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At least 415 records · Page 23

TRISO Spent Nuclear Fuel Recycling or Waste Reduction Using SRNL Vapor Digestion Technology – 25635

There is a renewed interest in advanced reactors, including high-temperature gas cooled reactors (HTGRs). Tri-structural isotropic (TRISO) fuel is being used in many HTGR designs, whether as SMRs or microreactors. However, TRISO-based HTGRs discharge the largest volume of used fuel per megawatt-hour of energy produced compared to other reactors. An order of magnitude reduction or more in the volume of SNF could be realized if the TRISO particles were separated from the graphite moderator and the carbon dispositioned as LLW. The Savannah River National Laboratory (SRNL) has a patented technology readiness level (TRL) 4/5 vapor digestion process for separating nuclear-grade graphite from HTGR SNF. The SRNL process is based on the reaction of NOx species with carbon to form CO2. Because NOx species are several orders of magnitude more reactive with graphite than oxygen, the process can operate at lower temperatures with uncrushed HTGR pebbles or prismatic blocks. Because the fuel elements do not need to be crushed and the graphite is digested using a vapor-based process, the potential for damaging the TRISO particles is much reduced. The DOE Office of Technology Transitions (OTT) is funding SRNL and the University of South Carolina at Columbia to close certain gaps that exist within the technology which impede its direct application to the processing of commercial TRISO-based SNF coming from HTGR advanced reactors.

Pierce, Robert [Savannah River National Laboratory↗

Boron nitride: Novel ceramic reductant for low‐activity waste vitrification

Abstract During vitrification of radioactive wastes, excessive foaming reduces processing rates within melters by hindering heat transfer from the molten glass pool to the reacting melter feed. Formulations of low‐activity waste (LAW) melter feeds, for vitrification at the Waste Treatment and Immobilization Plant at the Hanford Site, conventionally include the addition of sucrose to mitigate excessive foaming by hastening the denitration process. However, incomplete combustion of sucrose produces organics such as acetonitrile (C 2 H 3 N) that may exceed bounding limits of downstream effluent treatment facilities. Using boron nitride (BN) as an alternate reductant to sucrose, in a representative LAW melter feed reduced C 2 H 3 N production by 90% by preventing the low‐temperature sucrose–nitrate reactions. Furthermore, foaming was suppressed due to the higher decomposition temperature of BN than H 3 BO 3 meaning a delayed reaction of a large fraction of boron with the transient glass‐forming melt until above the foam onset temperature, thus reducing the quantity and viscosity of the connected melt and trapping less gas in the foam layer.

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Analysis of NuCycle® Process Waste Streams and Identification of Candidate Waste Forms (CRADA 719) Abstract

The objective of this project is to support collaboration between PNNL and Curio to address technical gaps in the immobilization of waste streams generated by the NuCycle® process, thereby improving the overall viability of the technology. This work will be carried out in two phases, described below. Phase 1 – Waste Management Study The first phase consists of a comprehensive waste management study focused on the waste streams produced during the NuCycle® process, including those containing long lived fission products (LLFPs, e.g., I 129) and transuranic elements (TRUs). This phase involves compiling waste compositions and characteristics based on recently completed NuCycle® testing and modeling efforts. Established waste forms and processing methods will be identified for streams where they are applicable.

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Preliminary Characterization and Evaluation on FFF Manufactured 316H and ODS Steels

The Advanced Materials and Manufacturing Technology (AMMT) program develops cross-cutting technologies in support of a broad range of nuclear reactor technologies and maintains U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of AMMT is to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy. Solid-state advanced manufacturing techniques can overcome some of the challenges in liquid-based additive manufacturing (AM) processes and should therefore be considered in material design and manufacturing as well. The work presented in this report forms part of a study on solid-state AM techniques of 316 stainless steels (SS) and oxide dispersion strengthened (ODS) steel components and supports the vision and goals of the AMMT program relevant to accelerate the development and deployment of advanced manufacturing processes. Achieving this can provide a safety improvement through larger safety margins, economic benefit for higher efficiency during operation, and a cost reduction through more effective manufacturing processes and less waste. This study provides preliminary information on development of the fused-filament fabrication (FFF) process using different powder types to demonstrate the sensitivity, and therefore the characterization of these sample components. The full solid-state manufacturing feasibility study will be completed and reported in a final feasibility evaluation during 2025. The study investigation used two 14YWT ODS powder batches, which provided information and the effect of different powder morphologies on manufacturability. The two 316SS powders demonstrated the effect of powder size on the manufacturability using FFF. Two product forms, namely a honeycomb structure and flat samples, were manufactured to demonstrate the flexibility of product form.

