Synthesis and characterization of sintered H–Y zeolite-derived waste forms for dehalogenated electrorefiner salt
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Gardner, Levi D.↗
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Investigation of potential causes of production limiting fouling within SWPF due to mercury solids formation during pH swings within the SWPF process
The Salt Waste Processing Facility (SWPF) goal to investigate the use a high sodium Salt Batch 5 Qualification sample as the radioactive salt solution for mixing tests meant to evaluate the propensity for solids formation between the 5.6 M sodium clarified salt solution (CSS) and scrub solutions (0.05 M HNO 3 and 0.025 M NaOH solutions) during pH changes has been completed at SRNL. The key results and conclusions from this investigation to evaluate the propensity for solids formation between CSS and scrub solutions because of pH fluctuations include the following: None of the 8 test matrices, involving the mixing/titrating of scrub solution with CSS and vice versa, showed any visible evidence for the formation of solids with pH changes. The average turbidity measurements for these 8 tests, excluding test 7, was 21.6 ± 0.4 NTU. This average turbidity value is within 2 sigma of the baseline turbidity value of 20.5 NTU for the CSS only without added scrub solutions. However, only in test 7, which involved the titrating of CSS into dilute nitric acid scrub solution (0.05 M HNO 3 solution), was a small measurable change in the turbidity of the mixture observed at 31.9 ºC. On the other hand, the equivalent test (instantaneous addition of CSS into 0.05 M HNO 3 ) gave an average turbidity measurement value of 21.6 ± 0.6 NTU at a temperature of 24.0 ºC. This difference in turbidity measurement has been attributed to either instrument performance errors or the effect of temperature. It is recommended that these CSS and scrub solution mixing tests be performed with corresponding simulant solutions in a non-radioactive environment. A simulant environment will aid in the full integration of a pH and turbidity meter equipment with the experimental set up. A new detection method, preferable a laser-based method, may be needed in detecting and confirming the precipitation of fine particulates at the right blending times and pH fluctuations in the solution mixtures. On the other hand, the differing free OH concentrations (~2 M in CSS), H + concentrations (0.05 M HNO 3 ) for the titrants and the bulk solutions, may make it almost impossible to observe a neutral pH or a pH in the range of 7-9, where the potential for solids formation is maximized and leads to possible precipitation of minerals like aluminum hydroxides.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Microbial Influence on the Mobility of +3 Actinides from a Salt-Based Nuclear Waste Repository
Biologically enhanced transport of radionuclides is one of several processes that can affect the performance of a nuclear waste repository. In this work, several microbial isolates from the Waste Isolation Pilot Plant (WIPP) were tested for their influence on the concentration of neodymium, as an analog for +3 actinides, in simple sodium chloride solutions and in anoxic WIPP brines. Batch sorption experiments were carried out over a period of 4–5 weeks. In many cases, the effect on neodymium in solution was immediate and extensive and assumed to be due to surface complexation. However, over time, the continued loss of Nd from the solution was more likely due to biologically induced precipitation and/or mineralization and possible entrapment in extracellular polymeric substances. The results showed no correlation between organism type and the extent of its influence on neodymium in solution. However, a correlation was observed between different test matrices (simple NaCl versus high-magnesium brine versus high-NaCl brine). Further experiments were conducted to test these matrix effects, and the results showed a significant effect of magnesium concentration on the ability of microorganisms to remove Nd from solution. Possible mechanisms include cation competition and the alteration of cell surface structures. This suggests that the aqueous chemistry of the WIPP environs could play a larger role in the final disposition of +3 actinides than the microbiology.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Modeling Interfaces to Support Low-Level Waste Disposal System Performance Assessments - 20366
In low-level waste (LLW) disposal sites, interfaces between cementitious materials, used as waste forms and/or engineered barriers, and the surrounding soil or backfill material are often encountered. Reactions across these interfaces may lead to chemical and structural alteration of the cementitious and backfill materials that may ultimately affect long-term performance. For example, the ingress of carbon dioxide from soil gas into a waste tank concrete shell at the Hanford Site or Savannah River Site (SRS) or a low-level waste disposal vault at SRS leads to carbonation of cement hydration products (e.g. portlandite, calcium-silicate hydrate or C-S-H, and ettringite). The result of carbonation is a decrease in pH in the cement paste portion of the tank wall or vault concrete structure that could ultimately lead to de-passivation of the embedded structural steel