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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Recent activities regarding the use of burnup credit in France - 14539

Burnup credit refers to the concept of taking credit for the reduction in reactivity due to irradiation of nuclear fuel. A working group, bringing together several French nuclear companies and institutions (AREVA, CEA, EDF and IRSN), was created in 1997 to analyze different propositions to introduce a limited number of fission products together with a more realistic axial burnup profile into the criticality safety studies for the nuclear fuel cycle. Initially, the discussions of the working group focused on PWR UOX fuel. The work and accomplishments performed by this working group on burnup credit regarding PWR UOX fuel assemblies has been presented in various publications. This paper gives an outlook of the French working group on burnup credit activities afterwards. These activities relate to industrial applications of burnup credit for PWR UOX fuel, burnup credit implementation for PWR MOX fuel allowing the plutonium recycling in French PWRs, as well as applications for BWR UOX fuel since it may be transported, stored and reprocessed in France. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Energy implications of mechanical and mechanical–biological treatment compared to direct waste-to-energy

Highlights: • Compared systems achieve primary energy savings between 34 and 140 MJ{sub primary}/100 MJ{sub input} {sub waste.} • Savings magnitude is foremost determined by chosen primary energy and materials production. • Energy consumption and process losses can be upset by increased technology efficiency. • Material recovery accounts for significant shares of primary energy savings. • Direct waste-to-energy is highly efficient if cogeneration (CHP) is possible. - Abstract: Primary energy savings potential is used to compare five residual municipal solid waste treatment systems, including configurations with mechanical (MT) and mechanical–biological (MBT) pre-treatment, which produce waste-derived fuels (RDF and SRF), biogas and/or recover additional materials for recycling, alongside a system based on conventional mass burn waste-to-energy and ash treatment. To examine the magnitude of potential savings we consider two energy efficiency levels (state-of-the-art and best available technology), the inclusion/exclusion of heat recovery (CHP vs. PP) and three different background end-use energy production systems (coal condensing electricity and natural gas heat, Nordic electricity mix and natural gas heat, and coal CHP energy quality allocation). The systems achieved net primary energy savings in a range between 34 and 140 MJ{sub primary}/100 MJ{sub input} {sub waste}, in the different scenario settings. The energy footprint of transportation needs, pre-treatment and reprocessing of recyclable materials was 3–9.5%, 1–18% and 1–8% respectively, relative to total energy savings. Mass combustion WtE achieved the highest savings in scenarios with CHP production, nonetheless, MBT-based systems had similarly high performance if SRF streams were co-combusted with coal. When RDF and SRF was only used in dedicated WtE plants, MBT-based systems totalled lower savings due to inherent system losses and additional energy costs. In scenarios without heat recovery, the biodrying MBS-based system achieved the highest savings, on the condition of SRF co-combustion. As a sensitivity scenario, alternative utilisation of SRF in cement kilns was modelled. It supported similar or higher net savings for all pre-treatment systems compared to mass combustion WtE, except when WtE CHP was possible in the first two background energy scenarios. Recovery of plastics for recycling before energy recovery increased net energy savings in most scenario variations, over those of full stream combustion. Sensitivity to assumptions regarding virgin plastic substitution was tested and was found to mostly favour plastic recovery.

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

Start-up fuel and power flattening of sodium-cooled candle core

The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic or nitride Th fuels relative to the U counterpart fuels. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Accelerating the Decommissioning of the Sellafield Pile Fuel Storage Pond Using Innovative Remote Tooling Developed with the Supply Chain - 16437

