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At least 73 records · Page 4

An Evaluation of the Impact of Glycolate and Glycolate Mitigation on the Defense Waste Processing Facility Recycle Diversion Project Flowsheet

The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) processes radioactive High Level Waste (HLW) sludge solids from the Concentration, Storage, and Transfer Facilities (CSTF); which includes the SRS Tank Farm and Evaporator facilities, and a concentrated Cs-137 laden stream and a Monosodium Titanate (MST) and sludge solids stream from the SRS Salt Waste Processing Facility (SWPF). The waste is chemically adjusted with acids and reductant (currently with 50 wt.% nitric acid and ~90 wt.% formic acid, but eventually formic acid will be substituted with ~70 wt.% glycolic acid), and frit is added so that a durable, borosilicate glass waste form can be produced when the material is vitrified in the melter. As a result of the evaporation of water during both the melter feed preparation and the melter feed vitrification steps in DWPF, a recycle waste stream is generated, neutralized, and sent back to the CSTF. The recycle waste is a dilute aqueous stream originating from the collection of condensate liquids containing some minor sludge, MST, and frit solids and other waste components resulting from melter feed entrainment during foamover events and transfer of volatile species into the condensate. The recycle stream volume is significant and is expected to approach 3 million gallons per year once SWPF reaches full operation, requiring the use of multiple large CSTF tanks for storage. The recycle waste is currently collected in the SRS Tank Farm and periodically evaporated in the 242-16H (2H) Evaporator to conserve storage space.

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Plutonium Solubility and Supernate Concentration for Neutralized Fast Critical Assembly Discards to Savannah River Site Tank Waste

The Savannah River Site (SRS) plans to dissolve non-irradiated stainless steel (SS)-clad bundles of Fast Critical Assembly (FCA) materials in eighteen batches.1 FCA dissolution is currently underway in the 6.3D dissolver by simultaneous chemical and electrolytic dissolution, which is required to generate the harsh conditions necessary for dissolution of metal-oxide (MOX) and non-aluminum spent nuclear fuels (NASNFs).2 Nitric acid and potassium fluoride are used to promote chemical dissolution.2 Gadolinium will be added during processing as a thermal neutron poison for criticality control. There are no plans for recovering plutonium from this waste stream. After FCA dissolution, the acidic (HNO3/KF) “discards” containing the dissolved metals will be neutralized by addition of 50 wt% sodium hydroxide to a final free hydroxide concentration of 1.2 M.1 Neutralization will precipitate a slurry of insoluble solids, predominantly metal oxides/hydroxides of plutonium, uranium, and SS components. Small fractions of the SS components, Pu, U, and Gd will remain dissolved in the supernate. The neutralized slurry will be composited to existing radioactive waste storage tanks within the SRS Concentration, Storage, and Transfer Facilities (CSTF) containing other similar sludge batch (SB) materials.1 The fate of soluble plutonium and freshly-precipitated, colloidal plutonium from this process are of concern since the total Pu can challenge the waste acceptance criteria (WAC) at the downstream SRS Liquid Waste (LW) facility. Supernate decants including the neutralized FCA discards (nFCAd) within the CSTF will be composited with salt batch (StB) materials and transferred to the SRS Salt Waste Processing Facility (SWPF), where total plutonium is also of concern.

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Application of Process Chemical Modeling to Optimize Radioactive Waste Disposal at the Savannah River Site - 24242

The Technical Optimization Model (TOM) is used by Savannah River Mission Completion (SRMC) to carry out facility-wide material balance and validate chemistrydependent processes for the purpose of their System Plan.The TOM simulates material movement and chemical reactions at the Tank Farm (TF), Salt Waste Processing Facility (SWPF) and Defense Waste Processing Facility (DWPF).

