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At least 559 records · Page 31

Saltstone Third Quarter Calendar Year 2021 (3QCY21) Toxicity Characteristic Leading Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled quarterly for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00023, Revision 0, two SDF waste form (saltstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the third quarter of calendar year 2021(3QCY21). One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the “Cement-Free grout sample”. Results from this technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Remediation (SRR). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The 3QCY21 saltstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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July 2022 Semiannual Saltstone Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Salstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00025, Revision 0, two SDF waste form (salstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the July 2022 Semiannual Toxicity Characteristic Leaching Procedure (TCLP) samples. One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the "Cement-Free grout sample." Results from this technical report support Task 2: 'Grout Leaching Analyses' of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using Toxicity Characteristic Leaching Procedure (TCLP). The July 2022 semiannual salstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for the non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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January 2022 Semiannual Saltstone Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00025, Revision 0, two SDF waste form (saltstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-Area premix material for the January 2022 semiannual Toxicity Characteristic Leaching Procedure (TCLP) samples. One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the “Cement-Free grout sample.” Results from this technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The January 2022 semiannual saltstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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January 2023 Semiannual Saltstone Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00027, Revision 0, two SDF waste form (saltstone) samples were prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample and Z-area premix material for January 2023 semiannual Toxicity Characteristic Leaching Procedure (TCLP) samples. One sample contained a Full Premix which included 10:45:45 (by weight) of cement, slag, and fly ash. The second sample contained 60:40 (by weight) of slag and fly ash only referred to as the “Cement-Free grout sample.” Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of each of the SDF waste forms was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The January 2023 semiannual saltstone (Full Premix) and the Cement-Free grout samples met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the SPF WAC that was in effect at the time of the tank sampling.

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Epsilon Metal Summary Report Fiscal Year 2011

The Epsilon-metal (ε-metal) phase was selected in FY 2009 as a potential waste form to for immobilizing the noble metals found in the undissolved solids + aqueous stream, and the soluble Tc from ion-exchange process, each resulting from proposed aqueous reprocessing. ε-metal phase is observed in used nuclear fuel and the natural reactors of Oklobono in Gabon, where the long-term corrosion behavior was demonstrated. This makes ε-metal a very attractive waste form. Last fiscal year, ε-metal was successfully fabricated by combining the five-metals, Mo, Ru, Rh, Pd and Re (surrogate for Tc), into pellets followed by consolidation with an arc melter. The arc melter produced fully dense samples with the epsilon structure. However, some chemistry differences were observed in the microstructure that resulted in regions rich in Re and Mo, and others rich in Pd, while Ru and Rh remained fairly constant throughout. This year, thermal stability (air), and corrosion testing of the samples fabricated by arc melting were the main focus for experimental work. Thermal stability was measured with a differential scanning calorimeter - thermogravimetric analyzer, by both ramp heating as well as step heating. There is clear evidence during the ramp heating experiment of an exothermic event + a weight loss peak both beginning at ≈700°C. Step heating showed an oxidation event at ≈690°C with minimal weight gain that occurs just before the weight loss event at 700°C. The conclusion being that the e-metal begins to oxidize and then become volatile. These findings are useful for considering the effects of voloxidation process. Three different pellets were subjected to electrochemical testing to study the corrosion behavior of the epsilon-metal phase in various conditions, namely acidic, basic, saline, and inert. Test was done according to an interim procedure developed for the alloy metal waste form. First an open circuit potential was measured, followed by linear polarization sweeps. The linear polarization sweep range was the Tafel equation was fit to the linear polarization sweep data to determine the corrosion rate of each pellet in each test solution. The average calculated corrosion rates of the three pellets according to solution conditions were: -1.91 x 10 -4 mm/yr (0.001 M NaOH), -1.48 x 10 -3 mm/yr (0.01 M NaCl), -8.77 x 10 -4 mm/yr (0.001 M H 2 SO 4 ), -2.09 x 10 -3 mm/yr (0.001 M NaOH + 0.01 M NaCl), and -1.54 x 10 -3 mm/yr (0.001 M H 2 SO 4 + 0.01 M NaCl). Three single-pass flow through (SPFT) test were conducted at a flow rate of 10 ml/day, at 90°C, and pH of 2.5, 7.0, and 9.0 for up to 322 days. Results of the tests indicate that dissolution rates were 5 x 10 -4 g m -2 d -1 at pH 9.0, 1.2 x 10 -4 g m -2 d -1 at pH 7.0, and 2 x 10 -4 g m -2 d -1 at pH 2.5. The sample used for the pH 7.0 SPFT test contains extra Re compared to samples used for the other two SPFT test, which came from a single pellet. The corrosion data measured this year indicate that the ε-metal phase is chemically durable. The two chemically different phases, but structurally the same, behave differently during dissolution according to the microstructure changes observed in both the electrochemical and in SPFT test. Characterization of the test specimens after testing suggests that the dissolution is complex and involves oxidative dissolution followed by precipitation of both oxide and metallic phases. These data suggest that the dissolution in the electrochemical and SPFT tests is different; a process that needs further investigation.

