SEARCH · Engineering Papers

Results for “WASTE PROCESSING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 343 records · Page 19

All sources NASA NTRS DOE OSTI

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 OSTI2015

Impact of glycolate anion on aqueous corrosion in DWPF and downstream facilities

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 other high level waste (HLW) facilities. The impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system has not been previously evaluated. A literature review had revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 OSTI2015

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 OSTI2016

Addressing the Legacy Waste Challenge - IAEA Activities in Support of Member States - 16304

The International Atomic Energy Agency (IAEA) promotes the safe and peaceful uses of nuclear energy including the safe management of radioactive waste. Although safe and effective radioactive waste processing technologies and practices are well established and widespread, the issue of legacy waste remains a challenge for many IAEA member states. The primary source of the world-wide legacy waste inventory has its origins in past nuclear weapons programs and associated research and development activities. Other sources include historical waste in old storage and disposal facilities that were either never intended to be a permanent end-state or are no longer considered adequate from a safety perspective. Additionally, many IAEA member states with less advanced nuclear energy programs, often have small inventories of legacy radioactive waste arising from medical, research and industrial activities but lack the processing capabilities to put these wastes into a safe form for storage and disposal. These wastes have the potential to create a safety and environmental hazard and pose a radioactive waste management challenge for many reasons including lack of characterization data and process knowledge, waste package degradation, or the presence of hazardous or mixed waste components. The Waste Technology Section (WTS) within the Department of Nuclear Energy at the IAEA has an extensive program of activities to support the safe retrieval, processing, storage and disposal of radioactive waste including legacy waste. These include publications, co-ordinated research projects, technical co-operation activities, and information exchange forums including workshops, technical meetings and virtual networks. This paper will present an overview of recent IAEA activities and accomplishments aimed as assisting member states with addressing their legacy waste inventories. (authors)

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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Accelerated Basin De-Inventory Maximum Fissile Estimate

The Accelerated Basin De-inventory (ABD) program increases processing of Spent Nuclear Fuel (SNF) in H-Canyon to accelerate the closure of L Basin. The dissolved SNF will be dispositioned to Sludge Batches that will be processing in the Defense Waste Processing Facility and converted to glass. The fissile glass loading directly impacts the amount of glass canisters made. This memorandum provides an estimated maximum quantity of total fissile expected to be transferred to a Sludge Batch to support increasing the fissile glass loading and thus reducing the number of canisters created across the lifetime of the ABD mission.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