The Pile Fuel Storage Pond (PFSP) at Sellafield and adjoining de-canning building provided the storage, cooling and de-canning facility for used fuel and isotopes from the two Windscale reactors and later some Magnox fuel from the Calder Hall reactors. Operations ceased in the 1970's and the facilities are now undergoing a programme of accelerated decommissioning. The pond is a sub-divided outdoor storage pond containing skips of irradiated fuel and miscellaneous waste items and the building contains a series of sub-ponds, known as bays, connected underwater to the main pond. The radiation environment and access limitations necessitate the use of remote technology for the decommissioning programme. This paper describes the philosophy and approach and the technology and equipment adopted for the removal and export of key plant equipment and waste items from within the PFSP. In order to progress the project cost effectively and speedily, the Sellafield project team looked for partners in the supply chain with the required experience, expertise and facilities to supply and test 'commercial off-the-shelf' (COTS) or modified proprietary equipment. This COTS approach minimised the need for multiple, expensive and lengthy design and manufacturing tasks associated with bespoke equipment. Sellafield partnered with specialist remote handling contractor, James Fisher Nuclear, and the respective project teams worked closely together at every stage of the project from qualifying requirements, identifying potential solutions, through to development, testing, training and final deployment. Optioneering of potential solutions involved early 'proof of principle' trials in JFN rig hall facilities, utilising existing equipment and plant mock-ups, along with other equipment manufactured in-house. Together the project teams selected the best solution to take forward, adopting a practical and trials based approach, which allowed rapid and cost effective modification of the solution throughout the project with minimal impact on the programme. When the final solution was validated, plant operators then gained familiarity and training on the equipment so they could execute the work efficiently on plant and so minimise their dose uptake. The removal of a Large De-canner is described in detail. This was a significant obstruction to the emptying, modification and re-use of the bay and a key milestone in the clean-up of the PFSP. The de-canners are large and heavy pieces of equipment which weigh 6.5 t each and approximately 6.5 m long. Their removal was further complicated by limited equipment and man access, radiological contamination from de-canning fuel, challenging material properties, unknown equipment condition and complex lifting requirements. Sellafield Ltd and JFN project teams worked closely together to rapidly develop, trial and deploy a succession of adapted 'commercial off-the-shelf' (COTS) tools whilst addressing emerging challenges and completing the overall task on time and budget. The Sellafield and JFN project and design teams provided an underwater: wheel stop removal tool; towing trolley; grinder on a carbon fibre pole, chute cutting 600 mm circular saw and hydraulic 6 t winch. The suite of tools will be reused on the remaining de-canners in the PFSP and First Generation Magnox Fuel Storage Pond (FGMSP) giving cost and programme savings. Each of the tools can also be used, and some have already been, for other tasks in the bay retrievals such as the size reduction of support frames, large steel sections and chemical de-canner preparatory works. The 'learning from experience' acquired during the project was shared by the Sellafield project with other decommissioning projects across the site. (authors)

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

Partitioning of fission products from irradiated nitride fuel using inductive vaporization

Irradiated nitride fuel (Pu{sub 0.3}Zr{sub 0.7})N fabricated at PSI in frame of the CONFIRM project and having a burn-up of 10.4 % FIMA (Fission per Initial Metal Atom) has been investigated by means of inductive vaporization. The study of thermal stability and release behavior of Pu, Am, Zr and fission products (FPs) was performed in a wide temperature range (up to 2300 C. degrees) and on different redox conditions. On-line monitoring by ICP-MS detected low nitride stability and significant loss of Pu and Am at T>1900 C. degrees during annealing under inert atmosphere (Ar). The oxidative pre-treatment of nitride fuel on air at 1000 C. degrees resulted in strong retention of Pu and Am in the solid, as well as of most FPs. Thermodynamic modelling of elemental speciation using GEM-Selektor v.3 code (Gibbs Energy Minimization Selektor), supported by a comprehensive literature review on thermodynamics of actinides and FPs, revealed a number of binary compounds of Cs, Mo, Te, Sr and Ba to occur in the solid. Speciation of some FPs in the fuel is discussed and compared to earlier results of electron probe microanalysis (EPMA). Predominant vapor species predicted by GEM-Selektor calculations were Pu(g), Am(g) and N{sub 2}. Nitrogen can be completely released from the fuel after complete oxidation at 1000 C. degrees. With regard to the irradiated nitride reprocessing technology, this result can have an important practical application as an alternative way for {sup 15}N recovery. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

