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66 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

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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DOE OSTI2025

Conceptual Model Testing Related to SDU 6 Drainwell Observations

From its inception in the early 1950s through the end of the Cold War in the early 1990s, the Savannah River Site (SRS) produced nuclear materials for national defense in five reactors. Additionally, irradiated reactor fuel and target tubes were dissolved in nitric acid to recover plutonium and uranium using the PUREX (Plutonium Uranium Reduction EXtraction) process. Liquid waste from these chemical separations processes was then stored onsite in 51 underground tanks. Eight waste storage tanks have been operationally closed (i.e. cleaned and grouted) and the remaining tanks hold a mixture of liquids, insoluble solids, and precipitated salts (SRMC-LWP-2022-00001), the latter generated by evaporating water from the liquid waste. Waste is currently being retrieved from tanks and separated into 1) high-radioactivity, low-volume, and 2) low-radioactivity, high-volume components, principally through the Salt Waste Processing Facility (SWPF) (SRMC-LWP-2023-00001). The former waste stream is vitrified in the Defense Waste Processing Facility (DWPF), stored onsite, and destined for offsite disposal in a deep geologic repository. The latter stream is mixed with dry cementitious materials in the Saltstone Production Facility (SPF) and the wet slurry placed in onsite Saltstone Disposal Units (SDUs) within the Saltstone Disposal Facility (SDF), where it hardens into a cement waste form termed saltstone. A low-infiltration surface cover system will be placed over the SDF at closure, where SDUs will then be in the subsurface post-closure (SRR-CWDA-2019-00001).

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DOE OSTI2024

Low Purge and Mercury Recovery Testing with Sludge Batch 10 Tank 40 Simulant

Researchers at the Savannah River National Laboratory (SRNL) have completed testing as requested by Savannah River Mission Completion (SRMC) to perform experiments to determine the impact of using a lower or inert purge in Sludge Batch (SB) 10 processing under the Nitric-Glycolic Acid (NGA) flowsheet. A key objective of this testing was also to determine the mercury speciation and recovery during each experiment. The testing was performed as part of the SB10 Technical Task Request (TTR) and Task Technical and Quality Assurance Plan (TTQAP). Two sets of tests were performed, and a Run Plan was approved prior to each set of experiments to document the planned testing. Three initial experiments were completed to determine whether a low air purge would be beneficial to CPC processing at higher acid stoichiometry (110%) based on the Koopman minimum acid (KMA) equation (116% Hsu). One of the tests, an inert nitrogen purge experiment, was also completed to demonstrate that excluding oxygen did not introduce any new hazards. The experiments were designed to be identical except for the change in purge gas and purge flowrate from run to run. After reviewing the results from the initial three experiments, six additional tests were proposed by SRNL to support the lower purge study and to look for processing alternatives for improving mercury recovery. These additional six tests were all completed at a very low acid stoichiometry to mimic the pH experienced during processing in DWPF (~7). DWPF is processing SB10 sludge at an acid stoichiometry of 90% based on the Hsu equation. The additional SRNL experiments were performed at an acid stoichiometry of 62.5% KMA stoichiometry (66.3% Hsu) to produce a Sludge Receipt and Adjustment Tank (SRAT) product with a pH of about 7. All experiments used simulants of both SWPF streams, although no entrained solvent was added during any of the experiments

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DOE OSTI2024

New Tank Mapping Method Improves Waste Removal Process

Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.

Mini, Melany↗
DOE OSTI2025

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↗
DOE OSTI2026

Improving Elemental Mercury Recovery in DWPF by pH and Ionic Strength Studies Using Laboratory Scale Replica – 26141

In order to safely disposition nuclear material at large scales, the Savannah River Site (SRS) constructed the Salt Waste Processing Facility (SWPF) for the removal of actinides, the Saltstone facility for preparation of a low-level cementitious grout, and the Defense Waste Processing Facility (DWPF) for the vitrification of high-level waste (HLW). A few years after processing began, high concentrations of mercury were discovered in cementitious waste at the Saltstone facility which initiated a search for the source of mercury during processing at these SRS facilities. Mercury serves as a catalyst in the dissolution of spent nuclear fuel, aluminum-actinide alloys, for actinide recovery, but not enough is known about the behavior and properties of mercury within these process streams.

