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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Response of soil bacteria to PUREX chemicals suggests biomarker utility and bioremediation potential

Chemicals involved in plutonium uranium reduction extraction (PUREX) can be released from nuclear reprocessing facilities and accumulate in the environment. We exposed chemically diverse soils to a range of concentrations of key chemicals used in the PUREX process. The responses of soil microbial communities are dependent on soil type, and tributyl phosphate exposure generates the most reproducible changes in microbial communities. We reconstructed the genomes of key bacteria and find several phosphotriesterase genes found only in Rhizobiaceae. The abundance of phosphotriesterase genes is significantly higher in samples exposed to tributyl phosphate. In conclusion, these phosphotriesterase genes may be involved in breakdown of tributyl phosphate, and a means of accessing phosphate for these bacteria.

Bioremediation↗
DOE OSTI2022

Neptunium redox speciation and determination of its total concentration in dissolved fuel simulant solutions by spectrophotometry

Here, two new approaches to measure Np concentration in dissolved used nuclear fuel simulant (aqueous feed for PUREX process) by spectrophotometry are developed. The first approach is based on chemical reduction of Np in the feed to its tetravalent state using ascorbic acid with simultaneous conversion of Pu(IV) to Pu(III). Interfering effects from light absorbing fission and corrosion products are accounted for by measuring optical absorbance spectrum of aqueous raffinate after extraction of U, Np, and Pu by tributyl phosphate in dodecane. The second approach uses no chemical treatment at all and relies on spontaneous valency adjustment of Np to Np(V) by dilution of the feed with water to reduce its acidity to low decimolar range of nitric acid concentration. Results of Np determination in the feed by spectrophotometry are in good agreement with its concentration measured by ICP-MS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Synergistic Extraction of Dysprosium and Aggregate Formation in Solvent Extraction Systems Combining TBP and HDBP

During treatment of nuclear fuel in the Plutonium/URanium EXtraction (PUREX) process, the extractant tri-n-butyl phosphate (TBP) is known to degrade to dibutylphosphoric acid (HDBP), which increases the extraction of metal ions, thereby inhibiting their stripping from the organic phase. To better understand this phenomenon, we investigated how mixtures of TBP and HDBP influenced the extraction of metal, nitric acid, and water, and correlated the results to aggregated structures in the organic phase. The mole ratios of TBP-HDBP mixtures had a non-linear effect on the extraction of Dy 3+ and water from 0.2 M HNO 3 , indicating synergism. In 2 M HNO 3 , the TBP:HDBP mole ratio had a more linear relation to Dy 3+ and water extraction, so the synergistic effect was less pronounced than in the low acid system. The extraction of nitric acid showed no synergistic effect and follows closely what would be expected in a system using TBP only. The small-angle X-ray scattering (SAXS) data of the 0.2 M acid system showed maximum contrast at a TBP:HDBP mole ratio of 0.25, so that the synergistic mixture is also the most aggregated at 0.2 M acid. The 2 M acid system also showed that the mixed system is more aggregated than the end members, although this does not result in peak extraction. Previous studies of synergistic extraction of metal cations explain the enhanced extraction by increased dehydration of the metal ion. Although our data do not rule out the formation of mixed complexes according to the classical mechanism of synergism, our evidence of increased water extraction and aggregate formation in systems combining TBP and HDBP are complementary to the metal-centric dehydration aspects of the process. The findings in this study give insights into the complex chemistry of solvent extraction, thus providing a possible link between formation of aggregates in the organic phase and synergistic extraction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Selective Partitioning of Ruthenium from Nitric Acid Media

Ruthenium removal from PUREX raffinate solutions is imperative for the application of some advanced actinide partitioning technologies, particularly partitioning of hexavalent americium. Ruthenium interferes with the oxidation of americium and suppresses its extraction. Selective extraction of ruthenium from these solutions is complicated by the existence of multiple ruthenium species in nitric acid. In this work, ruthenium is selectively extracted from nitric acid media through oxidation to ruthenium tetroxide and absorption into polymer beads. Extraction of ruthenium tetroxide into a variety of organic solvents is also explored. The use of UV-Visible spectroscopy to quantify extraction of ruthenium tetroxide was demonstrated. This method for extraction of ruthenium from nitric acid was applied to PUREX raffinate simulant solution and was found to simultaneously remove ruthenium (through oxidation and absorption) and zirconium (through precipitation). This method not only removes ruthenium and zirconium from solution but also separates ruthenium and zirconium from each other. Thus, the two most problematic fission product elements for the PUREX process are separated in a single step. Further, americium was not extracted or sorbed, which should allow it to be partitioned following ruthenium/zirconium removal using existing oxidative treatments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Analysis of Venting of a Resin Slurry

