Search NASA⌕ Search

SEARCH · Search NASA

Results for “Waste Forms”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,513 records · Page 84

Criticality Safety Evaluation for 700 and 350 Gram Moderated Fissionable Equivalent (MFE) Criticality Control Areas (CCAs)

This evaluation is intended to generically apply to facilities and/or areas designated as CCAs that store and/or handle greater than 15 grams of fissionable material but no more than 700 grams of MFE. This evaluation includes facilities and/or areas designated as CCAs limited to no more than 350 grams of MFE. Activities performed in these CCAs may include general laboratory activities to include work in radiological areas, chemical procurement, material handling, sampling, imaging, non-destructive analysis, and disposing of nuclear and radiological materials. Materials included in these activities may include radioactive-contaminated oil, polychlorinated biphenyls (PCBs), and/or equipment. Fissionable materials may consist of transuranic (TRU) waste, sources containing fissionable material, uranium in various forms such as metal, oxide and silicide, however other forms of uranium and plutonium may be present. Fissionable materials may consist of fines from cutting, be solid forms mounted on a ring and embedded in epoxy or be in small volumes of dilute solutions resulting from cleaning or other processes. Major equipment located in these CCAs may include inert gas environment gloveboxes, fume hoods, ultrasonic testing equipment, hot presses, rolling mills, furnaces, gamma irradiators, autoclaves, machining equipment, tensile testing equipment, shears, forklifts and cranes.

07 ISOTOPE AND RADIATION SOURCES↗

Reinforced radiological containment bag

Radiological containment bags for use in storing alpha particle emitting waste are described. The radiological containment bags are formed of a multilayer film with an inner layer of the film being a sacrificial layer that lies between the interior of the bag and a structural polymeric layer. The sacrificial layer includes a flexible composite material that in turn includes a polymeric matrix and an alpha particle energy absorber incorporated in the polymeric matrix. The sacrificial layer can include a chromophore that can detect early degradation of the sacrificial material.

Washington, II, Aaron L.↗

Characterization Techniques Investigated for Characterization of Anomalous Materials in Plutonium Oxide – 25666

The DOE has adopted a Dilute and Dispose approach for processing surplus plutonium which consists of blending plutonium oxide with adulterants and packaging it in a form which is acceptable for disposal at the Waste Isolation Pilot Plant. The feed material for the downblend process is intended to be pure plutonium oxide powder, however other objects are occasionally encountered within the oxide, particularly with legacy material. Potential technologies which could assist in the resolution of incidents where such anomalies are encountered have been investigated. Resolution of these incidents requires characterization of the anomalous object so that plutonium oxide processing can resume and so that a disposition pathway can be determined for the anomalous material. Technologies including gamma ray spectroscopy, alpha-gamma coincidence, LiDAR volumetric measurements, and surface conductivity measurements were investigated for this purpose, with a focus on systems which can be easily introduced into the processing environment when an anomaly is encountered, then removed from the environment after resolution to avoid impeding normal processing activities. This places an emphasis on small and portable measurement systems and systems that can operate in a high-background, oxide-processing environment. Gamma ray systems investigated include the GR1™a CZT detector and the MicroGe™a germanium detector, with a focus on detecting characteristic gamma rays from Pu-239 and other actinides. A gamma ray and alpha particle coincidence method was investigated with the goal of identifying actinides in the presence of a gamma ray background produced by adjacent plutonium oxide material. Leica™b BLK360 G1 and Keyence™c LJ-X8300 LiDAR systems were investigated for use in conjunction with mass measurements to gain accurate material density values, and a Foerster Sigmatest™d 2.070 was tested to determine surface conductivity. The combination of these properties would allow improved identification and characterization of a wide variety of potential anomalous material.

