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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Development of Bio-Compatible Fluorescent Sensor

The development for biocompatible fluorescent sensors is used to understand the movement and fracking of wild boar at SRS. Current fracking methods for wild boar include physical tagging, camera monitoring, and foot tracks in the soil. None of these methods are highly effective in determining the path and habits of the animals as it depends on live imaging, soil consistency (lack of rain), and potentially a gps system is tagged properly. This system allows fracking using an additive digested by the animal and released in the urine that will fluoresce when excited with a particular laser. This method helps determine the actual movements of the animals and even provide a potential prediction model based on their routine. We partnered with Ft. Gordon to utilize their animal testing expertise in determining the best material and pathway for these materials to be excreted.

47 OTHER INSTRUMENTATION↗
DOE OSTI2020

Oak Ridge Response to Versatile Test Reactor Environmental Impact Statement Data Request

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor being developed in the United States under the direction of the US Department of Energy Office of Nuclear Energy (DOE-NE). The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300 MWth sodium-cooled metallic-fueled pool-type fast reactor has been led by the US National Laboratories in collaboration with General Electric–Hitachi and Bechtel National, Inc. In support of the VTR project, DOE issued a Notice of Intent (NOI) in the Federal Register on August 5, 2019, announcing the intent to prepare an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act (NEPA) and its implementing regulations. The EIS will evaluate alternatives for a versatile reactor–based fast-neutron source facility and associated facilities for the preparation, irradiation, and post-irradiation examination (PIE) of test/experimental fuels and materials. Specifically, the NOI identified two siting alternatives for the VTR reactor facility: Idaho National Laboratory (INL) or Oak Ridge National Laboratory (ORNL). In addition, the NOI also specified two siting alternatives for VTR fuel fabrication: INL and the Savannah River Site (SRS). This report provides information in response to data requests made to ORNL to fill in site-specific knowledge gaps to develop a high-quality EIS. The responses provided are not required to provide full details in every aspect; instead, they adequately bound possible environmental impacts or provide sufficient information to adequately assess likely environmental impacts. This work is being performed under a subcontract from INL to ORNL using DOE-NE funds and is directed by DOE-NE and DOE-ID. Leidos has been contracted by DOE-NE to write the VTR EIS, so most data requests have come from Leidos but were often routed through INL or DOE-ID. DOE-ID is overseeing the NEPA and EIS processes for the VTR project. Leidos will use the information provided in this report to inform the VTR EIS and will also cite this document to establish a clear, publicly available source of the information. Section 2 of this report briefly describes the proposed ORNL VTR Alternative and illustrates the location of the proposed site for the ORNL VTR Alternative. Sections 3 through 7 provide direct responses to data requests received by ORNL. These sections use a tabular format in which data requests are divided into separate items to be addressed; the items are numbered, the data requests are restated with more topical information included, and then the responses are provided. Initial data requests and follow-on requests for additional information (RAIs) are combined under the original data request fields. Finally, Section 8 presents summarized conclusions and describes future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Rheology and Flow Evaluation of Neutralized Sodium Reactor Experiment Fuel with Manganous Nitrate

H-Canyon is preparing the Sodium Reactor Experiment (SRE) solutions in Tanks 16.3 and 16.4 for discard to the Savannah River Site (SRS) High Level Waste (HLW) Tanks into Sludge Batch (SB) 10. To meet HLW's criticality requirements, manganese will be added to the SRE solutions. The addition of manganese to the existing thorium and uranium in the SRE solution raised concerns with the flow of this neutralized material as it is discharged from H-Canyon through the gravity drain system to the H-Area Pump Pit (HPP). A neutralized Mn adjusted SRE stream will produce primarily Mn, Th, and U solids that can deter transfer. A 2012 rheology study was completed as part of the flowsheet development.1 However, the initial rheology study did not include the addition of manganous nitrate as a poison. Therefore, H-Canyon Engineering requested Savannah River National Laboratory (SRNL) to determine if the neutralized, Mn-adjusted SRE solutions will flow through the waste header to the HPP. The H-Canyon Technical Task Request (TTR) specified a target of 80 to 1 to bound the uncertainty in Mn target. Parallel studies were being performed to ensure that freshly precipitated Mn did not have a solubility that would result in challenging the DWPF WAC requirements of 70 to 1. This task was requested via a TTR and is governed by a Task Technical and Quality Assurance Plan (TTQAP).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Dilute and Dispose Cost Estimate for Equipment Installation per the LCCE

