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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Methodology of Qualification of CCIM Vitrification Process Applied to the High- Level Liquid Waste from Reprocessed Oxide Fuels - 12438

The vitrification of high-level liquid waste from reprocessed oxide fuels (UOX fuels) by Cold Crucible Induction Melter is planed by AREVA in 2013 in a production line of the R7 facility at La Hague plant. Therefore, the switch of the vitrification technology from the Joule Heated Metal Melter required a complete process qualification study. It involves three specialties, namely the matrix formulation, the glass long-term behavior and the vitrification process development on full-scale pilot. A new glass frit has been elaborated in order to adapt the redox properties and the thermal conductivity of the glass suitable for being vitrified with the Cold Crucible Induction Melter. The role of cobalt oxide on the long term behavior of the glass has been described in the range of the tested concentrations. Concerning the process qualification, the nominal tests, the sensitivity tests and the study of the transient modes allowed to define the nominal operating conditions. Degraded operating conditions tests allowed to identify means of detecting incidents leading to these conditions and allowed to define the procedures to preserve the process equipments protection and the material quality. Finally, the endurance test validated the nominal operating conditions over an extended time period. This global study allowed to draft the package qualification file. The qualification file of the UOX package is currently under approval by the French Nuclear Safety Authority. (authors)

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

Synroc Waste-forms for Immobilization of Advanced Reprocessing Wastes - 16167

EURO-GANEX, an advanced reprocessing technology, aims to separate both major (U, Pu) and minor (Np, Am, and Cm) actinides together from fission products. As the final waste composition is free from actinides - as opposed to current process waste, e.g. from PUREX - adapted immobilization matrices need to be developed. Synroc is a potential wasteform that has proven itself to be efficient to immobilize high-level nuclear wastes (HLW). In this study, a new composition of Synroc, 'Synroc-Z', is designed and characterized. The primary modification is to reduce the amount of zirconolite phase, which acts as the main host for actinides. The optimum hot-press conditions for Synroc-Z samples with a waste loading of 20 wt% were determined to be: 1200 deg. C and 20 MPa for 3 h. The durability of Synroc-Z was examined via the MCC-1 test at 90 deg. C. It shows similar leach resistance to Synroc-C, which is much better than current glass waste-forms. (authors)

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

Strategies for Decommissioning the Fuel Storage and Reprocessing Complex at Chalk River Laboratories - 16183

For over 60 years, Canadian Nuclear Laboratories (CNL), previously Atomic Energy of Canada Limited (AECL), has been a world leader in developing peaceful and innovative applications from nuclear technology. Resulting from these years of nuclear research there are legacy liabilities consisting of outdated and unused research facilities and buildings, a wide variety of buried and stored wastes, and affected lands. In 2006, the Government of Canada established the Nuclear Legacy Liabilities Program (NLLP) to safely and cost-effectively eliminate, to the extent practicable, the nuclear legacy liabilities. The National Research Experimental (NRX) fuel rod storage and handling facility, as well as two dedicated fuel reprocessing facilities at the Chalk River Laboratories (CRL) known as the Building 200 (B200) Complex, are examples of these legacy liabilities. It is planned to remove these facilities as part of a logical progression of activities to reduce CNL's liabilities under the NLLP and to prepare the property for the physical redevelopment of the CRL site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Modification of an Agitated Tube Reactor for use in the Reprocessing of Spent Nuclear Fuel