Pina, Jeanette [Savannah River National Laboratory↗
DOE OSTI2021

Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Sample Tank 202

An Extraction, Scrub, and Strip (ESS) test was performed on a sample from Tank-202. The purpose of testing this sample is to determine if the solvent is displaying the correct extraction and stripping behavior with cesium. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 14.2, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, the strip cesium distribution ratios fell within acceptable ranges.

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DOE OSTI2021

SB10 Frit Recommendation, and Evaluations of the Glass Variability Study and Cs-137 Concentrations in Strip Effluent Based on May 2021 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, and 26F, Alternate Feed Stock-2 (AFS-2) and Sodium Reactor Experiment (SRE) material from H-Canyon. In support of SB10 qualification, frit development using 2020 Tank 40 blend projections and experimental work for the glass variability study were previously conducted. Frit 473 and Frit 209 were identified as candidate frits and both were included in the development of the variability study test matrix; however, a final frit recommendation was postponed until more information could be determined about the composition of SB10 after washing. In May 2021, Savannah River Remediation (SRR) reprojected SB10 based on the analytical results from the Tank 51 qualification sample that was washed in the Savannah River National Laboratory (SRNL) Shielded Cells Facility.

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DOE OSTI2022

Impacts of Guanidine Degradation Products on Next Generation Solvent (NGS) Caustic Side Solvent Extraction (CSSX) Processing

Savannah River National Laboratory researchers have been requested to perform testing to assess the potential for build-up of guanidine degradation compounds in the Next Generation Solvent Caustic Side Solvent Extraction process. Testing was also requested to determine the impact of guanidine degradation compounds [3,7-dimethyloctylamine (iDA) and Bis-N,N’-(3,7-dimethyloctyl)urea (DiDU)] on cesium behavior in the flowsheet. Twelve partitioning experiments were performed to quantify the partitioning coefficient of identified guanidine degradation compounds in various organic-aqueous mixtures. Six Extraction, Scrub, and Strip (ESS) experiments were performed to quantify the impact of degradation products on cesium behavior.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Summary of Savannah River Site FY21 Salt Waste Qualification Data and Characterization Data from Salt Waste Batches Processed Since 2008

The Savannah River National Laboratory analyzed samples from Savannah River Site Waste Tank 41H and 21H to support qualification of Salt Waste Processing Facility Waste Batches 3 and 4 (the FY21 Salt Batch Qualification samples) for processing. Neither of the samples displayed any unusual or unexpected characteristics such as large amounts of solids, floating solids, or unusual color. Characterization of these samples confirmed similar chemical composition and characteristics to previous salt waste batches. These results were initially provided to Savannah River Remediation (Liquid Waste Operations Sub-contractor) as External Sample Results LIMS Reports. The analytical results (both expedited and routine) are now summarized and discussed in this technical report.

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DOE OSTI2022

Results of the Extraction-Scrub-Strip (ESS) Tests With Solvent Using IBC BOBCalix

Two Extraction, Scrub, and Strip (ESS) tests were performed on two BOBCalix solvent samples prepared using BOBCalix from IBC Advanced Technologies (“IBC”). The purpose of testing this sample is to confirm if the newly sourced BOBCalix displays the correct extraction and stripping behavior with cesium. The first batch of BOBCalix solvent was prepared using IBC lot#220310KC-1. The second batch was prepared using IBC lot#220615KC-3A. The ESS tests showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D (Cs) values for the tests were 12.9 and 13.6. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, the strip cesium distribution ratios (0.0588/0.0342/0.0258 and 0.0575/0.0335/0.0262) fell within acceptable ranges.

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DOE OSTI2022

Blend Pump Mixing in Tank 27 Evaluation

For future waste disposal operations, Savannah River Mission Completion (SRMC) plans to operate Tank 27 as a blend tank. It is desirable to avoid additional worker exposure by using the planned commercial submersible mixing pump (CSMP) installations in Risers B2 and B4 to function as the blend pumps for prolonged blend tank operations. The purpose of this task is to evaluate the impacts to mixing across a range of pump installation heights from 1 to 201 inches. Specifications include: A maximum pump diameter of 22.5 inches, dual nozzles placed tangentially opposed with nominal inside diameter of 2 inches, a total flow range (both nozzles combined) of 500 to 1950 gallons per minute (gpm) and a maximum revolutions per minute (RPM) of 1800. Tank 27 is expected to have maximum fill height of 360 inches.

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