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At least 1,243 records · Page 69

Application of machine learning for modeling brønsted-guggenheim-scatchard specific ion interaction theory (SIT) coefficients

Machine learning methodologies can provide insight into Brønsted-Guggenheim-Scatchard specific ion interaction theory (SIT) parameter values where experimental data availability may be limited. This study develops and executes machine learning frameworks to model the SIT interaction coefficient, ε. Key findings include successful estimations of ε via artificial neural networks using clustering and value prediction approaches. Additionally, applicability to other chemical parameters is also assessed briefly. Models developed here provide support for a use-case of machine learning in geologic nuclear waste disposal research applications, namely in predictions of chemical behaviors of high ionic strength solutions (i.e., subsurface brines).

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RANT Waste Container Receipt, Inspection, and Transfer

This procedure provides instructions for the receipt, inspection, and transfer of transuranic (TRU) waste containers within the Radioassay and Nondestructive Testing (RANT) Facility. Performance of this procedure meets the Surveillance Requirements (SRs) 4.1.1, 4.1.2, 4.1.3, 4.2.1, 4.2.2, 4.2.3 and 4.2.4.

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Summary of SRNL Support Activities to the DOE-ORP Enhanced Waste Glass Program for Fiscal Year 2024

In fiscal year 2024 (FY24) Savannah River National Laboratory (SRNL) continued tasked work for the Office of River Protection (ORP) to expand glass compositional regions accessible for low-activity waste (LAW) and high-activity waste (HLW) vitrification processing. Experimental work continued in four primary technical areas focused on processing and performance of glasses relevant to the Hanford missions. The data and results from this work will be used to expand and validate the glass models being developed at Pacific Northwest National Laboratory (PNNL) for waste processing and acceptance. This report summarizes the activities and deliverables associated with work performed in FY24.

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LDR Inorganic Spiked Simulants in Pozzolan-Substituted Cast Stone for Hanford Pretreated Low-Activity Tank Waste

Pretreated Low Activity Waste (PTW/LAW) at the Hanford Site, is currently projected for disposition through a cementitious waste form. The current standard formulation, Cast Stone, uses Class F fly ash, blast furnace slag, and Portland cement. However, the future availability of Class F fly ash is uncertain, as many coal-fired power plants—the source of this byproduct—are being decommissioned or converted to natural gas. This issue has prompted research into identifying and evaluating suitable alternative materials to replace fly ash in the cementitious waste forms used by the Department of Energy (DOE). The Savannah River National Laboratory (SRNL) successfully identified four natural pozzolans for the replacement of Class F fly ash in the current Savannah River Site (SRS) low-level waste (LLW) form, Saltstone.

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Ceramic waste form production from used nuclear fuel

According to one aspect of the invention, a method to create a ceramic waste form from used nuclear fuel. An active metal salt waste, a rare earth metal waste, and raw materials are received. The active metal salt waste is combined with the rare earth metal waste, forming a waste salt. The waste salt is then heated to approximately 500° C. The raw materials are also heated to approximately 500° C. The waste salt and raw materials are then blended to form a homogenous waste mixture. The homogenous waste mixture is heated to a first predetermined temperature for a predetermined amount of time, creating a ceramic waste form. The ceramic waste form is cooled to a second predetermined temperature.

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Influence of ethylenediaminetetraacetic acid on the long-term oxidation state distribution of plutonium

Here, spectrophotometry was used to study the effect of EDTA on plutonium oxidation state distribution as a function of time, pH, and ligand-to-metal ratio (L/M) under anoxic conditions. Novel Pu(V)-EDTA absorption bands were identified at 571, 993, 1105, and 1150 nm with molar absorption coefficients of 15 ± 1, 6 ± 1, 10 ± 1, and 10 ± 1 cm –1 M –1 , respectively. Pu(V)-EDTA spectral changes occurred at L/M < 1, indicating only Pu V O 2 (EDTA) 3- formed with logK = 3.6 ± 0.3. Time-resolved experiments showed EDTA drastically increased the Pu(V/VI) reduction rate, which we propose is driven by amine lone-pair electron donation and the oxidative decarboxylation of EDTA. Oxidation of Pu(III)-EDTA to Pu(IV)-EDTA occurred on a slower time scale (110–237 days) than previously reported (<15 min) and is hypothesized to be radiolysis driven. Pu(V/VI)-EDTA and Pu(III)-EDTA both approached Pu(IV)-EDTA stabilization over time, yet Pu(V/VI)-EDTA solubility data was ≥ 1.0 log 10 units higher than predicted by Pu(IV)-EDTA solubility models, indicating that current thermodynamic models are incomplete. Ultimately, the data show EDTA preferentially stabilizes Pu(IV) over time regardless of initial oxidation state, but Pu(V)-EDTA can persist under environmentally-relevant conditions, emphasizing the need to continue investigating redox reactions, speciation, and behavior of these complexes to support the transuranic waste disposal and surface remediation/containment efforts.

