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DOE OSTI2020

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Simulation of DWPF Chemical Process Cell Using a Reaction Calorimeter

The Defense Waste Processing Facility (DWPF) converts liquid radioactive waste into glass using a Sludge Receipt Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) in addition to other processes. The SRAT and SME processes can be simulated on a small-scale using the Mettler-Toledo Reaction Calorimeter (RC1). This equipment is able to automate the execution and monitoring of experiments involving nonradioactive chemical simulants. During the chemical processing of sludge in the SRAT process, foam begins to form and can foam over the reaction vessel. The current antifoam used at DWPF creates a risk of increased flammability and has a limited pH range. Usually, mercury is circulated back into the reaction vessel during the SRAT process, making it difficult to remove. To combat foam over, novel antifoams will be tested during the boil-up process using the RC1. To improve mercury extraction, 3D-printed coalescer columns will be tested within the Mercury Water Wash Tank (MWWT) of the RC1. Testing of Simulant Sludge Batch 6 (SB6) was performed to ensure: the sludge would form a stable foam at boiling during full SRAT/SME runs; the antifoam would reduce the amount of foam within the vessel The sludge and a Precipitate Reactor Feed Tank (PRFT) simulant with monosodium titanate (MST) were mixed separately, combined, and then heated to boiling. As shown below, the antifoam used was able to eliminate the foam within the vessel, as expected. The FormLabs Form2 3D Printer can be used to create roughly 3-inch tall and 1-inch diameter columns from clear and ceramic resins. The columns will utilize chemical and physical forces to draw the mercury out of a condensed offgas stream leaving the RC1 reaction vessel. These columns will be tested during full SRAT/SME runs on the basis of how well they are able to collect elemental mercury. In preparation for full-scale runs, the RC1 equipment was assembled onto a custom table and tested with water runs. The software used to control automation, iControl, has a user-friendly interface that allows the experiment to be altered as needed. An example of the data recorded during an experimental run is represented in Figure 1 below. Water runs were utilized as a way to prove if the RC1 system was functioning in the following areas: Boil-up rates, Chemical additions using pumps, Temperature readings, Balance readings, Probe readings, Offgas system, Agitation. Future Plans: Five novel antifoams will be tested during full SRAT/SME runs. The most effective antifoam will be utilized in DWPF's Chemical Process Cell. The design of the Mercury Extraction Coalescer can be further manipulated to increase the extraction of not only elemental mercury, but other heavy metals in environmental remediation processes as well.

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DOE OSTI2021

Cesium Exchange onto Crystalline Silicotitanate from Blended Hanford Tank Wastes

The Tank Side Cesium Removal (TSCR) system was developed to filter and remove cesium (Cs and 137 Cs) from Hanford tank waste supernate in preparation for vitrification. The Cs removal will be conducted with crystalline silicotitanate (CST) ion exchange media. Under the planned waste-processing strategy, the tank waste supernate will be queued for TSCR processing in tank 241-AP-107 (AP-107). Once AP-107 tank waste volume is sufficiently depleted, the waste supernate from tank 241-AP-105 (AP-105, the holding tank before transfer to AP-107) will be transferred to tank AP-107. Supernate from another tank will be transferred to the holding tank, AP-105, for eventual transfer to tank AP-107. These supernate streams will undergo blending in tanks AP-107 and AP-105; the volume blend ratios will be driven by how much the tank waste supernate volumes are depleted before the next tank waste is added. The consequence of tank waste blending on Cs uptake by CST was of interest and was tested via batch contacts; results are reported herein.