On the Use of Ionic Liquids for the Recycling of the Contaminated Material

Dealing with spent nuclear fuel (SNF) and other radioactive waste management is one of the main issues of the nuclear industry. Today, there is an increased demand for lower environmental affection and higher safety assurances within all industries dealing with radioactivity. The resulting increasing costs for dealing with radioactive waste, altogether with the pressure of public opinion, are strong motivation factors for the development of recycling schemes, softening the effects on the environment, and bringing economic advantages for the nuclear industry. This problem is even multiplied as a significant wave of decommissioning of nuclear facilities can be expected in the next decades and it makes the problem even more urgent. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency estimates that 400 nuclear power plants will have to be decommissioned between 2014 and 2050 and it will generate more than 5 million tons of radioactive scrap metal. If the decommissioning of other nuclear installations is taken into account, almost 30 million tons of scrap metal contaminated by radionuclides will need to be handled?1 According to an International Atomic Energy Agency report?2 during the nuclear facility decommissioning process, only 3 % of the material is considered contaminated, but costs for its disposal can make over 60 % of overall costs. In both of mentioned areas - SNF reprocessing and material decontamination - possible benefits of ionic liquids (ILs) are studied worldwide because of their interesting physical-chemical properties and chemical behavior. ILs (described as room-temperature molten salts in older literature) are commonly described as salts with organic cations associated with organic or inorganic anions that are liquid below 373 K and fully consisting of ions and offer advantages of nonaqueous processes like a safer environment because of substituting volatile and flammable organic solvents or less serious criticality issues. At the same time, ILs beat high-temperature molten salt in comparing energy costs. Also some other of ILs' unique properties are definitely attractive for areas of SNF reprocessing and solid waste decontamination. Among other major advantages of ionic liquids, negligible vapor pressure, non-flammability, chemical and mechanical stability, tunability as well as good electric conductivity or wide electrochemical window are usually being mentioned. In this contribution, the experimental verification of some principles needed for the development of the decontamination of material contaminated with uranium and europium are presented as model cases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Considerations for Managing DOE Standard Canisters within an Over-canister as Part of an Integrated Waste Management System PPT

To better enable informed decision making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management Program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring research into a comprehensive integrated waste management system (IWMS) that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). An important aspect of the IWMS is DOE-managed Spent Nuclear Fuel (SNF). DOE and its predecessor agencies have generated, transported, received, stored, and reprocessed SNF at DOE facilities nationwide, and DOE is responsible for managing the SNF currently in its possession. These fuels come from a wide range of reactor types that employ various cladding materials, fuel materials, and enrichments. To enable interim, road-ready dry storage (RRDS) of the wide variety of SNF types found in the DOE inventory, a standardized canister system (i.e., the DOE Standard Canister) was proposed for the packaging demonstration. This robust, welded canister system is designed to confine radionuclides, prevent criticality by precluding content moderation, and satisfy other requirements as part of a larger storage, transportation, and disposal system. While SNF has yet to be loaded into a DOE Standard Canister, DOE Standard Canister designs were included in past storage facility and disposal facility design licensing endeavors. In a renewed effort to evaluate packaging SNF at Idaho National Laboratory (INL) in a RRDS configuration, researchers are planning the RRDS Packaging Demonstration. This demonstration is supplemented by analytical structural, criticality, and material compatibility evaluations that support management of SNF in DOE Standard Canisters, taking advantage of past analysis work to the extent possible. One of the largest differences between the current Packaging Demonstration and past analytical evaluations is the inclusion of an over-canister containing multiple DOE Standard Canisters. For the Packaging Demonstration, DOE Standard Canisters loaded with SNF are planned to be placed in a larger diameter over-canister. The sealed over-canister could then be placed in a storage overpack for onsite storage, or in a transportation overpack for shipment to an offsite storage location or disposal site once one becomes available. This paper examines the relevant considerations and provides a preliminary evaluation of integrating the over-canister configuration into the storage, transportation, and disposal processes of the overall waste management system. For storage and transportation, the over-canister can be considered analogous to a multi-purpose canister (MPC) for commercial SNF. For disposal, the DOE Standard Canisters could be removed from the over-canister and placed in a co-disposal waste package with canisters containing vitrified high-level radioactive waste (HLW) similar to configurations examined previously, or the sealed over-canisters might be capable of direct disposal in a waste package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Analysis of H-Canyon Process Tanks in Preparation of Consolidation and Blending for HALEU Fuels

