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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Analysis Results for Sub-Samples of SRS Tanks 30, 32, and 39 to Support Evaluations of the I-129 Inventory

In order to appropriately model and predict the chemical integrity and performance of cementitious materials used for waste immobilization at the Savannah River Site (SRS), it is critical to understand the I-129 solubility and distribution within the tank farm. Iodine in radioactive waste and in environmental media is typically highly mobile and long lived. Iodine is ubiquitous in SRS tank waste and waste forms. The iodine is assumed to be soluble and present at low levels in Performance Assessments (PAs) for SRS Tank Farms, and is one of the dose drivers in the PAs for both the SRS Salt Disposal Facility (SDF) and the H-Area Tank Farm (HTF). Analysis of tank waste samples is critical to understanding the Tank Farm iodine inventory and reducing disposal uncertainty. Higher than expected iodine levels have recently been observed in residual solids isolated from some SRS tanks prior to closure, indicating uncertainty regarding the chemical species involved. If the iodine inventory uncertainty is larger than anticipated, future work may be necessary to reduce the uncertainty. This memorandum satisfies a portion of the work scope identified in Task Plan SRNL-RP-2016-00651. A separate memorandum issued previously, reported historical unpublished I-129 data, a significant portion of which was below detectable analytical limits. This memorandum includes iodine and general chemical analysis results for six archived SRNL samples which were previously reported to have I-129 concentrations below detectable limits. Lower sample dilution factors were used for the current analyses in order to obtain concentrations above detection. The samples analyzed included surface and depth samples from SRS tanks 30, 32, and 39.

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DOE OSTI2020

Summary of Analytical Results from Samples Supporting Tank Closure Cesium Removal (TCCR) Batch 3 and Modeling Results for Cs Loading on CST

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 3 for processing through the TCCR unit. Some results from analysis of those samples have been previously reported. This report documents the remaining analyses of the in-tank batch contact samples as well as the results of ZAM (Zheng, Anthony, Miller) isotherm modeling performed for comparison to the measured results. Results of the additional analyses include analysis of the loading of other radionuclides besides 137 Cs on the crystalline silicotitanate (CST) contained within the in-tank batch contact test samples. Results from those analyses revealed the next highest contributor to the activity on the CST was 90 Sr with an average loading of 2.76E+08 dpm/g CST compared to 3.56E+10 dpm/g CST for the 137Cs. Isotopes of plutonium were also detected on the samples. ZAM modeling was performed using the measured composition of the Tank 10H Batch 3 qualification samples. The modeling predicted a maximum Cs loading approximately 2.2x higher than the measured result. This is a slightly lower ratio (expected/measured) compared to what was observed for the prior TCCR in-tank batch contact testing performed for Batches 1A and 2 where the ZAM results were 2.7-2.8x higher than the measured values.

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DOE OSTI2025

Characterization of Tank 9H Annulus Sample in Support of Residual Material Inventory Determinations