As directed in the Consolidation Appropriations Act, 2016, the National Nuclear Security Administration (NNSA) initiated the preconceptual design and development of a Lifecycle Cost Estimate for the Surplus Plutonium Disposition (SPD) Dilute and Dispose Program. Based on August 2016 Program Requirements Document and subsequent supplemental guidance, LANL prepared the Lifecycle Cost Estimate under key assumptions that meet the program’s requirements. For Dilute and Dispose, the program would disposition surplus Pu by diluting oxide produced at LANL with inhibitor materials, packaging the materials in containers, and shipping the containers to a deep geologic repository for permanent disposal. The base assumption is that LANL would disassemble pits, convert the Pu metal to oxide, and characterize and package the material for shipment to SRS, where it would be diluted prior to geologic disposal at the WIPP site in New Mexico. Another major assumption for the Dilute and Dispose option is that LANL would increase the current oxide production rate (or throughput) to 1500 kg/year, 5 times higher than the maximum annual production of ~300 Kgs executed by the ARIES Oxide Production Program at LANL. Analysis based on the ARIES program’s throughput model revealed that 15 pieces of equipment would need to be installed within PF-4 and certain facility improvements would need to be accomplished in order to meet the desired throughput levels. The additional equipment would be essentially identical to equipment already used within PF-4 for existing operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

PDRD Project 19039 - All metal vacuum scroll pump development for tritium

Currently the United States relies on a single source, foreign company, Eumeca (formerly Normetex) for the supply of an all metal vacuum scroll pump to process tritium. This “one of a kind”, all metal, vacuum scroll pump utilizes no “wetted”, polymeric materials that degrade prematurely when exposed to tritium. U.S. based vacuum scroll pump suppliers typically utilize a polymer, Teflon®, as a tip seal material at the edge of the scroll impellers for proper pump function. Teflon® is known to degrade within a relatively short time frame (approximately 1 year) when continually exposed to tritium gas. This degradation results in “flaking” of the Teflon® material which results in particulate contamination of the process gas stream. While filtration of the particulate is possible, doing so requires additional maintenance which is costly in a glovebox environment. In addition, filtration can degrade pumping system performance as particulate collects in the filter media over time. To date, all of the well established U.S and European based vacuum scroll pump manufacturers have been uninterested in developing an all metal vacuum scroll pump that does not use a polymeric scroll tip seal. This position is taken as the vacuum scroll pump manufacturers do not see a large enough market to make it profitable for them to develop and market an all metal vacuum scroll pump. In the mid 2000s, one company, Air Squared, Inc., a small Original Equipment Manufacturer (OEM) specializing in design and fabrication of custom scroll vacuum pumps and compressors, was willing to pursue the effort. The first design effort incorporated Vespel® as a polymeric tip seal material. Results of this effort led to realizing the affect that the Met-Bel MB-601 metal bellows pump has on performance when used as a backing pump for an all metal vacuum scroll pump without a scroll tip seal. As a result, Air Squared subsequently designed an all metal vacuum scroll pump without a tip seal material for evaluation. Although the initial, all metal version of the Air Squared pump was found deficient in regard to scroll impeller stiffness, which resulted in scroll contact, general design and pump performance was verified to be adequate. SRS PDRD Project 19039 was launched in 2019 as a 3rd attempt with Air Squared, Inc. in regard to development of an all metal vacuum scroll pump for consideration as an equivalent to the Eumeca/ Normetex pump. The effort specified 300 series stainless steel for the scroll impellers, all metal sealing, and design to meet tritium throughput and ultimate vacuum requirements. In addition, a 60 psig design pressure rating requirement was specified for the pump containment boundary. The 3rd generation Air Squared pump has been designed, fabricated, and has received preliminary testing. Preliminary testing has indicated proper leak tightness and pump performance. However, work continues regarding development of a pumping chamber containment bellows that can be subjected to 66 psig without damage to the bellows. Work will continue in 2020 and 2021 to pursue a pumping chamber containment bellows that meets the pressure rating requirement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