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DOE OSTI1961

Laboratory Development of a Pot Calcination Process for Converting Liquid Wastes to Solids

Simulated Darex, Purex, Thorex, and TBP-25 wastes were batch-evaporated and calcined to solids. The weight of residue was decreased to ~ 8% of the original waste for Darex, to ~ 4% for Purex, to ~ 2% for Thorex, and ~ 7% for TBP-25 by calcination to 1200°C. Semicontinuous evaporation and calcination to solids at 900°C of Darex and Purex wastes gave volume reduction factors of ~ 8 and 7 to 10, respectively. The nitrate contents of the residues from batch calcination to 800°C were from ~ 0.07 to 0.5 wt % while the nitrate contents of the residues from semicontinuous calcination to 900°C were from ~ 0.1 to 0.7 wt %. Sodium, calcium, and magnesium additives to Purex waste decreased the percent of sulfate volatilized into the condensate from ~ 30% to <0.5%. Replacement of the atmosphere above the waste with nitric oxide decreased ruthenium volatility from 50 to 60% of that originally present in Purex waste to 0.5 to 3.5%. The thermal conductivities for the calcined wastes, measured in situ, were all >0.1 Btu/hr ft °F at ~ 400°F and increased almost linearly with increasing temperature to >0.3 Btu/hr ft °F at ~ 1600°F in all cases studied.

052001* -- Nuclear Fuels-- Waste Processing↗

DOE OSTI2020

Influence of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities – SCC Testing

Glycolic acid is being evaluated as an alternate reductant in the preparation of High Level Waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to waste facilities. The SRS liquid waste contractor requested an assessment of the influence of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this concern had not been previously evaluated. The influence of glycolic acid on the occurrence of stress corrosion cracking (SCC) had not been evaluated specifically in the previous testing. The present testing used electrochemical and coupon immersion testing to assess this influence. Additionally, the primary guidance on corrosion for operating DWPF with a formate-based flowsheet is based on testing that utilized test solutions with a small number of constituents. Recent testing used more complex solutions based on flowsheet development for DWPF. Additional electrochemical tests were also conducted to assess the influence of this difference on test results.

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DOE OSTI2019

Laboratory-Scale Melter Systems for Simulated and Real AP-105 Tank Waste - 19301

The Hanford Tank Waste Treatment and Immobilization Plant is being constructed to vitrify the nuclear waste stored in underground tanks on the Hanford Site. Through this conversion process, waste components become partitioned into one of two product streams: glass and off-gas, including condensate. A small-scale melter system is appropriate for analyzing the waste-to-glass process because it does not require the high volumes of input waste or resource commitment of a large-scale melter system while providing dynamic information that can be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM) was sized to collect the relevant process and chemical information from obtainable volumes of Hanford tank waste and designed to operate with a continuous feeding process while periodically pouring glass product and collecting off-gas. 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. Multiple waste simulants were first tested in the non-radioactive CLSM to determine the processability and desired operating window for actual tank waste. About 12 liters of real waste from Hanford tank 241-AP-105 were received, filtered, and ion exchanged for removal of Cs-137. This quantity of waste was vitrified into 9.5 kg of glass product over the course of ∼15 hours of feeding to the CLSM. (authors)

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DOE OSTI2023

Techno-economic and life cycle analyses for a supercritical biodiesel production process from waste cooking oil for a plant located in the Midwest United States

Existing literature lacks detailed techno-economic and environmental life cycle analyses (TEA-LCA) focused on supercritical biodiesel production pathways produced via waste products. Therefore, in this work, the TEA-LCA was conducted to generate supercritical biodiesel from waste cooking oil for plants located in the Midwest region of USA having annual production capacities of 10,600 and 128,000 t using Aspen Plus and GREET softwares. Here, two plant production capacities were chosen to capture the economies of scale impact on biodiesel manufacture cost. In the supercritical process, methanol and propane were used as a cosolvent to synthesize biodiesel from waste cooking oil (WCO) at 280 °C and 128 bar at a residence time of 8.4 min and a WCO conversion rate of 97%. Economic analysis revealed that the supercritical process was an economically attractive pathway with a 2-year payback period for the 10,600 t/year capacity along with a break-even selling price of $\$$ 2.42/gal of diesel. In case of the larger plant (128,000 t/year capacity), the payback period and the breakeven selling price were considerably lower at values of 0.4 years and $1.31/gal of diesel, respectively, due to economies of scale impact. The generated biodiesel from supercritical process for the both the plant capacities met European (EN14214) and US (ASTM D6751) fuel quality standards while the obtained commercially valuable side product, i.e., glycerol, adhered to a pharmaceutical grade of 99.7%. Cradle to gate life cycle analysis using GREET revealed that supercritical process possessed 17% lower CO2 emissions than alkali-catalyzed process and 4% lower CO2 emissions than the conventional diesel production process. Other biodiesel production pathways in the literature were also compared to the results of the TEA-LCA of supercritical biodiesel production pathway.

