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At least 127 records · Page 7

Nuclear waste reduction: Exploring new pathways one step at a time

In my home country of Venezuela, nuclear energy is not a topic that attracts much attention. The government briefly oversaw some nuclear energy programs during the 1950s, but currently there are no active nuclear power facilities in the country. In fact, the Venezuelan government signed and ratified the treaty of the prohibition of nuclear weapons in 2021, which states that Venezuela has never owned, possessed or controlled nuclear weapons or programs of any kind. When I moved to the United States, however, nuclear energy became an extremely relevant topic. In the 1940s, the U.S. government established and oversaw the Manhattan Project to build atomic bombs for use in World War II. After the war, the government encouraged scientists to use this information on nuclear reactions to develop nuclear energy for peaceful civilian purposes instead.1 During these early days of nuclear research, there were no formal regulatory standards for nuclear waste management. Policies usually were self-regulated and often created based on existing policies of disposal for non-nuclear waste.2 As a result, there were instances of nuclear waste leaching into the environment and affecting local communities. So, much research has been conducted since then to characterize and store nuclear waste safely and securely.3 I first became interested in nuclear energy during my undergraduate studies when I worked on a project involving ligand synthesis to help extract actinides from nuclear waste. I then studied electrochemistry in molten salt systems for nuclear energy applications during my Ph.D. As I approached graduation, I started looking into national laboratories that have programs involving nuclear energy and waste management. At Idaho National Laboratory (INL), the focus is more on applied processes and how nuclear energy can be innovated to realize next-generation reactor design and technologies. This focus led me to apply for a Seaborg distinguished postdoctoral position at INL, for which I was chosen based on my proposal of a way to improve nuclear waste recycling. To understand my proposal, we must familiarize ourselves with the makeup of nuclear waste. After uranium dioxide is used as nuclear fuel in a reactor, the fuel matrix is then characterized by various fission products, including rare earth elements, alkali and alkaline earths, and actinides. Some of these fission products can potentially be recovered through pyroprocessing, 4 which involves the electrochemical dissolution of the used nuclear fuel in a molten chloride salt mixture at high temperatures. Though some of the fission products can be easily recovered—for example, uranium is reduced onto an inert cathode by applied potentials—numerous other fission products such as rare earth elements are difficult to recover due to their multivalent oxidation states and side reactions.5 To improve the recovery efficiency of rare earth elements specifically, I proposed investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form (Figure 1). The kinetic pathways and the chemical reactions of these elements, which will be elucidated through spectro-electrochemistry at high temperatures, will give insights on how the recovery efficiency can be improved. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Halogenation of used aluminum matrix test reactor fuel – a bench-scale demonstration with surrogate materials

In this work, experiments with surrogate materials were performed at bench scale to demonstrate a halogenation technique applicable to treatment of used aluminum matrix test reactor fuel. The technique involves dissolution and separation of aluminum from used aluminum matrix test reactor fuel in molten-halide salt systems prior to treatment and disposition of the fuel’s uranium and fission products. Demonstration of the halogenation technique was performed with neodymium metal as a non-radiological surrogate for uranium metal. Experiments involved blending forms of aluminum and neodymium metal with ammonium and lithium chloride or ammonium and lithium bromide, which upon heating decomposed into ammonia gas and the respective hydrogen chloride or bromide gas. The latter reacted with the metals to form the respective aluminum and neodymium halides. At elevated temperatures, aluminum halides gasified away from the respective neodymium halides, which fused with their respective lithium halides. Samples of fused and distillate salts were collected and analyzed, yielding extents of aluminum removal that ranged from 94.5–98.2% for chlorination runs and 91.4–97.8% for bromination runs. No neodymium was detected in the distillate fractions. Some experiments were repeated with excess reactants, and a portion of aluminum chloride distillate was processed into a consolidated waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Finding ways to reduce nuclear waste: searching for the unknown one step at a time

