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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI1961

Laboratory Development of a Pot Calcination Process for Converting Liquid Wastes to Solids

Simulated Darex, Purex, Thorex, and TBP-25 wastes were batch-evaporated and calcined to solids. The weight of residue was decreased to ~ 8% of the original waste for Darex, to ~ 4% for Purex, to ~ 2% for Thorex, and ~ 7% for TBP-25 by calcination to 1200°C. Semicontinuous evaporation and calcination to solids at 900°C of Darex and Purex wastes gave volume reduction factors of ~ 8 and 7 to 10, respectively. The nitrate contents of the residues from batch calcination to 800°C were from ~ 0.07 to 0.5 wt % while the nitrate contents of the residues from semicontinuous calcination to 900°C were from ~ 0.1 to 0.7 wt %. Sodium, calcium, and magnesium additives to Purex waste decreased the percent of sulfate volatilized into the condensate from ~ 30% to <0.5%. Replacement of the atmosphere above the waste with nitric oxide decreased ruthenium volatility from 50 to 60% of that originally present in Purex waste to 0.5 to 3.5%. The thermal conductivities for the calcined wastes, measured in situ, were all >0.1 Btu/hr ft °F at ~ 400°F and increased almost linearly with increasing temperature to >0.3 Btu/hr ft °F at ~ 1600°F in all cases studied.

052001* -- Nuclear Fuels-- Waste Processing↗
DOE OSTI2013

Minor actinide separation: simplification of the DIAMEX-SANEX strategy by means of novel SANEX processes

The separation of An(III) from PUREX raffinate has previously been demonstrated by applying a DIAMEX process (i.e., co-extraction of An(III) and Ln(III) from HAR) followed by a SANEX process (i.e., selective extraction of An(III) from the DIAMEX product containing An(III) + Ln(III)). In line with process intensification issues, more compact processes have been developed: Recently, a 1c-SANEX process test was successfully performed, directly extracting An(III) from PUREX HAR. More recently, a new i-SANEX process was successfully tested. This process is based on the co-extraction of An(III) + Ln(III) into a TODGA solvent, followed by a selective back-extraction of An(III) by a water soluble complexing agent, in this case SO{sub 3}-Ph-BTP. In both cases, good recoveries were achieved, and very pure product solutions were obtained. However, both 1c-SANEX and i-SANEX used non-CHON chemicals. Nevertheless, these processes are a simplification to the DIAMEX + SANEX process as only one solvent is used. Finally, the new i-SANEX process is the most compact process. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Alternative methods for degradation studies by alpha radiolysis: tributyl phosphate and CMPO

Solvent extraction separation processes used in the recycling of used nuclear fuel are susceptible to radiolytic damage from radioactive isotopes present in used fuel. Studying the respective effects on matter of both low linear energy transfer (LET) radiation such as gamma radiation and high LET such as alpha radiation will allow for accurate prediction and modeling of process performance losses with respect to dose. The effects of gamma radiation on solvent extraction ligands have been more extensively studied than the effects of alpha radiation due to the inherent difficulty in producing a sufficient and confluent dose of alpha particles within a sample without leaving the sample contaminated with long lived radioactive isotopes. We have developed a method for studying the effects of high LET radiation in situ via {sup 10}B activation and the high LET particles that result from the {sup 10}B(n,a){sup 7}Li reaction which follows. In this study we applied this method to organic solutions of tributyl phosphate (TBP) and CMPO (compound octylphenyl-N, N-diisobutyl-carbamoyl methyl phosphine oxide) representing the PUREX and TRUEX processes respectively. Rates of degradation of TBP and CMPO and their respective degradation products in the presence of both high and low LET radiation are presented and compared to values reported in the literature. Preliminary data appears to show decreased degradation of CMPO in the presence of an aqueous acidic phase, which agrees with other studies performed on TBP solutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Managing Zirconium Chemistry and Phase Compatibility in Combined Process Separations for Minor Actinide Partitioning