36 MATERIALS SCIENCE↗

Nuclear Decontamination Evolution and Revolution - 20349

Decontamination, one of the oldest practices in the nuclear industry, is about to change. While no single approach to decontamination is appropriate in all situations, decontamination of highly contaminated surfaces is historically viewed as difficult and/or marginally effective and fixatives and strippable coatings are often called on to manage time sensitive contamination control issues. The deferral of proper decontamination can lead to accumulated concentrations of radioactive material increasing the risk of cross-contamination of workers and equipment while also increasing radiation levels, which further discourages decontamination. The results of the subject testing demonstrate that superior decontamination factors are readily achievable and the ALARA benefit of decontamination is viable through advanced decontamination technologies in conjunction with innovative application techniques. The collateral benefits will save countless radiological man-hours and personnel radiological exposure. The appropriate use of radiological decontamination techniques can prevent or limit the adverse effects of highly radioactive contamination in the work area. High levels of radiological contamination are typically associated with some of the most physically demanding work in the nuclear industry. Additionally, working in highly contaminated environments increases the risk of exposure to elevated levels of airborne radioactivity and radiation from the source term of the contamination, particularly in hard to reach areas or complex equipment. Veolia's Alaron Nuclear Services (Alaron), a fully integrated nuclear facility, has provided the nuclear industry waste treatment, consolidation, repackaging, and broad decontamination services for almost 35 years. In looking at solutions for its customers, as well as for their own facility, Alaron has recently collaborated with Environmental Alternatives, Inc. (EAI), which has provided innovative solutions addressing difficult nuclear decontamination and industrial cleaning challenges since 1989. This broad range of experience along with the products to support the work created an ideal partner for Alaron's needs at their facility. The benefits of collaboration were immediately recognizable to the management of both companies although the decontamination challenge was significant. Due to the nature of the services offered, contamination at the Alaron facility is routine. The recent decontamination experience at Alaron utilized an innovative surfactant process that quickly produces remarkably higher decontamination factors for both removable surface contamination and fixed contamination. Considering the time to decontaminate areas historically designated for high radiological hazard work, the results from this decontamination effort not only reduced contamination levels but also significantly lowered exposure rates in the working environment. The relative ease with which the surfactant is applied, combined with extraordinary decontamination test results, indicates the potential to alter current radiological work processes in a way that improves worker comfort, and removes radiological engineering barriers, allowing Alaron to complete complex radiological tasks more efficiently and effectively. In a series of two simple applications of the decontamination agent to contaminated high bay vertical surfaces, as well as a variety of horizontal and vertical surfaces with variable porousness and surface sealants, there was an overall reduction in removable contamination of approximately 73% with a reduction in area dose of approximately 93%. The results indicate additional reduction in fixed contamination with application of the surfactant. These values are much higher than experienced with more traditional decontamination agents. In conjunction with demonstrating EAI's surfactant on room surfaces, additional studies were conducted on complex geometry equipment including waste processing equipment, tools, and heavily contaminated parts normally handled from Alaron's customers. Evaluations were made with straight application of EAI's surfactant as well as adding additional techniques to the treatment which afforded revolutionary improvements in the levels of contamination removed. Successful decontamination with up to 95% reduction in removable contamination with just one application were demonstrated in several of the trials. This paper outlines the planned approach to use this technology, the variety of surfaces and equipment treated, and the results of the decontamination efforts. Furthermore, this paper discusses options for pretreatment of certain heavily contaminated equipment prior to employees handling and/or working with the equipment. The potential for dose saving and the reduced risk of cross-contamination with the added benefit of lower PPE requirements, produces an enormous potential for cost and time savings. Attendees will benefit from Alaron's experiences and more fully understand the capabilities of EAI's decontamination process. With the variety of contamination agents, forms and surfaces on which the technology was demonstrated, the information will be valuable to a broad cross-section of industry users. The significance of this report demonstrates superior decontamination factors are achievable utilizing the next evolution of decontamination technology and that the ALARA benefits of decontamination are available through advanced and modern decontamination efforts. The collateral benefits will save countless radiological man-hours and valuable personnel radiological exposure. (authors)

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Thermal Properties of Sodium Borosilicate Glasses as a Function of Sulfur Content