Aims: Convert an Agitated Tube Reactor (ATR) into a counter-current solvent extraction device. Assess effect of liquid fill level on liquid dispersion. The agitated tube reactor (ATR) is an array of connected tubes. The ATR laterally shakes in a sinusoidal path, moving the agitator bar to create mixing. Originally designed by AM Technology as a cocurrent reactor, the ATR has many features that are potentially useful in the nuclear industry: Tubes are connected by non-moving seals; Reduction in seal damage and reduced risk of expensive repairs in radioactive environments; Minimal axial dispersion within a tube; Plug flow; Narrower residence time distributions that traditional technology; Multiple tubes allow for multiple reactions in series; Converting to a counter-current design allows the ATR to be used for solvent extraction; Reprocessing or decommissioning. A 180 ml PERSPEX ATR tube with no liquid flow through was constructed and filled with a 1:1 solvent:aqueous ratio of kerosene and water. The tube was filled such that the total liquid filled 25 %, 50 %, 75 % or 100 % of the tube. A Photron Fastcam SA5, positioned at the side and the end of the tube was set to a frame rate of 5400 fps with a resolution of 1024 x 1024 and 6000 fps with a resolution of 1024 x 896 respectively. Droplets were recorded after 2 minutes of shaking to ensure good mixing. Droplet sizes are used to determine optimum fill level in the ATR. The shaking frequency follows a sinusoidal path and was set to 5 Hz. The images shown in the results section were taken at the peak of the sinusoidal motion. To convert to a counter-current design the two liquids need to be separated at the end of each tube. Liquids have been allowed to separate under gravity separation. Inclusion of baffles to reduce remixing by sloshing. Wall with height corresponding to 25 % fill volume allows kerosene to overflow into separate outlet. Increase in liquid fill level decreases the dispersion of the two liquid phases. Higher dispersion of droplets decreases surface area and therefore increases mass transfer in solvent extraction. 25 % should provide the greatest mass transfer though more work needs to completed to confirm this.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Modelling of the nitric acid reduction process: Application to materials behavior in reprocessing plants

In France, the recycling process of nuclear waste fuels involves the use of hot concentrated nitric acid. The understanding and the prediction of the structural materials (mainly austenitic stainless steels) behaviour requires the determination of the nitric acid reduction process. Nitric acid is indirectly reduced by an autocatalytic mechanism depending on the cathodic overpotential and the acid concentration. This mechanism has been widely studied. All the authors agree on its autocatalytic nature, characterized by the predominant role of the reduction products. It is also generally admitted that nitric acid or the nitrate ion are not the electro-active species. However, uncertainties remain concerning the nature of the electro-active species, the place where the catalytic species regenerates and the thermodynamic and kinetic behaviour of the reaction intermediates. The aim of this study is to clarify some of these uncertainties by performing an electrochemical investigation of the 4 mol.L -1 nitric acid reduction process at 40 deg. C occurring on an inert electrode (platinum or gold). An inert electrode was chosen as a working electrode in a first step in order to avoid its oxidation and focus the research on the reduction mechanism. This experimental work enabled to suggest a coherent sequence of electrochemical and chemical reactions. Then, a kinetic modelling of this sequence was carried out for a gold rotating disk system. In this objective, a thermodynamic study at 25 deg. C led to the evaluation of the composition of liquid and gaseous phases for nitric acid solutions from 0.5 to 22 mol.L -1. The kinetics of the reduction process of nitric acid 4 mol.L -1 was investigated by cyclic voltammetry and chrono-amperometry on an inert electrode at 40 deg. C. A coupling of chrono-amperometry and FTIR in gaseous phase led to the identification of the gaseous reduction products as a function of the cathodic overpotential. These different results showed that for potentials between 0.6 V/NHE and 1.15 V/NHE, the reduction process is autocatalytic. The regeneration of the electro-active species may take place at the surface of the electrode for lower over-potentials, and otherwise in solution. When the potential is lower than 0.6 V/ENH, the fast reduction of nitrogen oxide may lead to the rupture of the autocatalytic cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

NRC Perspectives on Waste Incidental to Reprocessing Consultations and Monitoring - 13398

Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the U.S. Department of Energy (DOE) to consult with the U.S. Nuclear Regulatory Commission (NRC) for certain non-high level waste (HLW) determinations. The NDAA also requires NRC to monitor DOE's disposal actions related to those determinations to assess compliance with NRC regulations in 10 CFR Part 61, Subpart C. The NDAA applies to DOE activities that will remain within the States of South Carolina and Idaho. DOE has chosen to, under DOE Order 435.1, engage in consultation with NRC for similar activities in the State of Washington and New York, however, the NRC has no monitoring responsibilities. In 2007, the NRC developed a draft Final Report for Interim Use entitled, NUREG-1854: NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste Determinations. Since the law was enacted, the DOE and NRC have consulted on three waste determinations within the affected States: (1) the Saltstone Disposal Facility at the Savannah River Site (SRS) within the State of South Carolina in 2005, (2) the INTEC Tank Farm at the Idaho National Laboratory within the State of Idaho in 2006, and (3) the F Tank Farm at SRS in 2011. After the end of consultation and issuance by DOE of the final waste determination, monitoring began at each of these sites, including the development of monitoring plans. In addition to the NDAA sites, DOE has requested NRC consultation support on both individual tanks and the entire C Tank Farm at the Hanford Nuclear Reservation in the State of Washington. DOE also requested consultation of waste determinations performed on the melter and related feed tanks at the West Valley site in New York that would be disposed offsite. In the next few years, NRC and DOE will consult on the last of the NDAA waste determinations for a while, the H Tank Farm waste determination at SRS. DOE may identify other activities in the future but largely NRC's role will change from doing both consultation and monitoring to being focused on monitoring activities within NDAA. DOE has identified other activities at the Hanford Nuclear Reservation that would continue consultation activities but outside of the NDAA in the future. During the past seven years of consultations and monitoring a number of lessons learned about the process, communication issues, and technical guidance have been identified. With the change in focus from reviewing initial performance assessments and draft waste determinations to long-term monitoring (e.g., individual waste tank closure, at F Tank Farm or complete tank farm closure at INTEC expected in the near future), the NRC is going to revise and update its guidance over the next few years to reflect the lessons learned and the change in focus. In addition to the lessons learned, improvements in the guidance will have to account possible rule and guidance changes underway within Part 61. This paper will discuss the initial plans, approaches, and time lines to revise the guidance within NUREG-1854, including opportunities for public involvement. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the second shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

New Design for an HLW Repository (for Spent Fuel and Waste from Reprocessing) in a Salt Formation in Germany - 12213

In autumn 2010, after a 10-year moratorium, exploration was resumed in Gorleben, the potential site for a German HLW repository. At the same time, the Federal Government launched a two-year preliminary safety analysis to assess whether the salt dome at Gorleben is suitable to host all heat-generating radioactive waste generated by German NPPs based on the waste amounts expected at that time. The revised Atomic Energy Act of June 2011 now stipulates a gradual phase-out of nuclear energy production by 2022, which is 13 years earlier than expected in 2010. A repository design was developed which took into account an updated set of data on the amounts and types of expected heat-generating waste, the documented results of the exploration of the Gorleben salt dome, and the new 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' of 30 September, 2010. The latter has a strong influence on the conceptual designs as it requires that retrievability of all waste containers is possible within the repository lifetime. One design considered that all waste containers will be disposed of in horizontal drifts of a geologic repository, while the other design considered that all waste containers will be disposed of in deep vertical boreholes. For both options (emplacement in drifts/emplacement in vertical boreholes), the respective design includes a selection of waste containers, the layout of drifts, respectively lined boreholes, a description of emplacement fields, and backfilling and sealing measures. The design results were described and displayed and the differences between the two main concepts were elaborated and discussed. For the first time in both repository designs the requirement was implemented to retrieve waste canisters during the operational phase. The measures to fulfill this requirement and eventually the consequences were highlighted. It was pointed out that there arises the need to keep transport- and storage casks in adequate numbers and interim storage facilities available until the repository is closed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

The development and initial testing of the Ice Pig cleaning method for nuclear reprocessing plants - 15363