Munson, Justin M.↗

Reactions of Studtite UO 4 ·4H 2 O in Alkali Hydroxides: Isolation of Single-Crystal Uranate Phases under Mild Hydrothermal Conditions

Uranyl peroxide complexes, particularly studtite (UO 4 ·4H 2 O), are important phases within the nuclear fuel cycle, forming through radiolysis-induced reactions on the surfaces of spent fuel and in waste environments. Studtite has been identified in Hanford’s irradiated fuel storage basins, Chornobyl’s corium lavas, and is anticipated on Fukushima’s fuel debris. While its formation and stability have been extensively studied, the reactivity of UO 4 ·4H 2 O in highly alkaline environments such as those encountered in high-level nuclear waste remains underexplored. These environments contain molar concentrations of [OH - ] and present a chemically dynamic and reactive environment where actinide behavior is not well understood. Reported here are investigations of uranium reactions in concentrated alkali hydroxides under mild hydrothermal conditions (<200 °C) that have resulted in the isolation of the alkali metal uranates Li 2 UO 4 , α-Na 2 UO 4 , γ-Na 2 U 2 O 7 , and K 2 U 2 O 7 . In contrast to conventional solid-state methods (>800 °C) that typically yield polycrystalline powders, our approach enables the isolation of these uranates as single crystals, allowing us to provide single-crystal structure solutions of certain uranates for the first time. Our findings match the results of high-level waste tank sampling, confirming that sodium diuranate (Na 2 U 2 O 7 ) is the most persistent uranium phase, potentially forming through reactions of uranyl peroxide intermediates with NaOH under radiolytic and highly alkaline conditions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

New Perspectives in Thermoelectric Energy Recovery System Design Optimization

Abstract: Large amounts of waste heat are generated worldwide in industrial processes, automotive transportation, diesel engine exhaust, military generators, and incinerators because 60-70% of the fuel energy is typically lost in these processes. There is a strong need to develop technologies that recover this waste heat to increase fuel efficiency and minimize fuel requirements in these industrial processes, automotive and heavy vehicle engines, diesel generators, and incinerators. There are additional requirements to reduce CO2 production and environmental footprints in many of these applications. Recent work with the Strategic Environmental Research and Development Program office has investigated new thermoelectric (TE) materials and systems that can operate at higher performance levels and show a viable pathway to lightweight, small form-factor, advanced thermoelectric generator (TEG) systems to recover waste heat in many of these applications. New TE materials include nano-composite materials such as lead-antimony-silver-telluride (LAST) and lead-antimony-silver-tin-telluride (LASTT) compounds. These new materials have created opportunities for high-performance, segmented-element TE devices. New higher-performance TE devices segmenting LAST/LASTT materials with bismuth telluride have been designed and fabricated. Sectioned TEG systems using these new TE devices and materials have been designed. Integrated heat exchanger/TE device system analyses of sectioned TE system designs have been performed creating unique efficiency-power maps that provide better understandings and comparisons of design tradeoffs and nominal and off-nominal system performance conditions. New design perspectives in optimization of sectioned TE design approaches are discussed that provide insight on how to optimize such sectioned TE systems. System performance analyses using ANSYS® TE modeling capabilities have integrated heat exchanger performance models with ANSYS® TE models to extend its analysis capabilities beyond simple constant hot-side and cold-side temperature conditions . Analysis results portray external resistance effects, matched load conditions, maximum power vs. maximum efficiency points simultaneously.

