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DOE OSTI2023

Characterization of the Soluble and Insoluble Portions of Solids from the Salt Waste Processing Facility Tank 201 and Tank 202 Samples

The Salt Waste Processing Facility (SWPF) Tank 201 and Tank 202 samples have been characterized using X-ray diffraction (XRD), scanning electron microscope/energy dispersive x-ray (SEM/EDX), particle size analysis (PSA-Microtrac), Fourier transform-infrared (FT-IR) and nuclear magnetic resonance ( 1 H NMR) spectroscopic techniques to characterize and identify organic and inorganic components in Tanks 201 and 202 as part of efforts to understand the problems occurring at SWPF due to the presence of solids in the recovered solvent.

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DOE OSTI2022

INL Site ARG-US Implementation – FY22 Activities and FY23 Plans

The Packaging Certification and Life Cycle Management program at Argonne National Laboratory (ANL) developed a suite of monitoring systems collectively referred to here as ARG-US. ARG-US provides necessary data for nuclear facility and system operation and maintenance, and has been previously demonstrated in hot cells, radioactive material (RAM) storage areas, and RAM shipment trucks. It has also been installed directly on RAM storage and shipment containers. ARG-US offers some unique advantages over other commercially available systems by using wireless data connections, battery power supplies, and customizable monitoring methods. Idaho National Laboratory (INL) has been tasked with investigating applications for ARG-US at INL site facilities, which are operated by several different contractors. The initial investigation scope centered on CPP-603 Irradiated Fuel Storage Facility in relation to the upcoming Department of Energy (DOE) Spent Nuclear Fuel (SNF) Packaging Demonstration. The investigation has been led by the Used Fuel Management Department in INL’s Nuclear Science and Technology (NS&T) directorate. This investigation is expected to recommend INL site processes or facilities in which to implement ARG-US systems. INL has engaged local site technical and oversight representatives, security, nuclear safeguards, and program management personnel to identify good candidates for ARG-US test implementations. INL has identified the following high-level goals for any INL site implementation of ARG-US: provide unique testing environments, prompt development of new monitoring methods/techniques for the ARG-US suite, and acquire useful monitoring for the user facility. As a result, INL suggests three program areas for further investigation: CPP-603 Fuel Handling Cave (FHC), legacy mixed waste storage systems at Idaho Nuclear Technology & Engineering Center (INTEC) known as the “Tank Farm,” and periodic and emergency environmental monitoring. This progress report relates activities undertaken in this investigation, describes the preliminary areas of interest for limited scope ARG-US testing or implementation, and relays expected actions for completing the task scope. This report offers an opportunity to the program sponsor, technical leads at ANL, and INL site representatives to give feedback on the initial assessment and make recommendations on the forthcoming activities.

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DOE OSTI2023

Modeling of Chemical Slurry Rheology in DWPF Sludge Batch (SB) 10 Simulants

The Defense Waste Processing Facility (DWPF) treats high-activity radionuclides from sludge through a process called vitrification. This process converts radioactive liquid waste currently stored in tank farms into a solid glass form that is suitable for long-term storage and disposal. Due to the complexities involved in vitrifying this waste within each operation of the Chemical Processing Cell (CPC), waste rheology is studied to characterize the fluid-mechanical properties as it passes through the CPC and into the Melter. To better understand the waste and validate flow behavior, slurry rheology of simulants that represents the waste was studied at various acid stoichiometry percentages and solids concentrations to determine the simulant’s yield stress and viscosity. This research work has been supported by the DOE-FIU Science & Technology Workforce Development Initiative, an innovative program developed by the U.S. Department of Energy’s Office of Environmental Management (DOE-EM) and Florida International University’s Applied Research Center (FIU-ARC). During the spring of 2022, a DOE Fellow intern, Brendon Cintas, spent 10 weeks doing a summer internship at Savannah River National Laboratory (SRS) under the supervision and guidance of Dan Lambert, Chemical Flowsheet Development. The intern’s project was initiated on June 6, 2022, and continued through August 11, 2022 with the objective of assisting scientists at SRNL’s Rheology and Grout Laboratory at Aiken Country Technology Lab (ACTL) better understand the sludge composition on the rheology of a simulant slurry using a HAAKE RheoStress 6000 rheometer and extrapolate the results to the real-waste data.