H-Canyon exhaust tunnel LiDAR data request for information review and release

This request is to release LiDAR data files acquired by SRNL R&D Engineering and H-Area Operations and Engineering while performing scans of the interior of the H-Canyon Exhaust tunnel. LiDAR data includes 3-dimensional (3D) point cloud data sets and panoramic digital images of the interior tunnel wall surfaces. Presently two deployments have been completed, the first in November 2019, Fig. 1, and the second in June of 2020. This request is for the release of the data collected during those two deployments. Detailed information on the deployment and data collected can be found in SRS document C-ESR-H-00072, “November 2019 Initial Deployment of LiDAR”. It is planned to perform ongoing scans at approximately 6-month intervals, the purpose of the deployments is to evaluate the usefulness of the data collected to enable quantitative measurements such as tunnel dimensions and rate of surface erosion and as a precursor to a potential deployment of a LiDAR system on the tunnel inspection crawler.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Data call questions related to potential NEPA/Data gaps

Based on review of the related NEPA documents, and consideration of the potential gaps in NEPA coverage, a number of questions have been generated. These questions are generally higher level (less detailed) questions that should help to establish the applicability of analyses in existing NEPA documents and the ability of these documents to provide NEPA coverage for the Proposed Action. These questions do not request detailed information about SRS facilities/activities and the Proposed Action like typically would be requested when performing impacts analyses for an EIS. We do not believe detail information is needed; if it is needed, it may be an indication that an SA is not the appropriate level of NEPA documentation for the Proposed Action.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Innovative mercury treatment technology options for the liquid waste system at the Savannah River Site: scoping studies

The Savannah River Site (SRS) Liquid Waste System (LWS) contains liquids, salts and sludges that are currently being processed into final wasteforms for disposition, specifically, waste glass for sludges, solids and liquids containing high levels of radioactivity, and saltstone grout for low activity decontaminated liquid solutions. The LWS also contains approximately 60,000 kg of mercury present in the following physical and chemical forms, specifically: a) ionic inorganic mercury, organomercury (e.g., methylmercury), and other minor components found in LWS fluids b) mercury solids such as oxides, hydroxides, amalgams sulfides and sorbed mercury, c) accumulations of dense liquid elemental mercury, and d) vapor phase elemental and organomercury mercury found in tank headspace gas and in evaporators. An effective and proactive management strategy of the mercury present in the LWS is needed to support processing of LWS wastes into glass and saltstone. Sustainable processing of the LAWS to completion requires mercury removal from the LWS at a rate of approximately 2,900 kg/yr. This removal can be accomplished through existing mercury treatment systems or newly implemented LWS “purge points”. The chemical speciation of mercury has emerged as the key factor that controls mercury behavior in the LWS. For example, past studies demonstrated that mercury speciation is adversely impacting the performance of existing removal systems so that significant levels of mercury are recycled from the Defense Waste Processing Facility (DWPF) back to the tank farm. Consequently, mercury concentrations have slowly increased in the LWS tank fluids over time. The presence of organo-mercury has also been identified as the cause of increased mercury leaching from saltstone. In response to these challenges, the Department of Energy (DOE) Office of Environmental Management (EM-TD) Technology Development Program has supported a series of scoping studies predicated on manipulating or controlling mercury speciation and mercury behaviors within the constraints of LWS waste chemistry and safety conditions. The intent of these studies was to rapidly triage potential technology options and develop a technically based go / no go recommendation for further work. This composite report presents the results of three scoping studies: 1) advanced photooxidation processes, 2) chemical reduction, and 3) mercury getters.