SO3 additions, up to 3.0 wt%, were systematically investigated for effects on the physical properties of sodium borosilicate glass melted in air, with a sulfur-free composition of 50SiO2-10Al2O3-12B2O3-21Na2O-7CaO (wt%). Solubility measurements, using electron microscopy chemical analysis, determined the maximum loading to be ~1.5 wt% SO3. It was found that sulfur (here as sulfate) in this glass increased the glass transition temperature, thermal diffusivity, heat capacity, and thermal conductivity, and decreased the mass density. Structural analysis, performed with Raman spectroscopy, indicated that the borosilicate network polymerized with sulfur additions, presumably due to Na2O being required to charge compensate the ionic SO42- additions, thus becoming unavailable to form non-bridging oxygen in the silicate network. It is postulated that this increased crosslinking of the borosilicate backbone lead to a structure with higher dimensionality and average bond energy. This increased the mean free paths and vibration frequency of the phonons, which resulted in the observed increase in thermal properties.

Thermal properties, Borosilicate glass, Sulfur sol↗

April 2024 Semiannual Composite Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from the Salt Waste Processing Facility (SWPF) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00027, Revision 0, one SDF waste form (saltstone) sample was prepared in the Savannah River National Laboratory (SRNL) from the SWPF Decontaminated Salt Solution (DSS) Waste Acceptance Criteria (WAC) sample and Z-area premix material for the April 2024 semiannual Toxicity Characteristic Leaching Procedure (TCLP) sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of the SDF waste form was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The April 2024 semiannual Cement-Free grout sample met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and met the SPF WAC that was in effect at the time of the tank sampling.

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Identification of opportunities for integrating chemical processes for carbon (dioxide) utilization to nuclear power plants

The energy and industrial sectors are responsible for more than 75% of global carbon dioxide emissions, which are primary contributors to climate change. Decarbonization strategies to reduce overall carbon dioxide emissions are urgently needed. Among the numerous decarbonization strategies that are being defined and implemented, a particular strategy of interest for the present work is “the Re-X or zero waste” strategy, which involves recycling, reuse, repurposing, and remanufacturing of industrial emissions, by-products, and waste in general. A circular value chain or Circular Economy is embedded within this strategy. In a circular carbon economy, efficient reuse of emitted carbon dioxide is considered. However, the stability of carbon dioxide implies the requirement of significant amounts of energy for its transformation into value-added chemicals or products, which must be supplied by low carbon emitting energy sources. Nuclear power plants are low carbon energy sources that, additionally to electricity, could also supply heat and radiation for chemical transformations. This work identifies the opportunities and challenges for the development of integrated energy systems to upgrade and transform carbon dioxide, involving chemical and nuclear energy. The analysis encompasses possible use of the different forms of energy that can be obtain from nuclear reactors (i.e., radiation, electricity, and heat), by reviewing the published literature on potential routes for its conversion. Overall, our review indicates that an universal technology of the one-fits-all solution-type is an utopic dream. Instead, a suite of (contaminants tolerant) technologies for processing various concentrations CO 2 and producing a variety of products is currently needed.

08 HYDROGEN↗

Quantifying model prediction sensitivity to model-form uncertainty

Computational and mathematical models are essential to understanding complex systems and phenomena. However, when developing such models, limited knowledge and/or resources necessitates the use of simplifying assumptions. It is therefore crucial to quantify the impact of such simplifying assumptions on the reliability and accuracy of resulting model predictions. This work develops a first-of-its-kind approach to quantify the impact of physics modeling assumptions on predictions. Here, we leverage the emerging field of model-form uncertainty (MFU) representations, which are parameterized modifications to modeling assumptions, in combination with grouped Sobol’ indices to quantitatively measure an assumption’s importance. Specifically, we compute the grouped Sobol’ index for the MFU representation’s parameters as a single importance measure of the assumption for which the MFU representation characterizes uncertainty. To ensure this approach is robust to the subjective choice of how to parameterize a MFU representation, we establish bounds for the difference between sensitivity results for two different MFU representations based on differences in model prediction statistics. The capabilities associated with this approach are demonstrated on three exemplar problems: an upscaled subsurface contaminant transport problem, ablation modeling for hypersonic flight, and nuclear waste repository modeling. We found that our grouped approach is able to assess the impact of modeling assumptions on predictions and offers computational advantages over classical Sobol’ index computation while providing more interpretable results.