In my home country, Venezuela, research has been stagnant. Due to the political turmoil and the crisis, many educated people have left the country in search for a better life. This has caused a deficit in any technological and scientific advances, making Venezuela one of the first South American countries to have its rate of publications decline by 29% in 2013. Currently, Venezuela lacks the infrastructure and the means to keep up with the research progress as compared to other countries in South America, such as Brazil. Since coming to the United States (US), and currently working for a national laboratory, the active research environment endorses a wide range of careers and engineering programs that allow researchers to thrive at any given field. Researchers have access to funds and tools to succeed in developing materials for the future. There are 17 national laboratories in the US, and all of these have a different research focus/objective. As examples, Los Alamos National Laboratory and Sandia National Laboratory focus is on national homeland security, weapon science, radiation effects, among others. Argonne National Laboratory focuses on nuclear energy, energy storage, high performance computing, etc. At Idaho National Laboratory (INL) the research focuses on innovating nuclear energy and clean energy resources, critical infrastructure materials, along with fuel cycle solutions to manage, dispose and find ways to recycle current and future radiological waste. Compared to other national laboratories, INL focuses slightly more on applied processes and how nuclear energy can be innovated to next reactor design and technologies. The research being conducted at INL made me apply for a Seaborg distinguished postdoctoral position. For the position itself, the researcher must submit a proposal related to actinide chemistry on a research field area. In this position, 50% of my time will be focused on my own proposal. The proposal that I am working on is focused on the innovation of nuclear energy and fuel cycle recycling, which is why I was mainly interested on working at this national laboratory. To give a bit more context of what my proposal is about, a little bit of background is necessary: After the nuclear fuel (UO2) is used in a reactor, the fuel matrix is then characterized by various fission products (FP). Among these FP (including rare earth elements, alkali/alkaline earths, and actinides), many can potentially be recovered through nuclear reprocessing technologies. In pyroprocessing, the used nuclear fuel undergoes electrochemical dissolution into a molten chloride salt mixture in an electrorefiner. Initially, uranium is reduced onto an inert cathode by applied potentials. However, numerous remaining FPs accumulate in the melt and pose challenges for recovery by an inert electrode, particularly the rare earth elements (e.g., Nd, Gd, Pr, Sm) due to their multivalent oxidation states and tendencies toward side reactions, leading to their dissolution in the electrolyte. These recovery challenges result in inefficiencies and necessitate the continual discarding of the molten chloride salt, thereby generating additional waste. Furthermore, the presence of rare earth elements and other fission products in the molten salt electrolyte alters its physical and chemical properties, affecting both uranium recovery efficiency and the longevity of the molten chloride salt. To improve the recovery efficiency of the FP, specifically rare earth elements, I am investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form. The kinetic pathways and the chemical reactions of these elements will give insights on how the recovery efficiency can be improved. The interactions and speciation of these elements are being studied by spectro-electrochemistry at high temperature environments in quartz and other ceramic materials (e.g., alumina crucibles). Some of the challenges I am facing specifically relates the reactivity of some of these elements with different glass and crucible materials. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for an efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Thermal storage for electric utilities

Applications of the thermal energy storage (TES) principle (storage of sensible heat or latent heat, or heat storage in reversible chemical reactions) in power systems are evaluated. Load leveling behind the meter, load following at conventional thermal power plants, solar thermal power generation, and waste heat utilization are the principal TES applications considered. Specific TES examples discussed include: storage heaters for electric-resistance space heating, air conditioning TES in the form of chilled water or eutectic salt baths, hot water TES, and trans-seasonal storage in heated water in confined aquifers.

Swet, C. J.↗

Supercritical Water Oxidation (SCWO) for the treatment of spacecraft waste.