Energy Conversion↗

THE EFFECT OF THE PRESENCE OF OZONE ON THE LOWER FLAMMABILITY LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE

The Enhanced Chemical Cleaning (ECC) process uses ozone to effect the oxidation of metal oxalates produced during the dissolution of sludge in the Savannah River Site (SRS) waste tanks. The ozone reacts with the metal oxalates to form metal oxide and hydroxide precipitants, and the CO{sub 2}, O{sub 2}, H{sub 2}O and any unreacted O{sub 3} gases are discharged into the vapor space. In addition to the non-radioactive metals in the waste, however, the SRS radioactive waste also contains a variety of radionuclides, hence, hydrogen gas is also present in the vapor space of the ECC system. Because hydrogen is flammable, the impact of this resultant gas stream on the Lower Flammability Limit (LFL) of hydrogen must be understood for all possible operating scenarios of both normal and off-normal situations, with particular emphasis at the elevated temperatures and pressures of the typical ECC operating conditions. Oxygen is a known accelerant in combustion reactions, but while there are data associated with the behavior of hydrogen/oxygen environments, recent, relevant studies addressing the effect of ozone on the flammability limit of hydrogen proved scarce. Further, discussions with industry experts verified the absence of data in this area and indicated that laboratory testing, specific to defined operating parameters, was needed to comprehensively address the issue. Testing was thus designed and commissioned to provide the data necessary to support safety related considerations for the ECC process. A test matrix was developed to envelope the bounding conditions considered credible during ECC processing. Each test consists of combining a gas stream of high purity hydrogen with a gas stream comprised of a specified mixture of ozone and oxygen in a temperature and pressure regulated chamber such that the relative compositions of the two streams are controlled. The gases are then stirred to obtain a homogeneous mixture and ignition attempted by applying 10J of energy to a fuse wire. A gas combination is considered flammable when a pressure rise of 7% of the initial absolute pressure is observed. The specified testing methodology is consistent with guidelines established in ASTM E-918-83 (2005) 'Standard Practices for Determining Limits of Flammability of Chemicals at Elevated Temperature and Pressure'.

08 HYDROGEN↗

The Effect of the Presence of Ozone on the Lower Flammability Limit (LFL) of Hydrogen in Vessels Containing Savannah River Site High Level Waste - 12387

The Enhanced Chemical Cleaning (ECC) process uses ozone to effect the oxidation of metal oxalates produced during the dissolution of sludge in the Savannah River Site (SRS) waste tanks. The ozone reacts with the metal oxalates to form metal oxide and hydroxide precipitants, and the CO{sub 2}, O{sub 2}, H{sub 2}O and any unreacted O{sub 3} gases are discharged into the vapor space. In addition to the non-radioactive metals in the waste, however, the SRS radioactive waste also contains a variety of radionuclides, hence, hydrogen gas is also present in the vapor space of the ECC system. Because hydrogen is flammable, the impact of this resultant gas stream on the Lower Flammability Limit (LFL) of hydrogen must be understood for all possible operating scenarios of both normal and off-normal situations, with particular emphasis at the elevated temperatures and pressures of the typical ECC operating conditions. Oxygen is a known accelerant in combustion reactions, but while there are data associated with the behavior of hydrogen/oxygen environments, recent, relevant studies addressing the effect of ozone on the flammability limit of hydrogen proved scarce. Further, discussions with industry experts verified the absence of data in this area and indicated that laboratory testing, specific to defined operating parameters, was needed to comprehensively address the issue. Testing was thus designed and commissioned to provide the data necessary to support safety related considerations for the ECC process. A test matrix was developed to envelope the bounding conditions considered credible during ECC processing. Each test consists of combining a gas stream of high purity hydrogen with a gas stream comprised of a specified mixture of ozone and oxygen in a temperature and pressure regulated chamber such that the relative compositions of the two streams are controlled. The gases are then stirred to obtain a homogeneous mixture and ignition attempted by applying 10J of energy to a fuse wire. A gas combination is considered flammable when a pressure rise of 7% of the initial absolute pressure is observed. The specified testing methodology is consistent with guidelines established in ASTM E-918-83 (2005) 'Standard Practices for Determining Limits of Flammability of Chemicals at Elevated Temperature and Pressure'. The LFL of hydrogen in air was determined and is in good agreement with the literature data. Ozone-oxygen mixtures were found to be flammable at concentrations above 8.3 vol.% based on the ASTM E918 7% pressure rise criteria for flame propagation. This result is lower than previously reported values which can be explained through the variations in the test setup and procedure. It is believed that the lower values obtained in this work are a result of improvements of the test methodology. Tests performed with hydrogen in various concentrations of ozone in oxygen have shown that the LFL of hydrogen decreases as the concentration of ozone in the mixture increases. This testing was designed to provide data under the conditions considered most optimal to produce deflagration. The geometry and materials of construction of the testing vessel; the location of the fuse wire; the magnitude of the supplied energy; the careful minimization of diluents and other contaminants; and meticulous procedural detail to maintain integrity of the ozone to the maximum extent practical, result in data that reflect not the expected process conditions, but those that enhance the possibility of flame propagation. For this reason, there is believed to be considerable conservatism in the indicated results. Per the vendor, the maximum possible ozone concentration producible by the planned ECC Ozone generator is 8 volume percent (the typical maximum operating setpoint concentration is 6.8 vol%), less than the 8.3 minimum volume % concentration shown to be flammable in a 99.999% pure O{sub 2} environment at the optimally conservative conditions established in this testing. Further, the feed to the ECC ozone generator is only 87% oxygen, the remainder, water vapor and nitrogen, both powerful diluents. It is, thus, believed not credible that deflagration can occur at this maximum feed concentration condition. In addition, once the ozone stream contacts the waste stream, the many simultaneous oxidizing reactions will rapidly decompose available ozone to well below flammable levels. Further, because the radiolytically generated hydrogen quantity is negligible compared to the supplied ozone/oxygen stream (0.0004 moles per minute H{sub 2} vs 76 moles per minute ozone/oxygen), even the total H{sub 2}/O{sub 3} mixture, without crediting decomposition reactions, does not approach flammable concentrations. Finally, even at the 'end' of the ECC batch cycle, when most of the metal oxalates have been decomposed, testing has indicated that the ozone concentration in the vapor space of the ECC process vessel reaches a concentration of no more than 3 vol%, remaining well below concentrations of concern. The major issue for the ECC operation established by this testing is the impact of the data when applied to off normal conditions. While it is possible to discontinue ozone addition to the reaction vessel at any time, the radiolytic hydrogen generation rate continues, varying slightly as ambient pressures and temperatures change. Relative to the data generated and analyzed in this testing, the ECC hazards analysis team will re-evaluate off normal conditions (e.g. those during which process exhaust ventilation is lost) such that issues involving mixtures of hydrogen and ozone in the vapor space can be appropriately controlled. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Determination of Pertechnetate and Non-Pertechnetate Species in Hanford Tanks - Phase 1