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Increased fissionable material loading of vitrified borosilicate glass in support of efficient disposition of nuclear materials at Savannah River - 15299

The US DOE is vitrifying liquid HLW, converting it into a glass form suitable for long term storage and disposal. The waste includes fissionable isotopes and, as a consequence, may possess an inadvertent criticality hazard. The technical basis for determining when an inadvertent criticality hazard exists is unclear. ANS criticality safety standards do not provide relevant subcritical safety limits and little experimental data exist to support development of such a limit. As a result, the DOE has implemented a conservative limit of 897 grams of fissile nuclides per cubic meter of glass. However, this limit has the potential to impede efficient disposal of various fissionable materials at the Savannah River Site. Work is being conducted to establish an alternate fissionable material limit which would be less restrictive, but still ensure criticality safety. A series of computer models are constructed and analyzed using a general purpose Monte Carlo N-Particle neutron transport code. First, a baseline model is established and analyzed. Next, numerous variations to the baseline model (i.e., perturbations) are analyzed with the code to understand the impact of the change. Based on the results, it is evident the atomic ratio of B-10 to equivalent Pu-239 would be an effective means to ensure subcriticality of HLW vitrified in a borosilicate glass matrix. The proposed limit, its technical basis, and all necessary application constraints are discussed. This proposed limit is expected to permit significantly higher quantities of fissile nuclides in the vitrified waste form. It will also account for several non-fissile, fissionable nuclides which are expected to add reactivity to the glass waste form. The proposed limit embodies large conservatisms intended to provide assurance of safety in spite of the significant uncertainties associated with its development. The single most important uncertainty arises from the inability to validate the calculations which form its technical basis. While they are done with the latest available nuclear data and well established analytical methods, they are, nonetheless, not validated as required for subcritical safety limits. (authors)

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Influence of multi-species solute transport on modeling of hydrated Portland cement leaching in strong nitrate solutions

The differences between the predictions of Fickian diffusion and Nernst-Planck electro-diffusion based reactive transport models are investigated in the context of an important practical problem – leaching of Portland cement in strong nitrate porewater solutions under saturated conditions. Numerical results are presented for three distinct porewater solutions: deionized water, ammonium nitrate solution, and diluted porewater from a cementitious low-activity nuclear waste form. Solute concentrations predicted by the two transport models show significant differences: as much as a factor of two for ammonium nitrate leaching and as much as a factor of four for waste form leaching; whereas, negligible differences are observed for deionized water leaching. The major conclusion is that Fickian diffusion may be accurate enough for deionized water leaching, but electro-diffusion should be considered in the reactive solute transport modeling of ammonium nitrate decalcification and in the degradation modeling of cementitious materials exposed to nuclear waste form solutions.

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Used Fuel Disposition Campaign - Features, Events, and Processes (FEPs) (Work Package FTLL11UF0329), Level 4 Milestone (M4): M41UF032901 (LLNL Input to SNL L3 Milestone): FEP Status Tool and LLNL FEPs

The combination of 208 FEPs with the 5 waste forms and 7 geologic settings leads to a truly large parameter space for evaluation. Fortunately, many evaluations can handle multiple waste form and/or geologic setting situations for a given FEP. Also, some evaluations of a given waste form and/or geologic setting can apply to multiple FEPs. The FEP Status Tool (an Excel workbook) was developed to track progress in these evaluations. The current version (12.0) of the Status Tool is being released on the SharePoint website (Ref. 1) concurrent with this deliverable. The SharePoint Status Tool workbook can be downloaded for use by the FEP Team and other UFD staff members. However, LLNL staff will be the only people uploading changed Status Tool workbooks to SharePoint, to ensure version control. FEP Team members with updates to the Status Tool should provide the information to LLNL (J. Blink and V. Chipman). The information may be in e-mail form, or in an e-mail attachment. The attachment would typically be an Excel workbook with one sheet for each FEP with changed or new information. LLNL staff will copy the modified sheets into the master workbook, update the version number, and upload the new Status Tool workbook onto SharePoint. Section 1.1 of the report discusses the heart of the Status Tool, the individual FEP Sheets. Section 1.2 discusses the Summary Sheets of the Status Tool. Finally, Section 1.3 discusses some of the overall status results that can be obtained from the Summary Sheets.