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Generation and Characteristics of Plutonium and Americium Contaminated Soils Underlying Waste Sites at Hanford

In 2012, a Hanford site-wide review on plutonium and americium geochemistry (Cantrell and Felmy 2012) identified remaining research challenges associated with defining scientifically defensible end-states and supporting remediation decisions. Three research challenges were identified in this review: 1. Determine the transformations of Hanford Site sediments in response to changes in waste/groundwater composition. Significant mineralogical transformations can take place when wastes contact Hanford sediments. These transformations can significantly impact the solubility and acidic adsorption of plutonium and americium and possibly result in the generation of pseudo colloids that could facilitate plutonium/americium migration. 2. Assess the impact of changes in waste/groundwater chemistry on the potential for plutonium/americium solubilization or colloid formation. Changes in waste or groundwater chemistry can greatly impact the chemical form or speciation of plutonium and americium. Knowledge of plutonium and americium speciation will be crucial in terms of evaluating the potential for solubilization, adsorption or remobilization of adsorbed complexes, colloid formation, and colloid interactions with sedimentary minerals. 3. Establish the role of organic complexants and/or non-aqueous solvents in the transport of plutonium/americium in the deep subsurface. Plutonium has been found to be associated with TBP and the presence of non-aqueous solvents, at least in certain Z-9 sediments. It will be important to establish the role of non-aqueous solvent in past movement of plutonium and its potential role in future mobility. This report provides the technical basis for interpreting data obtained from field samples and will allow waste chemistry impacts on plutonium/americium re-mobilization mechanisms to be assessed. This report helps address the three previously identified research challenges – sediment transformations, solution chemistry impacts, and the role of organic complexants and solvents – in plutonium and americium contaminated Hanford sediments. Ultimately, these and potential future studies will lead to a defensible description of plutonium and americium mobility in relation to waste site conditions that can be interpreted to balance the risks posed by their migration and envisioned remediation activities.

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Impacts of Fast Critical Assembly Fuel Discards on Liquid Waste Processes

The Savannah River Nuclear Solutions (SRNS) Fast Critical Assembly (FCA) mission is reestablishing the electrolytic dissolver for processing of Pu and Pu-U materials clad in stainless steel (SS). H-Canyon is planning to dissolve and neutralize FCA fuel without recovering the special nuclear material (i.e., Pu) prior to discarding to the Concentration, Storage, and Transfer Facilities (CSTF) operated by the Savannah River Mission Completion (SRMC) Liquid Waste (LW) Organization. The FCA discards will be combined with sludge in the CSTF after Low Temperature Aluminum Dissolution (LTAD) if needed. The combined waste will be washed, concentrated, and vitrified at the Defense Waste Processing Facility (DWPF). The high level waste canisters produced will be temporarily stored in the Glass Waste Storage Buildings prior to transfer to a future federal repository. Decants from LTAD and sludge washing will be combined with DWPF recycle and dissolved salt cake to prepare salt batches for processing in the Salt Waste Processing Facility. The resulting decontaminated salt solution will be processed in the Saltstone Production Facility.

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Reassessment of Dose Contributions from the Paducah Cask Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater and inadvertent human intruder screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in the addition of U-235 being brought back into the list of radionuclides requiring inventory limits within the Paducah Cask special waste form (SWF). In addition, while reviewing the earlier vadose zone transport analysis efforts, an inconsistency was observed in the initial U-235P inventory that was supplied to the Consolidated Waste Tracking System (CWTS) back in 2009.

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Impact of feed variability on cesium removal with multiple actual waste samples from the Hanford site

Here, the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions (WRPS), will send initial low-activity Hanford waste tank supernate feeds to 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 WTP. Crystalline silicotitanate (CST) ion exchange media, manufactured by Honeywell UOP (product IONSIV™ R9140-B), was selected as the ion exchange media at TSCR. This lot of material was found to have superior performance in comparison to historic batches of CST. Ion exchange column and batch contact testing with supernate from Hanford tanks AP-105, AP-107 and AW-102 was performed to assess the impact of feed variability on system performance. These tests demonstrated that batch contact measurements provide a reasonable prediction of column capacities with some deviation in performance attributed to column dynamics. In addition, the variability in CST capacity for cesium in the various actual waste samples does not track with historical understanding of the competing cations, suggesting that tank waste samples contain other components that may significantly impact cesium loading.