The Savannah River National Laboratory (SRNL) was requested by Savannah River Mission Completion (SRMC) to provide sample preparation and characterization of the Tank 9H annulus sample in support of Residual Material Inventory Determinations. One Tank 9H sample in three vials [HTF-9-25-13, HTF-9-25-14 and HTF-9-25-15], with each vial containing approximately 200 mL of the Tank 9H annulus salt solution, were delivered to the SRNL Shielded Cells for sample preparation and characterizations in February 2025. The density of the “as-received” solution contained in each of the three Tank 9H annulus sample vials were determined followed by a solid-liquid separation on each one using 0.45-micron Nalgene® nylon filter membranes. The resulting filtrates were combined to form the Tank 9H annulus sample with a total volume of about 600 mL. The combined wet solid fractions, about a total of 4.8 grams of salt material, remaining on the filter membranes were air-dried in the Shielded Cells for 72 hours. The total weight of the air-dried solids was 2.1 grams. These air-dried solids were washed with deionized water (DI water) at a phase ratio of 60 mL DI water/gram of solids to recover insoluble solids, if any. No visible or measurable quantity of insoluble solids were recovered after DI water washing of the air-dried solids because the air-dried solids completely dissolved in the DI water. The solid fraction-wash water was not combined with the 600 mL of the filtrate solution, and the resulting solution was not screened or analyzed for radionuclides. Aliquot sample volumes of the undiluted Tank 9H annulus sample were sent to the SRNL analytical services groups for radionuclides, elementals, anions and total mercury analysis by various methods including radiochemical separations/counting methods, inductively coupled plasma-atomic emission spectroscopy (ICP-AES), and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) and special preparations. All sample analyses were performed in triplicate. This report presents the analytical characterization results for the Tank 9H annulus sample. The results are also reported where analytical methods yielded additional analytes, other than those requested by SRMC. In the characterization of the Tank 9H annulus sample, the detection limits for all the analytes, as specified in the Technical Task Request (TTR) and Task Technical and Quality Assurance Plan (TTQAP), were met.

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DOE OSTI2023

Modeling of Stress Distribution in Molten Salt Thermal Energy Storage Tanks for In-Service Central Receiver Power Plants

Failures in molten nitrate salts thermal energy storage tanks (TES) have been occurring in several concentrating solar power (CSP) plants around the world after a few months or years of operation. These failures are mainly related to a combination of high stress, corrosion, large deformation, and thermal cycling. Most of these failures are a consequence of the infancy of this technology; technical standards for design, fabrication, implementation, and operation of molten salt tanks operating at temperatures up to 565 degrees Celsius do not exist and most of current in-service tanks have been designed using the recommendations from the American Petroleum Institute (API) 650 and the American Society of Mechanical Engineers ASME section II standards, which are applicable for TES tanks operating at lower temperatures. We have developed a 3-dimensional model of a representative molten salt hot tank design that uses typical CSP plant operation conditions and allows determining the temperature and stress distribution in the tank. The model also considers the initial floor residual stresses and distortion after its welding fabrication. We have identified critical thermal gradients in the tank floor and shell and locations in the tank where the stresses are high and could become susceptible points for failure.

concentrating solar power↗
DOE OSTI2015

Solubility predictions for sodium nitrate from water activity for Hanford tank concentrates - 15551

Space in Hanford waste tanks is very limited and therefore active evaporation of tank liquids continues to provide a means to conserve tank space. Prior to each evaporator campaign, a boil-down of a sample of tank liquid provides a basis for planning by measuring density, solids precipitation, and water activity as a function of concentration and temperature. We have found these datasets also provide opportunities to validate and calibrate various solubility models on actual waste. Previous work has shown the solubilities of minor electrolytes in these five boil-downs follow rather simple power law solubility products. Each of five boil-downs concentrated liquids from five waste tanks and tank blends and have been reported from Hanford's 222-S laboratory. All concentrates eventually showed the precipitation of the dominant electrolyte, NaNO{sub 3}, and solution activity largely drives this solubility. Although there are many well established techniques for calculating solution activity, the complexity of Hanford electrolyte mixtures has proven challenging for such solution activities. Accurate predictions of water activity for these concentrates provides an alternative method for calculating solution activity and previous work has validated a relatively simply model for predicting the water activity of even complex electrolyte mixtures. Given such a water activity, a Gibbs-Duhem integration of that water activity accurately predicts the solution activity and therefore the solubility of NaNO{sub 3} at 18 deg. C with a correlation coefficient of 0.94 over a wide range of liquid waste compositions for these boil-down studies. To our knowledge, no other solubility prediction has correlated as highly with measured solubility points of NaNO{sub 3} in any complex electrolyte mixture. Better predictions of NaNO{sub 3} solubility in complex waste mixtures improve planning for recovery of waste tank space and reduce the need for new tank space. (authors)