A resin slurry venting analysis was conducted to address safety issues associated with over-pressurization of ion exchange columns used in the plutonium uranium redox extraction (PUREX) process at the U. S. Department of Energy’s (DOE’s) Savannah River Site (SRS). If flow to these columns were inadvertently interrupted, an exothermic runaway reaction could occur between the ion exchange resin and the nitric acid used in the feed stream. The nitric acid-resin reaction generates significant quantities of noncondensable gases, which would pressurize the column. To prevent the column from rupturing during such events, rupture disks are installed on the column vent lines. The venting analysis models accelerating rate calorimeter (ARC) tests and data from tests that were performed in a vented test vessel with a rupture disk. The tests showed that the pressure inside the test vessel continued to increase after the rupture disk opened, though at a slower rate than prior to the rupture. The increase in the vessel pressure is modeled as a transient phenomenon associated with expansion of the resin slurry/gas mixture upon rupture of the disk. It is postulated that the maximum pressure at the end of this expansion is limited by energy minimization to approximately 1.5 times the rupture disk burst pressure. The magnitude of this pressure increase is consistent with the measured pressure transients. The results of this analysis demonstrate the need to allow for a margin between the design pressure and the rupture disk burst pressure in similar applications.

42 ENGINEERING↗
DOE OSTI2017

Tc-99 Decontamination From Heat Treated Gaseous Diffusion Membrane -Phase I

Uranium gaseous diffusion cascades represent a significant environmental challenge to dismantle, containerize and dispose as low-level radioactive waste. Baseline technologies rely on manual manipulations involving direct access to technetium-contaminated piping and materials. There is a potential to utilize novel thermal decontamination technologies to remove the technetium and allow for on-site disposal of the very large uranium converters. Technetium entered these gaseous diffusion cascades as a hexafluoride complex in the same fashion as uranium. Technetium, as the isotope Tc-99, is an impurity that follows uranium in the first cycle of the Plutonium and Uranium Extraction (PUREX) process. The technetium speciation or exact form in the gas diffusion cascades is not well defined. Several forms of Tc-99 compounds, mostly the fluorinated technetium compounds with varying degrees of volatility have been speculated by the scientific community to be present in these cascades. Therefore, there may be a possibility of using thermal desorption, which is independent of the technetium oxidation states, to perform an in situ removal of the technetium as a volatile species and trap the radionuclide on sorbent traps which could be disposed as low-level waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Reducing Actinide Production Using Inert Matrix Fuels

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Radioactive waste management approaches for developed countries

Nuclear power has demonstrated over the last 30 years its capacity to produce base-load electricity at a low, predictable and stable cost due to the very low economic dependence on the price of uranium. However the management of used nuclear fuel remains the “Achilles’ Heel” of this energy source since the storage of used nuclear fuel is increasing as evidenced by the following number with 2,000 tons of UNF produced each year by the 104 US nuclear reactor units which equates to a total of 62,000 spent fuel assemblies stored in dry cask and 88,000 stored in pools. Two options adopted by several countries will be presented. The first one adopted by Europe, Japan and Russia consists of recycling the used nuclear fuel after irradiation in a nuclear reactor. Ninety six percent of uranium and plutonium contained in the spent fuel could be reused to produce electricity and are worth recycling. The separation of uranium and plutonium from the wastes is realized through the industrial PUREX process so that they can be recycled for re-use in a nuclear reactor as a mixed oxide (MOX) fuel. The second option undertaken by Finland, Sweden and the United States implies the direct disposal of used nuclear fuel into a geologic formation. One has to remind that only 30% of the worldwide used nuclear fuel are currently recycled, the larger part being stored (70% in pool) waiting for scientific or political decisions. A third option is emerging with a closed fuel cycle which will improve the global sustainability of nuclear energy. This option will not only decrease the volume amount of nuclear waste but also the long-term radiotoxicity of the final waste, as well as improving the long-term safety and the heat-loading of the final repository. At the present time, numerous countries are focusing on the R&D recycling activities of the ultimate waste composed of fission products and minor actinides (americium and curium). Several new chemical extraction processes, such as TRUSPEAK, ALSEP, EXAM, or LUCA are pursued worldwide and their approaches will be highlighted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Oxidation and Extraction of Americium(VI) from Lanthanides