Biodiesel production↗

DOE OSTI2023

Optimization Efforts at DWPF to Ensure Waste Tank Closure by 2037 – 24404

At the Savannah River Site (SRS) in Aiken, South Carolina, the Defense Waste Processing Facility (DWPF) produces a glass-form product by processing high-level liquid waste (HLW) with borosilicate glass in a high-heated Melter, then putting the vitrified waste into stainless-steel canisters. Since 1996, DWPF has performed this vitrification process for the liquid waste mission at the SRS; the overall objective for this mission is to reduce the volume in the upstream waste tanks, so those tanks can be emptied and operationally closed. This mission seeks to eliminate the single biggest environmental risk in the state of South Carolina, which requires a robust processing strategy within the DWPF, along with optimal interface between the other facilities in the liquid waste organization (LWO). As of the end of July 2023, DWPF has safely remediated 63.7 MCi of HLW, and filled 4,319 canisters (We project that about 4,000 more canisters are needed to be filled to reach the closure goal.)

Armstead, III, Frank L.↗

DOE OSTI2019

Measurements of Hydrogen Generation Rates using a Sealed Reactor System - 19428

Sealed reactor tests were performed to measure the hydrogen generation rate (HGR) from stored radioactive waste at the Savannah River Site. HGR may be very low, on the order of 7.5 x 10{sup -13} (m{sup 3}/h)/L [that is, cubic meters of H{sub 2} generated per hour per liter of radioactive liquid waste, or 10{sup -10} (ft{sup 3}/h)/gal]. Because the estimated HGR limit of detection for an open, continuous flow, system is approximately 7.5 x 10{sup -10} (m{sup 3}/h)/L [or 10{sup -7} (ft{sup 3}/h)/gal], a more sensitive method was needed. By accumulating generated gasses in a sealed system without a continuous purge gas, the resulting concentrations can increase the sensitivity of measurement, which is based on time of heating. In 2017, Savannah River Remediation declared a Potential Inadequacy in the Safety Analysis related to the potential thermal decomposition of organics that generate hydrogen in each of three Liquid Waste facilities: Concentration, Storage and Transfer Facility, Saltstone Production Facility, and the Defense Waste Processing Facility. The research for this study concerns low activity waste of Tank 50, which feeds the Saltstone Production Facility. This study evaluated the impact of organics including formate on the generation of H{sub 2}. Samples of SRS Tank 50 waste, which is a nominal 5 molar sodium salt solution, were sealed in stainless steel reactors and held at varying temperatures, e.g., 50 deg. C, 75 deg. C, 100 deg. C, and 120 deg. C, to determine if organic compounds are decomposing into H2. The experiments included trace solids used to stabilize waste in grout, e.g., grout-premix and Blast Furnace Slag, to determine if any solid phase reactions are present, which would accelerate or inhibit the generation of H{sub 2} at these temperatures. This paper describes the sealed reactors designed and constructed at Savannah River National Laboratory and the radioactive tests that were performed to measure HGR in both oxic and anoxic atmospheres. The shakedown tests showed the sealed reactors to be leak tight and hydrogen concentrations from calibrated standards could be accurately measured. Previous simulant tests helped to perfect the test methods and better design the radioactive tests. The radioactive results showed that the rate of hydrogen generated from Tank 50 waste, over a range of temperatures from 30 deg. C up to 120 deg. C, is bounded by an empirical relation (Hu Equation) that can be used to approximate HGRs in Tank 50 radioactive waste. (authors)