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DOE OSTI2015

Testing of a 3-D Sonar for Future Deployment to Image Solids on the Floor of Hanford, HLW Conditioning Tanks - 15646

Florida International University (FIU) has tested the capability of 2-D and 3-D sonar systems to image solids settled on the floor of tanks, on the bottom of streams and small rivers, and in 55 gallon drums. FIU developed and successfully tested the Solid-Liquid Interface Monitor (SLIM) for mapping the interface between supernatant and settled solids in several test tanks. SLIM is comprised of: a sonar, a mechanical system for deploying the system into large high-level radioactive waste (HLW) tanks, and imaging programs and algorithms developed at and by FIU faculty and staff. Early testing had the objective to demonstrate the ability of SLIM to image and quantify the height and volume of settled solids on the floor of 1-million-gallon, HLW tanks at Hanford. Testing showed that SLIM: (1) functions accurately in liquid radioactive HLW tank conditions [i.e., caustic (pH>14), warm (up to 45 degrees Centigrade), and highly radioactive]; (2) provides accurate imaging of solids surfaces (1-2% relative error in the height of interfaces); and (3) provides an excellent volume estimation of the solids, thereby allowing engineers to fill tanks closer to the maximum allowed level and optimize solids retrieval; and (4) enables better placement of pumps inside tanks during retrieval to minimize the potential for plugging. The New Hanford technology need is for rapid imaging (even during vigorous mixing) of solids on the floor of conditioning (mixing) tanks. Results of recent successful testing show imaging times under 1 minute, solids volume estimates with <5% error and effective filtering and visualization. (authors)

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DOE OSTI2016

Detection and Mitigation of Hanford Site Radioactive Waste Tank Surface Water Intrusion - 16396

The last single-shell underground radioactive waste storage tanks at the U.S. Department of Energy's Hanford Site, near Richland, WA, were removed from service in November, 1980. At that time the single-shell tanks contained 153,300 m{sup 3} (40.5 million gallons) of waste, including 34,400 m{sup 3} (9.1 million gallons) of liquids. Subsequently the Interim Stabilization Program and Waste Retrieval/Closure Project have removed 24,600 m{sup 3} (6.5 million gallons) of liquid from the tanks to reduce the potential for leakage to the environment. After interim stabilization, each tank was isolated by cutting and capping waste transfer lines, filling drains with grout, and covering likely surface water entry points. The minimum expected lifetime for the isolation features was 20 years, consistent with planned retrieval of all single-shell tank wastes by 2018. Since then the retrieval completion date has been extended. In 2011 the Tank Operating Contractor began a systematic review of the aging isolation barriers to determine if surface water was entering the single-shell tanks. This paper summarizes the results of the 2011 - 2015 activities including single-shell tank inspection results, methodologies used to identify intrusions and waste leaks masked by intrusion, and eventual deployment of a prototype forced ventilation exhauster in 2015 to evaporate water intrusion pools on the waste surface. Further efforts to limit water intrusion, including deployment of surface barriers, are discussed. (authors)

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DOE OSTI2018

Robotic Cleaning of the Total Dissolved Solids Tanks at the Palo Verde Nuclear Generating Station - 18337

Westinghouse is executing a contract from Arizona Public Services to provide a system capable of removing highly radioactive sediment from eight Total Dissolved Solids (TDS) Tanks located at the Palo Verde Nuclear Generating Station. The scope of work covers the design, procurement, fabrication, and eventual operation of the system at the Palo Verde Site. It also includes the factory acceptance testing of the system, prior to operation at the site. Previous attempts to clean the tanks involved working inside and around the tanks for extended periods of time in a high radiation field. Spray hoses were used to wash down the sediment, and hand tools were used to scrape out the remaining sludge at the bottom. This methodology also required the lower manway to be removed. As a result, the worker radiation exposure limit was exceeded on the first of the nine tanks. As a solution to the high radiation environment of the TDS tanks, Westinghouse will provide and operate a pneumatically powered robotic crawler. This crawler traverses the bottom of the tank with a 4,000 psi high-pressure spray head affixed to the front of the crawler to blast sediment from the walls and floor of the TDS tanks. This paper will present the conceptual design of the TDS Tank cleaning system, noting the key benefits of specific components in the overall technology basis. (authors)