Georgiou, Andreas↗

Selection of Glasses to Confirm the 0.65 Weight Percent Sulfate Solubility Limit for Sludge Batch 10

In preparation for Sludge Batch 10 (SB10) processing, projections of sulfate (SO 4 2- ) in glass at 36% waste loading (WL) were calculated in May 2020 for Tank 40 blend projections representing 0.7M and 0.85M Na wash endpoints. The projected SO 4 2- ) concentrations for either sludge-only (SO) or coupled processing with the Salt Waste Processing Facility (SWPF) were either near or exceeded the current Sludge Batch 9 (SB9) limit of 0.65 weight percent (wt.%). Four nominal glass compositions were selected based on SO and coupled processing for the 0.85M Na wash endpoint Tank 40 blend projection to conduct an initial evaluation of the SB10 sulfate solubility behavior. A sulfate salt phase was absent from each of the prepared glasses, which provided preliminary results that supported the 0.65 wt.% SO 4 2- ) limit for SB10.

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Measurement Acceptance Region (MAR) Assessment Results Based on Confirmation Sample Sludge Batch 10 Projections

The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, 26F, 40H (heel only), and Alternate Feed Stock-2 and Sodium Reactor Experiment material from H-Canyon. Frit 473 was recommended for sludge-only (SO) and coupled processing with the Salt Waste Processing Facility (SWPF) based on previous assessments of SB10 projections with the DWPF Product Composition Control System (PCCS) glass property models and their associated Measurement Acceptance Region (MAR) constraints. Savannah River Mission Completion (SRMC) subsequently pursued Wash Cycle Y to further reduce the total sulfur in the sludge batch and increase processing flexibility at DWPF. In October 2022, SRMC System Planning provided an updated SB10 Tank 40H blend baseline projection based on the Tank 51 confirmation sample results. Due to the reduced Na concentration relative to previous SB10 projections, two additional projections based on the addition of 7,000 gallons of caustic were provided in November 2022.

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Evaluation of the 0.65 Weight Percent Sulfate Solubility Limit for Sludge Batch 10 Based on November 2022 Projections

The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, 26F, 40H (heel only), and Alternate Feed Stock-2 and Sodium Reactor Experiment material from H-Canyon. Frit 473 was recommended for sludge-only (SO) and coupled processing with the Salt Waste Processing Facility (SWPF) based on previous assessments of SB10 projections with the DWPF Product Composition Control System glass property models and their associated Measurement Acceptance Region constraints. Savannah River Mission Completion (SRMC) subsequently pursued Wash Cycle Y to further reduce the total sulfur in the sludge batch and increase processing flexibility at DWPF. In October 2022, SRMC System Planning provided an updated SB10 Tank 40H blend baseline projection based on the Tank 51 confirmation sample results. Due to the reduced Na concentration relative to previous SB10 projections, two additional projections based on the addition of 7,000 gallons of caustic were provided in November 2022.

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Measurement Acceptance Region (MAR) Assessment Results Based On Sludge Batch 10 Projections from December 2021, February 2022 and August 2022

