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At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Uncertainty improvement of 22 Na based radioactive tracer dilution for determining total mass of pyroprocessing molten salt systems by 154 Eu removal

To determine the total salt mass of the molten salt systems for pyroprocessing spent nuclear fuels, a 22 Na based radioactive tracer dilution was studied in Idaho National Laboratory in recent years. This 22 Na based RTD technique was deemed feasible, but due to the gamma energy peak of 22 Na coinciding with one of the energy peaks of 154 Eu radioisotope in the molten salt, the uncertainty of the 22 Na radioactivity in the 22 Na-spiked salt samples was quite high. To improve the uncertainty of the 22 Na based RTD technique, we proposed to chemically remove the 154 Eu of the salt samples by DGA resin for gamma spectroscopy. The effectiveness of removing 154 Eu on uncertainty improvement was evaluated. Furthermore, it was found that (1) the 154 Eu fission product effect on the uncertainty and detection limit can be effectively eliminated by chemically removing the 154 Eu during the salt sample preparation and (2) the uncertainty of 22 Na radioactivity in the salt samples for electrorefining was significantly improved from 13% to 2%, showing the potential of practical engineering application of 22 Na based RTD as a safeguards technique for molten salt systems for pyroprocessing spent nuclear fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes PE for high-temperature processing. After removal of the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using an Fe-based catalytic process. The PE-derived graphite anode in a lithium-ion coin cell showed a specific capacity of 345 mAh/g at 0.05C with an initial Coulombic efficiency of 87% and reversible capacity retention of ~100% at different current rates. It also showed a specific capacity of up to 313 mAh/g at 0.5 discharge/charge cycles per hour (0.5C) and Coulombic efficiency of 99.9% after 250 cycles, indicating excellent electrochemical performance as an anode material for lithium-ion batteries. This method illustrates that there are opportunities for upcycling large quantities of PE waste to produce graphite powders suitable for use in LIBs.

25 ENERGY STORAGE↗
DOE OSTI2019

Three Years of Multilevel Monitoring Data at the Riverton, Wyoming, Processing Site That Show Contaminant Increases After River Flooding Events and a Large Recharge Event

The tailings pile at the Riverton, Wyoming, Processing Site has been removed, but contaminant concentrations in the shallow aquifer increase after large recharge events. The Riverton site experienced flooding near the Little Wind River in early May 2016, February 2017, and June 2017. The May 2016 flooding was from a large rain event. The February 2017 flooding was due to an ice jam, and the June 2017 flooding was due to mountain snowmelt runoff. In addition, the site experienced high water tables in April and May 2017, as this was the second wettest spring in Riverton since records began in 1918. Past reports indicated the accumulation of solid-phase uranium and molybdenum in evaporites above the respective contaminant groundwater plumes. Given these data, multilevel monitoring wells were installed in 2015 to monitor seasonal changes in groundwater quality. Data from three years of groundwater monitoring are summarized in this report and compared with the flooding and large recharge events. The results indicate that evaporite-related constituents (especially sodium, sulfate, and chloride) present across the site in a silt layer can be released to the shallow sand and gravel aquifer during flooding and large recharge events. The solid-phase and water-phase data confirm that these evaporites are mainly sodium sulfate and sodium chloride salts with some calcite and gypsum.

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

Fundamental Study of Uranium Dendrites Electrochemically Separated in Molten Salt with the Presence of Cerium

In this study, uranium was electrochemically separated in the presence of cerium in eutectic LiCl-KCl at varying concentrations of cerium and varying overpotentials. Harvested dendrites were then examined and analyzed for their morphological features and elemental composition using various analytical techniques. Conclusions: Increasing the applied overpotential causes the dendrites to evolve from relatively simple and large platelet-like crystals (25 mV), to a grid-like network with crystals having jagged and smooth features (50 mV, 100 mV), and finally into a very chaotic and complex crystal morphology (200 mV) for all uranium-cerium compositions explored. Varying the uranium:cerium concentration ratio did not have much effect on the uranium dendrite morphologies. EDS results show that cerium is not co-deposited on the surface of the dendrite crystals, ICP-MS results do show a presence of cerium in the volumetric composition. This is likely due to salt occlusion during the dendrite formation. XRD has been performed on several samples at the 25 mV overpotential with 1 wt% U and 0 wt% Ce in the molten salt eutectic. These results indicate that there is not salt occlusion within the dendrites during electrochemical separation. Additionally, the uranium metal quickly oxidizes during preparation for XRD.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Normal Pulse Voltammetry for Real Time Analysis of Electro-refiner Salt with High Concentrations of Analytes

Spent nuclear fuel pyro-processing could be a preferred alternative for aqueous processing. However, there will be a large hold up of actinides in the molten salt media that must be safeguarded in commercial facilities. A real or near real time monitoring method for electro-refiner salt needs to be developed to meet this requirement. In addition, real time monitoring method can also be applied to optimize process control. Voltammetry would be an ideal method to monitor the ER salt composition because of the low cost and robustness. The equipment required to carry out voltammetry analysis include only a potentiostat and three electrodes, which can be placed in-situ.