An analytical method is required to distinguish between the pertechnetate and non-pertechnetate forms of technetium; currently, the methods available only report the total technetium present in a sample. The overall objective of this effort is to develop a method for routinely analyzing Hanford tank waste for technetium in the pertechnetate and the non-pertechnetate forms. A phased approach will be deployed to accomplish this objective: Phase I Comparison of existing technetium analysis methods with modification; Phase II Demonstration of modified methods using non-pertechnetate spiked simulants; and, Phase III Demonstration of chosen method on Hanford tank sample containing non-pertechnetate. This report describes the Phase I work, providing a comparison of Aliquat 336 and TEVA(R)1 in the removal of pertechnetate and discussing the subsequent analysis for technetium in both alkaline and acidic environments without oxidation. The effort was executed under LAB-PLN-13-00004, The Determination of Pertechnetate and Non-Pertechnetate Species in Hanford Tanks Phase I.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor

The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Acetonitrile Destruction in ETF Feed Solutions by UV/OX and Persulfate (Final Report)

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams that include condensates from the 242-A evaporator, leachate from the Environmental Restoration Disposal Facility (ERDF), as well as laboratory wastes and, in the future, will treat liquid effluents from the Hanford Waste Treatment and Immobilization Plant (WTP) and Integrated Disposal Facility (IDF) leachate. Liquid effluents from the WTP will have significant concentrations of acetonitrile. Acetonitrile is formed by reaction of nitrates and sugar in the WTP low activity waste (LAW) melters and is prevalent in the submerged bed scrubber (SBS) and wet electrostatic precipitator (WESP) liquid effluents from WTP off-gas treatment. When these liquids are concentrated in the WTP Effluent Management Facility (EMF) evaporator in the direct feed low activity waste (DFLAW) flow-sheet, testing has shown that the majority of the acetonitrile partitions to the evaporator condensate. Since the evaporator condensate is directed to the ETF, this creates a potential issue with the ETF waste acceptance criteria and, specifically, the ETF Delisting Modification treatability envelope limits. Consequently, there is a need to assess possible mitigation approaches, one of which involves identification and assessment of potential treatment processes to destroy the acetonitrile, which is the subject of the present report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Testing of UV Oxidation Process for Acetonitrile Destruction