There are numerous situations throughout the nuclear industry where it would be advantageous to clean out enclosed systems, e.g. ducts and pipes, to remove sediments and radioactive contamination. However, conventional flushing methods result in the production of significant volumes of contaminated liquor. The wall Shear achieved with water flushing is insufficient to remove and carry dense deposits often found in nuclear facilities. Conventional mechanical pigs can be propelled through the duct by application of a driving medium such as compressed air or water. These are able to achieve very high wall shears, but can also get physically stuck where there is variation in the pipe / duct diameter or small radius bends. The field of application is best suited to straight constant diameter systems with no discontinuities such as Tee pieces, valves and in-pipe instrumentation. Driven by the desire to clean and displace radioactive materials from complex topology ducts and pipes, without the risk of having conventional pigs becoming lodged in a plant and coupled with the need to minimize effluent volumes, the UK Nuclear Industry led by Sellafield Limited (on behalf of the UK Nuclear Decommissioning Authority), has worked to demonstrate and underpin the use of the ice pigging technology for the nuclear industry. It is found that high ice fraction slurries flow as viscous liquids that exhibit a number of mechanical properties that offer much higher shear rates. The 'Ice Pig' exerts shear rates on pipe walls which are typically 1000 times greater than those achieved with water flowing at the same velocity. The 'Ice Pig' is able to negotiate almost any topology, flow through pipe networks of vastly differing diameter, through heat exchangers and in-line instrumentation etc. Experimental data and operational experience from other industries demonstrates that whilst there is high shear, the 'Ice Pig' does not damage the surface of the pipe wall. A further attractive characteristic of the 'Ice Pig' is that it eventually melts rendering it a simple effluent for processing by downstream processes. Due to the special nature of ice slurry, its unique rheological and physical characteristics, together with the complementary underpinning experimental work undertaken by Sellafield Limited enables the Ice Pigging process to move from an 'interesting' technology to one suitable for use by the nuclear industry for operations, Post Operational Clean Out (POCO) and decommissioning. (authors)

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

Adsorption of Iodine and Tritium from UNF Reprocessing Off-Gas Streams

The need to control emissions of the gaseous radionuclides to the air during nuclear fuel recycling is widely recognized. Gaseous radionuclides can partition to off-gas streams emanating from the various individual processes used in the recycle plant, such as from the mechanical head end (shear), tritium pretreatment, fuel dissolver, vessel vents, cell ventilation, and conversion operations. In traditional recycle plants, the bulk of the four primary volatile fission products (i.e., {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I) is released in the head-end processes, for example, the dissolution process. The usual off-gas treatment sequence of operations targets removal of tritium (as {sup 3}HHO), iodine, carbon (as {sup 14}CO{sub 2}), and the noble gases, in that order. Water and iodine are captured first because the other species are relatively unimpeded by the sorbents used for water and iodine, so they flow downstream to additional treatment processes. Water and iodine do react with one another, and many techniques used to collect one of these species also collect some fraction of the other. It is quite desirable to sequester the iodine and water separately because the tritiated water may be disposed of as a low level waste. If not cleanly separated, contamination of the low-level waste by iodine may elevate its classification to high-level waste and result in greater costs for disposal. The proposed sorption sequence was tested in the laboratory. Use of a second column of AgZ permits the detection of breakthrough from the primary AgZ column while protecting the downstream 3AMS column from collecting iodine. (This arrangement is equivalent to enabling intra-column sampling of the primary column.) Neutron activation analysis of the 3AMS revealed that iodine was below, or at, the level of detection. The results successfully demonstrated that iodine can be sequestered in the AgZ while the water is collected free of iodine in the 3AMS column. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Radiation Effects in Used Next Generation Nuclear Fuel Reprocessing Strategies

Given global commitments to significantly increase nuclear energy capacity, it is now more important than ever to develop efficient used nuclear fuel (UNF) management strategies to encourage widespread adoption of closed fuel cycles. To achieve this ambitious goal, a comprehensive understanding of radiation effects is essential for these next generation technologies, as radiolysis often limits longevity and performance. Here, we present new findings on: (i) the radiation robustness and performance of advanced sulfur chloride-based chlorination processes in the presence of nuclear materials (Fig. 1A); and (ii) the impacts of voloxidized uranium and rhenium complexation on monoamide-based UNF direct dissolution strategies (Fig 1B). These studies employed a combination of time-resolved electron pulse and dose accumulation gamma and electron beam irradiation techniques.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2016

Zurückgezogen: Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing

Abstract Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt‐free complexing and reducing reagent, glutarimidedioxime (H 2 A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO 3 with H 2 A. The complexation‐reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu IV complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu IV , through the complexation‐reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

Xian, Liang↗
DOE OSTI2016

Glutarimidedioxime: A Complexing and Reducing Reagent for Plutonium Recovery from Spent Nuclear Fuel Reprocessing

Abstract Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt‐free complexing and reducing reagent, glutarimidedioxime (H 2 A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO 3 with H 2 A. The complexation‐reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu IV complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu IV , through the complexation‐reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

Xian, Liang↗