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

Ion separations with membranes

Ion separations are important for resource recovery, water treatment, and energy production and storage. Techniques such as chemical precipitation, selective adsorption, and solvent extraction are effective, but membranes may separate ions continuously with less waste and lower energy costs. Separation of monovalent and multivalent ions with nanofiltration or electrodialysis membranes already enables water softening and edible salt purification. Similar membranes are attractive as separators in vanadium redox flow batteries. Selective partitioning of divalent counter-ions into ion-exchange membranes even allows transport of these ions against their concentration gradients in salt mixtures. However, separations of ions with the same charge is more challenging. Recent research demonstrated highly selective ion “sieving” at small scales. Here, separations using electrical potentials and differences in ion electrophoretic mobilities are promising, but relatively unexplored. Carrier-mediated transport affords high selectivity in liquid membranes, but these systems are not very stable, and selective transport via hopping between anchored carriers has proven elusive. Finally, this paper discusses how concentration polarization decreases selectivities in many membrane processes. Although development of selective, inexpensive ion-separation membranes is a work in progress, successes in water softening and edible salt purification suggests that future selective membranes will serve as complementary methods to traditional purification techniques.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Planning for Material Control and Accountancy at Liquid Fueled Molten Salt Reactors

The purpose of this report is to provide molten salt reactor (MSR) developers and future US Nuclear Regulatory Commission (NRC) license applicants with recommendations for developing an effective and practical material control and accounting (MC&A) plan, focused primarily on MSR designs that use circulating liquid fuel. Because of the breadth of MSR designs, there is no single, generic, detailed MC&A plan that will work for every design. The wide variation of fresh fuel salts, the method and frequency of loading fresh fuel, the reactor system design components (e.g., tanks, filtration systems, chemical processing streams), and waste streams will determine the specific measurement locations and instrumentation that can best meet MC&A objectives throughout an MSR facility. Additionally, MSR designs are rapidly evolving, and new design features and deployment scenarios that will affect MC&A are being explored and pursued. This report defines a generic MC&A approach that was developed for terrestrial (as opposed to maritime) deployments to meet the intent of NRC domestic safeguards and MC&A. MSR license applicants should consider nuclear safeguards (both domestic and international) and security throughout the design, as early as the preconceptual design phase. MC&A of special nuclear material (SNM) is an aspect of the NRC’s domestic safeguards program, alongside physical protection. Because liquid-fueled MSRs are reactors with SNM in nondiscrete (or item) form, it is likely that the NRC may require liquid-fueled MSR license applicants to submit a formal MC&A plan as a part of their license application. Currently, the NRC licensing protocol presents a challenge because the NRC MC&A regulations have not been updated to accommodate advanced reactors, including types of MSRs. Because no liquid-fueled MSR has been licensed for operation at the time of this report, no template or precedence for a successfully licensed MSR MC&A plan exists. However, the MSR license applicant can take advantage of the NRC’s published commitments to performance-based regulations. The authors recommend that the license applicant, or MSR designers, develop an MC&A plan throughout the design lifecycle and plan to submit a detailed MC&A program description, or MC&A plan, to the NRC as a part of a license application. No MC&A plan template or guidance exists that is specific to liquid-fueled MSRs. The authors recommend that license applicants discuss the topic of MC&A during preapplication engagement. Because of the uniqueness of MC&A for liquid fueled MSRs, the authors recommend that liquid fueled MSR developers engage with the NRC on the topic of MC&A in the early phases of its design development and follow up any time there are significant modifications in design plans that would affect MC&A. For example, topics like modifications in fuel handling processes, changes in uranium enrichment, or additional chemical processing streams added to the design could be discussed with the NRC specifically on the topic of MC&A.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Sodium and Lithium Hydride Shields At the Department of Energy's Oak Ridge Reservation - 18412