Supercritical water oxidation (SCWO) is a powerful waste treatment that has been considered as a substitute or complement to other potential spacecraft waste handling methods including jettisoning, compacted sealed storage, drying, and others. Combustible trash and garbage is mixed with water and then heated and pressurized above the critical point of water (705 °F and 218 atmospheres). The waste is nearly all oxidized to produce carbon dioxide, water and nitrogen. NOx is not formed. The combustion energy can be used to heat the input stream. SCWO is relatively complex, expensive, and hazardous compared to other methods. Oxidation of compounds containing chlorine and sulfur produces hydrochloric and sulfuric acid which can cause corrosion in the reactor. Neutralizing these acids produces salts which can precipitate and cause clogging. The major engineering challenges to implementing SCWO are chemical corrosion and fouling due to the deposition of salts. SCWO was first proposed for general waste treatment and for spacecraft in the 1980’s and extensive research has been carried out.

Harry W Jones↗

Implications of Non-Bingham Rheology

One of the key challenges now facing the US Department of Energy (DOE) is the fate of radioactive waste remaining from World War II and the Cold War which is stored underground in tanks some 75 feet in diameter and over 30 feet tall. Over time, the waste has segmented into multiple layers with sludges and slurries at the bottom with salt crust layers often at the top and liquid in between. DOE’s current official baseline plans call for remaining sludges and slurries to be removed from the tanks and converted into a more stable glass waste form. Minimizing worker exposure to radiation drives DOE to use slurry processing techniques to suspend, mobilize, transport, mix, and process the waste. Therefore, a clear and quantitative understanding of Hanford waste rheology is essential for the success of the DOE mission. Historically much of the waste has been characterized using Eugene Bingham’s model more than a century old that provides a straight line fit to higher shear rate data with the intercept suggesting a yield stress and the slope providing the consistency. Yet, Bingham fits overestimate the shear stress at a given shear rate for low to intermediate shear rates, exactly the range of shear rates typically encountered in pipe flow, where shear rates peak along the pipe wall and vanish in the center. This model produces a fictitious yield stress for some of the wastes that do not exhibit yield phenomena. While over estimating the yield stress may be prudent, safe, and conservative for some applications (e.g., pump sizing to ensure that pumps can handle yield stresses), over estimating the rheology may be inaccurate and non-conservative for other applications (e.g., eroding settled particle beds). Therefore, this paper evaluates the slurry rheology of Hanford and Savannah River wastes using a more modern rheological model that fits the full range of experimental data. Although a bias has been recognized and alternative models proposed, the magnitude of this bias and the implications for tank waste have only been qualitatively suggested. The purpose of this paper is to evaluate quantitatively implications of the poor quality of fit between a Bingham model for rheology and experimental data at modest shear rates. We first demonstrate the magnitude of the bias between the data and the Bingham extrapolation. We then evaluate quantitatively the velocity profile under laminar conditions. This analysis shows that the bias may be large (hundreds of percent or more) at modest shear rates and that modest shear rates dominate pipe velocity profiles.

Rheology, Radioactive waste↗

Effect of sucrose on the oxidation-reduction conditions and retention of rhenium during vitrification of low-activity waste

Technetium speciation and solubility during nuclear waste vitrification are affected by oxygen partial pressure in the melt. Thus, reducing agents such as sucrose, which are routinely added to nuclear waste feeds to repress foaming, can significantly affect the volatilization of Tc and increase its retention in glass. Using Re as a surrogate for Tc, we measured the effect of sucrose addition to two low-activity waste melter feeds on the Re retention, oxygen partial pressure, and iron redox as functions of temperature. As expected, Re retention in glass increased as the sucrose content increased. Furthermore, this is attributed to sucrose reaction with nitrates and nitrites in the feed. Destruction of nitrates and nitrites reduces the salt phase fraction and thus likely promotes Re diffusion into the glass-forming melt. In agreement with available literature, we discuss that a modest increase in the sucrose content (compared to current practice) can significantly improve the Tc retention without reducing Fe 2+ to metal.