In response to the NEUP Program Supporting Fuel Cycle R&D Separations and Waste Forms call DEFOA- 0000799, this report describes the results of an R&D project focusing on streamlining separation processes for advanced fuel cycles. An example of such a process relevant to the U.S. DOE FCR&D program would be one combining the functions of the TRUEX process for partitioning of lanthanides and minor actinides from PUREX(UREX) raffinates with that of the TALSPEAK process for separating transplutonium actinides from fission product lanthanides. A fully-developed PUREX(UREX)/TRUEX/TALSPEAK suite would generate actinides as product(s) for reuse (or transmutation) and fission products as waste. As standalone, consecutive unit-operations, TRUEX and TALSPEAK employ different extractant solutions (solvating (CMPO, octyl(phenyl)-N,Ndiisobutylcarbamoylmethylphosphine oxide) vs. cation exchanging (HDEHP, di-2(ethyl)hexylphosphoric acid) extractants), and distinct aqueous phases (2-4 M HNO 3 vs. concentrated pH 3.5 carboxylic acid buffers containing actinide selective chelating agents). The separate processes may also operate with different phase transfer kinetic constraints. Experience teaches (and it has been demonstrated at the lab scale) that, with proper control, multiple process separation systems can operate successfully. However, it is also recognized that considerable economies of scale could be achieved if multiple operations could be merged into a single process based on a combined extractant solvent. The task of accountability of nuclear materials through the process(es) also becomes more robust with fewer steps, providing that the processes can be accurately modeled. Work is underway in the U.S. and Europe on developing several new options for combined processes (TRUSPEAK, ALSEP, SANEX, GANEX, ExAm are examples). There are unique challenges associated with the operation of such processes, some relating to organic phase chemistry, others arising from the variable composition of the aqueous medium. This project targets in particular two problematic issues in designing combined process systems: managing the chemistry of challenging aqueous species (like Zr 4+ ) and optimizing the composition and properties of combined extractant organic phases.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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 OSTI2012

Separations Technology for Clean Water and Energy

Providing clean water and energy for about nine billion people on the earth by midcentury is a daunting challenge. Major investments in efficiency of energy and water use and deployment of all economical energy sources will be needed. Separations technology has an important role to play in producing both clean energy and water. Some examples are carbon dioxide capture and sequestration from fossil energy power plants and advanced nuclear fuel cycle scemes. Membrane separations systems are under development to improve the economics of carbon capture that would be required at a huge scale. For nuclear fuel cycles, only the PUREX liquid-liquid extraction process has been deployed on a large scale to recover uranium and plutonium from used fuel. Most current R and D on separations technology for used nuclear fuel focuses on ehhancements to a PUREX-type plant to recover the minor actinides (neptunium, americiu, and curium) and more efficiently disposition the fission products. Are there more efficient routes to recycle the actinides on the horizon? Some new approaches and barriers to development will be briefly reviewed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Electrochemistry of PUREX: R is for Reduction and Ion Transfer

The PUREX (Plutonium, Uranium, Reduction, EXtraction) process is the foundation of all industrial activities that involve the recycling of plutonium and uranium from used nuclear fuel. With over 70 years of research, engineering, and operations, it can be argued that little opportunity is left for further discovery and new fundamental understandings regarding PUREX, in general, and tri-n-butyl phosphate (TBP), in particular. Through use of an advanced electroanalytical approach, molecular-level insights regarding the back-extraction of Pu(IV) by reductive separation is reported. Studies of model PUREX systems, with lanthanide ions extracted into 20 % TBP in n-dodecane, are used to demonstrate the approach. These studies reveal knowledge pertinent to mass transfer aspects of the biphasic reductive stripping reaction central to the separation of U and Pu in PUREX. Two lanthanide ions—Ce(IV) and Yb(III)—were extracted from their concentrated solutions in aqueous 3 M HNO 3 to expressly provide electrically conducting third phases to facilitate electrochemical data acquisition. To simplify the diabolically-complicated reductive separation aspects of PUREX, the controlled polarization of electrode surfaces was used to reduce Ce(IV) and Yb(III) in place of chemical reductants. Electroanalyses of the biphasic systems with Ce(IV) and Yb(III) using three-phase electrode differential pulse voltammetry demonstrate the effects of mass transfer across water-oil interfaces on the electrode potentials for the reduction to Ce(III) and Yb(II). These reductions trigger nitrate transfer out of the oil phase for Ce and proton transfer into the organic phase for Yb. This is in correlation with the transfer of the charge-neutral ion pairs Ce·3NO 3 and Yb·2NO 3 out of the oil phases. Here, the differences in ion transfer steps reflect significant variations in the equilibrium speciation of multinuclear Ce(IV) and mononuclear Yb(III) solvates of TBP in the oil phases and the loading of the organic phase.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS1980