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LANL's Waste Management Transportation Program [Slides]

Los Alamos National Laboratory (LANL) has developed a safe, compliant, and cost-effective transportation program that moves over 5,000 m3 of regulated waste off site each fiscal year. This program handles all waste types including universal waste, New Mexico Special Waste, TSCA and RCRA wastes, low level and mixed low-level wastes, and TRU and MTRU wastes. This program allows for the treatment and disposal of hazardous wastes while maintaining the protection of people, facilities, and the environment This program provides critical support to a wide variety of missions across the laboratory site.

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Fluorine Limits and Impacts in High-Level Waste Glass Compositions

The impact of elevated fluorine (F) content on Hanford high-level waste (HLW) glasses has not previously been studied in detail. This effort represents the first systematic study to determine what F concentration limits should be used for the design of alkali-borosilicate-based Hanford Waste Treatment and Immobilization Plant (WTP) HLW glasses, and to document the technical basis for that limit. If alkali borosilicate glass made from Hanford HLW can accommodate a large amount of F, the large capital costs for complex sludge washing facilities may be avoided, as would much of the operational costs and negative schedule impacts associated with handling the large volumes of water required to dissolve these salts. In order to determine a limit for F in likely HLW glass compositions, an evaluation was conducted on glasses with F ≤ 0.90 mass% from previous nuclear waste glass studies. The collected dataset contains 239 glasses (232 HLW glasses and 7 LAW glasses) including 109 glasses with 0.9 ≤ F mass% ≤ 2.5, 116 with 2.5 < F mass% ≤ 8.0, and 14 with F mass% ≥ 8 (max. F mass% = 17.42). The collected composition and property data were analyzed to determine the basis for the F tolerance, i.e. the maximum F concentration that can be processed without potential issues. Fluorine volatility, product consistency test (PCT) response, liquidus temperature (T L ), glass melt viscosity, and crystallinity have been investigated. No limits for F concentration can be made based on F volatility, T L , or glass melt viscosity, because the data show that high F in glasses do not indicate, with high probability, being restricted by those property constrains. However, crystallinity and PCT response were used to estimate the F tolerance. The results show that glasses with high F (≥ 0.90 mass%) are more likely to form large fractions of F-containing crystal phases which may increase PCT responses, i.e. decrease the glass durability. Based on the results of crystallinity and PCT data, the F tolerance of 4.5 mass% is estimated. There is no evidence of other glass components, such as calcium oxides and alkali metal oxides have combined impacts with F on the glass properties. Overall, the available high-F glass data is limited, especially in the designed HLW glass composition regions. Future work on formulation and testing of HLW glasses with F ≥ 0.9 mass% will close the data gaps and expand operational flexibility with respect to the fluoride tolerances. Volatility of F from melters and corrosion of materials in contact with glass melts are important for processing of high-F wastes; yet no test data are currently available. It is recommended tests be conducted to address these two potential issues.

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Supplemental Information Regarding the Application for Remediation of the Flanged Tritium Waste Containers at Los Alamos National Laboratory

This letter is in response to the October 18, 2021 letter from David Gray, Acting Regional Administrator, to Mr. Theodore Wyka, Manager of the National Nuclear Security Administration’s Los Alamos Field Office (NA-LA). That letter requested additional information regarding the Application for remediation of the Flanged Tritium Waste Containers (FTWCs) at Los Alamos National Laboratory (LANL) Technical Area 54, Area G. The Emissions Management Plan for the FTWC Venting Project is attached.

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Enhanced prediction of Cs removal by CST from Hanford tank waste with K accountability

Abstract The treatment of Hanford tank waste is one of the most challenging environmental cleanup activities to date. To expedite the processing of liquid waste stored in underground tanks in Washington State it is necessary to remove the significant dose contributor, 137 Cs. Crystalline silicotitanate ion exchanger is currently used to remove 137 Cs from the aqueous phase of Hanford tank wastes in preparation for vitrification at the Waste Treatment and Immobilization Plant (WTP). Improving the understanding of potassium impacts on ion exchange behavior of Cs will help in the operation of a critical component of one of the most complex treatment processes in the world today. Optimization of this process can result in significant cost savings and less waste production. Toward this effort, a series of batch contact tests varied in potassium concentration were conducted to look at the impact of potassium concentration on Cs distribution. Experimental distribution ratios ( K d ) were compared to the distribution ratios predicted using the ZAM model. A significant underprediction of Cs capacity in the presence of potassium was seen with the existing model. A revision of the equilibrium constants was determined and provided a statistically better fit for determining the Cs K d values in tank waste matrices.

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