Partitioning and transmutation of americium and curium has been the subject of significant research in the nuclear fuel cycle over the past two decades. Numerous approaches have been studied to separate these elements from the lanthanide elements to enable their recycle and transmutation. Americium is one of the primary long-term heat sources in used nuclear fuel, which may impact repository loading. The lanthanides are both a) neutron poisons, reducing reactor performance, if recycled, and b) can chemically interact with fuel cladding to cause early failure, if present in sufficient concentration. Chemical separation approaches for heterogeneous fuel recycle, typically utilize proven technology (PUREX or modifications to the PUREX process) to remove uranium and plutonium from the dissolved fuel (and if desired, can remove neptunium as well). Follow-on processing is then required to separate the americium and curium, or the americium alone (for target fabrication). In a homogeneous recycle approach, the bulk of the uranium is typically separated first, followed by additional processing to remove the transuranic elements as a group. Because of the chemical similarity of the americium with the lanthanides, and particularly with curium, these separate processes are difficult. Utilizing the unique valence states of americium (primarily the (VI) oxidation state), opens up new opportunities for the separation of americium for either heterogeneous recycle (enabling americium separation alone) or homogeneous recycle (separating americium together with plutonium and neptunium). The US Department of Energy Fuel Cycle Technologies program has been researching methods to oxidize americium and to extract it from lanthanides as americium(VI), preferably from molar nitric acid solutions. Early efforts identified sodium bismuthate as an effective oxidant, which enabled testing of the extraction of americium(VI) by tributyl phosphate and diamylamylphosphonate. Successful extraction of americium(VI) was demonstrated in this system, in both batch contact laboratory tests and in centrifugal contactors; however, the sodium bismuthate is a solid, and must be in excess of saturation with the solution. This creates some engineering challenges in the process. A significant effort to identify alternative oxidants has led to a number of promising approaches. Silver-catalyzed ozone, copper(III) periodate, and electrochemical oxidation of americium from the trivalent to hexavalent state have all been successfully demonstrated in laboratory experiments. These approaches would enable oxidation and extraction of americium without having to filter solids. Once the americium is oxidized, a key issue is how long it will remain in the hexavalent state. Investigations have shown that some materials of construction (e.g., tantalum) do not reduce americium as quickly as stainless steel. Also, alternative extractants, such as amides, do not appear to reduce the americium as rapidly as the phosphate or phosphonate extractants. While still early in their development, these advances in the technology offer some positive indications that a workable extraction process for the separation of americium(VI) can be realized in the near future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Mock Nuclear Processing Facility-Safeguards Training Requirements

This document outlines specific training requirements in the topical areas of Material Control and Accounting (MC&A) and Physical Protection(PP) which are to be used as technical input for designing a mock Integrated Security Facility (ISF) at Sandia National Laboratories (SNL). The overall project objective for these requirements is to enhance the ability to deliver training on Material Protection Control and Accounting (MC&A) concepts regarding hazardous material such as irradiated materials with respect to bulk processing facilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Vibrational spectroscopy for online monitoring of extraction solvent degradation products

In our research, we are exploring the potential of online monitoring of the organic solvents for the flowsheets relevant to the used nuclear fuel reprocessing and tributyl phosphate (TBP)- based extraction processes in particular. Utilization of vibrational spectroscopic techniques permits the discrimination of the degradation products from the primary constituents of the loaded extraction solvent. Multivariate analysis of the spectral data facilitates development of the regression models for their quantification in real time and potentially enables online implementation of a monitoring system. Raman and FTIR spectral databases were created and used to develop the regression partial least squares (PLS) chemometric models for the quantitative prediction of HDBP (dibutyl phosphoric acid) degradation product, TBP, and UO{sub 2}{sup 2+} extraction organic product phase. It was demonstrated that both these spectroscopic techniques are suitable for the quantification of the Purex solvent components in the presence of UO{sub 2}(NO{sub 3}){sub 2}. Developed PLS models successfully predicted HDBP and TBP organic concentrations in simulated Purex solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗
DOE OSTI2016

Annual Report, Fall 2016: Alternative Chemical Cleaning of Radioactive 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 exposed to two proposed chemical cleaning solutions: acidic permanganate (0.18 M nitric acid and 0.05M sodium permanganate) and caustic permanganate. (10 M sodium hydroxide and 0.05M sodium permanganate). These solutions have been proposed as a chemical cleaning solution for the retrieval of actinides in the sludge in the waste tanks, and were tested with both HM and PUREX sludge simulants at a 20:1 ratio.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Automation Toward Tether-less Robotic Operations in Nuclear Facilities - 17474

Automation capabilities that enable communication-denied nuclear robotic operations have come of age. These layer over tethered operation or enable un-tethered systems for some operations. Tele-operated, tethered devices are traditionally preferred for nuclear remote systems. Tethers provide comm, power, and a means for mechanical extraction. They also restrict mobility and create additional failure points, and in some cases, circumstances preclude mechanical recovery by tether. Tethering has superb advantages where it is viable, but a subset of nuclear EM challenges compel mobile remote systems that forgo tethering. Tethering was essential at a time when autonomy or wireless comm could not accomplish viable operations, but now an emerging class of un-tethered exploration and service robots are compelling for situations where tethers compound waste and exposure, or are not viable. Technical advances in localization, modeling, planning, autonomy, integration and reliability have made it possible to augment tethered operations or operate altogether without tether in selective operations. Tether preclusion technologies include localization, perception, navigation, safeguarding and task prescription. These effectively accomplish tasks like way-point following, wall-following, coverage patterning, navigate-to-goal, next-best view and many others. Although these are not yet general for the broad agenda of the nuclear complex, these are already useful for many operations. These provide a foundational base that can be built upon over time for more complex operations yet to come. As a context and analog study, the technologies are exhibited by robotically (without tether) modeling a coal mine with tracks and train cars. This is analogous to technology and operation that might apply for inspection of waste storage at WIPP or the PUREX tunnels. Autonomous exploration of this nature is within state-of-art and possible with or without tether. (authors)