URS-CH2M Oak Ridge LLC (UCOR) is responsible for cleanup of the U.S. Department of Energy's (DOE) East Tennessee Technology Park (ETTP) in Oak Ridge, Tennessee. Formerly known as the K-25 site, the facility enriched uranium for major defense and energy programs for more than 40 years. DOE is seeking the safest, most environmentally protective and fiscally responsible treatment for its large inventory of radioactive-contaminated metals known as 'shields'. The 66 sodium and lithium shields were constructed for use in experiments at the Tower Shielding Facility (TSF), Oak Ridge National Laboratory (ORNL). The TSF operated from 1954 to 1992 and was designed and built for radiation-shielding studies. The shield containers are constructed of aluminum or stainless steel of varying sizes and shapes that were then filled with either sodium metal or lithium hydride material. The historical records indicate the sodium shields were constructed at the Molten Salt Reactor for use in experiments in the 1971 time frame, while the lithium shields were constructed at the Y-12 site and were used in experiments as early as 1961. A subset of the sodium and lithium shields contain depleted uranium. The sodium and lithium shields were initially used at the TSF to perform in-depth measurements of the neutron transport through the shield materials. Based on process knowledge, the shields are managed as radiologically contaminated containers. Efforts have been made over the last fifteen years to find recycle markets for both the sodium and lithium hydride shields. This includes an onsite operation that was initiated at the East Tennessee Technology Park (ETTP) in 2004. After an operational exothermic reaction incident occurred soon after startup, the onsite ETTP recycling action was discontinued and a search for offsite recycling facilities was initiated. In 2016, the Y-12 National Security Complex was contacted regarding recycle/repurpose of the lithium shields; however, the lithium was determined to not be acceptable for their current repurpose process. To date, no offsite recycling outlets have been identified. Recent progress on the waste disposal actions, a shipment of 22 small sodium shields have been shipped and treated offsite. Therefore, the current shield inventory consists of 29 sodium shields located at ETTP and 15 sodium and lithium hydride shields staged at ORNL. The time period needed to reasonably identify safe disposition or recycling outlets for the shields was expected to exceed the one-year Land Disposal Restrictions storage time limit; therefore, approval was received to add the shields to the Site Treatment Plan (STP) in March, 2017. Until a disposition path is found, the shields will continue to be stored in a safe configuration that is protective of human health and the environment. This engineering evaluation identifies and screens the alternatives for dispositioning the shields at the Oak Ridge(ORR). Four alternatives were evaluated including; (1) macro-encapsulation, (2) deactivation, (3) recycle and (4) leave-in-place. An initial screening was performed, and the alternative passing the initial screening are then evaluated in more detail. The evaluation criteria used included environmental risk, technical feasibility, administrative feasibility, worker safety, nuclear safety, transportation safety, and cost. (author)

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

The controlling role of atmosphere in dawsonite versus gibbsite precipitation from tetrahedral aluminate species

In highly alkaline solution, aluminum speciates as the tetrahedrally coordinated aluminate monomer, Al(OH)4- and/or dimer Al2O(OH)62-, yet precipitates as octahedrally coordinated gibbsite (Al(OH)3). This tetrahedral to octahedral transformation governs Al precipitation, which is crucial to worldwide Al production, and to the processing of caustic high-level radioactive wastes. Despite its significance, the transformation pathway remains unknown. Here we explore the roles of atmospheric water and carbon dioxide in mediating the transformation of the tetrahedrally coordinated potassium aluminate dimer salt (K2Al2O(OH)6) to gibbsite versus potassium dawsonite (KAl(CO3)(OH)2). A combination of in-situ attenuated total reflection infrared spectroscopy, ex-situ micro X-ray diffraction, and multivariate curve resolution-alternating least squares chemometrics analysis reveals that humidity plays a key role in the transformation by limiting the amount of alkalinity neutralization by dissolved CO2. Lower humidity favors higher alkalinity and incorporation of carbonate species in the final Al product to form KAl(CO3)(OH)2. Higher humidity enables more acid generation that destabilizes dawsonite and favors gibbsite as the solubility limiting phase. Because the transformation was restricted to occur in thin water films, the results suggest that transition from tetra- to octahedrally coordinated Al does not have to occur in bulk solution, as has often been hypothesized, but may instead appear on the source mineral surface.