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DOE OSTI2020

RANT Waste Container Receipt, Inspection, and Transfer

This procedure provides instructions for the receipt, inspection, and transfer of transuranic (TRU) waste containers within the Radioassay and Nondestructive Testing (RANT) Facility. Performance of this procedure meets the Surveillance Requirements (SRs) 4.1.1, 4.1.2, 4.1.3, 4.2.1, 4.2.2, 4.2.3 and 4.2.4.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Selection of Well Locations for Saturated Zone Monitoring at the E-Area Low Level Waste Facility

Savannah River Site (SRS) has adopted a revised approach to performance monitoring of the E- Area Low Level Waste Facility (ELLWF) that adds a saturated zone component to the existing vadose zone monitoring program (Kubilius, 2019). The revised monitoring approach specifies that eight additional water table wells be installed near Slit Trench 1 (ST1) and Engineered Trench 2 (ET2). Six wells will be installed at locations characterized in 2017 and 2018. These include four wells near ET2 (ELF001D, ELF002D, ELF003D, and ELF004D) at locations identified by Kubilius (2019) and two wells north of ST1 (ELF005D and ELF006D) at locations finalized in this report. Additionally, two wells (ELF007D and ELF008D) are to be installed at locations south of ST1 based on characterization work completed in 2020 and documented in this report. This area is near well cluster BGX-2 between the Mixed Waste Management Facility (MWMF) and the ELLWF. The area is upgradient of ELLWF and, contamination in the water table aquifer is sourced from MWMF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

CST R9120-B 30x60 Media Pretreatment and Digestion Standard Development to Support Tank Closure Cesium Removal 1A (TCCR-1A) Operations

In support of continued operations of Savannah River Remediation’s (SRR) ion exchange process to remove radioactive cesium from Savannah River Site (SRS) dissolved saltcake, referred to as the Tank Closure Cesium Removal 1A process (TCCR-1A; containing the second set of at-tank ion exchange columns), SRNL was tasked with: 1) validating the field protocol planned for pretreating the crystalline silicotitanate (CST) ion exchange media to be used in the columns, 2) preparing CST media for use in in-tank batch contact testing, and 3) developing a digestion standard to verify complete CST dissolution prior to characterization following each in-tank batch contact test. A procedure (SRNL L29 Manual, ITS-0229) was developed previously in support of the TCCR Demonstration (which contained the first set of ion exchange columns) based on evaluations of a different CST production lot with a larger particle size range to document the protocol for pretreating CST utilizing conditions similar to what will be performed in the field.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Calculated Roughness Changes due to Clear-Cutting At Savannah River Site

At the Savannah River Site (SRS), multiple facilities are routinely producing and working with radiological constituents which require safety basis modeling by the National Nuclear Safety Administration. The models used to determine the safety basis employ roughness length as an input parameter which describes the roughness of the ground surface which influences momentum exchange near the surface and the subsequent generation of mechanical turbulence. Because roughness length is dependent on surface characteristics, including vegetation, any changes to the surface would have a subsequent impact on the roughness length and require a new safety basis analysis to be performed. An analysis of measurements prior to and after the dear-cuts and thinning was performed and roughness lengths were calculated for each of sixteen wind sectors. Not all sectors contained dear-cuts, so this provides comparison of natural variations in roughness. Based on the results of the estimated changes, consideration will be given to how these changes may affect atmospheric dispersion modeling. It was found that clear-cutting activities generally led to a reduction in roughness. Reductions in roughness length also occurred in some of the wind sectors containing only undisturbed forest, but those in sectors with dear-cuts were found to be slightly greater in magnitude than natural variation. Only in one instance was the change in roughness length found to exceed a decrease of 10%, indicating that there is no significant safety basis to the site as a whole based on previous clear-cutting activities.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

A novel approach to mitigating the potential release of radioisotopes under fire conditions - enhancing fire resiliency of radiological contamination fixatives during deactivation & decommissioning activities

