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DOE OSTI2023

Results for the September Bimonthly Calendar Year 2022 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the 2022 September bimonthly sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the 2022 September bimonthly sampling of Tank 50 were requested by SRMC personnel via a Task Technical Request (TTR). Details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is the TTR deliverable relating Salt Solution Analysis from the SRMC request. The following facts pertaining to the WAC are drawn from the analytical results, including analytical uncertainty, provided in this report. 1) WAC LIMITS and TARGETS were met for all analyzed chemical and radioactive contaminants for which the detection limits are below the WAC LIMITS and TARGETS; 2) Measured average concentrations of nitrate, nitrite and total mercury are approximately 24%, 9% and 12% of the WAC LIMITS, respectively; 3) Measured average concentrations of Tc-99 and I-129 are approximately 19% and 20% of the WAC LIMITS, respectively; 4) All other radionuclide average concentrations are at 4% or less of the WAC LIMITS and TARGETS.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes PE for high-temperature processing. After removal of the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using an Fe-based catalytic process. The PE-derived graphite anode in a lithium-ion coin cell showed a specific capacity of 345 mAh/g at 0.05C with an initial Coulombic efficiency of 87% and reversible capacity retention of ~100% at different current rates. It also showed a specific capacity of up to 313 mAh/g at 0.5 discharge/charge cycles per hour (0.5C) and Coulombic efficiency of 99.9% after 250 cycles, indicating excellent electrochemical performance as an anode material for lithium-ion batteries. This method illustrates that there are opportunities for upcycling large quantities of PE waste to produce graphite powders suitable for use in LIBs.

25 ENERGY STORAGE↗

DOE OSTI2022

Results for the March Bimonthly Calendar Year 2022 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the 2022 March bimonthly sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the 2022 March bimonthly sampling of Tank 50 were requested by SRMC personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is the TTR deliverable relating Salt Solution Analysis from the SRMC request.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing, and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create a stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes the PE for high-temperature processing. After removing the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using a Fe-based low-temperature (<1500 °C) catalytic process. The PE-derived graphite powder showed excellent electrochemical performance as an anode material for lithium-ion batteries (LIBs) with a capacity of up to 302 mAh/g at 0.5 discharge/charge cycles per hour (0.5 C) and capacity retention of 100% after 415 cycles. This method illustrates there are opportunities for upcycling large quantities of PE waste to produce graphite powders.

Huynh, Ngoc Tien [NETL Site Support Contractor, Na↗

DOE OSTI2020

Ion-Exchange Modeling of Crystalline Silicotitanate for Cesium Removal - 20283

The Tank Closure Cesium Removal (TCCR) system is a Savannah River Site (SRS) demonstration 'at-tank' process designed to remove {sup 137}Cs from the high-level aqueous tank waste so that the decontaminated solution can be disposed as low-level waste. Cesium is removed by ion exchange (IX) columns using engineered IONSIV{sup TM} R9120-B form of the Crystalline Silicotitanate (CST) media. The TCCR system is deployed at Tanks 10 and Tank 11 in the SRS H Tank Farm. Water is added to the salt-cake in Tank 10 H to dissolve it. The dissolved salt solution waste is pumped out of Tank 10H (feed tank), through filters and IX columns. The decontaminated salt solution is transferred to Tank 11 (receipt tank), and on to Tank 50H for final disposal in the Saltstone Production Facility. The current TCCR can accommodate lead-lag (two-column) or lead-lag-guard (three-column) configurations to optimize media utilization and achieve the target decontamination. To assist the TCCR operations, a parametric study was conducted to evaluate the impact of different parameters (e.g., column configurations (single column, two columns or three columns in series), waste characteristics, operating temperature, process flow rate, CST average particle size) on the IX column performance including CST bed utilization. The initial results indicate that the IX column performance is improved at slower process flow rate, at lower operating temperature, and with smaller CST average particle size. Multi-column configurations are recommended, because the single-column configuration does not utilize CST bed effectively. This paper demonstrates the versatility of the ion exchange modeling to evaluate the effects of CST characteristics and operational parameters on IX column performances. (authors)

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DOE OSTI2020

Multivariate Linear Models for In Situ Monitoring of Low-Activity Nuclear Waste Using Infrared and Raman Spectroscopy - 20218

This paper presents the use of in situ monitoring tools for molecular analysis of non-radioactive species found in the supernatant liquid and dissolvable salt cake of nuclear waste at Hanford. We have selected two in situ spectroscopic techniques that do not require any sample preparation or consumables: infrared (in the form of attenuated total reflectance Fourier transform infrared or ATR-FTIR) and Raman spectroscopy. This work aims to evaluate their performance for liquid waste simulants representative of the Hanford DFLAW (Direct Feed Low-Activity Waste) feed. In addition, we have evaluated the performance of four multivariate linear models that extract concentrations from spectral measurements for a data set that was designed to include key molecular species found in liquid low-activity waste. (authors)