Dembowski, Mateusz↗
DOE OSTI2022

Results of Long-Term Analyses of the Salt (Macro) Batch 5 Tank 41h Qualification Sample (Resample)

Savannah River National Laboratory (SRNL) analyzed samples from Tank 41H in support of qualification of Salt Water Processing Facility (SWPF) Salt Batch 5. This document reports the long-term results of the analyses of samples of Tank 41H. Analysis of the Tank 41H Salt Batch 5 sample indicated that the material does not display any unusual characteristics or observations, such as floating solids, the presence of large amounts of solids, or unusual colors. This memo satisfies Deliverable 3 of the Technical Task Request (TTR). After a data review it was found that the original results reports in the revision 0 of this document were actually from Salt Batch 4 (Tank 21H). A new set of samples was sent for analysis to satisfy Deliverable 3 of the TTR. The results from the original analyses have been replaced with the recent analyses of the actual Salt Batch 5 sample.

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

The role of international collaboration in the United States geologic disposal research program

Abstract. More than a decade ago, the United States disposal program discontinued all research activities focused on the unsaturated fractured tuff formation at Yucca Mountain as the geologic disposal site for spent fuel and high-level radioactive waste. A new research and development (R&D) program was initiated to provide a sound technical basis for alternative disposal options across clay, crystalline, and salt rocks. The goals of this broad program were (and still are) to (1) increase confidence in the robustness of generic disposal concepts, (2) develop the science and engineering tools needed to support disposal concept implementation, and (3) conduct R&D on the direct disposal of existing dual-purpose (storage and transportation) canisters. Recognizing the benefits of international collaboration toward the common goal of safely and efficiently managing the back end of the nuclear fuel cycle, the program emphasized international cooperation as an effective strategy for sharing information and knowledge. In a multi-laboratory effort coordinated by Lawrence Berkeley National Laboratory, the United States Department of Energy (DOE) program established formal and informal cooperation partnerships with several international initiatives and institutions and developed a number of collaborative R&D activities in important research areas, such as engineered barrier integrity, near-field perturbations, radionuclide transport, performance assessment, and methods for characterization and monitoring of engineered and natural barriers. This presentation gives an overview of these R&D activities, with a specific focus on activities that improve our current understanding of the coupled thermal–hydrological–mechanical–chemical (THMC) processes occurring in engineered and natural barriers. We start with a brief review of selected international collaboration initiatives and then describe a few specific collaboration projects. We focus specifically on such studies that use experimental data sets provided by international research cooperation for joint modeling work to increase confidence in performance-relevant predictions of coupled processes. Overall, the focus on international collaboration has allowed deep engagement of US researchers with the international waste management R&D community in terms of best practices, new scientific advances, state-of-the-art simulation tools, new monitoring and performance confirmation approaches, and lessons learned. The joint R&D with international researchers, worldwide sharing of knowledge and experience, and access to relevant data and experiments from a variety of host rocks have helped our researchers to significantly improve their understanding of the current technical basis for disposal in a range of potential host rock environments. International collaboration also provides ample opportunity for training and educating junior staff that are well suited to move the United States disposal research program forward into the next decades, a promising avenue for developing a next-generation workforce of disposal scientists.

Birkholzer, Jens T.↗
DOE OSTI2022

Frustrated Coulombic and Cation Size Effects on Nanoscale Boehmite Aggregation: A Tumbler Small- and Ultra-Small-Angle Neutron Scattering Study

To better understand the effects of solution chemistry on particle aggregation in the complex legacy tank wastes at the Hanford (WA) and Savannah River (SC) sites, we have performed a series of tumbler small- and ultra-small-angle neutron scattering experiments on 20 wt % solid slurries of nanoparticulate aluminum oxyhydroxide (boehmite) with M 1+ nitrates of various concentrations and radii. In this study, the solutes consisted of H, Li, Na, K, and Rb nitrates at 10 –5 , 10 –3 , 10 –1 , 2, and 4 molal (m) concentrations, as well as in pure H 2 O. Synthetic boehmite nanoparticles were used with a size range from ~20 to 30 nm. Tumbler cells were used to keep the solids from settling. Although particles initially form individual rhombohedral platelets, once placed in solution, they quickly form well-bonded stacks, primary aggregates, up to ~1500 Å long, and a second level of aggregates whose concentration and structure vary as a function of cation type and concentration. Aggregation generally increases with increased solute concentration and with cation radius up to a concentration somewhat above 10 –1 m, at which point the trend reverses. Primary aggregates become more rodlike and larger. The Kirkwood-like reversal probably reflects a change from Derjaguin–Landau–Verwey–Overbeek (DLVO)/Debye behavior controlled by surface chemistry to a frustrated Coulombic system controlled by the solution structure. These data suggest that an understanding of the effects of salt concentration and chemistry on nanoparticle aggregate structures provides useful physical insights into the microscopic origin of slurry rheology in the Hanford and Savannah River legacy wastes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Developments of solid materials for UF 6 sampling