High-Assay Low- Enriched Uranium (HALEU) fuels are being developed to support the replacement of Highly Enriched Uranium (HEU) fuels used in U.S. High-Performance Research Reactors (USHPRR) as well as advanced nuclear power reactor designs. The projected demand for HALEU far exceeds the supply and studies are underway to assess various options to partially mitigate the potential short supply. The H Canyon facility at the Savannah River Site (SRS) Low Enriched Uranium (LEU) containing 4.95% U-235 from the reprocessing of highly enriched foreign and domestic research reactor fuel for the Tennessee Valley Authority’s (TVA) commercial power reactor market for several decades. The production of LEU at the H-Canyon facility can be readily transitioned to produce 19.75% HALEU solutions from the current separated inventory of purified HEU solutions in H-Canyon storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Uranium Ion Exchange Resin Performance - A Case Study at 200 West Pump and Treat - 18103

The 200 West Pump and Treat Facility (200W P and T), located on the U.S. Department of Energy's Hanford site, is a 2,500 gallon per minute (GPM) plant designed to capture and treat an eight square kilometer groundwater plume. The groundwater contamination is largely a result of process liquids disposed from operations involving the separation of nuclear materials and storage of spent nuclear fuel. This document will report on the performance of a full scale uranium ion exchange system used to treat contaminated groundwater and other waste streams. The uranium ion exchange system at 200W P and T was designed as a three vessel operation with a lead, lag and polish vessel, which contain 300 cubic feet of resin in each vessel. The ion exchange system treats a portion of the total plant flow with a maximum capacity of 400 GPM. Prior to start-up, one key issue was the potential to treat feed streams with a wide range of uranium concentrations (22 to 150,000 microgram per liter [μg/L]). The resin change-out frequency was a primary concern. A literature review was performed, resulting in resin capacity and change out frequency that varied by two orders of magnitude (1.2 milligram per gram [mg/g] to 104 mg/g capacity). Since startup in September 2015, the ion exchange system has experienced a dynamic feed stream as wells were brought online, which changed the contaminant concentrations. The plant also treats leachate liquid waste from on-site low-level waste site, known as the Environmental Restoration Disposal Facility (ERDF), which contains uranium as well as other competing anions, such as sulfate. The uranium breakthrough is closely monitored throughout the resin beds with weekly sampling events at the influent, effluent, and intermediate ports in the first vessel. In the influent to the ion exchange system, the uranium concentration fluctuated from 24 μg/L up to 910 μg/L. The information on influent and effluent was used to determine the change out frequency and the capacity of the resin for uranium under these operating conditions. (authors)

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

Micro-Pocket Fission Detectors (MPFD) For Fuel Assembly Analysis

Neutron sensors capable of real-time measurement of thermal flux, fast flux, and temperature in a single miniaturized probe are needed in irradiation tests required to demonstrate the performance of candidate new fuels, and cladding materials. In-core ceramic-based miniature neutron detectors or “Micro-Pocket Fission Detectors” (MPFDs) have been studied at Kansas State University (KSU). The first MPFD prototypes were tested in various neutron fields at the KSU TRIGA research reactor with successful results. Currently, a United States Department of Energy-sponsored joint KSU/Idaho National Laboratory (INL) effort is underway to develop a high-temperature, high-pressure version of the MPFD using radiation-resistant, high temperature materials, which would be capable of withstanding irradiation test conditions in high performance material and test reactors (MTRs). Ultimately, this more compact, more accurate, and longer lifetime flux sensor for critical mock-ups, existing and advanced reactor designs, high performance MTRs, and transient test reactors has the potential to lead to higher accuracy and resolution data from irradiation testing, more detailed core flux measurements and enhanced fuel assembly processing. Prior evaluations by KSU indicate that these sensors could also be used to monitor burn-up of nuclear fuel. If integrated into nuclear fuel assemblies, MPFDs offer several advantages to current spent fuel management systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fabrication and Properties of Technetium-Bearing Pyrochlores and Perovskites as Potential Waste Forms - 13222