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Vitrification Process for Idaho HLW Calcine - 18433

The majority of the liquid wastes produced from reprocessing of several types of spent nuclear fuel at the Department of energy's (DOE) Idaho site have been volume reduced by calcination. The resulting ∼155,600 ft{sup 3} of calcined waste is stored in 43 shielded stainless steel bins, which have a 500-year design life. The calcined material is both a high-level waste and a RCRA hazardous waste, which complicates treatment and disposal options. In addition to a potential direct disposal option, treatment options that have been considered include steam reforming, hot isostatic pressing (HIP), and vitrification. This paper describes testing and evaluation of a vitrification option for the calcine. Vitrification to produce a borosilicate glass has the advantage that it is the only treatment that provides a waste form that meets the current HLW repository waste acceptance criteria. It is therefore the lowest risk option in terms of regulatory and legislative concerns. The present work addressed several key questions: (i) Calcine waste loadings in glass, which determine the number of HLW canisters produced and the required glass production rate (throughput) for the system design; (ii) Melt rate per unit glass pool surface area of the melter, which determines the sizing of the vitrification system; and (iii) The ability to meet waste form acceptance requirements (PCT and TCLP) while achieving reasonable waste loadings. These are issues of fundamental importance to the sizing estimates for the unit operations and overall flow-sheet and they, therefore, directly impact the cost and schedule estimates. Fully compliant, high-waste-loading glasses were developed for two of the four Idaho calcine wastes. The results demonstrate that waste loadings of 35 wt% waste oxides are achievable for the two alumina calcines and about 30 wt% is achievable for the two zirconia calcines. Testing of these formulations in a joule-heated ceramic melter showed glass production rates of over 5000 kg/(m2/d) for the Alumina calcine and over 8500 kg/(m{sup 2}/d) for the Zirconium-Na Blend calcine. These extremely high production rates translate directly into smaller system size and shorter mission duration and, therefore, reduced costs. Based on these results, a preliminary design for the full-scale vitrification system was developed. The system consists of waste receipt, vitrification, off-gas treatment, and canister sealing and decontamination equipment. The calcine is blended with the appropriate glass forming chemicals (or glass frit) and fed to the melter as a solid stream. The melter has a glass pool surface area of about 2.5 m{sup 2} and is equipped with a bubbling system to increase processing rates. Glass production rates of about 6 metric tons per day are projected. The vitrification system would be installed into the existing Idaho Waste Treatment Unit Facility (IWTU) so that no new facilities are required to house the treatment system, thus greatly reducing overall cost, duration, and risk. The system would be designed to treat all of the calcine waste within a period of 12 years at an operating efficiency of 80%, producing a total of about 8500 canisters each weighing two metric tons. (authors)

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Analyses of iron phosphate glasses for dehalogenated salt waste

Specimens cut from series of glasses made at PNNL to represent a range of salt loadings in prototype composition DPF-5 and a range of iron oxide loadings in prototype composition DPF-3 were subjected to modified ASTM C1308 tests to assess the relative durabilities and retention of salt cations. Test results show both the intrinsic durability and salt cation retention decrease as the total amount of salt cations is increased, but waste forms made with up to 34 mass % salt waste loadings on a salt-cation basis are more durable than glass-bonded sodalite waste forms made with optimal salt waste loadings (about 6% on a salt-cation basis). Other tests are in progress to assess the effect of solution feedback on the long-term durability of these materials.