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams that include condensates from the 242-A evaporator, leachate from the Environmental Restoration Disposal Facility (ERDF), as well as laboratory wastes and, in the future, will treat liquid effluents from the Hanford Waste Treatment and Immobilization Plant (WTP) and Integrated Disposal Facility (IDF) leachate. Liquid effluents from the WTP will have significant concentrations of acetonitrile. Acetonitrile is formed by reaction of nitrates and sugar in the WTP low activity waste (LAW) melters and is prevalent in the submerged bed scrubber (SBS) and wet electrostatic precipitator (WESP) liquid effluents from WTP off-gas treatment. When these liquids are concentrated in the WTP Effluent Management Facility (EMF) evaporator in the direct feed low activity waste (DFLAW) flow-sheet, testing has shown that the majority of the acetonitrile partitions to the evaporator condensate. Since the evaporator condensate is directed to the ETF, this creates a potential issue with the ETF waste acceptance criteria and, specifically, the ETF Delisting Modification treatability envelope limits. Consequently, there is a need to assess possible mitigation approaches, one of which involves identification and assessment of potential treatment processes to destroy the acetonitrile, which is the subject of the present report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Acetonitrile Destruction and Fate of Organics in the Reverse Osmosis System at the ETF

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams that include condensates from the 242-A evaporator, leachate from the Environmental Restoration Disposal Facility (ERDF), as well as laboratory wastes and, in the future, will treat liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) and Integrated Disposal Facility (IDF) leachate. Liquid effluents from the WTP will have significant concentrations of acetonitrile. Acetonitrile is formed by reaction of nitrates and sugar in the WTP low activity waste (LAW) melters and is prevalent in the submerged bed scrubber (SBS) and wet electrostatic precipitator (WESP) liquid effluents from WTP off-gas treatment. When these liquids are concentrated in the WTP Effluent Management Facility (EMF) evaporator in the direct feed low activity waste (DFLAW) flow-sheet, testing has shown that the majority of the acetonitrile partitions to the evaporator condensate. Since the evaporator condensate is directed to the ETF, this creates a potential issue with the ETF waste acceptance criteria. Consequently, there is a need to validate flow-sheet assumptions on the fate of acetonitrile and other organics within the ETF. The present plan includes the addition of a steam stripper to the ETF to remove acetonitrile. There is, therefore, also a need to determine a suitable method to destroy acetonitrile in the overhead condensate stream from the new steam stripper. Washington River Protection Solutions, LLC (WRPS) previously contracted with Atkins and the Vitreous State Laboratory (VSL) of The Catholic University of America (CUA) to perform development and testing work to evaluate potential methods for destruction of acetonitrile in WTP secondary liquid effluents. Based on the results of that work, WRPS requested that follow-on testing be conducted to further evaluate acetonitrile destruction in the steam stripper condensate using ultraviolet oxidation (UV/OX) with persulfate. WRPS also requested testing to assess the rejection rate of organics in the reverse osmosis (RO) system installed in the ETF. This report presents the results from testing to address those needs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Flowsheet for Conversion of Anion Resin from Nitrate to Sulfate Form for Disposal

Nitrate form anion exchange resins are specifically prohibited for disposal in TRU waste. At some DOE sites, spent resin was mixed with cement to immobilize the resin in a monolith with varying success. At SRS starting in the 1980s, spent anion resin has generally been converted to a more chemically stable form or it was digested with permanganate and transferred as a slurry to the HLW tank farm. In the early years of the site it is reported that over 20 metric tons of organic resin were transferred directly to the tank farm without any attempt to digest it. It appears that in the future conversion to a stable sulfate form could well be the favored disposal path for spent nitrate-form anion resin. This report will document the site history of converting between ionic forms (nitrate to sulfate) and will document process flowsheet conditions for performing that conversion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Picolinate-based acyclic ligand for rare earth element extraction and separation