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Insoluble Solids from Salt Dissolution: Characterization and Testing

Savannah River National Laboratory (SRNL) has further characterized insoluble solids that were observed in a variable depth sample from a salt dissolution campaign in Tank 9H. The insoluble solids were determined to be predominately gibbsite, a mineral form of aluminum hydroxide. From a review of salt dissolution testing and field experience, SRNL provided a realistic estimate of 8 vol% for solids of this type is formed per volume of saltcake dissolved. This estimate was doubled to 16 vol% to account for dissolution test uncertainty and differences between in-tank settling and laboratory testing. Savannah River Mission Completion (SRMC) is currently assessing the solids formed during salt dissolution as slurried sludge for hydrogen retention and release, which is driving flammability controls during salt dissolution activities. SRMC has requested SRNL perform a gas retention and release study to better understand the impact of the insoluble solids on waste tank flammability, and to provide a more accurate estimate of their ability to retain and release flammable gases.

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An Evaluation of the Impact of Glycolate and Glycolate Mitigation on the Defense Waste Processing Facility Recycle Diversion Project Flowsheet

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes radioactive High Level Waste (HLW) sludge solids from the Concentration, Storage, and Transfer Facilities (CSTF); which includes the SRS Tank Farm and Evaporator facilities, and a concentrated Cs-137 laden stream and a Monosodium Titanate (MST) and sludge solids stream from the SRS Salt Waste Processing Facility (SWPF). The waste is chemically adjusted with acids and reductant (currently with 50 wt.% nitric acid and ~90 wt.% formic acid, but eventually formic acid will be substituted with ~70 wt.% glycolic acid), and frit is added so that a durable, borosilicate glass waste form can be produced when the material is vitrified in the melter. As a result of the evaporation of water during both the melter feed preparation and the melter feed vitrification steps in DWPF, a recycle waste stream is generated, neutralized, and sent back to the CSTF. The recycle waste is a dilute aqueous stream originating from the collection of condensate liquids containing some minor sludge, MST, and frit solids and other waste components resulting from melter feed entrainment during foamover events and transfer of volatile species into the condensate. The recycle stream volume is significant and is expected to approach 3 million gallons per year once SWPF reaches full operation, requiring the use of multiple large CSTF tanks for storage. The recycle waste is currently collected in the SRS Tank Farm and periodically evaporated in the 242-16H (2H) Evaporator to conserve storage space.

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Chlorination of HALEU regulus casting dross

A scoping study was performed for chlorinating dross formed during uranium casting operations. Here, the purpose is to minimize the losses of uranium to dross wastes. The dross is primarily a mixture of uranium metal and uranium oxide with a minor fraction of crucible and crucible coating materials. The reaction chemistries were performed in a carrier salt of LiCl-KCl eutectic at 500 °C. The addition of FeCl 2 chlorinated the uranium metal to UCl 3 , by reducing the FeCl 2 to iron metal. After the uranium metal was chlorinated, zirconium metal was added to the salt. The residual FeCl 2 chlorinated the zirconium metal to ZrCl 4 , by reducing the FeCl 2 to iron metal. In turn, the ZrCl 4 chlorinated the uranium oxide to UCl 3 , by converting the ZrCl 4 to zirconium oxide. The effectiveness of the chlorination reactions was qualitatively verified by cyclic voltammograms that indicated the presence or absence of FeCl 2 and UCl 3 in the salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

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Evaluating Liquid Waste Transfers and their Impacts to the SRS Tank Farm to Support Operations and Closure

The Liquid Waste (LW) contractor at the Savannah River Site, Savannah River Mission Completion (SRMC), supports the storage, processing, and safe disposition of legacy, radioactive liquid waste. The LW Tank Farms contain approximately 127 million liters (33.5 million gallons) of liquid waste within 43 active, underground waste tanks. To meet mission critical milestones for the closure of waste tanks and processing of 34 million liters (9 million gallons) of salt waste per year by the LW Salt Waste Processing Facility (SWPF), an increase in Tank Farm operations, including waste tank transfers, is required. Waste is compiled in salt and sludge batches in the Tank Farms and transferred to SWPF and the Defense Waste Processing Facility (DWPF) for treatment. All waste tank transfers, such as waste removal and batch compilation transfers, must be pre-evaluated to ensure Documented Safety Analysis (DSA) requirements are met via Evaluated Transfer Approval Forms (ETAFs). Facility conditions and configurations may change as a result of a waste transfer. These changes must be reflected in the Tank Farms Emergency Response Datasheet (ERD), which contains data utilized for operation and emergency situations.