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DOE OSTI2018

Monitoring Waste Tanks Using Guided Wave Technology - 18576

There are many challenges associated with the inspections of double-shell storage tanks containing radioactive waste, including limited space for accessibility and maneuverability, environments with high radiation, temperature and toxicity, and unknown physical conditions. Contents of the tank and limited space may prevent access of scanning devices that employ conventional NDE techniques such as ultrasonics, eddy current, or magnetic based methods. Resolving these challenges will become more urgent as the nuclear industry is forced to address the complications of the aging nuclear infrastructure - specifically at tank storage locations on laboratory sites across the DOE complex. A powerful, new approach to inspecting and monitoring the structural integrity of large structures, such as double-shelled tanks, is being developed using ultrasonic guided waves. Acoustic waves that propagate in a plate or other structure where the wavelength is larger than the thickness of the structure are known as ultrasonic guided waves. Ultrasonic guided waves can potentially propagate distances of tens to hundreds of feet in plate-like structures, such as those that comprise the walls and bottom of a storage tank. Guided waves enable the examination of relatively large inspection volumes from the outside surface of the structure. In double-shell tanks, the technology offers promise for examining the floor of the structure or the walls by placing guided wave sensors at accessible locations in the annulus. This guided wave technology has the capability to detect and localize areas of damage or degradation. While guided wave has recently been used in the nuclear and oil and gas industries for inspecting one-dimensional piping systems, current research is developing methods to apply this technology to the two-dimensional surfaces of double-shell storage tanks. Permanently installed sensors are being developed which would remotely collect data that would be analyzed to monitor the structural condition of the tank. Initial tests have been conducted on single plate structures to study guided wave interaction with various shapes and sizes of pertinent simulated damage mechanisms, and the experimental results were compared with hybrid finite-element numerical models. Dispersion curves were calculated to study guided wave modes and their properties in carbon steel plates and to evaluate behaviors at joints between plates. The next steps in the research process include applying the transmitters and sensors to full scale mockups, developing more mature data interpretation, and then demonstrations at an actual site. (authors)

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DOE OSTI2019

Characterization of Tc-99 Speciation and Chemical Composition of 241-AN-102 Tank Waste Supernatant - 19201

Technetium (Tc) in its isotopic form {sup 99}Tc is generated in high yields from the fission of {sup 235}U and {sup 239}Pu. A significant inventory of {sup 99}Tc remains at the U.S. Department of Energy's Hanford Site, and it is one of the most difficult contaminants to address. In the strongly alkaline solutions that are typical of Hanford tank waste, Tc exists as both pertechnetate (TcO{sub 4}{sup -} in oxidation state VII) and in reduced forms (oxidation state <VII) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized anionic Tc species that can be removed from low-activity waste by anion exchange or other methods. There is no definitive information on the origin or a comprehensive description of the non-pertechnetate species in Hanford tanks. Designing strategies for effective management of Tc, including separation and immobilization, requires an understanding of the molecular structure of the non-pertechnetate species and their identification in actual tank waste samples. Developing this advanced understanding would facilitate development of new technologies for treating dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO){sub 3}]{sup +} and Tc(IV, VI, VII) compounds was generated using a range of techniques and applied to the characterization of actual 241-AN-102 tank waste supernatant. A sample of the 241-AN-102 tank waste (collected in 2013) was processed to remove {sup 137}Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin after the sodium concentration was adjusted to 5.6 M. The Cs-loaded sRF column was eluted with 0.45 M HNO{sub 3}. As-received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using {sup 99}Tc nuclear magnetic resonance (NMR) spectroscopy and X-ray absorption spectroscopy. The results of these analyses demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO){sub 3}L{sub 3}]{sup +}, Tc(IV), and possibly Tc(VI) compounds. This is the second experimental observation of the [fac-Tc(CO){sub 3}L{sub 3}]{sup +} species in the Hanford tank waste that highlights their importance for waste processing. In addition, the chemical composition of the 241-AN-102 tank waste was characterized with the focus on small chelators. (authors)