U.S. program assessing nuclear waste disposal in space - A status report

Various concepts for the space disposal of nuclear waste are discussed, with attention given to the destinations now being considered (high earth orbit, lunar orbit, lunar surface, solar orbit, solar system escape, sun). Waste mixes are considered in the context of the 'Purex' (Plutonium and Uranium extraction) process and the potential forms for nuclear waste disposal (ORNL cermet, Boro-silicate glass, Metal matrix, Hot-pressed supercalcine) are described. Preliminary estimates of the energy required and the cost surcharge needed to support the space disposal of nuclear waste are presented (8 metric tons/year, requiring three Shuttle launches). When Purex is employed, the generated electrical energy needed to support the Shuttle launches is shown to be less than 1%, and the projected surcharge to electrical users is shown to be slightly more than two mills/kW-hour.

Rice, E. E.↗
DOE OSTI2016

Re-evaluation of the Gamma Radiolysis Radiation Chemistry of Neptunium

The reprocessing of spent nuclear fuel by solvent extraction techniques starts with dissolving fuel in concentrated nitric acid (HNO{sub 3}). The resulting radioactive liquor is subjected to multiple stages of solvent extraction to partition desired metal ions (e.g. uranium, plutonium, and neptunium) into an organic phase using specialized extractants. Under process conditions the solvent system is continuously exposed to a high intensity radiation field, which induces radiolytic degradation of both the diluent and extractants. This degradation occurs through the formation of radicals, as well as degradation products. Moreover, both radicals and degradation products can undergo redox chemistry with metal ions within the solvent system. This can be particularly problematic for metals such as neptunium which depend on the maintenance of a particular valence state amenable to complexation by extractants such as tributylphosphate in the PUREX (Plutonium Uranium Redox Extraction) process. Neptunium (Np) is well extracted when it is Np(IV) and Np(VI) but almost unextractable when Np(V). A recently study showed that the radiolysis of Np (VI) in acidic (4 M HNO{sub 3}) solution underwent reduction to Np(V). This reduction only occurred after sufficient production of nitrous acid (HNO{sub 2}) which countered the action of radiolytically-produced oxidizing free radicals (nitrate radical, NO{sub 3}{sup .} and hydroxyl radical, {sup .}OH). This radiation-induced chemistry was successfully modeling using a simplified kinetic model, that included a reaction set for water radiolysis, plus published rate constants for the reactions of free radicals with Np(VI) and Np(V). The most sensitive reaction rate constants were identified in this model, and optimization of the five most sensitive was performed to get excellent model agreement with the experimental data. No Np(IV) was observed or predicted by the model under these conditions. The present research re-evaluates this radiation chemistry of neptunium in aerated aqueous solutions of HNO; by using a far more comprehensive FACSIMILE code. We have also extended this modeling to study the radiolytically-induced Np redox chemistry occurring under multiple other nitric acid acidities, to fully characterize the importance of the oxidizing radical and reducing nitrous acid reactions under a range of conditions relevant to large-scale process conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