Peterson, Shelby R.↗

Evaluation of a Sulfate Solubility Limit Greater Than 0.65 Weight Percent in Sludge Batch 10 Glasses

Previous laboratory-scale crucible testing with batch chemicals confirmed that the sulfate (SO 4 2- ) limit for Sludge Batch 10 (SB10) was 0.65 weight percent (wt.%) in glass. This limit signifies that 0.65 wt.% SO 4 2- can be retained in the glass without the formation of a sulfate phase. The Defense Waste Processing Facility imposes this constraint in the Material Tracking Program. Based on preliminary calculations to support the Material Tracking Program, it was anticipated that transfer volumes of the monosodium titanate/sludge solids (MST/SS) stream from the Salt Waste Processing Facility (SWPF) may need to be reduced to maintain projected sulfate concentrations below 0.65 wt.% in glass. Savannah River Mission Completion requested that the Savannah River National Laboratory perform additional sulfate testing to determine whether a sulfate solubility limit greater than 0.65 wt.% is feasible for SB10, which could allow for higher transfer volumes of the MST/SS stream. This report documents the results of the testing at higher sulfate concentrations for the glass composition region defined by the most recent SB10 projection (November 2022) and Frits 473 and 625. Frit 473 was recommended for SB10 and Frit 625 was used during SB9 processing and the SB9 to SB10 transition. A total of twenty-one glass compositions were developed based on the expected compositional variables, which include sludge-only (SO) and coupled processing with the SWPF, waste loading (WL), and frit composition. The target sulfate concentrations were varied from 0.65-0.85 wt.% at 32 and 40% WL. Each glass was prepared from reagent grade chemicals and melted at 1150 °C. Visual observations were used to confirm the presence of a sulfate salt phase on the cooled glass surfaces. Representative samples of each glass were submitted for chemical composition analysis by inductively coupled plasma-optical emission spectroscopy and Cs analysis by inductively coupled plasma-mass spectrometry. Overall the majority of mean measured values are consistent with the target values for each major oxide of interest with less than 5% error. The percent errors for the measured SO 4 2- concentrations are generally less than 10%, which is comparable with previous sulfate solubility study measurements and acceptable. Only the SO glasses based on Frit 625 formed a sulfate phase at a 0.80 wt.% SO 4 2- target concentration at both 32 and 40% WL. The remainder of the glasses did not form a sulfate layer. Due to the formation of the sulfate phase, the limit is conservatively set at 0.70 wt.% based on the measured sulfate concentrations of 0.71 wt.% and 0.75 wt.% for these two glasses. None of the SO or coupled operation glasses based on Frit 473 formed a sulfate salt phase, which supports a sulfate limit of 0.80 wt.%. The following SO 4 2- concentration limits are recommended during SO and coupled SB10 processing: (1) 0.70 wt.% during processing with Frit 625, and (2) 0.80 wt.% during processing with Frit 473.

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Dehalogenation reactions between halide salts and phosphate compounds