The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, and 26F, Alternate Feed Stock-2 and Sodium Reactor Experiment material from H-Canyon. Frit 473 (8B 2 O 3 -8Li 2 O-5Na 2 O-79SiO 2 , in weight percent) was recommended for sludge-only (SO) and coupled processing with the Salt Waste Processing Facility (SWPF) based on previous assessments of SB10 projections with the DWPF Product Composition Control System (PCCS) glass property models and their associated Measurement Acceptance Region (MAR) constraints. Due to the lower processing rate of Sludge Batch 9 (SB9), the heel in Tank 40 is anticipated to be greater than 40 inches at the projected start of SB10 processing to meet the upcoming Accelerated Basin Deinventory addition dates in Tank 51 for Sludge Batch 11. In December 2021 and February 2022, Savannah River Remediation System Planning provided updated SB10 Tank 40 blend projections based on heels of 74 inches, 84 inches, 94 inches and 103.4 inches. Savannah River Mission Completion (SRMC) subsequently pursued Wash Cycle Y to further reduce the total sulfur in the sludge batch and increase processing flexibility at DWPF. In August 2022, SRMC System Planning provided an updated SB10 Tank 40 blend projection with a 76 inch-heel representing Decant Y2 and 60 kilogallons of bearing water inleakage that is anticipated during the Tank 51 to Tank 40 transfer. The objectives of this task were to: • Determine the impact on the operating windows for SO and coupled processing • Determine whether any composition gaps exist between the already completed SB10 variability study and the reprojected SB10 glass composition region • Compare the SB10 reprojected glass composition region to the DWPF PCCS model development and validation ranges to ensure that compositional gaps do not exist between the data sets. This report documents the results of these evaluations. Calculation-based frit assessments were performed using the DWPF PCCS glass property models and their associated MAR constraints. Evaluated parameters for coupled processing included the following transfer volumes per Sludge Receipt and Adjustment Tank batch: 5700 gallons of Tank 40 sludge, 2400-4500 gallons of the SWPF monosodium titanate (MST) and sludge solids stream, and 15,000 gallons of strip effluent. Based on these MAR assessment results, Frit 473 remains viable for SB10 processing. A target waste loading (WL) of 36% is possible for SO operation and single strike (0.4 g/L MST) coupled processing up to 600 mg/L of SB9 insoluble sludge solids at a nominal transfer volume of ~2800 gallons. Increasing the single strike transfer volume to 4500 gallons may reduce the maximum WL below 40%. Operating windows are 12 percentage points for SO processing and 14-16 percentage points for coupled processing. Frit 625 allows for a target WL of 36% and is acceptable for use during the SB9 to SB10 transition to deplete remaining inventory as needed. Operating windows are 8 percentage points for SO processing and 12-14 percentage points for coupled processing. Predictive PCCS evaluations performed at DWPF will provide insight into batch-specific acceptability at desired WLs for compositions having expected oxide ratios during processing versus the extreme vertices (corner points) evaluated in this study. The reprojected SB10 glass composition region generally overlaps the previously evaluated SB10 variability study composition region. Thus, the minor composition shift of these updated SB10 projections indicates that no additional glasses are necessary to demonstrate acceptability relative to the chemical durability of the Environmental Assessment benchmark glass and predictability using the current PCCS models for durability. Based on a comparison of the PCCS model development and validation data to the reprojected SB10 glass composition region, the viscosity and liquidus temperature models will reliably predict SB10 compositions. No additional glasses are necessary to demonstrate predictability of these models.

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Uranium Solubility and Supernate Enrichment Testing for Accelerated Basin De-Inventory Discards to Concentration, Storage, and Transfer Facilities

After future Accelerated Basin De-inventory (ABD) enriched uranium discards into sludge batches (SB), portions of supernate decants during SB preparation will be blended into salt batches (StB) and the resulting feeds must meet the nuclear safety requirements for the Salt Waste Processing Facility (SWPF). During previous SB10 sampling and testing, which involved H-Canyon material containing enriched uranium being mixed with Tank 51 sludge shortly after the H-Canyon stream was neutralized, it was identified that the uranium isotopic enrichment in the supernate deviated from the uranium isotopic enrichment in the slurry. The higher uranium isotopic enrichment in the supernate introduced the risk of challenging the feed requirements of SWPF. The ABD material added to SB11 was isotopically diluted with depleted uranium, mitigating any downstream impacts. However, H-Canyon desires to eliminate or minimize future depleted uranium additions in order to meet the mission schedule.

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Investigation of Mercury Compound Partitioning Through the Solvent Extraction System

Savannah River National Laboratory (SRNL) has been requested by Savannah River Mission Completion (SRMC)/ Salt Waste Processing Facility (SWPF) personnel to investigate the potential routes for titanium and mercury accumulations within the SWPF Flowsheet. This report documents work performed to examine how various forms of mercury can migrate through the multiple flowsheets of a solvent extraction system. This entailed 29 single stage distribution tests, a multi-stage Extraction-Scrub-Strip (ESS) test, and a detailed speciation analysis of several salt batch feed samples.