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DOE OSTI2024

Bonding Environments and Radiation Stabilities of Phosphate Glasses

This report is a summary of some key take-aways presented during the Phosphate Round Table Workshop held in 2020, which was led by PNNL and ANL and funded by DOE-NE. Phosphate glasses have a wide range of commercial and industrial uses due to their unique optical, chemical, and physical properties. They are a candidate material for use as an advanced waste form matrix for immobilizing radioactive waste due to their unique ability to immobilize high fractions of alkali- and sulfur-rich streams in chemically durable glasses. This literature review provides an overview of the structure of phosphate glasses, waste-form related properties of interest (e.g., chemical durability and radiation stability), a summary of how specific composition ratios affect chemical durability (e.g., [Fe]/[P] ratio, [O]/[P] ratio, Fe 2+ /Fe 3+ ratio), as well different ways that phosphate glasses can be affected by radiation. All of these are of interest to the waste form community as phosphate glasses, primarily Fe-P-O glasses, are investigated for usage in immobilizing various types of nuclear waste including U.S. legacy wastes as well as a method for treating salt-based high-level wastes from molten salt reactors and pyroprocessing of used nuclear fuels.

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DOE OSTI2026

Online Alpha Monitoring of High Cs-137 Hanford and SRS High Level Waste with Tensioned Metastable Fluid Detectors

The Department of Energy’s Hanford and Savannah River Sites maintain millions of gallons of caustic supernate and salt high activity waste in their high-level waste (HLW) tank farm inventories. The Savannah River Site is currently treating this waste with a calixarene-based solvent extraction of Cs-137 to reduce these inventories. Hanford is employing an at-tank crystalline silico titanate (CST) based solid phase extraction methodology to reduce their liquid HLW inventories. Due to the high solubility of Cs-137 and the relative insolubility of the actinides in these caustic waste forms, the beta to alpha radioactivity ratio can often exceed six orders of magnitude in the feed solutions to these treatment processes. This unique characteristic leads to significant technical challenges in making rapid gross alpha measurements in the presence of the overwhelming beta, gamma, as well as dissolved sodium salt in these HLW matrices. Conventional radioanalytical techniques, such as liquid scintillation analysis or gas flow proportional counting require significant radiochemistry preparation prior to the radiometric measurements for gross alpha activity. These required pretreatments render these technologies untenable for rapid quantification of gross alpha activity that could be required to support on or at-line measurements ensuring a waste stream will meet regulatory requirements. The radiation measurement properties of Tensioned Metastable Fluid Detectors (TMFDs) have been studied by Purdue University’s Taleyarkhan research group for well over a decade. Fluids tensioned to the appropriate degree will rupture when struck by radiation, resulting in a measurable cavitation event. The negative pressure generating this tension can be adjusted by centrifugal rotation or by acoustic means in such a way that these cavitation events can be generated from alpha radiation but will not be generated by beta or gamma radiation. Purdue University and the Savannah River National Laboratory are currently collaborating to develop a gamma/beta blind, spectroscopic alpha measurement system based on the Tensioned Metastable Fluid Detector technology to provide a potential solution for performing rapid gross alpha measurements on these high gamma/beta sample matrices. Measurements using the Indirect Drive Acoustically Tensioned Metastable Fluid Detectors developed as part of this collaboration were performed with an alpha emitting radionuclide. Successful determination of gross alpha activity was observed, indicating a potential pathway for rapid gross alpha measurements in remote-handled shielded cells or in process situations requiring online alpha monitoring. Measurements using this system have been conducted on high beta activity solutions, demonstrating the beta blind capability of this system. Measurements are currently underway to test the system’s capability to measure gross alpha activity on Savannah River Site high level waste high Cs-137 samples that have been measured by the SRNL radiochemistry team. This work was supported by the DOE EM Technology Development program.