Reactions between phosphoric acid [H 3 PO 4 ] or ammonium hydrogen phosphates [i.e., NH 4 H 2 PO 4 , (NH 4 ) 2 HPO 4 ] and halide salts can be used to dehalogenate (remove halides from) salt-based waste streams, where the process of removing halides yields products that have more efficient disposal pathways for repository storage. In this context, the term efficiency is defined as higher waste loadings and simplified immobilization processes with potential for recycle of certain salt components (e.g., 37 Cl as H 37 Cl or NH 4 37 Cl). The main streams identified for these processes are nuclear wastes generated during electrochemical reprocessing of used nuclear fuel as well as used halide salts from molten salt reactor operation. The potential byproducts of these reactions are fairly consistent across the range of halide species (i.e., F, Cl, Br, I) where the most common are hydrogen halides [e.g., HCl (g) ] or ammonium halides (e.g., NH 4 Cl). However, trihalide compounds (e.g., NCl 3 ), nitrogen triiodide ammine adducts [NI 3 ·(NH 3 ) x ], and ammonium triiodide (NH 4 I 3 ) are also possible. Several of these byproducts (i.e., NCl 3 , NBr 3 , NI 3 , and NH 4 I 3 ) are shock-sensitive contact explosives so their production in these processes must be tracked and carefully controlled, which includes methods of immediate neutralization upon production such as direct transport to a caustic scrubber for dissolution. Several benefits arise from utilizing H 3 PO 4 as the phosphate additive during dehalogenation reactions for making iron phosphate waste forms including more oxidized iron (higher Fe 3+ :Fe 2+ ratios), higher chemical durabilities, and the avoidance of trihalides, but the byproducts are hydrogen halides, which are corrosive and require special handling.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Innovative mercury treatment technology options for the liquid waste system at the Savannah River Site: scoping studies

The Savannah River Site (SRS) Liquid Waste System (LWS) contains liquids, salts and sludges that are currently being processed into final wasteforms for disposition, specifically, waste glass for sludges, solids and liquids containing high levels of radioactivity, and saltstone grout for low activity decontaminated liquid solutions. The LWS also contains approximately 60,000 kg of mercury present in the following physical and chemical forms, specifically: a) ionic inorganic mercury, organomercury (e.g., methylmercury), and other minor components found in LWS fluids b) mercury solids such as oxides, hydroxides, amalgams sulfides and sorbed mercury, c) accumulations of dense liquid elemental mercury, and d) vapor phase elemental and organomercury mercury found in tank headspace gas and in evaporators. An effective and proactive management strategy of the mercury present in the LWS is needed to support processing of LWS wastes into glass and saltstone. Sustainable processing of the LAWS to completion requires mercury removal from the LWS at a rate of approximately 2,900 kg/yr. This removal can be accomplished through existing mercury treatment systems or newly implemented LWS “purge points”. The chemical speciation of mercury has emerged as the key factor that controls mercury behavior in the LWS. For example, past studies demonstrated that mercury speciation is adversely impacting the performance of existing removal systems so that significant levels of mercury are recycled from the Defense Waste Processing Facility (DWPF) back to the tank farm. Consequently, mercury concentrations have slowly increased in the LWS tank fluids over time. The presence of organo-mercury has also been identified as the cause of increased mercury leaching from saltstone. In response to these challenges, the Department of Energy (DOE) Office of Environmental Management (EM-TD) Technology Development Program has supported a series of scoping studies predicated on manipulating or controlling mercury speciation and mercury behaviors within the constraints of LWS waste chemistry and safety conditions. The intent of these studies was to rapidly triage potential technology options and develop a technically based go / no go recommendation for further work. This composite report presents the results of three scoping studies: 1) advanced photooxidation processes, 2) chemical reduction, and 3) mercury getters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The S solubility of Cr and Al containing simulated low-activity waste glass

Here, the presence of sulfur-containing molten salt phases during vitrification is a particular problem which can be detrimental to the melter. Borosilicate compositions were designed to explore both the effects of a fixed amount of Cr, as Cr 2 O 3 or Na 2 CrO 4 , and the impact of Al 2 O 3 , on sulfur solubility as determined by progressively adding elemental S. The NABS series (with Al 2 O 3 ) has a higher SO 3 solubility of ~3.5 mol% while the NBS series (without Al 2 O 3 ) has a SO 3 solubility of ~2 mol%. Crystalline Cr 2 O 3 formed in the NABS series upon the addition of S but is not present in the NBS series. At the highest S addition, a separate salt phase always forms, and was identified as Na 2 (S,Cr)O 4 via electron dispersive spectroscopy, ultraviolet-visible spectroscopy, x-ray diffraction, and thermal analysis. Magic angle spinning nuclear magnetic resonance identified no change in the B coordination in either series as S is added, suggesting that the Na is scavenged from other locations, such as the Si sites, to charge compensate sulfate, whether in the glass or in the salt. The Cr precursor identity did not affect the observed overall behavior regarding sulfur, but did impact Cr redox as identified with Raman spectroscopy.