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DOE OSTI2023

Refinement of Pitting Factor Basis to Support the Corrosion Control Program (Interim Report)

At Savannah River Site (SRS), High-Level Waste is stored in below-grade tanks constructed of carbon steel. This waste is composed of sludge, salt cake, and/or supernate. In part, preparation of this waste for future processing involves dissolution of the salt cake layer. The salt dissolution process can create conditions that leave the carbon steel tanks susceptible to a number of corrosion processes. The salt to be dissolved contains high concentrations of nitrate, that once released, create an environment that may be conducive to pitting corrosion or stress corrosion cracking (SCC) of carbon steel. Additionally, during the salt dissolution process, in the absence of mixing, stratification of the supernatant liquid may occur. This can result in less dense, more dilute waste layers occurring higher in the tank. In these more dilute waste layers, the susceptibility to localized corrosion could potentially differ from that of the more concentrated salt solutions evaluated in previous testing, as the amount of inhibiting and aggressive species, not just the ratios, can affect susceptibility. Evaluation of the susceptibility to localized corrosion in these more dilute waste chemistries could provide insight into the amount of inhibitors required to effectively mitigate pitting corrosion in stagnant salt dissolution environments, as well as other tank farm operations involving dilute waste streams.

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DOE OSTI2022

Characterization of Contaminants and Mobility in WMA A-AX Boreholes D0006 and D0008

This study was initiated to determine if subsurface contaminants in the A-AX vadose zone at the Hanford site (boreholes D0006 and D0008) caused corrosion of nearby stainless steel well casings (299-E-24-19, 299-E25-46, and 299-E25-236) and, based on the type and distribution of contaminants, determine if contaminants are from one or more sources including a) the A-104 and/or A-105 tanks, b) the 242A evaporator, and c) the 200-E-286 ditch. A tiered approach was used for characterizing contaminants in vadose zone sediments starting with mobile and total contaminant concentrations (Tier 1), characterizing contaminant speciation (Tier 2), and measuring contaminant leaching (Tier 3). Total and mobile contaminant measurements at seven depths showed elevated sulfate and nitrate in all samples (130 to 280 ft depth), elevated chloride in five depths, and no elevated radionuclides except I-129 in D0006 at 206 to 208 ft depth. If these contaminants were initially acidic, stainless-steel corrosion would occur. However, because sediments likely have a high acid neutralization capacity, an acidic spill would have been neutralized within tens of feet, so casing corrosion would likely occur in shallow sediments. GEL labs data also shows low (but above natural) chromate from shallow to 170 ft depth, then higher chromate to 280 ft, suggesting a surface Cr spill or casing corrosion. GEL labs data also showed evidence of more than one plume, with deep nitrate, sulfate, and chloride migration, but only shallow tritium migration. Tritium should migrate nearly unretarded in the subsurface like nitrate and chloride. Elevated aqueous cations and anions in the D0006 206 to 208' depth pore water (SO 4 2- , NO 3 - , Na + , K + , Ca 2+ , Mg 2+ , Si) could be the result of acid disposal and calcite dissolution at shallow depth, with deeper migration of SO 4 2- , NO 3 - from acids and Ca 2+ and Mg 2+ from calcite. Elevated Si (160 mg/L) indicates dissolution of clays or other silicates, which can also occur in acids. Acid extractable metals did show elevated Cr (but not Fe, Ni, Mo, and Mn) above natural levels, which may indicate stainless-steel well corrosion or transport of metals from a shallow spill. The elevated I-129 in D0006 at 206 to 208' depth was low with 0.34 μg/g in pore water and 1.5 μg/g total extractable I-129 in the sediment, which indicates significant fraction of I-129 was bound in one or more precipitates. Although total I-129 was below detection limits in most extractions due to high molybdate interference, I-127 analysis provided some insight into potential precipitate phases. Total I-127 was mainly extracted by acetate or acetic acid (67%), which may be iodine in calcite. Inorganic carbon measurement identified 1.65% calcite, and total organic carbon (TOC) was below detection limits, so iodine was not associated with organic matter. Nearly all the I-127 was present as iodide. In 1-D leach experiments, I-127 as iodide leached from the sediment within 2 pore volumes. Due to high and dynamic molybdate leaching from the sediment and interference with I-129 analysis, there were few quantifiable I-129 measurements. The calculated molybdate release rates from the sediment suggested Mo was being released from the same precipitate phase, such as molybdate incorporated into calcite. Overall, because there was little radioactive contamination measured in D0006 and D0008 boreholes, potential releases from A-104 and A-105 did not reach this location south of the A tank farm. Elevated cations and anions found in D0006 and D0008 at depth as the result of acidic disposal could cause shallow stainless steel casing corrosion. The nearby 242A evaporator generally contained alkaline waste and tritium, but the nearby 200-E-286 ditch generally contained high Cl - liquid effluent, so these are unlikely to be acidic sources.