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DOE OSTI2019

Progress on Advancing Robotic Ultrasonic Volumetric Inspection Technology for Hanford Under-tank Inspection - 19474

The Hanford River Protection Project is reliant on safe interim storage of high-level waste at the Hanford site and will continue to depend on it until the waste has been treated and vitrified by the Hanford Waste Treatment and Immobilization Plant (WTP). The Hanford double-shell tanks (DSTs) have provided interim storage since the DSTs were put into service (1971 - 1986) and the RPP will continue to rely on their service until least 2050, at which time the 27 in-service DSTs will be 14-54 years beyond their original design lives. As the DSTs continue to age and be operated beyond design life, reliable leak integrity assessments of primary tanks and secondary liners will become increasingly important in order to provide situational awareness to tank operations personnel and to support risk-informed decisions on the selection of DSTs as waste feed staging tanks and the order and timing of future DST refurbishment or, alternatively, DST retrieval sequence. This paper summarizes progress made in fiscal year (FY) 2018 on the engineering of a robotic air-slot volumetric inspection system (RAVIS) that is designed to perform ultrasonic (volumetric) leak integrity inspections of physically inaccessible primary tank bottom plates and welds. Leak integrity inspections performed with the RAVIS will supplement today's primary tank leak integrity inspections, which are currently limited to physically accessible sidewall plates and welds. The RAVIS will be deployed into air slots of the insulating concrete (refractory) pad beneath the primary tanks and use its ultrasonic guided wave phased-array (GWPA) sensor to electronically steer an ultrasonic beam 360-degrees around the sensor to volumetrically examine adjacent carbon steel plates and welds covered by refractory pad material. Deployments of RAVIS down multiple air-slots will be performed to determine which regions of a tank bottom have sound leak integrity or which regions are in need of future wall thickness restoration or weld repair due to pitting corrosion, general corrosion, or heat/stress-induced cracks. The initial RAVIS design and the flaw detection test results for the first air-slot GWPA sensor that was designed to fit inside the air-slots will be presented in this paper. (authors)

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DOE OSTI2020

DOE Hanford Tank Side Cesium Removal System (TSCR) Update: Fabrication and Factory Acceptance Testing - 20300