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Evaluation Of Glass Density to Support the Estimation of Fissile Mass Loadings in Sludge Batch 10 Glasses

Per a directive from the Department of Energy Savannah River Operations Office (DOE-SR) in 2008, the fissile mass loading concentration must remain below 897 g/m 3 in each high-level waste (HLW) glass canister produced by the Defense Waste Processing Facility (DWPF). To support Sludge Batch 5 (SB5) processing, the Savannah River National Laboratory (SRNL) developed a technical basis that facilitates the evaluation of fissile mass loading of the glass product. The calculation is based on the iron (Fe) concentration in the glass as determined by measurements from the Slurry Mix Evaporator acceptability analysis as well as the glass density. In April 2022, a subsequent DOE-SR directive increased the fissile mass loading limit to 2500 g/m 3 beginning with Sludge Batch 11. Thus, the 897 g/m 3 limit still applies to Sludge Batch 10 (SB10) processing. For SB5 through initial Sludge Batch 9 (SB9) processing prior to coupled operation with the Salt Waste Processing Facility (SWPF), SRNL provided DWPF a bounding glass density value that was based on a statistical evaluation of density measurements. To eliminate the need for experimental work, a composition-based density model for HLW glasses was developed at SRNL in 2019. The objective of this report is to present the bounding glass density determined with the composition-based density model for SB10 sludge-only (SO) and coupled processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Antifoam Development for Eliminating Flammability Hazards and Decreasing Cycle Time in the Defense Waste Processing Facility

Foam, due to the high gas generation rates of boiling and chemical reaction offgasing, requires control measures to prevent the foam from contaminating the condensate and to facilitate efficient plant operation. Antifoam was utilized to minimize foam production during chemical processing in the DWPF and during High-Level Waste (HLW) evaporation at SRS and Hanford. However, the current antifoam used in the SRS DWPF increases flammability risk during chemical processing (generates three flammable degradation products) and while feeding the melter (can decompose to CO/hydrogen). It is also the likely source of methyl functional groups for the organo-mercury present in the tank farm and excessive mercury in Saltstone. Additionally, the planned startup of Salt Waste Processing Facility (SWPF), with much higher throughput, will challenge DWPF to process at higher gas generation rates. DWPF employs Antifoam 747, a superspreader produced by Momentive Performance Materials, as an antifoaming agent during waste processing. During DWPF chemical processing, antifoam must be effective up to boiling (i.e., up to 103°C) and between a pH of 3-13. Antifoam 747 is most effective at a pH range of 6-8 and degrades as pH deviates. In addition, SRNL identified three flammable antifoam degradation products using mass spectrometer (MS) and fourier transform infrared (FTIR) offgas analyzers during simulations. A new antifoam or a new method to control foam is needed to minimize DWPF processing time and reduce the risk of contamination. In addition, testing should be completed to ensure that other antifoams used in HLW processing do not have similar flammability hazards or cause unintended impacts in downstream processing.

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Implementation of the Glycolate Destruction Process with Sodium Permanganate in the Defense Waste Processing Facility (DWPF) - 20141