Savannah River National Laboratory (SRNL), in close collaboration with the Florida International University Applied Research Center (FIU ARC), successfully executed a technology development activity on behalf of the Department of Energy, Office of Environmental Management (DOE EM). The purpose of the activity was to improve the capability of fixative technologies in immobilizing residual contamination when exposed to thermal stressors as postulated in accident scenarios in Basis for Interim Operations (BIO) documents across the DOE EM complex. The effort resulted in the test and evaluation of a down-selected, commercial-off-the-shelf (COTS) intumescent technology in a radioactive environment at the Savannah River Site (SRS) Building 235-F Plutonium Fuel Form (PuFF) Facility and highlighted the potential use of this technology as a stand-alone, fire retardant fixative. Furthermore, this activity highlighted significant shortfalls in common fixatives currently used to support decommissioning activities and brought to the forefront the need for a methodical, uniformed approach to certifying fixative technologies for operational use in Deactivation and Decommissioning (D&D) activities. An essential component of this research was to evaluate the state of industry fixatives currently in use and set the foundation for comparison to alternative technological solutions. A baseline of five (5) commonly used fixatives and decontamination gels (hereafter collectively referred to as fixatives) was conducted, and notable shortfalls and deficiencies were revealed, particularly when exposed to thermal, water, and other environmental stressors. At temperatures as low as 300 °F - 400 °F (148.89 °C – 204.44 °C), all commonly used fixatives melted from the substrates within 3-5 minutes of exposure, resulting in contaminant transport. Significant degradation in terms of mass loss, desiccation, and off-gassing occurred, and the chemical breakdown of the polymer was so complete researchers assessed there would likely have been a release of residual contamination. Additional vulnerabilities became evident when exposed to water immersion and high humidity; existing fixatives took up water, swelling and frequently delaminating from the substrate, once again increasing the likelihood of a contaminant release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Body Mass Variation in an Introduced Wild Pig Population with Changing Ancestry

Total body mass variation was studied in a single wild pig (Sus scrofa) population that originated from three introductions of varying ancestry. Body mass data from 12,257 wild pigs collected over more than five decades were analyzed for differences due to the effects of sex, age class, reproductive status, season, and historic subpopulation groupings. Total body mass data from this population averaged 38 kg and ranged from 0.511 to 189 kg. The overall observed range of body mass data was consistent with previous studies. Very heavy individuals (i.e., >150 kg) made up only a small percentage of the overall population. Males were significantly larger than females, being on average 1.09 times heavier. This significant sexual dimorphic difference was seen in all but the piglet age class. This percent dimorphic difference increased with age. Body mass variation and overlap between the age classes also increased with age. Males were more variable than females within the same age class. The study area’s wild pigs were on average born at slightly less than one percent of their adult body mass. Percent weight gain toward the adult body mass increased with age and was slightly greater in females than in males. In contrast, absolute weight gain between age classes was greater in males. Pregnant sows were significantly heavier than non-pregnant females in all age class comparisons; however, this difference decreased with age. There was a seasonal shift in body mass with age, with younger age classes being heaviest in the fall-winter period and the older age classes being heaviest in spring-summer. Having originated from different ancestral stocks, two major subpopulations of wild pigs on the study area differed significantly in body mass in all sex and age class comparisons. The SRS wild pig population showed a decrease in body mass during the 1990s that was significant. This change may have been the result of both ancestral and population density variation, as well as unspecified environmental factors

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Corrosion of Steel During Long-Term Exposure to Evolving Cementitious Environments

SRS is proceeding with closure of the H-Area Tank Farm (HTF) and F-Area Tank Farm (FTF). Closure consists of removing the bulk waste, heel removal, and filling the tank with tailored grout formulations. The long-term performance of the concrete and steel materials of construction and the closure grout is an important consideration for Performance Assessment (PA) of closed tanks in FTF and HTF. Over PA timeframes of hundreds, thousands, to tens of thousands of years, the chemical and physical properties of the concrete and steel materials will slowly degrade due to environmental exposure and material aging. In the interim these materials will provide a barrier to the leaching of radionuclides into the soil. Savannah River Remediation (SRR) is updating relevant portions of these analysis inputs to support an imminent revision to the HTF PA and a future update to the FTF PA. This analysis provides an update to the previous PA inputs for steel corrosion. The Central Scenario for the PA analysis includes three postulated modeling cases: 1) Realistic Case, 2) Compliance Case, and 3) Pessimistic Case. This analysis also included a Fast Flow Path Case. This latter case considered the circumstance where the initial condition of the concrete was a completely degraded state and that grout shrinkage exposed the steel to the soil environment immediately. In effect, the steel was unprotected by the concrete and grout. The cases have various degrees of conservatism considered. Chemical, physical and tank configuration parameters were investigated to understand their effects on the predicted time to release of the contaminants. The assessment reviewed the initial tank and steel configuration, service life degradation of the steel, and potential corrosion mechanisms associated with degradation of the concrete materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers'’ for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field's Metal eutectic alloy with a melting point of 62°C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions. It is recommended that glass vessels of various volumes ranging from 125 mL to 1 gallon with modified tops containing the specified Field's Metal eutectic alloy which are appropriately designed to avoid flashing spray material releases be added to the DSA container list. These vessels will be utilized in the Shielded Cells for the storage of radioactive liquid samples of sludge slurries. Use of these vessels will also be advantageous for other liquid samples containing accountable amounts of nuclear material throughout other SRNL laboratories. Similarly constructed paraffin vessel lids could also potentially be added to the approved list, though additional testing would be required.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Saltstone Waste Characterization Analysis - Salt Waste Processing Facility (SWPF) Waste Stream (Q4 CY2020)