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DOE OSTI2018

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 18327

The Waste Treatment and Immobilization Plant (WTP) Project is vital to the U.S. Department of Energy's (DOE) mission to cleanup radioactive waste at the Hanford Site, located in southeastern Washington State. The overall WTP Project objective is to design, build, and commission the facilities and systems that will treat and immobilize approximately 56 million gallons of radioactive waste stored in Hanford's 177 underground storage tanks. The combination of technologies in the WTP makes it a first-of-a-kind facility at a scale larger than ever, built for radioactive waste processing. DOE is proceeding with its approach to complete the WTP Project in phases, which includes the following principal actions: - Completing construction of the Low-Activity Waste (LAW) Facility, Balance of Facilities (BOF), and Analytical Laboratory (LAB) (collectively referred to as LBL); - Implementing a direct-feed low-activity waste (DFLAW) process to treat low-activity waste independent of the Pretreatment (PT) Facility; - Resolving the technical issues associated with primarily the PT Facility; - Completing the design and construction of the High-Level Waste (HLW) and PT facilities to enable integrated WTP operations. In March 2016, a federal judge amended the 2010 Consent Decree 1 completion milestones for the Hanford WTP, which generally aligns with DOE's plans for completing the WTP facilities in sequence: - Complete LAW Facility hot commissioning: December 31, 2023; - Complete HLW Facility hot commissioning: December 31, 2033; - Complete PT Facility hot commissioning: December 31, 2033; - Achieve initial plant operations for the WTP: December 31, 2036. Having established milestones for completing the WTP facilities has enabled DOE to formally implement its plans for completing the WTP facilities in sequence, starting with the LAW Facility. In December 2016, DOE executed a contract modification and approved a change to the project's performance baseline to complete the LBL facilities and incorporate modifications needed to allow direct feed of low-activity waste to the LAW Facility. Significant progress has been made toward completing the design, construction, and planning for commissioning of the facilities needed to enable the DFLAW operations to commence as early as January 2022. In 2012, DOE partially suspended production work on the WTP's PT Facility and to a lesser degree the HLW Facility because of unresolved technical issues associated with those facilities. In September 2017, DOE acknowledged the WTP contractor had met all the criteria for full resumption of engineering, procurement, and construction (EPC) activities for the HLW Facility. By December 2017, DOE had completed the actions needed to resolve the six of the eight primary technical issues associated with the PT Facility, with the remaining two issues expected to be resolved in the first quarter of 2018. Although significant progress has been made to prepare for resuming production activities for the HLW and PT facilities, DOE has made the decision to constrain EPC work on those facilities in the near term based on DOE's higher priority of completing the WTP facilities needed to begin DFLAW operations. (authors)

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DOE OSTI2021

Current status of anaerobic digestion of food waste in the United States

Utilization of food waste (FW) and food processing waste streams in waste-to-energy technologies has increased over the years in United States (US). This review paper compiled information from published literature and open data sources to obtain a holistic view of the current state of FW availability and its management using anaerobic digestion (AD). The review examined the usage of FW as a feedstock for AD by considering the feedstock composition and availability of FW in the US as well as challenges and opportunities with implementing FW in mono- and co-digestion. Also presented are global trends in FW digestion and the current status of FW codigestion in the US. FW in the US was characterized with regard to volume and composition to show a need for waste management and an opportunity for valuable resource recovery. Challenges and opportunities of processing via AcoD were presented to review best practices for successful operations. Environmental standards, economic incentives, and governmental policies in the US and around the world regarding waste management and utilization were discussed to identify barriers and sources of opportunity for large-scale implementation of AcoD. Additionally, a range of facilities throughout the US was examined for information regarding digestion type, feedstock (s), biogas production and utilization, and energy self-sufficiency to better understand the current status of FW digestion. Several US facilities were outlined as case studies that co-digest FW and produce and utilize biogas for various applications. Future strategies, perspectives, and roadmaps were discussed to elucidate the shaping of future FW management practices.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2020

Effect of Evaporation on Vapor Space Flammability in DWPF Process Vessels - 20278

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes liquid high-level waste (HLW) and encapsulates it in glass within stainless-steel containers suitable for safe, long-term storage and disposal. Radiolytic and chemical decomposition of the liquid waste generates hydrogen and other flammable gases creating a hazard. The rate of flammable gas generation increases by an order of magnitude as temperature increases from sub-boiling to boiling. DWPF controls this hazard in process tanks by ensuring a minimum purge flow through each tank's head space. The minimum purge flow rate is established assuming dry air and neglecting evaporation. However, this study determined that steam from boiling in a representative process tank is sufficient to dilute flammable gases and, thereby, preclude flammability. For sub-boiling, water vapor is an insufficient diluent considering hydrogen's lower flammable limit (LFL) in air. However, the upper and lower flammable limits approach each other in a hydrogen-air-water-vapor environment as the water vapor concentration increases. This effect may preclude flammable conditions if water vapor concentration is about 60 volume percent or higher. A saturated, constant-pressure air-water system would achieve 60 volume percent water vapor at 86 deg. C. (authors)