and, eventually, to cracking. Cracking can lead to increased ingress of water into the structure, corresponding increased release of constituents of concern, and to subsequent increased transport of these constituents to the surrounding environment. The mobility of trace constituents (e.g., radionuclides of concern) in waste forms may increase in response to changes in pH, pore structure, and mineralogical gradients. Depletion of portlandite results in subsequent decalcification of C-S-H that leads to changes in the cement strength and may lead to structural failure. At the boundary between waste forms and concrete vaults, the migration of sulfate ions from the salt waste form (e.g., the saltstone waste form used for LLW disposal at the Savannah River Site in Saltstone Disposal Units, or SDUs) into the barrier or vault concrete has been predicted to result in expansive mineral phase formation (or 'sulfate attack'). Expansive mineral formation could result in cracking and potential loss of structural integrity in the vault concrete. In the performance assessment (PA) for the SRS SDUs, prediction of both the carbonation and sulfate attack fronts are critical to assessing the long-term performance of these LLW disposal vaults. A general purpose geochemical reactive transport model has been developed in LeachXS/ORCHESTRA to evaluate typical interfaces for LLW disposal environments, including SDUs at SRS. The model may be used to predict the long-term performance of interfaces between cementitious materials and saltstone or backfill materials with respect to primary phases by considering mass transport, cement chemistry, geochemical speciation, and multi-ionic diffusion over interfaces between different materials. The present study compares the performance and interactions between vault concrete and saltstone waste form for a representative SDU scenario. The materials were characterized using pH-dependence leaching test (EPA 1313) and semi-dynamic transport tests (EPA 1315) with test results used to parameterize the interface model. The development of a credible inter model provides an important basis for supporting future PAs where carbonation and sulfate attack are important aging and degradation mechanisms. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Dehalogenation of electrochemical processing salt simulants with ammonium phosphates and immobilization of salt cations in an iron phosphate glass waste form
This study demonstrates the conversion of simulated pyroprocessing salt wastes [i.e., ER(SF) and ERV2] into chemically durable iron phosphate glass waste forms through reactions with ammonium hydrogen phosphate precursors [i.e., NH4H2PO4 and (NH4)2HPO4] and added Fe2O3. The reaction of the salt simulants with the phosphate precursors resulted in dehalogenation of the salt through the production of NH4Cl and NH4I (along with water), of which the NH4Cl can be used to produce UCl3 oxidant for use in the electrorefiner during pyroprocessing. These iron phosphate glasses contain a notably higher waste loading than the baseline glass-bonded sodalite ceramic waste form, which requires about 70 vol% sodalite to accommodate Cl from the salt, because they are dehalogenated prior to vitrification. The high waste loadings coupled with higher bulk densities (~3×103 kg/m3) than glass-bonded sodalite (~2.3×103 kg/m3) mean that these glasses have a higher storage volume, i.e., the volume required to immobilize a given mass of waste. The iron phosphate glass waste forms were characterized using a variety of techniques including physical property measurements and chemical durability testing. Following initial testing, a proof-of-concept demonstration was conducted to show that the NH4Cl generated from this process could be used to produce UCl3 using depleted uranium dendrites as a feedstock.
phosphate glass, pyroprocessing, waste forms, salt↗
An Assessment of the Materials of Construction for the Transfer Lines and Unit Operations Equipment Associated with the Recycle Diversion Process
The Defense Waste Processing Facility (DWPF) processes and vitrifies radioactive waste that it receives from the Concentration, Storage, and Transfer Facility (CSTF) and Salt Waste Processing Facility (SWPF). 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 water stream originating from the collection of condensate liquids containing some minor sludge and frit solids and other waste components primarily resulting from entrainment into the condensate during foam-over events. The recycle stream volume is significant and is expected to approach 3 million gallons per year once the SWPF reaches full operation. Diverting the bulk of the recycle waste stream from the CSTF is essential for the eventual closure of the waste tanks, and hence the completion of the SRS liquid waste mission.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Electrochemical Salt Wasteform Development: A Review of Salt Treatment and Immobilization Options