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DOE OSTI2024

Application of Process Chemical Modeling to Optimize Radioactive Waste Disposal at the Savannah River Site - 24242

The Technical Optimization Model (TOM) is used by Savannah River Mission Completion (SRMC) to carry out facility-wide material balance and validate chemistrydependent processes for the purpose of their System Plan.The TOM simulates material movement and chemical reactions at the Tank Farm (TF), Salt Waste Processing Facility (SWPF) and Defense Waste Processing Facility (DWPF).

Georgiou, Andreas↗
DOE OSTI2024

Recommendations for Updating Liquid Discharged Inventory and Transport Modeling Parameters for Cumulative Impacts Evaluation of Hexavalent Chromium in the 200 West Area

The purpose of this environmental calculation file (ECF) is to document information regarding hexavalent chromium (Cr(VI)) inventory discharged in 200 West Area at the Hanford Site and provide data to support predictive transport through the vadose zone and saturated zone for modeling efforts. This document provides a focused evaluation of historical waste stream data and studies to develop estimates of Cr(VI) inventory, discharge fractions, and transport parameters for the 200 West Area waste sites and tank farms during discharge events and for long-term contaminant releases.

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DOE OSTI2025

Creating a Salt Batch: Challenges and Lessons Learned

CONCLUSION • The overall Salt Batch system has been honed through each iteration of the process • Increased communication across facilities and work groups has enabled the salt team to react swiftly to emergent items and identify improvement opportunities • Even the best plans require backup options – Emergent items can appear at any time during the process – Flexibility to adapt to the unexpected is a must-have team mindset • To date, every salt batch compiled has been qualified and approved as feed for SWPF – No feed break has occurred between the Tank Farm and SWPF

Goldblatt, Joshua B. [Savannah River National Labo↗
DOE OSTI2025

In-Tank Processing Using Low Temperature Aluminum Dissolution: Year 1 Progress Report

The Hanford Site stores approximately 56 million gallons of radioactive legacy defense waste in underground tanks. This study investigates the feasibility of utilizing in-tank low-temperature aluminum dissolution to favorably alter the transport properties of aluminum-rich southeast quadrant waste sludges, thereby de-risking sludge delivery for Direct Feed of High-Level Waste. Specifically, this study aims to assess the rate and extent of gibbsite dissolution under various conditions, including different particle sizes (~10 to 90 µm), NaOH concentrations (1 to 6 M), and the presence of specific analytes such as NO 2 - , NO 3 - , and other aluminum-rich waste background analytes, using simulated waste solids and liquids. Experimental tests were designed to highlight the impact of aluminum leaching on simulated waste transport properties, quantified by the just-suspended mixing speed (NJS), and were conducted under well-mixed, turbulent conditions, ensuring full suspension of gibbsite particles using an overhead mixer. The results suggest that the presence of NO 2 - and NO 3 - , in conjunction with leaching at relatively low free hydroxide contents, improved transport requirements (i.e., lowered NJS) for pure gibbsite slurries. These findings will inform the design of potential in-tank processing systems, optimizing aluminum dissolution for improved waste management in the Hanford Tank Farms.