The Savannah River Site (SRS) seeks to replace formic acid with glycolic acid as a chemical reductant in the Defense Waste Processing Facility (DWPF), where borosilicate glass is mixed with high-level radioactive waste, melted at high temperatures, then poured into stainless steel canisters for safe storage. Prior research and development have demonstrated the feasibility and advantages of the new nitric-glycolic acid flowsheet over the current nitric-formic acid flowsheet to chemically adjust the radioactive sludge slurry waste in the Sludge Receipt and Adjustment Tank (SRAT) prior to vitrification. Use of glycolic acid reduces hydrogen and ammonia generation during the reduction process, thus reducing flammability hazards and purge requirements in the DWPF processing vessels. In addition, glycolic acid will support the new Salt Waste Processing Facility (SWPF) by allowing for receipt of higher volumes of waste due to reduction of flammable gases, increased boil-up rates in the vessels, and easing the processing of future sludge batches containing high levels of mercury. However, during DWPF operations, trace amounts of glycolate are anticipated to be entrained in the recycle stream and sent to the Concentration, Storage, and Transfer Facilities (CSTF) via the Recycle Collection Tank (RCT). The RCT receives condensate from the Slurry Mix Evaporator Condensate Tank (SMECT) and Off-gas Condensate Tank (OGCT) that may contain unreacted glycolate. A literature review found the potential for hydrogen generation due to thermolysis of glycolate in caustic CSTF conditions. Savannah River National Laboratory (SRNL) conducted several tests on simulant and radioactive waste from various tanks at SRS and confirmed that glycolate added to waste at concentrations of approximately 1000 mg/L and higher generates hydrogen through thermolysis. There are potential operating scenarios in the future where glycolate concentration could exceed 1000 mg/L in the 2H Evaporator Drop Tank, the highest accumulation point for DWPF organics received via recycle. These findings presented a need to develop another process to mitigate the impact of glycolic acid to the CSTF. Several DWPF compatible options were evaluated for their ability to destroy glycolate in the DWPF recycle stream under various processing conditions, thereby mitigating its introduction to the CSTF. Downselection testing identified sodium permanganate as the most promising and simplest option to oxidize glycolate. It has been demonstrated to be effective in various RCT simulant compositions, operating temperatures, oxidant addition rates, solution pH, and initial glycolate concentrations. A feasibility study identified existing tanks within the facility that could be used for receiving and storing the commercially available sodium permanganate, as well as feeding it to the RCT for processing. The use of existing equipment as well as the ability to purchase the pre-mixed chemical reduces implementation and operational complexity. The downstream impacts of the sodium permanganate reaction were also evaluated and demonstrated to have minimal impact on other chemical species present in the RCT and CSTF, as well as on the materials of construction of the RCT and CSTF vessels and associated piping and instrumentation. Future work is in progress to validate the glycolate destruction process with testing in actual radioactive RCT waste before implementation in DWPF. (authors)

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Impacts of the Addition of Sodium Reactor Experiment (SRE) and DR-3 Fuel from H-Canyon to Tank 40 on Acid Stoichiometry and Hydrogen Production

H-Canyon plans to discard Sodium Reactor Experiment (SRE) material that is currently stored in Tanks 16.3 and 16.4. Savannah River Remediation (SRR) determined that a portion of the SRE material can be added to Sludge Batch 9 (SB9) in Tank 40 and the remainder will be added to Sludge Batch 10 (SB10) that is currently being prepared in Tank 51. A processing flowsheet for both the nitric-formic acid flowsheet and the nitric-glycolic acid flowsheet were previously developed for SB9 sludge-only and coupled operation with the Actinide Removal Process – Modular Caustic Side Solvent Extraction Unit (ARP-MCU) and the Salt Waste Processing Facility (SWPF). Due to the compositional change in Tank 40 after the addition of SRE material, it is necessary to determine the influence on the acid stoichiometry operating window and hydrogen production for SB9 processing. An assessment was completed to address the influence of an addition of SRE and DR-3 Fuel from H-Canyon on the chemistry of Tank 40, the defined acid stoichiometry window (with and without SWPF streams) for the nitric-formic and nitric-glycolic acid flowsheets, and the resulting influence of the hydrogen production. Hydrogen production during the SRAT/SME cycles consists of thermolytic, radiolytic, and catalytic hydrogen. For the nitric-formic acid flowsheet, catalytic hydrogen dominates the hydrogen production during DWPF operation mode for the SRAT and SME vessels. This report was written to summarize that assessment. No SRNL testing is needed prior to adding up to 7,100 additional gallons of SRE and DR-3 Fuel from H-Canyon to Tank 40. Also, the addition of SRE will not exceed the Shielded Cells reported values for catalytic hydrogen for either the nitric-formic acid flowsheet (Technical Safety Requirement (TSR) limit of 0.15 lb/h) or the nitric-glycolic acid flowsheet (TSR limit 0.024 lb/h). The addition of the ARP stream from SWPF and the use of the nitric-glycolic acid flowsheet are both expected to decrease hydrogen generation, leading to a larger safety margin in the CPC.