Ebert, W. L.↗

Time-Temperature-Transformation (TTT) Diagram for a Sludge Batch 9 Glass Composition Based on Coupled-Operation with the Salt Waste Processing Facility

The amorphous structure of a glass waste form has the potential to rearrange into crystalline phases at temperatures between the liquidus temperature and the glass transition temperature (Tg). Certain phases that can form will be detrimental to the durability of the glass and it is important to know the conditions that promote devitrification. The canister-centerline-cooling (CCC) profile is used to replicate the area within the center of the Defense Waste Processing Facility (DWPF) canister with the slowest cooling during the initial cool down after pouring, which has the greatest potential for crystallization. Other time-temperature conditions that cause significant changes in either phase structure or phase composition are identified by a time-temperature-transformation (TTT) study. The phase stability of a waste form must be determined as a part of the Waste Acceptance Product Specifications (WAPS) if it is to eventually be stored in a geologic repository. This requires the creation of a TTT diagram and analysis of the Tg, as defined by the Department of Energy (DOE). The previous TTT study for a DWPF glass waste form was completed in 2010 prior to coupled operation with the Salt Waste Processing Facility (SWPF). SWPF transfers two high activity waste streams to DWPF for vitrification: a cesium-containing strip effluent and a stream containing monosodium titanate/sludge solids. These SWPF streams were first transferred to DWPF for vitrification during Sludge Batch 9 (SB9) in 2021. The impact of these SWPF streams on crystallization behavior was not determined in previous studies and the need for data to satisfy WAPS Specification 1.4 for SB9 was identified.

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Design and Application of Materials for Sequestration and Immobilization of 99 Tc

99 Technetium ( 99 Tc) is a hazardous radionuclide generated by the nuclear industry that poses a serious environmental threat. The wide variation and complex chemistries of nuclear waste streams containing 99 Tc often create unique, site specific challenges when sequestering and immobilizing the waste in a matrix suitable for long-term storage and disposal. Therefore, an effective management plan for 99 Tc containing radioactive waste will likely require a variety of suitable materials/matrices capable of adapting to and addressing these challenges. In this review, we discuss and highlight the key developments for effective removal and immobilization of 99 Tc in inorganic waste forms. Specifically, we review the synthesis, characterization, and application of materials for targeted removal of 99 Tc from (simulated) waste solutions under various experimental conditions. These materials include: i) layered double hydroxides (LDHs), ii) metal-organic frameworks (MOFs), iii) ion-exchange resins (IERs) as well as cationic organic polymers (COPs), and iv) surface modified natural clay materials (SMCMs). Secondly, we discuss some of the major and recent developments towards 99 Tc immobilization in i) glass, ii) cement, and iii) iron mineral waste forms. Finally, we present future challenges that need to be addressed for the design, synthesis, and selection of suitable matrices for the efficient sequestration and immobilization of 99 Tc from targeted wastes. Finally, the purpose of this review is to inspire the researches on the design and application of various suitable materials/matrices for selective removal of 99 Tc present globally in different radioactive wastes and its immobilization in stable/durable waste forms.

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Application of Analytical Heat Transfer Models of Multi-layered Natural and Engineered Barriers in Potential High-Level Nuclear Waste Repositories - 12435

A combination of transient heat transfer analytical solutions for a finite line source, a series of point sources, and a series of parallel infinite line sources were combined with a quasi-steady-state multi-layered cylindrical solution to simulate the temperature response of a deep geologic radioactive waste repository with multi-layered natural and engineered barriers. This evaluation was performed to provide information to scientists and decision makers to compare candidate geologic media for a repository (crystalline rock [granite], clay, salt, and deep borehole), and to provide input for the future evaluation of the trade-off between pre-emplacement surface storage time, waste package size, and repository footprint. This approach was selected in favor of the finite element solution typically used to analyze the temperature response because it allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. The analytical solution approach used to analyze the repository temperature response allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. This approach allowed investigation of the sensitivity of the results to combinations of parameters that show that there is much flexibility to be gained in terms of spent fuel management options by varying a few key parameters. This initial analysis used representative design concepts and thermal constraints based on international design concepts, and it also included waste forms representing future fuel cycles with high burnup fuels. Unlike repository designs with large open tunnels and pre-closure ventilation, all of the disposal concepts analyzed in this study used enclosed emplacement modes, where the waste packages were in direct contact with encapsulating engineered or natural materials. The deep borehole repository concept limits the size of the SNF waste packages and may require rod consolidation to fit within the drill casing diameter. A single assembly waste package, assuming rod consolidation, was evaluated in the current analysis. Similar size restrictions apply for the HLW canisters. At this time no thermal constraints have been defined for the deep borehole repository concept. Representative EBS materials and properties were evaluated. However, changes in EBS design concepts and materials can also have significant effects on the maximum waste package surface temperature. Increased thermal conductivity of the buffer layer can be achieved by using an engineered buffer consisting of a mixture of graphite, sand, and bentonite [14]. One of the advantages of the analytical model is that it highlights the sensitivity of the results to the parameters that define the repository layout, including spacing between axial and lateral neighboring waste packages and drifts. It is clear that repository layout adjustments can be made to reduce the calculated peak temperatures. The results also show that significant reductions in required surface storage times can be achieved if higher thermal constraints can be justified Additional studies are planned to evaluate the trade-offs between surface storage times, repository layout parameters, and variations in EBS design concepts. Model validation and uncertainties will also be addressed. It is expected that shorter surface storage times and more optimized repository design configurations may be achieved. (authors)