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DOE OSTI2021

Influences on Subsurface Plutonium and Americium Migration

Plutonium (Pu) has been released to the environment worldwide, including approximately 1.85 × 1015 Bq (200 kg) of Pu from process waste solutions to unconfined soil structures at the Hanford Site in Washington State. The subsurface mobility of Pu is influenced by complex interactions with sediments, groundwater, and any co-contaminants within the waste stream. Previous investigations at Hanford have shown that Pu exists as discrete PuO 2 particles forming before or after disposal, as secondary solid phases formed from waste interactions with sediments as adsorbed/incorporated species, and/or as dissolved species. In this research, new evidence is presented for the existence of PuO 2 , PuO 2 -Bi 2 O 3 composites, and particles from burnt Pu metal in near-surface sediments where Pu-laden acidic process waste was disposed to sediments. Pu and americium (Am) L3 X-ray absorption spectroscopy and density functional theory suggest that, in larger, more crystalline PuO 2 particles, Am formed from radioactive decay is retained in the Pu IV O 2 structure as Am IV . The Pu and Am that were disposed of in an acidic waste stream have since migrated deeper into the subsurface with detection to at least 37 meters below ground surface. In contrast, Pu deposited near the ground surface from neutral pH waste is found to be homogeneously distributed and relatively immobile. Groundwater extractions performed on contaminated sediments indicate that both Pu and Am are recalcitrant, with Am being fractionally less extractable than Pu on a molar basis. These results suggest that the more mobile fraction of Am has migrated from the near-surface and may be present in the deeper sediments as a different phase than Pu. From these results, it is suggested that Pu and Am deposited from acidic wastes were initially mobile and became significantly less mobile as wastes were neutralized within the soil profile.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

Development of a Testbed for Pipeline Flushing - 20425

Pipelines which carry high-level waste within the DOE complex should be properly flushed in the cases where stationary or moving beds of solid sediment occur, or lines are prone to hydrogen gas buildup. Present guidelines establish a minimum for the amount of water (flush volume) and flush velocity values used for post-transfer flushing operations to achieve satisfactory cleaning of pipelines. However, further studies are needed to find optimal operating velocity modes/values in order to minimizing the flush volume and consequent downstream waste. These efforts are significantly helpful to DOE waste remediation sites by preserving tank storage, preventing additional waste processing, and minimizing dilution and changes in waste chemistry. An experimental test loop was recently developed for study of non-Newtonian slurry flushing at the Florida International University (FIU). This loop was designed to create sediment beds of various materials and bed heights and investigate parameters that effect the efficiency of flushing operations. The objective is to find flush velocity values/modes which lead to satisfactory cleaning of transport lines with a minimum amount of water usage. A 165 ft experimental test loop made of 3-inch carbon steel pipes was constructed and used in initial testing campaigns. This pipe loop was equipped with real-time monitoring instruments to record variation of pressure, density, mass flow rate, and temperature during testing. Post-flush in-situ no-flow evaluations were performed using ultrasonic, endoscopy, and visual inspection. Additional post-flush evaluations were conducted using filtration and density monitoring of residuals within a designated circulations loop. These provisions enable comparison of pipeline cleanness before the start and after termination of flushing operations for efficiency evaluation. This paper presents efforts associated with flushing of kaolin-water mixtures at different concentrations and with various flush volume and flush velocity values/modes. In initial testing, efforts focused on creating repeatable sediment beds inside the pipeline in both fully-flooded (no pre-flush drainage) and gravity-drained (pre-flush drained) systems. Bed characterization (i.e., measurement of sediment height and solids concentration) was conducted to ensure about consistency of initial conditions between tests. Preliminary results for flushing study of 10% vol. kaolin-water mixtures starting from a fully-flooded initial condition showed that flushing with 1-line volume (65 gallons) could remove up to 72 percent of solids from the pipeline. (authors)

