplifies reactor control, and allows infinite recycle of MOX plutonium in LWRs. Eliminating fertile Pu-240 and Pu-242 reduces the plutonium content in MOX fuel and simplifies fabrication. Reducing production of Pu-241 reduces production of Am-241 - the primary heat generator in spent nuclear fuels after several decades. Reducing heat generating Am-241 would reduce repository cost and waste toxicity. Avoiding Am- 241 avoids its decay product Np-237, a nuclide that partly controls long-term oxidizing repository performance. Most of these benefits also apply to LWR plutonium recycled into fast reactors. There are benefits for plutonium isotopic separation in fast reactor fuel cycles (particularly removal of Pu-242) but the benefits are less. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Acidic dissolution behavior of U containing ZrO2–MgO ceramics

This study explores the possibility of dissolving zirconia-magnesia inert matrix fuel containing uranium oxide as a fissile material and plutonium homolog and erbium oxide as a burnable poison with nitric and sulfuric acid as a potential first step in a reprocessing scheme. The progress of the dissolution is followed by monitoring the amount of material in solution by inductively coupled plasma-atomic emission spectroscopy, assessing the speciation of the material by time resolved laser fluorescence spectroscopy, and determining and quantifying the crystalline phases present in the remaining residue by X-ray diffraction. This study has shown a linear incongruent dissolution of the cubic zirconia phase in concentrated nitric acid under certain chemical compositions, while the magnesium oxide phase is completely soluble. In sulfuric acid uranium, erbium, and magnesium are soluble to different extents while zirconium forms a colloidal suspension that conglomerates and settles out of solution. The feasibility of the dissolution of zirconia-magnesia inert matrix fuel with nitric and sulfuric acid for reprocessing is discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Radiation-Induced Plutonium Redox Chemistry

Plutonium plays a key role in global actinide research and nuclear fuel cycle technologies, and yet, our fundamental understanding of its inherent radiation-induced chemical behavior is limited. These radiation-induced processes cannot simply be switched off, as they are as fundamentally inherent to plutonium as the impact of relativistic effects on its f-electrons. In less chemically complex actinide systems, such as aqueous solutions of neptunium and americium, , radiolysis products play a significant role in the redox cycling of their oxidation states. However, plutonium's multiple, coexisting, and chemically active oxidation states, which comprise of bare ions and dioxo cations, provide additional redox pathways that complicate radiation-induced processes. Oxidation state control is critical for the manipulation of plutonium, especially in used nuclear fuel reprocessing technologies, where oxidation specific states are successfully extracted, and others rejected. Consequently, mechanistically understanding the behavior of plutonium’s multiple oxidation states in the presence of intense ionizing radiation fields is essential for predicting the behavior of this element under multiple conditions that support the development and innovation of nuclear fuel cycle technologies. Here, we present recent advances in our understanding of plutonium radiation chemistry, including the first-ever multiscale model for predicting gamma radiation-induced plutonium redox chemistry, and new chemical kinetics for the reaction of plutonium and its complexes of tributyl phosphate (TBP), N,N-di-(2-ethylhexyl)butyramide (DEHBA), and N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) with transients radiolysis products, a measured using electron pulse radiolysis.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