36 MATERIALS SCIENCE↗

Fiscal Year 2020 Filtration of Hanford Tank Waste 241-AP-105

Bench-scale filtration testing of 15 liters of 241-AP-105 supernatant was conducted using a Mott inline filter Model 6610 (media grade 5) backpulsed dead-end filter (BDEF) in the hot cells of the Radiochemical Processing Laboratory at Pacific Northwest National Laboratory. The as-received samples were diluted with raw water from the Hanford Site to assess the propensity for solids to form upon dilution and to assess the impact those solids have on filtration performance. The feed was split into two batches. The first batch was settled and decanted as planned in Tank Side Cesium Removal operations. The BDEF was used to filter this AP-105 feed at a targeted flux of 0.065 gpm/ft 2 and exhibited no measurable pressure increase during filtration. The second batch was fed to the BDEF with solids resuspended. This was also operated at the targeted flux of 0.065 gpm/ft 2 . In this case, the target pressure differential of 2 psid (resistance of 6.24 × 10 10 m -1 ) was reached in slightly more than 12 h. A backpulse restored the filter flux; however, the time to the next backpulse was slightly shorter. These tests indicate that after the addition of raw water, settle and decant effectively removed solids from the feed. It is not possible to assess the time required for these solids to settle based on the result from these tests. However, if sufficient time is allowed for the solids to settle, no measurable filter fouling was observed in current testing. In contrast, if insufficient time is allowed for solids to settle, filter fouling will occur relatively quickly and will likely require backpulses more frequently than every 12 hours (1.91 m 3 /m 2 volume filtered). Solids concentrated from backpulse solutions displayed large variability in composition with Al-silicates and mixed chromium-aluminum oxide phases. Amounts of Mg, Fe, and Ca were also found in the presence of Na and O. These solids could be grouped in three categories: (1) soluble Na-salts that formed large block-like particles but were made up of fibrous materials, (2) amorphous Al-phases, and (3) colloidal iron oxides with an average particle diameter of 4.2 µm.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Brine Availability Test in Salt, a Heated Borehole Experiment at the Waste Isolation Pilot Plant, New Mexico, USA - 20233

The Brine Availability Tests in Salt is a US Department of Energy Office of Nuclear Energy supported activity as part of the Spent Fuel and Waste Disposition Campaign where we are performing simultaneous heated and control experiments in adjacent horizontal boreholes in the WIPP Salt Disposal Investigation drifts as part of experimental Phase 1. Phase 1a of BATS, a low-cost, small spatial-scale experiment performed in order to design instrumentation and methods for further experimentation in the URL, has been completed (June 2018 - May 2019) and is described elsewhere. In coincidence with multiple physical parameters that will be measured over the duration of the experiment, we will be measuring the liquid and vapor phases of H{sub 2}O as well as its isotopic content (not measured in Phase 1a). Three different forms of water occur in natural salts: as fluid in macro scale (mm size) inclusions, as intergranular free H{sub 2}O and as H{sub 2}O incorporated in hydrous minerals. Each of these forms has potentially different stable isotope signatures and will be influenced differentially by heat induced transport and by interaction with introduced tracers. Continuous measurement of vapor phase H{sub 2}O isotopes during borehole heating will be impacted by several effects: 1) evaporation as liquid brine flows into the borehole and evaporates completely or partially into the dry N{sub 2} stream; 2) steam from dehydration of hydrous minerals, that may enter the borehole; and 3) hydrous minerals that may form in the borehole (preferentially incorporating heavier isotopes). The continuous time series collected over the course of the heating test will be combined with analyses of pre-test and post-test collected samples of test site salt to investigate the contributions of each of the three water types as H{sub 2}O is mobilized due to the effects of heating. (authors)

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