Electrochemical reprocessing, also referred to as pyroprocessing, is a technique for recycling actinides from used nuclear fuel (UNF) to produce fuel for future reactors. Here, UNF is dissolved in a molten salt (e.g., LiCl-KCl eutectic) within an electrorefiner. After UNF dissolution, fission products are released into the electrolyte salt and converted to chlorides. This paper discusses wasteform options for processing the base electrolyte salt with the fission product salts as well as just the rare-earth fission products (as RECl 3 , REOCl, or REO x ) with the intent of finding optimal methods for reducing total waste salt volumes or partitioning the salt for alternate wasteform options. Furthermore, two of the more detailed partitioning options discussed herein include halide removal from the salt (dehalogenation), which accounts for more than half of the salt on a molar basis, and RE fission product removal for wasteforms with high-RE loadings. Wasteform properties are compared with emphasis on wasteform volume starting from a given amount of (1) total salt cations or (2) RE cations. Comparisons are also made of wasteform chemical durabilities, with the data available from like testing methods. A main conclusion from this work is the justification of subsequent salt processing after electrorefiner operations for achieving significant wasteform volume reduction.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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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Evaluating Liquid Waste Transfers and their Impacts to the SRS Tank Farms to Support Operations and Closure
BACKGROUND • SRMC mission critical milestones – 34 million liters (9 million gallons) Salt Waste Processing Facility (SWPF) Processing Rate – Accelerated Waste Tank Closures • Results in an increase in Tank Farm transfers • Barriers – Equipment failures, replacements, and adjustments – Weather – Procedure development – Outside facility changes How has the Tank Farm adapted to meet these milestones? By improving the transfer evaluation process
B↗
Sulfonamide and Sulfonamido-phenol Ligands for Extraction of f-Elements from Alkaline High-Level Waste
Alkaline High Level Waste (HLW) has been accumulated at Hanford and Savannah River Sites as a result of reprocessing for nuclear weapons production during the cold war. A large volume (∼100 MGal) has been accumulated in carbon steel tanks at Savannah River (SRS) and Hanford. The tank waste contains three separate phases generated when NaOH was added to previously acidic Purex raffinates: 1-2 Supernatant liquid, salt-cake, and sludge. The sludge consists mainly of insoluble hydroxides of transition metals while the supernate and salt-cake contain caustic-soluble materials, including salts of highly radioactive fission products Cs(I) and Sr(II). Current treatment of alkaline HLW in SRS includes: i) The Actinide Removal Process (ARP), which is based on sorption of {sup 90}Sr and Actinides (An) on monosodium titanate (MST), also known as 'alpha-strike' process, followed by ii) Caustic Side Solvent Extraction (CSSX),4 which is used for the extraction of {sup 137}Cs by modified calixarenes in a hydrocarbon diluent. Residual actinides in some tanks are removed after CSSX by an additional ARP process commonly referred to as 'alpha-finishing'. Despite the success of ARP for Sr and An removal, as it is a sorption process, it represents the kinetic bottleneck of integrated salt waste processing. Hence potential introduction of additional organic ligands for actinide extraction (in a modified CSSX process) could simplify the overall integrated process, making it more efficient and economical, with less titanate needed and shorter sorption time, as some of the actinide component would be removed during CSSX. In this study tri-sulfonamide and o-sulfonamido-phenol (mono-sulfonamide) ligands have been studied as extractants for Sm(III), which is being used as an Am(III) surrogate. Our prior studies in the group using a tri-sulfonamide (iPr-tsa-B) showed favorable extraction for Sm(III) nitrate salts from alkaline solutions. Mono-sulfonamides possess similar orientation of binding sites to pyrocatechols, which have been found to be good ligands for Am(III) binding and extraction from alkaline media. Tri-sulfonamide of the type iPr-tsa-B6 (1 mM in CH{sub 2}Cl{sub 2} solution) was studied for Ln{sup 3+} extraction using Sm(NO{sub 3}){sub 3}.6H{sub 2}0 (10 and 25 μM) in alkaline solution of NaOH (0.05, 0.1, 0.2, 0.3 mM) / 0.1 M NaNO{sub 3}. Stripping of the organic phase was done using 0.1 M HNO{sub 3} and quantification of Sm{sup 3+} was done using ICP-OES at 359.3 nm. The need to improve stability of the complex led to synthesis of compounds with N-donor site closer to the central benzene ring to facilitate cation-π interactions. Synthesis of tri-sulfonamide type A: a) Chloromethyl methyl ether, SnCl{sub 4}, CH{sub 2}Cl{sub 2}, 0 deg. C, N{sub 2}, 4 h, 57%; b) NaN{sub 3}, reflux in H{sub 2}O/acetone for 22 h, 80%; c) PPh{sub 3}, THF/H{sub 2}O, 22 h, 79%; d) p-toluene sulfonyl chloride, Et{sub 3}N, 1,2-DCE, 22 h. Extraction: Sm(NO{sub 3}){sub 3}.6H{sub 2}O (2 mM) in 5 ml of aqueous NaOH (pH 10.5 - 14) + 6 ml of CH{sub 2}Cl{sub 2} solution of msa (20 mole equiv.) were rotated on a wheel (60 rpm; 20 h). Stripping: 5 ml of 0.1 M HNO{sub 3} + CH{sub 2}Cl{sub 2} solution of msa (after extraction, centrifugation and filtration) was rotated on the wheel (60 rpm; 20 h). Sm{sup 3+} was quantified using UV-Visible spectrophotometry.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Evaluation of a Sulfate Solubility Limit Greater Than 0.65 Weight Percent in Sludge Batch 10 Glasses