The Tank Side Cesium Removal (TSCR) System is a modular, quickly-deployable, and passively safe system for efficiently removing solids and cesium from alkaline liquid waste raffinate created by fuel processing on the Hanford site. This paper provides an overview of the project with particular emphasis on completion of fabrication and factory acceptance testing (FAT). All TSCR System components are inside three enclosures known as the Process Enclosure, Ancillary Enclosure, and Control Enclosure. The Process Enclosure is seismic design category II (SDC-II) and built to withstand the rigors of transportation and lifting, high winds, volcanic ash fall, and heavy precipitation (snow) over a temperature range of -25 to 115 deg. F. The Process Enclosure and Control Enclosure were fabricated in AVANTech's Columbia, SC facility, while the Ancillary Enclosure was fabricated in AVANTech's Richland, WA facility. The design and preliminary documented safety analysis were completed in December 2019, with all work being performed in accordance with ASME NQA-1 quality assurance requirements. The Process Enclosure, Ion Exchange Columns (IXCs), and several other TSCR components have a nuclear safety function and are designated as safety significant structures, systems, and components (SS-SSCs). The NQA-1 Commercial Grade Dedication (CGD) process was used to identify critical characteristics in order to accept materials and items used for fabrication of several safety significant TSCR components. This paper also addresses unique aspects and challenges of implementing CGD in DOE facilities and systems. After fabrication and preliminary testing, the Process and Control Enclosures were shipped to AVANTech's Richland facility where the complete TSCR System was setup and connected in a represented layout and configuration that will be used on the Hanford site. Test tanks and pumps were put in place and connected to the TSCR Process Enclosure in a manner that mimicked interfacing Double Shell Tanks in the 241-AP Tank Farm. The TSCR System was then tested/operated under conditions equivalent to those planned for onsite operation. Testing was separated into five separate phases as defined below: - Phase 1: Confirmation of TSCR System readiness for the FAT; - Phase 2: Verification of TSCR System operation under standard conditions; - Phase 3: Demonstration of TSCR System operation under non-standard conditions; - Phase 4: Demonstration of ion exchange media removal from an IXC; and - Phase 5: Completion of TSCR System lift plans and site mobilization. The Hanford Tank Operations Contractor, Washington River Protection Solutions (WRPS), participated in all phases of testing. Testing at AVANTech's Richland Facility allowed WRPS personnel to familiarize themselves with TSCR operations in a radiologically clean facility prior to onsite mobilization. Engineering and operations personnel gained valuable experience that allowed them to better refine onsite operational protocols and procedures - including handling and movement of IXCs with a forklift. This paper describes the technologies used by the TSCR System as well as how these technologies are deployed and operated to advance the Hanford mission of safely, efficiently, and effectively treating tank waste. It also provides detailed information on fabrication and FAT, including operational data from simulated tank waste runs. A project schedule for WRPS' planned construction acceptance testing, operational acceptance testing, and actual tank waste treatment is provided. This presentation should be of great interest to parties responsible for the design, testing, and mobilization of modular NQA-1 technologies for the treatment of raffinate and similar liquid wastes containing high concentrations of radioactive cesium. (authors)

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DOE OSTI2012

Relationship Between Flowability And Tank Closure Grout Quality

After completion of waste removal and chemical cleaning operations, Tanks 5-F and 6-F await final closure. The project will proceed with completing operational closure by stabilizing the tanks with grout. Savannah River Remediation's (SRR) experience with grouting Tanks 18-F and 19-F showed that slump-flow values were correlated with flow/spread inside these tanks. Less mounding was observed when using grouts with higher slump-flow. Therefore, SRNL was requested to evaluate the relationship between flowability and cured properties to determine whether the slump-flow maximum spread of Mix LP#8-16 could be increased from 28 inches to 30 inches without impacting the grout quality. A request was also made to evaluate increasing the drop height from 5 feet to 10 feet with the objective of enhancing the flow inside the tank by imparting more kinetic energy to the placement. Based on a review of the grout property data for Mix LP#8-16 collected from Tank 18-F and 19-F quality control samples, the upper limit for slump-flow measured per ASTM C 1611 can be increased from 28 to 30 inches without affecting grout quality. However, testing should be performed prior to increasing the drop height from 5 to 10 feet or observations should be made during initial filling operations to determine whether segregation occurs as a function of drop heights between 5 and 10 feet. Segregation will negatively impact grout quality. Additionally, increasing the delivery rate of grout into Tanks 5-F and 6-F by using a higher capacity concrete/grout pump will result in better grout spread/flow inside the tanks.