Technetium-99 (t{sub 1/2}= 2.13x10{sup 5} years) is important from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes in used nuclear fuel (UNF). As such, it is targeted in UNF separation strategies such as UREX+, for isolation and encapsulation in solid waste forms for storage in a nuclear repository. We report here results regarding the incorporation of Tc-99 into ternary oxides of different structure types: pyrochlore (Nd{sub 2}Tc{sub 2}O{sub 7}), perovskite (SrTcO{sub 3}), and layered perovskite (Sr{sub 2}TcO{sub 4}). The goal was to determine synthesis conditions of these potential waste forms to immobilize Tc-99 as tetravalent technetium and to harvest crystallographic, thermophysical and hydrodynamic data. The objective of this research is to provide fundamental crystallographic and thermophysical data on advanced ceramic Tc-99 waste forms such as pyrochlore, perovskite, and layered perovskite in support of our current efforts on the corrosion of technetium-bearing waste forms. The ceramic Tc-99-bearing waste forms exhibit good crystallinity. The lattice parameters and crystal structures of the technetium host phases could be refined with high accuracies of ±3, ±4, and ±7 fm (10{sup -15} m), for Nd{sub 2}Tc{sub 2}O{sub 7}, SrTcO{sub 3}, and Sr{sub 2}TcO{sub 4}, respectively. The associated refinement residuals (R{sub Wp}) for the patterns are 4.1 %, 4.7 % and 6.7 %, and the refinement residuals for the individual phases (R{sub Bragg}) are 2.0 %, 2.4 % and 3.9 %, respectively. Thermophysical properties of the oxides SrTcO{sub 3}, Sr{sub 2}TcO{sub 4}, and Nd{sub 2}Tc{sub 2}O{sub 7} were analyzed using AC magnetic susceptibility measurements to further harvest information on the critical temperature (T{sub c}) for superconductivity. In our experiments the strontium technetates, SrTcO{sub 3} and Sr{sub 2}TcO{sub 4}, show superconductivity at rather high critical temperatures of T{sub c} = 7.8 K and 7 K, respectively. On the other hand Nd{sub 2}Tc{sub 2}O{sub 7} did not show any changes in magnetic properties above 3 K. (authors)

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

Magnetic separation - Advanced nanotechnology for future nuclear fuel recycle

The unique properties of magnetic nanoparticles (MNPs), such as their extremely small size and high surface area to volume ratio, provide better kinetics for the adsorption of metal ions from aqueous solutions. In this work, we demonstrated the separation of minor actinides using complex conjugates of MNPs with diethylenetriamine-pentaacetic acid (DTPA) chelator. The sorption results show the strong affinity of DTPA towards Am (III) and Pu (IV) by extracting 97% and 80% of actinides, respectively. It is shown that the extraction process is highly dependent on the pH of the solution. If these long-term heat generating actinides can be efficiently removed from the used fuel raffinates, the volume of material that can be placed in a given amount of repository space can be significantly increased. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Pile Fuel Cladding Silo Deflector Plate Removal Project Using Long Reach Tooling and Suspended Abrasive Water Jet Cutting System - 17541

The Pile Fuel Cladding Silo (PFCS) at the Sellafield decommissioning, reprocessing and nuclear waste management facility in the UK was constructed in the late 1940-s and used for the storage of Intermediate Level Waste (ILW) arising from the Windscale Piles, Calderhall and Chapelcross Reactors until the 1960's. The PFCS is 21 metres high and comprises 6 identical tall compartments with the waste added to the silo via a charge-hole at the top of each compartment; each compartment is split by a central dividing wall that runs from the base of the silo to a few metres below the underside of the silo roof. A deflector plate assembly (DPA) supported on two horizontal beams is located between the top of the dividing wall and the roof and allowed the waste to spread as it fell into the silo and prevent excessive mounding. Sellafield Ltd is currently engaged in an overall programme of works to remove the waste for storage within modern ILW containers. Prior to the commencement of these works the Deflector Plate Assemblies (DPA) and beam structures within the silos are required to be removed to allow access for the waste retrievals equipment through the side of the silo at the point where the DPAs are situated. James Fisher Nuclear (JFN) and Sellafield Ltd have developed a system using suspended water jet cutting equipment combined with custom long reach tooling, installed through penetrations made in the side of the silo, to enable the remote size reduction of the DPA and beams within each compartment. The use of a low pressure suspended water jet cutting system has enabled a single water/abrasive hose to be routed down a tool deployment pole and allowed long hose lengths up to 80 m from pump to nozzle enabling siting of equipment on a congested and difficult access site. The use of lightweight, yet high strength, carbon fibre poles within the tools has enabled the deployment of tool end effectors at distances of up to 6.5 m from the compartment walls using a relatively small 60 mm diameter pole. This maximised articulation of the pole within the 300 mm wall penetrations to reach all areas of the DPA. Capability of the operators to install the tooling and size reduce DPA and beam assemblies have been demonstrated through the training on a Full Scale Test Rig (FSTR). The FSTR was integral to the development of the equipment and techniques and to providing assurance to the client that the project was technically achievable given the tight building restraints and safety issues. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Use of Sensors and Machine Learning for Signal Discovery in a Solvent Extraction Process – 23280