MONITORING AND MODELLING OF RADIOLYTIC DEGRADATION PRODUCTS OF TBP/n-DODECANE

The Plutonium Uranium Extraction (PUREX) solvent system was developed for the separation of plutonium and uranium from irradiated fuel. Since the implementation of this process, the degradation chemistry associated with the irradiated solvent system, tributyl phosphate (TBP)/n-dodecane/nitric acid has been extensively studied as the integrity of the organic solvent is paramount for reproducible performance of the separation flowsheet (extraction/scrub/strip) during multiple cycles. In PUREX-like processes, the extent of decomposition is dependent not only upon the solvent, but also upon the presence of constituents, such as nitric acid, that interact with TBP and increase its susceptibility to radiolytic degradation. The build-up of degradation products in the organic phase alters process flowsheet performance via modification of the metal ions speciation, loss of solvent components, and enhanced water transport into the organic phase. On-line identification and quantification of the solvent degradation products would provide the necessary information for more detailed process control as well as providing the basis for timing solvent washing or replacement. In our research, we are exploring the potential of on-line monitoring for the PUREX solvent radiolytic degradation products. To identify degradation products, TBP/n-dodecane solvent, contacted with aqueous nitric acid solutions of variable concentrations are subjected to various gamma radiation external doses then analyzed by electro-spray ionization mass spectrometry (ESMS). In addition, vibrational spectroscopy is utilized to monitor and quantify major degradation products including dibutyl phosphoric acid (HDBP) and monobutyl phosphoric acid (H2MBP) in TBP/n-dodecane solvent. The compiled spectroscopic databases serve for developing interpretive and predictive chemometric models for the quantification of the PUREX solvent degradation products.

Peterson, James M.↗
DOE OSTI2025

Conceptual Model Testing Related to SDU 6 Drainwell Observations

From its inception in the early 1950s through the end of the Cold War in the early 1990s, the Savannah River Site (SRS) produced nuclear materials for national defense in five reactors. Additionally, irradiated reactor fuel and target tubes were dissolved in nitric acid to recover plutonium and uranium using the PUREX (Plutonium Uranium Reduction EXtraction) process. Liquid waste from these chemical separations processes was then stored onsite in 51 underground tanks. Eight waste storage tanks have been operationally closed (i.e. cleaned and grouted) and the remaining tanks hold a mixture of liquids, insoluble solids, and precipitated salts (SRMC-LWP-2022-00001), the latter generated by evaporating water from the liquid waste. Waste is currently being retrieved from tanks and separated into 1) high-radioactivity, low-volume, and 2) low-radioactivity, high-volume components, principally through the Salt Waste Processing Facility (SWPF) (SRMC-LWP-2023-00001). The former waste stream is vitrified in the Defense Waste Processing Facility (DWPF), stored onsite, and destined for offsite disposal in a deep geologic repository. The latter stream is mixed with dry cementitious materials in the Saltstone Production Facility (SPF) and the wet slurry placed in onsite Saltstone Disposal Units (SDUs) within the Saltstone Disposal Facility (SDF), where it hardens into a cement waste form termed saltstone. A low-infiltration surface cover system will be placed over the SDF at closure, where SDUs will then be in the subsurface post-closure (SRR-CWDA-2019-00001).

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

Lessons Learned from Hanford's Purex Tunnel 2 Structural Stabilization Project - 20509

In 2019, U.S. Department of Energy (DOE) contractor CH2M Hill Plateau Remediation Company (CHPRC) completed structurally stabilizing the Plutonium Uranium Extraction Plant (Purex) waste storage Tunnel 2 with engineered grout. This action significantly reduced risk to Hanford workers, the surrounding community and environment by preventing future collapse of the tunnel and possible release of radioactive materials. Decades of plutonium production at the Purex facility required construction of two large underground tunnels for storage of highly contaminated plutonium processing equipment that had failed during the operation of Purex and other facilities across the Hanford Site. Following the partial collapse of Purex Tunnel 1 on May 9, 2017, DOE and CHPRC took immediate action to safely stabilize the collapsed section of the tunnel. The hole in the tunnel was filled the next day as a temporary stabilization measure, and no contamination was detected as a result of the partial collapse of the tunnel. Workers then stabilized Tunnel 1 in its entirety by filling with engineered grout, which also encapsulated the radioactive equipment and railcars within the tunnel. Grouting eliminated the risk of further collapse and stabilized the tunnel contents, all while not precluding future remedial actions or final closure decisions. While conducting an integrity analysis of Tunnel 1 by using construction drawings, photos and other historical documents, project teams also analyzed the much larger Purex Tunnel 2. The analysis found Tunnel 2 did not meet current codes for structural integrity, and it may not be able to bear the weight of the soil above the tunnel. Therefore, it was determined that Tunnel 2 was also at risk of collapse. DoE's independent panel of experts determined grouting to be the best choice for stabilizing Tunnel 2 because it provides the highest level of stability and protection, and does not preclude future remedial actions. DOE authorized CHPRC to commence work and the Purex Tunnel 2 Structural Stabilization Project was launched in December of 2017. The paper will discuss the process of recognizing the following critical aspects and risks associated with stabilizing Tunnel 2, and the actions taken by the project team to mitigate these risks, including: - Tunnel investigation and data collection to assist with future remediation; - On-site batching and mock-ups to provide the large quantity of grout required for stabilizing the tunnel; - Workforce involvement and innovative actions that mitigated the project's most significant potential hazards, including traffic control; - Project team high level of execution translated into successful project execution. Sharing lessons learned from this critical risk-reduction effort can enhance the safety and efficiency of other complex and hazardous projects across the DOE complex. (authors)