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DOE OSTI2022

WASTE TREATMENT PLANT (WTP) LIQUID EFFLUENT TREATABILITY EVALUATION (SEPTEMBER 2004)

A forecast of the radioactive, dangerous liquid effluents expected to be produced by the Waste Treatment Plant (WTP) was provided by Bechtel National, Inc. (BNI 2004). The forecast represents the liquid effluents generated from the processing of Tank Farm waste through the end-of-mission for the WTP. The WTP forecast is provided in the Appendices. The WTP liquid effluents will be stored, treated, and disposed of in the Liquid Effluent Retention Facility (LERF) and the Effluent Treatment Facility (ETF). Both facilities are located in the 200 East Area and are operated by Fluor Hanford, Inc. (FH) for the US. Department of Energy (DOE). The treatability of the WTP liquid effluents in the LERF/ETF was evaluated. The evaluation was conducted by comparing the forecast to the LERF/ETF treatability envelope (Aromi 1997), which provides information on the items which determine if a liquid effluent is acceptable for receipt and treatment at the LERF/ETF. The format of the evaluation corresponds directly to the outline of the treatability envelope document. Except where noted, the maximum annual average concentrations over the range of the 27 year forecast was evaluated against the treatability envelope. This is an acceptable approach because the volume capacity in the LERF Basin will equalize the minimum and maximum peaks. Background information on the LERF/ETF design basis is provided in the treatability envelope document.

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DOE OSTI2020

Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

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DOE OSTI2020

Issue Resolution During the Development of the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20125

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high-activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance (i.e., 100 m from the SDUs). In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. A number of issues were overcome during the development of these simulations. These issues were identified as part of internal technical reviews. Simulations are developed by people and people make mistakes, so the internal technical review process is a vital step in PA development. Specific examples of resolved issues include a unit-conversion error, an inappropriate definition for a model boundary condition, a model time-stepping issue, and an error in the calculation for the buildup of contaminants in soil. Actions taken to address these issues resulted in an improved product with a better supported technical basis and more defensible results. The identification and correction of these issues are discussed. By understanding these issues, model developers and technical reviewers working on PAs in the future may avoid repeating these types of mistakes. Transparency with respect to these mistakes builds trust between waste management sites, regulators, and stakeholders. (authors)

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DOE OSTI2020

Updates to a Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 20420

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The U.S. Department of Energy (DOE) was required to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980 including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the West Valley Site that evaluates potential future consequences for human health and the environment. This forms the core of the spatial domain of the West Valley PPA Model. The PPA Model, developed using the GoldSim system modeling software, is a tool intended to provide support for decision making that evaluates uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-grouped contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of high level radioactive waste resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The PPA Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife could be exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the PPA Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. PPA Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes. With this information, the decision makers can drive informed, defensible decisions regarding decommissioning of the Site. This paper includes an update of the information presented at WM2019 [1]. (authors)

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DOE OSTI2024

Effect of Drying on Corrosion Mitigation of Hanford Transfer Lines

Radioactive waste is stored in underground, carbon-steel double-shell tanks at the Department of Energy Hanford site. The waste is transferred between the tanks and other assets using the transfer lines spanning throughout the various tank farms at Hanford. The transfer lines consist of a pipe-in-pipe design, small diameter pipes, and are not piggable. Recent inspection data of the transfer lines have shown areas with corrosion on both interior of the encasements and exterior of the primary pipes, with nearly 50 percent wall loss on the primary pipes and nearly 25% wall loss on the encasement pipes due to pitting corrosion. The visual inspections of the transfer lines have shown presence of corrosion products near the pipeline risers and beyond. It has been hypothesized that the corrosion is predominantly due to the high humidity conditions and in some cases is driven by the presence of residual hydrotest water in the encasement and the associated contact with the safety significant primary pipe. Therefore, drying of the transfer lines could lead to corrosion mitigation. Experimental studies are being conducted to understand the effect of environmental conditions, especially, relative humidity and temperature, on transfer line grade carbon steel corrosion and on mitigating corrosion. The experimental conditions are selected based on the seasonal temperature changes, and relative humidity conditions ranging from 30 to 100 percent. The experimental data will be used as guidance for maintaining a dry environment that will help mitigate the transfer-line corrosion caused by the high humidity conditions.