97 MATHEMATICS AND COMPUTING↗

Bounding Pressure and Flammability Evaluations for a Department of Energy Standard Canister Loaded with Aluminum-Clad Spent Fuel

This report presents bounding pressurization and flammability evaluations from the radiolytic gas generation expected during extended (>50 years) dry storage of aluminum-clad spent nuclear fuel (ASNF) elements in a sealed Department of Energy (DOE) Standard Canister. The primary questions involving extended ASNF dry storage center around the adequacy of dry storage conditioning processes (i.e., drying) and the behavior of residual hydrated aluminum oxides on the cladding—specifically, the radiolytic breakdown of chemically bound water in these corrosion products. The objectives of the presented work include providing a bounding assessment of the pressure with respect to the DOE Standard Canister’s structural integrity limits and identifying the potential for forming flammable or explosive gas mixtures (i.e., exceedance of the lower flammability limit of the molecular hydrogen [H2] and oxygen [O2] concentrations). The evaluation results confirm the findings of previous, more complex M&S work. That is, the structural integrity of the canister remains unchallenged by a wide margin. Nevertheless, it is important to recognize that the presented pressure calculations consider a full breakdown of the chemisorbed water, including a consequent release of all available H2. In reality, the breakdown of water in these systems will likely remain incomplete, due to competing chemical and radiolytic reactions, thereby attaining an equilibrium in the storage environment.

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Damaged Fuel in the United States - 20325

Throughout the history of commercial nuclear power plant operations in the U.S., many fuel assemblies have lost the capability to perform all of their desired functions. Since they can no longer be handled, stored, or transported in accordance with established regulations, they are classified as damaged fuel. The causes of these failed assemblies are diverse and plant-specific. The majority of these failed assemblies are contained in damaged fuel cans to be used in conjunction with storage and/or transportation systems. These systems have a limited number of slots that can be filled with damaged fuel cans. A damaged fuel can is generally a stainless-steel container that confines damaged spent nuclear fuel (SNF) and is closed at one end by mesh endpoints that allow gaseous and liquid media to escape but minimize the dispersal of gross particulate material. Out of extreme caution, a few reactors have loaded high burnup fuel into damaged fuel cans. Damaged fuel is not licensed for storage or transport in the U.S. because relevant regulations do not specify exactly how to classify damaged fuel. Instead, these regulations license/certify packages that specify approved contents. Damaged fuel must be included among the approved contents to be considered acceptable. In many cases, damaged SNF is encapsulated in damaged fuel cans to ensure it can confine gross fuel particles, debris, and/or damaged assemblies to known volumes within loaded casks. A damaged fuel can may then be utilized in the same way as an assembly in a storage and transportation system. Some storage cask systems utilize top and bottom plugs to confine debris in damaged fuel. The most recent domestic documentation on damaged SNF was published by the U.S. Energy Information Administration (EIA), which used data from U.S. reactors compiled from 1968 to June 30, 2013, to produce Form GC-859, 'Nuclear Fuel Data Survey.' According to this form, there were 136,821 boiling water reactor (BWR) SNF assemblies and 104,647 pressurized water reactor (PWR) SNF assemblies, for a combined total of 241,468. Of these, 4,521 were classified as failed. Some were also disassembled and the fuel rods or pieces of fuel rods combined to make consolidated assemblies. These consolidated assemblies may include damaged fuel or were perhaps consolidated as part of a demonstration project. The GC-859 data includes 2,550 consolidated assemblies containing 0 - 264 entire fuel rods. These consolidated assemblies could be placed in single assembly canisters and stored in the spent fuel pool. For dry storage and transportation, a single assembly canister is generally placed in each damaged fuel can. In addition to the consolidated assemblies, 2,391 un-canistered fuel rod pieces exist, which were removed from 494 assemblies. (authors)

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Tank 50 Simulant Grout and Toxicity Characteristic Leaching Procedure (TCLP) Results for Methyl Mercury Waste Acceptance Criteria (WAC) Limit