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

Chemical Disposition of Plutonium in Hanford Site Tank Wastes

This report examines the chemical disposition of plutonium (Pu) in Hanford Site tank wastes, by itself and in its observed and potential interactions with the neutron absorbers aluminum (Al), cadmium (Cd), chromium (Cr), iron (Fe), manganese (Mn), nickel (Ni), and sodium (Na). Consideration also is given to the interactions of plutonium with uranium (U). No consideration of the disposition of uranium itself as an element with fissile isotopes is considered except tangentially with respect to its interaction as an absorber for plutonium. The report begins with a brief review of Hanford Site plutonium processes, examining the various means used to recover plutonium from irradiated fuel and from scrap, and also examines the intermediate processing of plutonium to prepare useful chemical forms. The paper provides an overview of Hanford tank defined-waste–type compositions and some calculations of the ratios of plutonium to absorber elements in these waste types and in individual waste analyses. These assessments are based on Hanford tank waste inventory data derived from separately published, expert assessments of tank disposal records, process flowsheets, and chemical/radiochemical analyses. This work also investigates the distribution and expected speciation of plutonium in tank waste solution and solid phases. For the solid phases, both pure plutonium compounds and plutonium interactions with absorber elements are considered. These assessments of plutonium chemistry are based largely on analyses of idealized or simulated tank waste or strongly alkaline systems. The very limited information available on plutonium behavior, disposition, and speciation in genuine tank waste also is discussed. The assessments show that plutonium coprecipitates strongly with chromium, iron, manganese and uranium absorbers. Plutonium’s chemical interactions with aluminum, nickel, and sodium are minimal to non-existent. Credit for neutronic interaction of plutonium with these absorbers occurs only if they are physically proximal in solution or the plutonium present in the solid phase is intimately mixed with compounds or solutions of these absorbers. No information on the potential chemical interaction of plutonium with cadmium was found in the technical literature. Definitive evidence of sorption or adsorption of plutonium onto various solid phases from strongly alkaline media is less clear-cut, perhaps owing to fewer studies and to some well-attributed tests run under conditions exceeding the very low solubility of plutonium. The several studies that are well-founded show that only about half of the plutonium is adsorbed from waste solutions onto sludge solid phases. The organic complexants found in many Hanford tank waste solutions seem to decrease plutonium uptake onto solids. A number of studies show plutonium sorbs effectively onto sodium titanate. Finally, this report presents findings describing the behavior of plutonium vis-à-vis other elements during sludge dissolution in nitric acid based on Hanford tank waste experience gained by lab-scale tests, chemical and radiochemical sample characterization, and full-scale processing in preparation for strontium-90 recovery from PUREX sludges.

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

Literature Review: Crud Formation at the Liquid/Liquid Interface of TBP-Based Solvent-Extraction Processes

This report summarizes the literature reviewed on crud formation at the liquid:liquid interface of solvent extraction processes. The review is focused both on classic PUREX extraction for industrial reprocessing, especially as practiced at the Hanford Site, and for those steps specific to plutonium purification that were used at the Plutonium Reclamation Facility (PRF) within the Plutonium Finishing Plant (PFP) at the Hanford Site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

The H-Modified Process

Construction of the H-Area canyon separation facility was completed in the early 1950's. Processing of irradiated depleted Uranium fuel using the Purex flowsheet began in July 1955. The introduction of enriched uranium fuels into Savannah River reactor operations was to produce a greater variety of products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Autonomous 3D and Radiation Mapping in Tunnel Environments Using Aerial Robots - 18156