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Tank 49H Solids Disturbance Analysis

Tank 49H will serve as the feed tank for the Salt Waste Processing Facility (SWPF). Transfers into Tank 49H may disturb solids that have settled to the bottom of the tank, resulting in feed that may exceed the insoluble solids content limit of 1200 mg/L from the SWPF Waste Acceptance Criteria (WAC). During a transfer into Tank 49H, material that free falls from the Tank 49H B4 downcomer and passes through the supernate could potentially disturb the solids on the bottom of the tank and scour or suspend solids from a settled solids layer or turbid region. The scope of this task is to perform fluid flow analysis to determine the impact of a “plunging jet” through the B4 downcomer on solids in Tank 49H and to determine a minimum liquid level to be maintained in Tank 49H prior to transfers that will minimize disturbing the solids and exceeding the SWPF WAC limits for insoluble solids carryover to SWPF.

54 ENVIRONMENTAL SCIENCES↗

Tank 49H Solids Disturbance Analysis

Tank 49H will serve as the feed tank for the Salt Waste Processing Facility (SWPF). Transfers into Tank 49H may disturb solids that have settled to the bottom of the tank, resulting in feed that may exceed the insoluble solids content limit of 1,200 mg/L from the SWPF Waste Acceptance Criteria (WAC). During a transfer into Tank 49H, material that free falls into Tank 49H through a downcomer and passes through the supernate could potentially disturb the solids on the bottom of the tank and scour or suspend solids from a settled solids layer or turbid region. A previous analysis and report evaluated the potential to disturb solids when transferring into Tank 49H through the B4 riser and recommended a minimum tank level of 120 inches to prevent disturbing any solids in the bottom of the tank. The scope of this task is to perform additional fluid flow analysis to determine the impact of a “plunging jet” through the C3 downcomer on solids in Tank 49H and to determine a minimum liquid level to be maintained in Tank 49H prior to transfers that will minimize disturbing the solids and exceeding the SWPF WAC limits for insoluble solids carryover to SWPF since the C3 riser downcomer is at a lower elevation and farther away from the B5 transfer pump than the B4 downcomer.

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Tank 49H Solids Disturbance Analysis

Tank 49H will serve as the feed tank for the Salt Waste Processing Facility (SWPF). Transfers into Tank 49H may disturb solids that have settled to the bottom of the tank, resulting in feed that may exceed the insoluble solids content limit of 1,200 mg/L from the SWPF Waste Acceptance Criteria (WAC). During a transfer into Tank 49H, material that free falls into Tank 49H through a downcomer and passes through the supernate could potentially disturb the solids on the bottom of the tank and scour or suspend solids from a settled solids layer or turbid region. A previous analysis and report evaluated the potential to disturb solids when transferring into Tank 49H through the B4 riser and recommended a minimum tank level of 120 inches to prevent disturbing any solids in the bottom of the tank. The scope of this task is to perform additional fluid flow analysis to determine the impact of a “plunging jet” through the C3 downcomer on solids in Tank 49H and to determine a minimum liquid level to be maintained in Tank 49H prior to transfers that will minimize disturbing the solids and exceeding the SWPF WAC limits for insoluble solids carryover to SWPF since the C3 riser downcomer is at a lower elevation and farther away from the B5 transfer pump than the B4 downcomer.

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Literature Review on the Impact of Thermolytic Hydrogen Generation Rate from the Next Generation Solvent

A literature review was undertaken to assess the thermolytic hydrogen generation potential of the Next Generation Solvent (NGS) system, which is planned to replace the Original Caustic-Side Solvent Extraction (CSSX) Solvent at the Salt Waste Processing Facility. NGS uses a new extractant and suppressor and the same modifier and organic carrier solvent as the Original CSSX Solvent at different concentrations.

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A Review of the Literature on the Radiolytic Stability of the Next-Generation Solvent

The Original Caustic-Side Solvent Extraction (CSSX) solvent (based on the BOBCalixC6 extractant) currently in use at the Salt Waste Processing Facility (SWPF) is set to be replaced by the Next-Generation Solvent (NGS, based on the MaxCalix extractant). Current SWPF flammability safety controls use G-values for volatile gases obtained from radiolysis studies with the Original CSSX solvent. A review of past irradiation tests of both solvent systems was performed to assess the radiolytic stability of NGS with respect to flammable gas generation from radiolysis.

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