Testing of simulated Tank 50 salt solution grout, spiked with methyl mercury (MeHg) and ionic mercury, used the Toxicity Characteristic Leaching Procedure (TCLP) to evaluate leaching performance with respect to Hg and the Waste Acceptance Criteria (WAC) limits for the Saltstone Production Facility (SPF). This report describes the test parameters for simulant preparation and confirmatory analysis, grout preparation, TCLP leach testing and subsequent leachate analyses. These scoping tests were designed to provide an initial evaluation of the effect of increased MeHg and ionic Hg on the TCLP and were requested by Savannah River Remediation LLC (SRR) through a Technical Assistance Request (TAR). It is expected that this preliminary testing will likely lead to future, more detailed, testing involving the mercury (Hg) WAC and Hg limits with respect to the Tank 50 TCLP for mercury. The TCLP concentrations for total Hg have approached 0.06 mg/L in recent years for Tank 50 radioactive solutions containing up to 105 mg/L of Hg in the form of ionic (Hg(I/II), elemental (Hg(0)) and organo-mercury such as MeHg. In discussions with SRR personnel, it was decided for this non-baseline work to target Hg spike concentrations, using predominately MeHg, in the range of 150 to 350 mg/L total Hg to investigate corresponding Hg TCLP leaching behaviors.

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Effects of SRPPF ARS Effluent Variations on Magnesium Oxysulfate Solidification

A magnesium oxysulfate (MOS) grout formulation has been identified to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). The formulation uses reactive, light burnt MgO (28 wt%), anhydrous MgSO 4 (10 wt%), and dead burnt MgO (62 wt%), resulting in a grout with good mixability, acceptable density, acceptable setting times, and a leachate pH around 9.4, within the assumed Waste Isolation Pilot Plant (WIPP) brine pH range. A series of tests were then designed to observe how this formula reacted to variations in the liquid effluent. Liquid effluent temperature, the caustic pH level, concentration of NaNO 3 , and the presence of neutralization products, trace metals, and/or sodium sulfate (all potential residuals found in the effluent from upstream processing), were evaluated to see how each impacts the grout mixing time, peak temperature, setting time, presence of bleed water, leachate pH, and density. The standard mix has an average mixing time of 18 minutes, peak temperature of 100.5 °C, and set time of 60 minutes. Increasing the temperature of the liquid effluent increased the reaction rate of the mix and reduced mixing and setting times. Between 30 and 40°C the change was relatively small, raising no more than 6 °C compared to the standard mix maximum temperature. Increasing the pH of the liquid effluent had no significant impact on the mix. The addition of neutralization products and trace metals had an overall impact of decreasing the reactivity of the mix. The addition of more ions in the liquid effluent, such as an increase in the concentration of NaNO 3 and the addition of Na 2 SO 4 , generally decreased the reactivity of the mix. None of the variations to the liquid effluent hindered solidification as indicated by the lack of bleed water found on all the samples. The leachate pH values for all mixes tested did not significantly vary from the expected pH of 9.4 and none of the density values fell below 1.8 g/cm 3 . Overall, changes to the liquid effluent were found to have a minimal impact on the formulation suggesting the grout's ability to properly form despite increased temperatures and the presence of residuals typically found in the liquid effluent from the ARS.

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Thermodynamics of Minerals Stable near Earth's Surface

Project “Thermodynamics of Minerals Stable Near the Earth’s Surface” uses specialized calorimetric techniques to obtain new thermochemical data for solid phases of interest in the Earth’s “critical zone”, the near surface regions of the planet most directly affecting and affected by human activity. Major activity was in three areas: manganese oxide materials, aluminum clusters which are precursors to aluminum oxyhydroxide precipitates, and zeolites. All these classes of minerals are potential carriers of heavy metals and radionuclides in the environment, both within their crystal structures and as adsorbed layers or colloids. Their thermodynamic stability, as well as the sequence of stability of competing phases, governs the aqueous solubility of contaminants (and nutrients) in soil, groundwater, and sediments. The thermodynamics of these phases is strongly affected by particle size and crystallinity, since they often occur as very fine particles which do not coarsen in the near-surface environment. Many of these materials exhibit extensive polymorphism, multicomponent solid solution, and variable hydration states. Thus, one must define and systematize important thermodynamic variables, since one cannot make measurements on each and every possible phase, particle size, and composition. A major finding of the work on manganese oxides was labile water-induced oxidation-reduction reactions. These will be studied further by a combination of thermodynamic and synchrotron-based spectroscopic methods, the latter in collaboration with Peggy O’Day at UC Merced. The adsorption of other species, including simple organics, on such redox-active surfaces will also be investigated by a combination of thermochemical and spectroscopic studies. Studies of the energetics of water adsorption on mineral surfaces will be expanded to emphasize the adsorption of CO2 and small organic molecules, relevant to soil, groundwater, and the subsurface environment. Gas adsorption calorimetry, immersion calorimetry, and solution calorimetry will be combined with appropriate analytical and spectroscopic techniques to obtain a comprehensive picture of the interactions and their energetics. The evolution of aqueous aluminum clusters to form aluminum oxyhydroxide precipitates will be studied in detail by a combination of calorimetric and structural methods applied to aqueous samples aged at room temperature. The energetics of new classes of clusters will be explored. The project addresses several common themes relevant to the broad mission of the Department of Energy. Thermochemical data for minerals stable near the Earth’s surface, though still incomplete, are essential for understanding both natural and anthropogenic processes that affect our very existence on this planet. Such data contribute to the fundamental underpinning for mitigation and management strategies for the extraction and use of energy. Fundamental thermodynamic and mechanistic understanding is required to model processes ranging from heavy metal pollution to nuclear waste disposal to climate change.