The Saltstone Production and Disposal Facility is designed and permitted by the state of South Carolina Department of Health and Environmental Control (SCDHEC) to treat and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LEW) aqueous streams from the Effluent Treatment Project (ETP) and decontaminated solutions from the Tank Closure Cesium Removal Unit (TCCR) and the Salt Waste Processing Facility (SWPF) are stored in Tank 50 until the LEW can be transferred to the Saltstone Facility for treatment and disposal. In the past, decontaminated solution from the Modular Caustic Side Solvent Extraction Unit (MCU) was stored in Tank 50 until the LEW could be transferred to the Saltstone Facility for treatment and disposal. MCU is currently in a suspended operations state. LEW that meets the Waste Acceptance Criteria (WAC) can be transferred, stored, and treated in the Saltstone Production Facility (SPF) for subsequent disposal as saltstone grout in the Saltstone Disposal Facility (SDF). Sampling will be conducted as new waste streams are identified for treatment and disposal at the Saltstone Industrial Wastewater Treatment Facility (IWTF) and Z-Area Industrial Solid Waste Landfill (ISWLF), Facility ID# 025500-1603, General Condition B.9 or every six years in accordance with South Carolina (SC) Regulation 61-107.19 Parti C, “Solid Waste Management: Solid Waste Landfills and Structural Fill - General Requirements.”

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Tank 48H Tetraphenylborate Mitigation: Simulant Studies using Sodium Permanganate

Tank 48H currently holds legacy material containing organic tetraphenylborate (TPB) compounds from the operation of the In-Tank Precipitation (ITP) process. TPB was added during the ITP process to precipitate the otherwise soluble cesium as insoluble cesium TPB (CsTPB), but excessive benzene generation from TPB degradation curtailed this treatment method. The contents of Tank 48H, which include approximately 26.000 kg of potassium TPB (KTPB) and trace CsTPB, are not compatible with the waste treatment facilities at the Savannah River Site (SRS) since the organic content and the associated flammability issues pose a challenge to the salt processing and sludge processing facilities within the liquid waste system. An in-tank process to remove (or decompose) TPB safely would be of great value. Previous testing at Savannah River National Laboratory (SRNL) demonstrated the destruction of glycolate via chemical oxidation using sodium permanganate with simulated and radioactive waste. Scoping tests were performed to study the destruction of TPB to determine if the contents of Tank 48H would be amenable to the same type of destruction. Partial destruction of TPB was observed in Tank 48H simulants under mild conditions (e g., pH 11, room temperature) with no definitive indication of benzene generation. To build upon the success of the scoping tests, an additional study was requested to provide a better understanding of the underlying chemistry for Tank 48H content destruction using sodium permanganate. Three experiments were performed with Tank 48H simulants at 40 °C to determine the efficacy of using sodium permanganate for TPB destruction. Three starting pH values were selected: 1) pH 11 for comparison with the previous work at room temperature, 2) pH 10 as the minimum pH recommended by the Corrosion Control Program (CCP) for in-tank processing, and 3) pH 8 to determine the effectiveness of TPB destruction at near neutral pH. While below the allowable pH for the CCP, the experiment at pH 8 was performed to study the TPB-Permanganate reaction under more extreme conditions and further verify the potential for out-of-tank processing.

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