In this paper we present the algorithms and systems developed to enable autonomous 3D and radiation mapping, as well as radiological source localization in tunnel environments using aerial robots. The developed technologies aim to address the need of autonomous robotic inspection of nuclear sites such as facilities in the process of being decommissioned. As a motivational example, the PUREX tunnels at the Hanford Site are considered and the robot is designed to be able to conduct multi-modal 3D/radiation characterization of the tunnel environment without any prior knowledge or map provided beforehand. In order to enable the necessary GPS-denied simultaneous localization and mapping in dark, visually-degraded tunnel environments, a multi-modal mapping unit was developed which fuses the data from stereo vision cameras synchronized with flashing LEDs, inertial sensors, and three time-of-flight 3D depth sensors. The robot's pose and surrounding map are estimated at 20 Hz through a depth-enhanced visual- inertial localization and mapping framework. Provided this ability to estimate the robot's pose and develop a dense map in real-time, a receding horizon sampling-based autonomous exploration path planning algorithm was utilized and equips the robot with the ability to explore its environment given no prior map. The aforementioned key robotic functionalities are further extended through the integration of miniaturized fully solid-state scintillation detectors, utilizing advanced scintillators and Silicon Photomultipliers. The derived sensor data is employed for radiation mapping, which when fused with the 3D maps and the spectroscopic source localization methods, allow the aerial robotic platform to localize the radiation source and identify the specific source isotope. The source location is estimated through a geometric algorithm which can localize a source given only three measurements and short dwell times. The developed systems were tested using two small aerial robot platforms of the F450 and F550 frame classes with weights of 2.6 kg and 2.9 kg respectively. All localization and planning as well as high-level control algorithms are executed onboard. The scintillation detector is integrated on the robot and its data is collected through a dedicated microcontroller. A set of field experiments were conducted inside the dark and GPS-denied environment of a 4.25 m wide, 7.5 m tall, and 137 m long railroad tunnel. A Cs-137 source was placed within the tunnel, the associated 3D/radiation maps were derived, and the source location was estimated. Through these efforts, the goal is to develop the robotic technology that would enable the autonomous systematic inspection or emergency response in nuclear waste decommissioning facilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Flowsheet Evaluation of Dissolving Used Nuclear Fuel in PUREX Solvent

This study explores the potential benefit to the used nuclear fuel actinide recovery flowsheet of a process simplification and establish the experimental work needed to develop the simplified flowsheet. The simplification is based on replacing acid dissolution of used nuclear fuel with dissolution in the tri-butyl phosphate solvent used in the industrially mature Plutonium Uranium Reduction Extraction flowsheet. Though characterized by considerable uncertainty, simplified flowsheets appear feasible and potentially offer significant reductions in process complexity, nitrate inventory, secondary liquid effluent generation and plant footprint. Initial development of the technology should focus on key fission product and actinide dissolution as a function of process conditions and understanding the fundamental chemistry of tri-butyl phosphate solvent chemistry. Development of ancillary technologies for minor actinide separations and fuel pretreatment for tritium management is also recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Annual Report, Spring 2015: Alternative Chemical Cleaning Methods for High Level Waste Tanks – Corrosion Test Results

The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel when interacted with the chemical cleaning solution composed of 0.18 M nitric acid and 0.5 wt. % oxalic acid. This solution has been proposed as a dissolution solution that would be used to remove the remaining hard heel portion of the sludge in the waste tanks. This solution was combined with the HM and PUREX simulated sludge with dilution ratios that represent the bulk oxalic cleaning process (20:1 ratio, acid solution to simulant) and the cumulative volume associated with multiple acid strikes (50:1 ratio). The testing was conducted over 28 days at 50°C and deployed two methods to invest the corrosion conditions; passive weight loss coupon and an active electrochemical probe were used to collect data on the corrosion rate and material performance. In addition to investigating the chemical cleaning solutions, electrochemical corrosion testing was performed on acidic and basic solutions containing sodium permanganate at room temperature to explore the corrosion impacts if these solutions were to be implemented to retrieve remaining actinides that are currently in the sludge of the tank.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