The rare earth elements (REEs) play an important role in many modern technologies, particularly those relevant to clean energy. Despite their increasing importance, obtaining them in elementally pure forms suitable for downstream applications is challenging due to their similar chemical properties. This problem has impeded efforts to efficiently and selectively extract them from end-of-life materials and electronic waste. Here, we report a cost-efficient acyclic picolinate-based chelator H 4 aapa. The REE stability constants (log K ML ) of this chelator were measured via pH potentiometric and UV-Visible spectrophotometric titrations, revealing it to preferably bind light over heavy REEs like many recently reported 18-membered macrocycles. Its REE complexes were characterized by X-ray crystallography and NMR spectroscopy, demonstrating that this chelator can attain different conformations. The unique properties of aapa were subsequently used to separate REEs via the dissolution of insoluble REE oxalate mixtures. This dissolution-based separation led to large separation factors, the most significant being that for the Ce 3+ /Lu 3+ pair (38.6) at pH 4. Leveraging the strong REE binding affinity of aapa, we further demonstrated this chelator can leach REEs from authentic end-of-life materials in the form of magnet waste and autocatalyst smelting (autocat) slag. With this approach, exposure of these materials to a 20 mM solution of aapa at neutral pH generates a metal-containing solution enriched in Nd 3+ and Dy 3+ by 56.9 wt% and 3.0 wt%, marking a 4-fold improvement over the use of 4 M HNO 3 .

Gao, Yangyang [University of California, Santa Bar↗

Results for the October 2023 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the October 2023 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the SWPF DSS sample was obtained. The October 2023 Semiannual sample of the SWPF DSS is a composite from the six months of SWPF processing during the Third Quarter Fiscal Year 2023 (3QFY2023) and the Fourth Quarter Fiscal Year 2023 (4QFY2023).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Results for the April 2024 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the April 2024 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the SWPF DSS sample was obtained.1 The April 2024 Semiannual sample of the SWPF DSS is a composite from the six months of SWPF processing during the First Quarter Fiscal Year 2024 (1QFY2024) and the Second Quarter Fiscal Year 2024 (2QFY2024). The following facts pertaining to the WAC are drawn from the analytical results provided in this report. WAC TARGETS and LIMITS were met for all analyzed chemical and radioactive contaminants for which the detection limits are below the WAC TARGETS and LIMITS. Nitrosamines were not detected in the SWPF DSS salt solution sample above the instrument detection limits of <1 mg/L. The minimum detection limit (<3.33E-01 pCi/mL) is reported for 94 Nb as determined from the minimum detectable activity associated with the radiochemical method used for this radionuclide. The reported detection limit is above the requested SRMC target minimum detection limit concentration. However, the minimum detection limit reported for the April 2024 semiannual SWPF DSS sample for 94 Nb is lower than the estimated detection limit of 4.38E-01 pCi/mL initially established by SRNL in 2009. Thus, per guidance from SRMC, 2 SRNL continues to achieve as low as practical detection limits for this radionuclide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Results for the April 2025 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the April 2025 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the SWPF DSS sample was obtained. The April 2025 Semiannual sample of the SWPF DSS is a composite from the six months of SWPF processing during the First Quarter Fiscal Year 2025 (1QFY2025) and the Second Quarter Fiscal Year 2025 (2QFY2025).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Results for the October 2025 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the October 2025 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TAR GETS that were established at the time the SWPF DSS sample was obtained. 1 The October 2025 Semiannual sample of the SWPF DSS is a composite from one month of SWPF processing during the Third Quarter Fiscal Year 2025 (3QFY2025) and two months of SWPF processing during the Fourth Quarter Fiscal Year 2025 (4QFY2025).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