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DOE OSTI2017

Progress Updates for the Hanford Waste Treatment and Immobilization Plant Project - 17275

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DOE OSTI2023

Cesium Removal from 5.5 and 7.0 M Na AP-105 Using Crystalline Silicotitanate

The Tank Side Cesium Removal (TSCR) system, currently operational by Washington River Protection Solutions LLC (WRPS), prepares initial low-activity Hanford waste tank supernate feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility. In addition to entrained solids removal from the supernate, the primary goal of TSCR is to remove cesium-137 ( 137 Cs) by ion exchange, allowing contact handling of the liquid effluent product at the WTP as governed by a waste acceptance criterion (WAC). Specific to 137 Cs, this requirement is < 3.18E-5 Ci 137 Cs/mole of Na. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B), has been selected as the ion exchange media for TSCR. CST is a non-elutable inorganic material that has demonstrated robust chemical, physical, and radiation tolerance while maintaining functionality. However, testing to date on actual tank waste samples has been limited to Na concentrations between 5 and 6 M Na (Fiskum et al. 2019a, 2021b and Westesen et al. 2021a, 2021b, 2022) while actual tank conditions can reach upwards of 9 M Na. Testing with feed from Hanford tank AP-105 incorporated testing at both 5.5 and 7 M Na in order to evaluate the impact of Na concentration on volume of waste processed before reaching the WAC. A 6-L volume of 5.5 M Na AP-105 and an 8-volume of 7 M Na AP-105 was processed through the Radioactive Waste Test Platform system, established at Pacific Northwest National Laboratory to support small-scale waste qualification efforts. The columns consisted of 6- to 9-mL CST beds (CST Lot 2002009604, sieved to screen out >30-mesh particles) placed in 1.5-cm-inner-diameter columns. Feed was processed at 1.9 bed volumes (BV) per hour; the flowrate, in terms of contact time with the CST bed, matched the expected flowrate at TSCR. Table S.1 and Figure S.1 summarize the measured AP-105 Cs load performance for each feed condition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Bacterial Productivity in a Ferrocyanide-Contaminated Aquifer at a Nuclear Waste Site

This study examined potential microbial impacts of cyanide contamination in an aquifer affected by ferrocyanide disposal from nuclear waste processing at the US Department of Energy’s Hanford Site in south-eastern Washington State (USA). We examined bacterial productivity and microbial cell density in groundwater (GW) from wells with varying levels of recent and historical total cyanide concentrations. We used tritiated leucine (3H-Leu) uptake as a proxy for heterotrophic, aerobic bacterial productivity in the GW, and we measured cell density via nucleic acid staining followed by epifluorescence microscopy. Bacterial productivity varied widely, both among wells that had high historical and recent total cyanide (CN−) concentrations and among wells that had low total CN− values. Standing microbial biomass varied less, and was generally greater than that observed in a similar study of uranium-contaminated hyporheic-zone groundwater at the Hanford Site. Our results showed no correlation between 3H-Leu uptake and recent or historical cyanide concentrations in the wells, consistent with what is known about cyanide toxicity with respect to iron speciation. However, additional sampling of the CN− affected groundwater, both in space and time, would be needed to confirm that the CN− contamination is not affecting the GW biota.

Plymale, Andrew (ORCID:0000000333070155)↗

DOE OSTI2024

Analytical Results of the Tank 51H Sludge Batch 11 As-Received Qualification Sample

Savannah River National Laboratory (SRNL) has been requested by Savannah River Mission Completion (SRMC) to qualify the next sludge batch, Sludge Batch 11 (SB11), for processing at the Defense Waste Processing Facility (DWPF). This report documents the first steps of the qualification process, characterization of the Tank 51H as-received qualification material. SRMC sent SRNL two 200 mL samples received from Tank 51H on the 12th of August 2024. The combined sample, HTF-51-24-85/86, was analyzed for the following: supernate and slurry density, weight percent solids, settling, chemical composition, and radionuclides.

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