58 GEOSCIENCES↗

Improving Efficiency of DWPF Operations via Automating Process Calculations and Vitrifying High-Curie Feed- 24529

The Liquid Waste Organization (LWO) at the Savannah River Site (SRS) uses a “Power As One®” motto to process and dispose of radioactive waste. The Defense Waste and Processing Facility (DWPF) treats the high-level waste through a process of vitrification. The DWPF receives three incoming waste streams that are added to the Sludge Receipt and Adjustment Tank (SRAT): sludge, Monosodium Titanate/Sludge Solids (MST/SS), and Strip Effluent (SE). The liquid waste is mixed with pre-fabricated frit and treated with high temperatures in the melter prior to being poured into stainless steel canisters. The liquid cools to form solid glass within these canisters that are suitable for long-term storage and disposal. Savannah River Mission Completion (SRMC), the SRS liquid waste contractor for the U.S. Department of Energy, has implemented several facility improvements to further enhance the operations to support the Liquid Waste Operation (LWO) mission. Two of these operational enhancements include implementing the electronic Material Tracking Program Calculator (eMTPC)software and increasing the DWPF canister heat rate limit.

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October 2024 Semiannual Composite Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from the Salt Waste Processing Facility (SWPF) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00027, Revision 0, one SDF waste form (saltstone) sample was prepared in the Savannah River National Laboratory (SRNL) from the SWPF Decontaminated Salt Solution (DSS) Waste Acceptance Criteria (WAC) sample and Z-area premix material for the October 2024 semiannual Toxicity Characteristic Leaching Procedure (TCLP) sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). At 64 days cured, a sample of the SDF waste form was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The October 2024 semiannual Cement-Free grout sample met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and met the SPF WAC that was in effect at the time of the sample collection at SWPF.

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Characterization of the Sulfur-Saturated Melt Versions of the HFG1 Study Glasses

The U.S. Department of Energy (DOE) is responsible for building the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is temporarily stored in 177 underground tanks. The Office of River Protection (ORP) has requested that the Savannah River National Laboratory (SRNL) contribute to current glass studies to support successful startup of the WTP, due to SRNL’s recognized capabilities and expertise for glass waste form development. As part of this effort, the Pacific Northwest National Laboratory (PNNL) is developing, batching, and fabricating simulated high-level waste (HLW) glasses to generate the associated property data needed to validate or identify areas of potential characterization improvements in the current glass property/ composition models. Currently, the models cover only a small fraction of the waste compositions projected in the Hanford tank farm; therefore, the models must be expanded to include high fluoride compositions to successfully complete the WTP mission. SRNL support of this work is defined in the Task Technical and Quality Assurance Plan (TTQAP). This report provides results from the chemical analyses of a series of sulfur-saturated melt (SSM) versions of simulated nuclear waste glasses fabricated at PNNL. The glasses were selected as part of a broader study of the influence of glass composition on chemical durability, sulfur retention, and other properties. The glasses were designated the High Fluoride Glasses-1 (HLG1).

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Actinide (III/IV) - Silica Colloids in the WIPP

During the completeness process for the 2019 Compliance Recertification Application (CRA) for the Waste Isolation Pilot Plant (WIPP), the US Environmental Protection Agency (EPA) raised concerns about the possible presence of actinide-silica colloids in the WIPP. Several publications that studied the An(IV)- silica system indicated that these colloids could form under conditions similar to WIPP. Colloids are important for WIPP Performance Assessment (PA) as these are a mobile fraction of actinides that contribute to release. As the previously published papers did not explore the formation of actinide-silica colloids under the high ionic strength environment expected in WIPP, an experimental strategy to study these systems was developed.

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