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

Vitrification of High-Level Waste: how to Select the Best Suited Technology - 18117

Since more than 30 years now, a number of countries have selected and operated vitrification as the best industrial practice for safely and durably immobilizing the most active liquid waste. This now significant worldwide experience covers a wide range of waste compositions, from concentrated alkaline defense waste sludges with a variety of salts and chemicals (Na, Fe, Al, Cr, sulphate,... ) to relatively clear acidic HLLW from commercial, low salt, PUREX-type processes, waste that include dissolved cladding material (Al, Mg, stainless steel), waste that include high levels of noble metals and fines... The activity levels vary from 40- year old defense waste to relatively young waste from high burn-up commercial fuel. The current throughputs vary from about 5 kg/h of glass to about 90 kg/h of glass per melter. Even higher throughputs are currently being considered for Hanford facility for instance. In order to cope with the range of waste compositions, glass compositions have also evolved from low melting temperature borosilicate or phosphate glass to glass or glass ceramics with higher melting temperatures. More complex matrices are currently under study in a number of countries, to immobilize more complex waste and / or increase waste loadings. Each challenge is unique and, for each challenge, it is necessary to select the best combination of waste form and technology. For instance for very active solutions with high levels of noble metal particles that could settle, increase apparent glass viscosity, and impact electrical conductivity, one would select borosilicate glass in the mechanically-stirred hot wall induction melter with bottom pouring that has proven its efficiency in industrial conditions at La Hague. For high glass throughputs with low solids in a borosilicate melt (typically Defense waste), ceramic melters have proven their reliability and efficiency. For matrices that need higher temperatures, or with very high corrosiveness, Cold-Crucible Induction Melters are now considered, with one first industrial implementation at La Hague to produce borosilicate glass and a glass-ceramic. All these technologies have been improved over the years to increase throughput, waste loading, melting temperature, reliability... Other types of waste are now being studied worldwide, and for each, it is necessary to carefully select the adapted waste form and technology. The technologies studied will range from traditional technologies (such as hot-wall induction melter technology and liquid-fed ceramic melter technologies) to the more recent advanced technologies (such as Cold Crucible Induction Melter or Advanced Ceramic melters). In addition, the field of vitrification is extending to other types of waste; the current advances in this field will also be described. (authors)

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

Countercurrent Actinide Lanthanide Separation Process (ALSEP) Demonstration Test with a Simulated PUREX Raffinate in Centrifugal Contactors on the Laboratory Scale

An Actinide Lanthanide Separation Process (ALSEP) for the separation of trivalent actinides (An(III)) from simulated raffinate solution was successfully demonstrated using a 32-stage 1 cm annular centrifugal contactor setup. The ALSEP solvent was composed of a mixture of 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and N,N,N',N'-tetra-(2-ethylhexyl)-diglycolamide (T2EHDGA) in n-dodecane. Flowsheet calculations and evaluation of the results were done using the Argonne’s Model for Universal Solvent Extraction (AMUSE) code using single-stage distribution data. The co-extraction of Zr(IV) and Pd(II) was prevented using CDTA (trans-1,2-diaminocyclohexane-N,N,N',N'-tetraacetic acid) as a masking agent in the feed. For the scrubbing of co-extracted Mo; citrate-buffered acetohydroxamic acid was used. The separation of An(III) from the trivalent lanthanides (Ln(III)) was achieved using citrate-buffered diethylene-triamine-N,N,N',N",N"-pentaacetic acid (DTPA), and Ln(III) were efficiently back extracted using N,N,N',N'-tetraethyl-diglycolamide (TEDGA). A clean An(III) product was obtained with a recovery of 95% americium and curium. The Ln(III) were efficiently stripped; but the Ln(III) product contained 5% of the co-stripped An(III). The carryover of Am and Cm into the Ln(III) product is attributed to too few actinide stripping stages, which was constrained by the number of centrifugal contactors available. Improved separation would be achieved by increasing the number of An strip stages. The heavier lanthanides (Pr, Nd, Sm, Eu, and Gd) and yttrium were mainly routed to the Ln product, whereas the lighter lanthanides (La and Ce) were mostly routed to the raffinate.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

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

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

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility, the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