Previous laboratory-scale crucible testing with batch chemicals confirmed that the sulfate (SO 4 2- ) limit for Sludge Batch 10 (SB10) was 0.65 weight percent (wt.%) in glass. This limit signifies that 0.65 wt.% SO 4 2- can be retained in the glass without the formation of a sulfate phase. The Defense Waste Processing Facility imposes this constraint in the Material Tracking Program. Based on preliminary calculations to support the Material Tracking Program, it was anticipated that transfer volumes of the monosodium titanate/sludge solids (MST/SS) stream from the Salt Waste Processing Facility (SWPF) may need to be reduced to maintain projected sulfate concentrations below 0.65 wt.% in glass. Savannah River Mission Completion requested that the Savannah River National Laboratory perform additional sulfate testing to determine whether a sulfate solubility limit greater than 0.65 wt.% is feasible for SB10, which could allow for higher transfer volumes of the MST/SS stream. This report documents the results of the testing at higher sulfate concentrations for the glass composition region defined by the most recent SB10 projection (November 2022) and Frits 473 and 625. Frit 473 was recommended for SB10 and Frit 625 was used during SB9 processing and the SB9 to SB10 transition. A total of twenty-one glass compositions were developed based on the expected compositional variables, which include sludge-only (SO) and coupled processing with the SWPF, waste loading (WL), and frit composition. The target sulfate concentrations were varied from 0.65-0.85 wt.% at 32 and 40% WL. Each glass was prepared from reagent grade chemicals and melted at 1150 °C. Visual observations were used to confirm the presence of a sulfate salt phase on the cooled glass surfaces. Representative samples of each glass were submitted for chemical composition analysis by inductively coupled plasma-optical emission spectroscopy and Cs analysis by inductively coupled plasma-mass spectrometry. Overall the majority of mean measured values are consistent with the target values for each major oxide of interest with less than 5% error. The percent errors for the measured SO 4 2- concentrations are generally less than 10%, which is comparable with previous sulfate solubility study measurements and acceptable. Only the SO glasses based on Frit 625 formed a sulfate phase at a 0.80 wt.% SO 4 2- target concentration at both 32 and 40% WL. The remainder of the glasses did not form a sulfate layer. Due to the formation of the sulfate phase, the limit is conservatively set at 0.70 wt.% based on the measured sulfate concentrations of 0.71 wt.% and 0.75 wt.% for these two glasses. None of the SO or coupled operation glasses based on Frit 473 formed a sulfate salt phase, which supports a sulfate limit of 0.80 wt.%. The following SO 4 2- concentration limits are recommended during SO and coupled SB10 processing: (1) 0.70 wt.% during processing with Frit 625, and (2) 0.80 wt.% during processing with Frit 473.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Improbability of Nuclear Criticality in Compacted Criticality Control Overpacks after Room Closure by Salt Creep at Waste Isolation Pilot Plant
Based on the rationale presented, nuclear criticality is improbable after salt creep causes compaction of criticality control overpacks (CCOs) disposed at the Waste Isolation Pilot Plant, an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality potential has been updated here because of the introduction of CCOs, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCOs during two representative conditions: (1) large salt block fall, and (2) gradual salt compaction (without brine seepage and subsequent gas generation to permit maximum room closure). Geomechanical models of rock fall demonstrate three tiers of CCOs are not greatly disrupted. Geomechanical models of gradual room closure from salt creep predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Criticality models of spheres and cylinders of 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular spacing demonstrate that an array of CCOs is not critical when surrounded by salt and magnesium oxide, provided the amount of hydrogenous material shipped in the CCO (usually water and plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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↗
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.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
AI/ML-assisted Design of Phosphate Glass and Ceramic Nuclear Waste Forms
Borosilicate glass is the widely accepted waste form for immobilization of high and medium level nuclear wastes. Advances in nuclear energies and new reactor designs require the development of new waste forms. For example, wastes from molten salt reactors and reprocessing of nuclear fuels lead to salt-based wastes that are difficult to be immobilized by conventional borosilicate glasses due to limited solubility and waste loading. In designing new waste forms, machine learning (ML) and artificial intelligence (AI) based approaches are much needed and can be beneficial in enabling a more efficient design in large parameter spaces as compared to traditional Edisonian trial-and-error approaches. Here, we report in this paper the rationale and latest progress of our ML/AI-based design of phosphate-based waste forms.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