DiPrete, David [Savannah River National Laboratory↗

DOE OSTI2024

Simulant Development of Potential 200 West Area Waste Feeds

Preliminary planning for retrieval, qualification, and pretreatment of waste in Hanford’s 200 West Area (200W) has begun as part of the West Area Risk Management project. Experimental studies to technically mature pretreatment process operations will likely be needed because of the uniqueness of 200W waste. Pacific Northwest National Laboratory formulated five simulants to represent 200W-qualified feed based on the preliminary flowsheet provided by Washington River Protection Solutions, LLC. The simulant recipes were devised using applicable historical information as a reference point to support the use of the flowsheet waste vectors, which were combined into five distinct groups. These five groups formed the basis for the liquid composition targets that were adapted into recipes using charged-balanced salt species. The liquid phase recipes were batched in 1-L quantities and analyzed at Pacific Northwest National Laboratory. Once confirmed to be stable, the liquid solutions were tested for compatibility with candidate solid components. Specific solid components were recommended based on cross-examining the proposed solid phases in the flowsheet with relevant data from the literature. Mixtures of solid components were added to aliquots of the liquid batches and sub-sampled to measure particle size distribution. The measured distribution was compared to independently created benchmark distributions appropriate for each simulant. This process was iterated until a solid phase composition that resulted in a representative particle size distribution was found. After the final compositions were confirmed, a suite of chemical and physical characterization data was collected. This report describes the simulant basis, formulation methodology, laboratory measurements, and data collected for the recipes recommended to represent 200W waste feeds.

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DOE OSTI2021

Results for the Fourth Quarter Calendar Year 2020 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2020 (CY20) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) Limits and Targets. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter CY20 sampling of Tank 50 were requested by SRR personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals per Task 2 from the SRR request, will be obtained semi-annually for the 1QCY21 and 3QCY21 Tank 50 samples. However, data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals for this current 4QCY20 Tank 50 sample will be obtained and documented per the special analysis associated with the vault classification sample as described in the associated TTQAP.

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DOE OSTI2021

Tank 50 Salt Solution Sample (2QCY21)

In this Technical Report, the chemical and radionuclide contaminant results from the Second Quarter Calendar Year 2021 (CY21) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) Limits and Targets that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the Second Quarter CY21 sampling of Tank 50 were requested by SRR personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals per Task 2 from the SRR request, will be obtained semi-annually for the 1QCY21 and 3QCY21 Tank 50 samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Results for the Third Quarter Calendar Year 2021 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2021 (CY21) sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) Limits and Targets that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY21 sampling of Tank 50 were requested by SRR personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals per Task 2 from the SRR request, will be obtained semi-annually for the 1QCY21 and 3QCY21 Tank 50 samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Results for the Second Quarter Calendar Year 2021 Tank 50 Salt Solution Sample

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Cradle to grave: the importance of the fuel cycle to molten salt reactor sustainability

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

A newly proposed isotherm model to predict Cs exchange with crystalline silicotitanate in tank waste simulants

The Zheng Anthony Miller (ZAM) computer model, a multicomponent ion exchange model used to predict the exchange of Group I metals onto crystalline silicotitanate (CST), has historically been used to predict Cs distribution coefficients from Hanford and Savannah River Site (SRS) tank waste simulants. Comparison of experimentally determined Cs distribution coefficients from tank waste simulants with ZAM isotherm model predictions indicate overprediction of Cs and K distribution coefficients for simple and complex simulants with the engineered form of CST. Additionally, recent changes in chemical composition/manufacturing of IONSIV TM R9140-B have resulted in increased Cs capacity from high-salt, highly alkaline solutions. Here, this work served to assess different isotherm models and refine equilibrium parameters to develop a model that can be applied to Hanford and SRS tank waste Cs removal efforts. Toward this goal, the Campbell Westesen Peterson (CWP) model was developed. This model utilized the experimentally determined Cs capacity, and simplified ZAM equilibria expressions to include only the binary substitution of Cs + or K + on the Na + sites. Equilibrium constants for these equations were refined using experimentally determined distribution coefficients. Overall, the CWP model significantly improved our ability to predict both Cs and K loading capacity from complex matrices.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Results for the January 2022 Semiannual Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the January 2022 Semiannual sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the January 2022 semiannual sampling of Tank 50 were requested by SRMC personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRMC request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals per Task 2 from the RMC request, will be obtained semiannually for the January 2022 and July 2022 Tank 50 samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Results for the July 2022 Semiannual Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the July 2022 Semiannual sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the July 2022 semiannual sampling of Tank 50 were requested by SRMC personnel via a Task Technical Request (TTR) and details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is part of Deliverable 2 relating to Task 1 from the SRMC request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals per Task 2 from the SRMC request, will be obtained semiannually for the January 2022 and July 2022 Tank 50 samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing, and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create a stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes the PE for high-temperature processing. After removing the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using a Fe-based low-temperature (<1500 °C) catalytic process. The PE-derived graphite powder showed excellent electrochemical performance as an anode material for lithium-ion batteries (LIBs) with a capacity of up to 302 mAh/g at 0.5 discharge/charge cycles per hour (0.5 C) and capacity retention of 100% after 415 cycles.

Huynh, Ngoc Tien↗