Shukla, Pavan K.↗
DOE OSTI2024

Application of Process Chemical Modeling to Optimize Radioactive Waste Disposal at the Savannah River Site – 24242

The Liquid Waste Program (LWP) managed by Savannah River Mission Completion (SRMC) is responsible for the treatment and disposal of waste at the Savannah River Site (SRS). Radioactive waste at SRS is stored and processed at four key facilities – each with their respective functions to store, blend, grout, or vitrify waste. The tank farm, where waste is stored, consists primarily of legacy waste with new material incoming from the Accelerated Basin De-inventory program (ABD), which is managed by Savannah River Nuclear Solutions (SRNS). System planning is done by SRNS and SRMC to optimize ABD and LWP operations, respectively.

Georgiou, Andreas↗
DOE OSTI2025

Determining Drying Conditions to Mitigate Hanford Transfer Line Corrosion

Radioactive waste is stored in underground, carbon-steel double-shell tanks at the Department of Energy Hanford site.1,2,3 The underground transfer lines are used to transfer the waste between the tanks and other assets at the Hanford tank farm facilities. The transfer lines constructed before 1990 are pipe-in-pipe design with most of the lines having carbon steel carrier pipe and carbon steel encasements. The lines constructed after 1990 have a stainless-steel carrier with a carbon steel encasement. The carrier lines, which are in contact with the waste, provide a critical confinement function, while the encasements provide secondary containment to reduce the risk of a release and contamination to the environment. In 2022, a visual inspection of the annulus between the encasement and the carrier for one of the pre-1990 transfer lines revealed standing liquid in contact with the carrier pipe. Additionally, a significant build-up of corrosion products on the exterior of the carrier pipe and the interior of the encasement pipe was observed. To assess the extent of condition, borescope inspections were performed in the encasement space of additional transfer lines in the facility to provide a baseline. The inspection revealed that of the lines inspected, 29 had either moisture present or evidence that moisture had been present in the past (e.g., waterline corrosion where the pipe had been exposed to standing water). Additional inspections indicated several instances of pitting corrosion on the interior of the encasement and exterior the carrier transfer lines. An image of the carrier pipe with several pits is presented in Figure 1(a) and the profiled image of the transfer line section along with the pit depths are presented Figure 1(b). As seen in the figure, the deepest pit was 73 mils on the pipe section. These lines were designated as “do-not-use” until further compensatory actions are implemented to return the lines to service.

Shukla, Pavan K. [Savannah River National Laborato↗
DOE OSTI2025

Development of a Robust Reference Electrode in Aggressive Chemical and Radiation Environments in the Hanford Waste Tanks