Reprocessing is an important step in the nuclear fuel cycle where usable nuclear materials are extracted from spent fuel for recycling. The extraction of materials for reuse simultaneously reduces not only the volume of nuclear waste, but its decay time to radioactivity levels similar to that of the originating uranium ore. As part of an initiative to steward research, development, and innovation into the nuclear fuel cycle, Idaho National Laboratory is designing and constructing a solvent extraction testbed named Beartooth. Beartooth will allow researchers to refine extraction processes, test innovative extraction processes, and give early career scientists opportunities to gain skills in performing separations chemistry that uses centrifugal contactors. In addition, the Beartooth testbed is being uniquely designed to enable novel technologies including machine learning capabilities for the characterization of chemical process operations in near real-time. To aid in the design of Beartooth, a team of researchers are installing a variety of atypical sensors into a system of contactors for signal discovery. The team will implement machine learning methods on sensor data to determine signal features with the goal of providing a process operator with a deeper understanding of the chemical process and equipment usage. This work will summarize sensors utilized and preliminary results from an infrared camera.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Use of Sensors and Machine Learning for Signal Discovery in a Solvent Extraction Process

Reprocessing is an important step in the nuclear fuel cycle where usable nuclear materials are extracted from used fuel for recycling. The separation of materials for reuse simultaneously reduces not only the volume of nuclear waste, but its decay time to radioactivity levels similar to that of the originating uranium ore. As part of an initiative to steward research, development, and innovation into the nuclear fuel cycle, Idaho National Laboratory is designing and constructing a solvent extraction testbed named Beartooth. This testbed will allow researchers to refine separation processes, test innovative extraction processes, and give early career scientists opportunities to gain skills in performing separations chemistry utilizing centrifugal contactors. In addition, the Beartooth testbed is being uniquely designed to enable novel technologies including machine learning capabilities for the characterization of chemical process operations in near real-time. To aid in the design of Beartooth, a team of researchers are installing a variety of atypical sensors into a system of contactors for signal discovery. The team will implement machine learning methods on acquired sensor data to extract signal features. The goal is to provide a process operator with a deeper understanding of the chemical process and equipment usage. This work will summarize sensors utilized and preliminary results from an infrared camera.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Non-Equilibrium Actinide Radiation Chemistry and the Nuclear Fuel Cycle

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

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2013

Demonstrating the Safety of Long-Term Dry Storage - 13468

Commercial nuclear plants in the United States were originally designed with the expectation that used nuclear fuel would be moved directly from the reactor pools and transported off site for either reprocessing or direct geologic disposal. However, Federal programs intended to meet this expectation were never able to develop the capability to remove used fuel from reactor sites - and these programs remain stalled to this day. Therefore, in the 1980's, with reactor pools reaching capacity limits, industry began developing dry cask storage technology to provide for additional on-site storage. Use of this technology has expanded significantly since then, and has today become a standard part of plant operations at most US nuclear sites. As this expansion was underway, Federal programs remained stalled, and it became evident that dry cask systems would be in use longer than originally envisioned. In response to this challenge, a strong technical basis supporting the long term dry storage safety has been developed. However, this is not a static situation. The technical basis must be able to address future challenges. Industry is responding to one such challenge - the increasing prevalence of high burnup (HBU) used fuel and the need to provide long term storage assurance for these fuels equivalent to that which has existed for lower burnup fuels over the past 25 years. This response includes a confirmatory demonstration program designed to address the aging characteristics of HBU fuel and set a precedent for a learning approach to aging management that will have broad applicability across the used fuel storage landscape. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