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DOE OSTI2014

ACTUAL WASTE TESTING OF GYCOLATE IMPACTS ON THE SRS TANK FARM

Glycolic acid is being studied as a replacement for formic acid in the Defense Waste Processing Facility (DWPF) feed preparation process. After implementation, the recycle stream from DWPF back to the high-level waste Tank Farm will contain soluble sodium glycolate. Most of the potential impacts of glycolate in the Tank Farm were addressed via a literature review and simulant testing, but several outstanding issues remained. This report documents the actual-waste tests to determine the impacts of glycolate on storage and evaporation of Savannah River Site high-level waste. The objectives of this study are to address the following: Determine the extent to which sludge constituents (Pu, U, Fe, etc.) dissolve (the solubility of sludge constituents) in the glycolate-containing 2H-evaporator feed. Determine the impact of glycolate on the sorption of fissile (Pu, U, etc.) components onto sodium aluminosilicate solids. The first objective was accomplished through actual-waste testing using Tank 43H and 38H supernatant and Tank 51H sludge at Tank Farm storage conditions. The second objective was accomplished by contacting actual 2H-evaporator scale with the products from the testing for the first objective. There is no anticipated impact of up to 10 g/L of glycolate in DWPF recycle to the Tank Farm on tank waste component solubilities as investigated in this test. Most components were not influenced by glycolate during solubility tests, including major components such as aluminum, sodium, and most salt anions. There was potentially a slight increase in soluble iron with added glycolate, but the soluble iron concentration remained so low (on the order of 10 mg/L) as to not impact the iron to fissile ratio in sludge. Uranium and plutonium appear to have been supersaturated in 2H-evaporator feed solution mixture used for this testing. As a result, there was a reduction of soluble uranium and plutonium as a function of time. The change in soluble uranium concentration was independent of added glycolate concentration. The change in soluble plutonium content was dependent on the added glycolate concentration, with higher levels of glycolate (5 g/L and 10 g/L) appearing to suppress the plutonium solubility. The inclusion of glycolate did not change the dissolution of or sorption onto actual-waste 2H-evaporator pot scale to an extent that will impact Tank Farm storage and concentration. The effects that were noted involved dissolution of components from evaporator scale and precipitation of components onto evaporator scale that were independent of the level of added glycolate.

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DOE OSTI2015

Evaluation of Settler Tank Thermal Stability during Solidification and Disposition to ERDF

Ten 16-foot-long and 20-inch diameter horizontal tanks currently reside in a stacked 2×5 (high) array in the ~20,000-gallon water-filled Weasel Pit of the 105-KW Fuel Storage Basin on the US-DOE Hanford Site. These ten tanks are part of the Integrated Water Treatment System used to manage water quality in the KW Basin and are called “settler” tanks because of their application in removing particles from the KW Basin waters. Based on process knowledge, the settler tanks are estimated to contain about 124 kilograms of finely divided uranium metal, 22 kg of uranium dioxide, and another 55 kg of other radioactive sludge. The Sludge Treatment Project (STP), managed by CH2MHill Plateau Remediation Company (CHPRC) is charged with managing the settler tanks and arranging for their ultimate disposal by burial in ERDF. The presence of finely divided uranium metal in the sludge is of concern because of the potential for thermal runaway reaction of the uranium metal with water and the formation of flammable hydrogen gas as a product of the uranium-water reaction. Thermal runaway can be instigated by external heating. The STP commissioned a formal Decision Support Board (DSB) to consider options and provide recommendations to manage and dispose of the settler tanks and their contents. Decision criteria included consideration of the project schedule and longer-term deactivation, decontamination, decommissioning, and demolition (D4) of the KW Basin. The DSB compared the alternatives and recommended in-situ grouting, size-reduction, and ERDF disposal as the best of six candidate options for settler tank treatment and disposal. It is important to note that most grouts contain a complement of Portland cement as the binding agent and that Portland cement curing reactions generate heat. Therefore, concern is raised that the grouting of the settler tank contents may produce heating sufficient to instigate thermal runaway reactions in the contained uranium metal sludge.