The Hanford site stores more than 200 million liters of radioactive and chemically hazardous wastes from the production of weapons materials. The wastes are stored in 177 underground carbon-steel storage tanks, separated between 149 single shell tanks (SSTs) and 28 double shell tanks (DSTs). The DSTs provide critical retrieval and interim storage before the waste is vitrified in the Waste Treatment and Isolation Plant (WTP). The tanks will need to remain in-service far beyond the initial 40-year design life, and effective corrosion control practices must remain in force to extend the tanks’ lifespans. This effort includes direct measurements of corrosion rate (e.g., ultrasonic measurements and corrosion coupons) and electrochemical processes (e.g., linear polarization measurements and open circuit potential measurements). The Hanford site began monitoring the corrosion potential in select DSTs in 2008. Of the 45 reference electrodes that have been installed, 29 have failed and 6 others provided unreliable results. DOE-EM is supporting a 3-year program to develop a chemical and radiation resistant reference electrode for application in the Hanford tanks. The first year of the program focused on understanding the failure mechanism for the reference electrodes and identification of candidate construction materials that would mitigate degradation of the electrodes in the waste environment. During the second year of the program, the objectives were to: 1) test candidate materials under simulated waste conditions, 2) design components that will extend the service life of the electrode, 3) fabricate materials for prototype reference electrodes, and 4) assemble prototype reference electrodes for accelerated testing. The reference electrode is constructed of four principal parts: 1) junction, 2) casing, 3) inner chamber backfill materials, and 4) the sensing wire. Principally, improvements of the junction, casing, and inner chamber backfill materials are being pursued. The junction material at the interface between the waste and the inner chamber of the reference electrode was identified as a critical component in the failure of the reference electrodes. Nine candidate replacement junction materials were tested under simulated waste conditions to evaluate permeation rate. These materials included a variety of polymeric and ceramic materials, some of which were 3-D printed. Thus far, porous polyvinylidene fluoride materials have performed satisfactorily and are being considered for prototype development. The commercial electrode casing materials in general have performed well. Additionally, 3-D printing of chemically and mechanically stable materials is being investigated as a means for further improvement in fabrication consistency. SRNL has also investigated altering the reference electrode design to extend the service life. The new design of the interior of the reference electrode casing creates a longer, more tortuous path between the junction material and the electrode sensing wire. A finite element model was used to optimize the design without adversely impacting the circuit resistance of the electrode during the measurements, thus preserving the measurement accuracy while enhancing the service life. The inner chamber back fill materials are also critical to the performance of the reference electrode. Materials that are resistant to intruding tank waste and provide a conductive path to the sensing wire were investigated. Gel and powder materials that are interspersed with a conductive chloride bearing material were tested for their influence on diffusion and electrode resistance. All the investigated materials and components will be assembled, with collaboration from commercial vendors, to fabricate the initial prototypes. Accelerated testing of the prototypes will be initiated in Year 2 of the program and will be completed in Year 3. A recommendation on the materials of construction and the design of the new robust reference electrode will be presented to the Hanford tank farm facility.

Sykes, Kiana [Savannah River National Laboratory (↗
DOE OSTI2026

Determining the Solubility Behavior of Kogarkoite in Simulated Nuclear Waste

Kogarkoite (Na 3 FSO 4 ) is a sparingly soluble fluoride–sulfate double salt that has been identified in high level nuclear waste sludge at the Hanford Site and, more recently, in sludge batch compilation samples at the Savannah River Site (SRS). Due to its complex dissolution behavior, which exhibits an inverse dependence on sodium ion activity, the presence of this mineral poses significant challenges to waste retrieval and processing. Incomplete dissolution during sludge washing can lead to the retention of fluoride and sulfate in the high-level waste feed, potentially causing the formation of corrosive, immiscible molten salt layers, known as "glass gall,” in vitrification melters. Current efforts to optimize flowsheet parameters and wash-water volumes are hindered by the absence of a commercially available, certified reference material, which prevents the accurate calibration of analytical methods and the verification of dissolution kinetics. To address this critical gap, this research focuses on the laboratory synthesis of pure Kogarkoite to serve as a standard for comprehensive solubility and washing performance testing. A coupled synthesis and simulant campaign was executed using an evaporative crystallization protocol designed to replicate the dynamic concentration effects observed in tank farm operations. Thirteen simulant matrices were prepared by dissolving systematically varied ratios of sodium fluoride (NaF) and sodium sulfate (Na 2 SO 4 ) in deionized water under three distinct caustic regimes: 0.0 g (control), 4.0 g (~1 M), and 12.0 g (~3 M) sodium hydroxide (NaOH). While thermodynamic equilibrium models suggest that high-caustic environments should favor the stability of the double salt7, results from this evaporative study at 25 0 C revealed a distinct kinetic divergence. Simulants with high hydroxide loading predominantly yielded large, blocky crystals of sodium sulfate decahydrate (Na 2 SO 4 .10H 2 O). Successful synthesis of pure Kogarkoite was achieved exclusively in specific NaOH-free compositional windows, where the precipitate manifested as fine, opaque granular aggregates. Ion chromatography (IC) analysis confirmed phase purity through the simultaneous stoichiometric depletion of both fluoride and sulfate from the supernatant. This successful synthesis establishes a reproducible route to generate bulk Kogarkoite, enabling the subsequent phase of quantitative dissolution testing using inhibited water to optimize sludge-batch assembly.

Sarker, Md Sharif [Florida International Univ. (FI↗