This project demonstrated that a device using majority Commercial Off the Shelf (COTS) components could be used to collect uranium hexafluoride samples safely from gaseous or solid sources. The device was based on the successful Cristallini method developed by ABACC over the past 10 years. The system was designed to capture and store the UF 6 as an inert fluoride salt to ease transportation regulations. In addition, the method was considerably faster than traditional cryogenic methods, collected enough material to perform analyses without undue waste, and could be used either inside a facility or in the storage yard.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Stabilization of Uranium Metal by Chloride Conversion-Economical, Safe, Scalable [Factsheet]

Researchers at Los Alamos National Laboratory have developed SUCCESS, a process to convert uranium metal to uranium tetrachloride enabling production of these valuable compounds for actinide science, molten salt nuclear reactors, and energy research. Conventional methods are problematic due to the oxide reduction processes use rare reagents and generate significant amounts of waste and hydride conversion process using chlorine at high temperature is hazardous. This new method using inexpensive reagents chemically transform uranium metal into uranium halide compounds under mild conditions. The facile conversion to stable uranium tetrachloride could reduce the need to mine uranium for uses other than nuclear, by recycling the uranium without oxidation. The Laboratory is interested in supporting further scale-up efforts using a Cooperative Research and Development Agreement (CRADA) or by granting a license to a qualified entity

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Simulant Development of Potential 200 West Area Waste Feeds

Preliminary planning for retrieval, qualification, and pretreatment of waste in Hanford’s 200 West Area (200W) has begun as part of the West Area Risk Management project. Experimental studies to technically mature pretreatment process operations will likely be needed because of the uniqueness of 200W waste. Pacific Northwest National Laboratory formulated five simulants to represent 200W-qualified feed based on the preliminary flowsheet provided by Washington River Protection Solutions, LLC. The simulant recipes were devised using applicable historical information as a reference point to support the use of the flowsheet waste vectors, which were combined into five distinct groups. These five groups formed the basis for the liquid composition targets that were adapted into recipes using charged-balanced salt species. The liquid phase recipes were batched in 1-L quantities and analyzed at Pacific Northwest National Laboratory. Once confirmed to be stable, the liquid solutions were tested for compatibility with candidate solid components. Specific solid components were recommended based on cross-examining the proposed solid phases in the flowsheet with relevant data from the literature. Mixtures of solid components were added to aliquots of the liquid batches and sub-sampled to measure particle size distribution. The measured distribution was compared to independently created benchmark distributions appropriate for each simulant. This process was iterated until a solid phase composition that resulted in a representative particle size distribution was found. After the final compositions were confirmed, a suite of chemical and physical characterization data was collected. This report describes the simulant basis, formulation methodology, laboratory measurements, and data collected for the recipes recommended to represent 200W waste feeds.

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

Thin-Layer Electrodeposition of Uranium Metal from Molten LiCl-KCl

A uranium metal electrodeposition extension to the previously proposed analytical technique for determining actinides in molten salts is presented. The technique combines information from electrodeposition with the output of an alpha particle detector. This technique requires a thin actinide layer to be deposited on a metal substrate. Electrodeposition of a thorium layer has been reported previously. Electrodeposition of a thin uranium layer from molten LiCl-KCl onto nickel plates is reported in this paper using repeating chronoamperometry. Alloy formation was encountered which caused unusual deposit morphology. Furthermore, the effect of uranium concentration on time to deposit a fixed amount was determined. Re-usability of the detectors was shown via demonstrating the anodic stripping of the deposit verified by surface analysis (SEM-EDS).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Results for the First Quarter Calendar Year 2018 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the First Quarter Calendar Year 2018 (CY18) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. The chemical and radionuclide contaminant results from the characterization of the First Quarter CY18 sampling of Tank 50 were requested by SRR personnel via a Task Technical Request (TTR) 2 and details of the testing are presented in the Savannah River National Laboratory (SRNL)Task Technical and Quality Assurance Plan (TT QAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRR request. 2 Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the TTQAP for the Tank 50 Saltstone task.

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

Results for the Second Quarter Calendar Year 2018 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the Second Quarter Calendar Year 2018 (CY18) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility, where the waste will be treated and disposed in the Saltstone Disposal Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