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April 2024 Semiannual Composite Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from the Salt Waste Processing Facility (SWPF) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00027, Revision 0, one SDF waste form (saltstone) sample was prepared in the Savannah River National Laboratory (SRNL) from the SWPF Decontaminated Salt Solution (DSS) Waste Acceptance Criteria (WAC) sample and Z-area premix material for the April 2024 semiannual Toxicity Characteristic Leaching Procedure (TCLP) sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). After at least 28 days cured, a sample of the SDF waste form was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The April 2024 semiannual Cement-Free grout sample met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and met the SPF WAC that was in effect at the time of the tank sampling.

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October 2024 Semiannual Composite Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from the Salt Waste Processing Facility (SWPF) is sampled semiannually for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). Per request of customer, X-TTR-Z-00027, Revision 0, one SDF waste form (saltstone) sample was prepared in the Savannah River National Laboratory (SRNL) from the SWPF Decontaminated Salt Solution (DSS) Waste Acceptance Criteria (WAC) sample and Z-area premix material for the October 2024 semiannual Toxicity Characteristic Leaching Procedure (TCLP) sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR) prepared by Savannah River Mission Completion (SRMC). At 64 days cured, a sample of the SDF waste form was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). The October 2024 semiannual Cement-Free grout sample met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and met the SPF WAC that was in effect at the time of the sample collection at SWPF.

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Thermal investigation of nuclear waste disposal in space

A thermal analysis has been conducted to determine the allowable size and response of bare and shielded nuclear waste forms in both low earth orbit and at 0.85 astronomical units. Contingency conditions of re-entry with a 45 deg and 60 deg aeroshell are examined as well as re-entry of a spherical shielded waste form. A variety of shielded schemes were examined and the waste form thermal response for each determined. Two optimum configurations were selected. The thermal response of these two shielded waste configurations to indefinite exposure to ground conditions following controlled and uncontrolled re-entry is determined. In all cases the prime criterion is that waste containment must be maintained.

Wilkinson, C. L.↗

IMPACT OF THE SMALL COLUMN ION EXCHANGE PROCESS ON THE DEFENSE WASTE PROCESSING FACILITY - 12112

The Savannah River Site (SRS) is investigating the deployment of a parallel technology to the Salt Waste Processing Facility (SWPF, presently under construction) to accelerate high activity salt waste processing. The proposed technology combines large waste tank strikes of monosodium titanate (MST) to sorb strontium and actinides with two ion exchange columns packed with crystalline silicotitanate (CST) resin to sorb cesium. The new process was designated Small Column Ion Exchange (SCIX), since the ion exchange columns were sized to fit within a waste storage tank riser. Loaded resins are to be combined with high activity sludge waste and fed to the Defense Waste Processing Facility (DWPF) for incorporation into the current glass waste form. Decontaminated salt solution produced by SCIX will be fed to the SRS Saltstone Facility for on-site immobilization as a grout waste form. Determining the potential impact of SCIX resins on DWPF processing was the basis for this study. Accelerated salt waste treatment is projected to produce a significant savings in the overall life cycle cost of waste treatment at SRS.

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Milestone Report - M3FT-15OR03120215 - Recommend HIP Conditions for AgZ

The purpose of this study was to continue research to determine if HIPing could directly convert I-Ag 0 Z into a suitable waste form. Fiscal year (FY) 2015 work completed studies of Phase IIA, IIB, and IIC samples. Product consistency testing (PCT) of Phase IIA samples resulted in iodine release below detection limit for six of twelve samples. This is promising and indicates that a durable waste form may be produced through HIPing even if transformation of the zeolite to a distinct mineral phase does not occur. From PCT results of Phase IIA samples, it was determined that future pressing should be conducted at a temperature of 900°C. Phase IIC testing continued production of samples to examine the effects of multiple source materials, compositional variations, and an expanded temperature range. The density of each sample was determined and x-ray diffraction (XRD) patterns were obtained. In all cases, there was nothing in the XRD analyses to indicate the creation of any AgI-containing silicon phase; the samples were found to be largely amorphous.

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