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DOE OSTI2018

Vitrification of Hanford Tank Waste 241-AP-105 in a Continuous Laboratory-Scale Melter

Low-activity waste (LAW) stored in underground tanks on the Hanford site in Washington State is planned to be filtered for solids removal and processed through ion exchange columns for cesium removal at the Low-Activity Waste Pretreatment System facility. These pretreatment steps will allow the waste to be transferred to the Hanford Tank Waste Treatment and Immobilization Plant LAW facility for immobilization into glass. The liquid waste will be combined with glass-forming chemicals (GFCs) to form a waste feed slurry that can be fed to electric melters for vitrification. The process of continuously converting the aqueous feed slurry into a melt is dynamic and includes multiple reactions, degassing, and dissolution processes that depend on heat from the melt below. In this conversion process, waste components are partitioned into one of two streams: glass and off-gas. Processing information and chemical information associated with these waste product streams has been requested by Washington River Protection Solutions for actual tank waste from tank 241-AP-105 (referred to as AP-105). To acquire this type of information, a small-scale melter system was desired that would not require high volumes of input waste or the large resource commitment of a full-scale melter system, while also providing dynamic information that would be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM), has been designed to operate with a continuous feeding process, while periodically pouring glass product and collecting off-gas. The CLSM vessel has been sized to collect the relevant process and chemical information from obtainable volumes of AP-105 waste samples. Two CLSMs were constructed: one in a non-radioactive environment for processing waste simulants and another in a fume hood capable of handling radioactive material for processing actual Hanford tank waste. This dual-system setup allowed for comparison between simulated and real tank waste with essentially identical test equipment. AP-105 waste simulant was first tested in the non-radioactive CLSM to determine the processability and desired operating window for the actual tank waste. Rhenium (Re) was added to the simulant to act as a surrogate for technetium-99 ( 99 Tc), a key waste component found in the actual tank waste. The partitioning of Re into the glass and off-gas products during the continuous feeding of simulant was determined, which allowed the Re retention, the mass flow rate of Re out of the CLSM via the glass product with respect to the mass flow rate of Re into the CLSM via the feed slurry, to be calculated. Key processing results and the average Re retention, from the CLSM test run with AP-105 simulant feed slurry are shown in Table ES.1.

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DOE OSTI2019

Analysis of Tank 23H (HTF-23-19-23, -24, -25) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 23H to support ECP and CCP. The results of the analysis indicate the concentrations of species in Tank 23H increase with depth in the tank. The sample obtained from the lowest level of 60” from the bottom of the tank shows the highest concentrations for most species analyzed. The surface sample and the 140” sub-surface sample show similar compositions with the sub-surface sample being slightly more concentrated, and both are more dilute than the lowest 60” sub-surface sample for most species analyzed. The total uranium concentration in the three Tank 23H samples show a slight increase from the surface sample to the next lower sub-surface sample, but then a slight decrease in the uranium concentration in the lowest sub-surface sample. The U-235 weight fraction ranges from 0.94-1.2% in the Tank 23H samples. The sum of the major cations versus the sum of the major anions shows a difference of ~5% or less for all three of the Tank 23H samples indicating good data quality for the non-radioactive analytes in the table. Small %RSD values for the replicate analyses with the exception of species close to the detection limit (e.g., Fe) provide another indication of good data quality for the non-radioactive species. The silicon concentrations measured in the Tank 23H samples were all below detection.

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DOE OSTI2021

End-of-Life Evaluation of Compressed Natural Gas Vehicle Fuel Tanks

This study evaluated the structural integrity of Type III and Type IV CNG fuel tanks from the Los Angeles (LA) County Metropolitan Transportation Authority to characterize the fuel tank conditions after experiencing a full service life of 15 years in transit bus application. The data produced provide insight about the condition of the CNG fuel tanks at the conclusion of their defined EOL and potential risks of continued operation of the CNG vehicle without replacement of the expired fuel tanks. In addition to physical testing performed on the fuel tanks, a nondestructive evaluation of modal acoustic emission (MAE) was utilized to assess the structural integrity of the tanks. The CNG fuel tanks were evaluated in their received condition and after experiencing artificial damage. The MAE results were then compared to visual inspection results of the procedures defined in the Compressed Gas Association’s (CGA’s) C-6.2 and C-6.4 standards (Compressed Gas Association 2013; 2012) to better understand the effectiveness of the visual inspection and potential safety risks of continued use of the tanks.

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