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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.

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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

SOLIDS PRECIPITATION EVENT IN MCU CAUSAL ANALYSIS AND RECOMMENDATIONS FROM SOLIDS RECOVERY TEAM

A process upset occurred in the Modular Caustic-Side Solvent Extraction Unit (MCU) facility on April 6th, 2014. During recovery efforts, a significant amount of solids were found in the Salt Solution Feed Tank (SSFT), Salt Solution Receipt Tanks (SSRTs), two extraction contactors, and scrub contactors. The solids were identified by Savannah River National Laboratory (SRNL) as primarily sodium oxalate and sodium alumina silicate (NAS) with the presence of some aluminum hydroxide. NAS solids have been present in the SSFT since simulant runs during cold chemical startup of MCU in 2007, and have not hindered operations since that time. During the process upset in April 2014, the oxalate solids partially blocked the aqueous outlet of the extraction contactors, causing salt solution to exit through the contactor organic outlet to the scrub contactors with the organic phase. This salt solution overwhelmed the scrub contactors and passed with the organic phase to the strip section of MCU. The partially reversed flow of salt solution resulted in a Strip Effluent (SE) stream that was high in Isopar™ L, pH and sodium. The primary cause of the excessive solids accumulation in the SSRTs and SSFT at MCU is attributed to an increase in the frequency of oxalic acid cleaning of the 512-S primary filter. Agitation in the SSRTs at MCU in response to cold weather likely provided the primary mechanism to transfer the solids to the contactors. Sources of the sodium oxalate solids are attributed to the oxalic acid cleaning solution used to clean the primary filter at the Actinide Removal Process (ARP) filtration at 512-S, as well as precipitation from the salt batch feed, which is at or near oxalate saturation. The Solids Recovery Team was formed to determine the cause of the solids formation and develop recommendations to prevent or mitigate this event in the future. A total of 53 recommendations were generated. These recommendations were organized into 4 focus areas: • Improve understanding of oxalate equilibrium and kinetics in salt solutions • Reduction/elimination of oxalic acid cleaning in 512-S • Flowsheet optimization • Improving diagnostic capability The recommendations implemented prior to resumption of MCU operations provide a risk mitigation or detection function through additional sampling and observation. The longer term recommendations provide a framework to increase the basic process knowledge of both oxalate chemistry and filtration behavior and then facilitate decisions that improve the salt flowsheet as a system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

RH-TRU Waste Streams: HFDA Crucibles Create New Complex Waste Stream

This project focuses on the removal of the Hot Fuel Dissolution Apparatus (HFDA) crucibles from the Hot Fuel Examination Facility’s (HFEF) hot cells to create space for upcoming projects. The project team has created a crucible and pyro-salt non-debris waste stream that can be disposed of at Waste Isolation Pilot Plant (WIPP). Chemical Compatibility Examinations (CCEs) must be performed on candidate waste by reviewing the types of chemicals used in the experiments, information found in INL Process Knowledge Summary Reports (PKSRs), and Environmental Protection Agency guides on chemical compatibility. The CCEs identify which chemicals can be safely combined to ensure adverse reactions do not occur in the waste packages. The CCEs are then used to update the PKSR, Basis of Knowledge, and Acceptable Knowledge documents and demonstrate compliance with the WIPP Waste Acceptance Criteria. Once documents are finalized, they are reviewed and approved by the WIPP Central Characterization Project (CCP) to ensure proper waste stream characterization. After the documents are approved, the crucibles can be packaged and shipped to the Idaho Nuclear Technical and Engineering Center for storage awaiting certification. CCP will certify the waste using the approved documents and a series of confirmatory testing before being shipped and disposed in WIPP’s deep geological repository. Currently, the CCE is in progress and the PKSRs are being updated.

Basis of Knowledge↗
DOE OSTI2013

Public involvement on closure of Asse II radioactive waste repository in Germany

From 1967 to 1978, about 125,800 barrels of low- and intermediate level waste were disposed of - nominally for research purposes - in the former 'Asse' salt mine which had before been used for the production of potash for many years. Since 1988 an inflow of brine is being observed which will cause dangers of flooding and of a collapse due to salt weakening and dissolution if it should increase. Since several years the closure of the Asse repository is planned with the objective to prevent the flooding and collapse of the mine and the release of radioactive substances to the biosphere. The first concept that was presented by the former operator, however, seemed completely unacceptable to regional representatives from politics and NGOs. Their activities against these plans made the project a top issue on the political agenda from the federal to the local level. The paper traces the main reasons which lead to the severe safety problems in the past as well as relevant changes in the governance system today. A focus is put on the process for public involvement in which the Citizens' Advisory Group 'A2B' forms the core measure. Its structure and framework, experience and results, expectations from inside and outside perspectives are presented. Furthermore the question is tackled how far this process can serve as an example for a participatory approach in a siting process for a geological repository for high active waste which can be expected to be highly contested in the affected regions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

SuperLig® 639 Resin Performance using 8 Molar LAW Feed

All prior testing with SuperLig® 639 has been done with the aqueous concentration of LAW at ~5 M [Na+], where the resin sinks, and can be used in a conventional down-flow column orientation. However, the aqueous LAW stream from the Waste Treatment Plant is expected to be ~8 M [Na+]. The resin would float in this higher density liquid, potentially disrupting the ability to achieve a good decontamination due to poor packing of the resin that leads to channeling. Testing was completed with a higher salt concentration in the feed simulant (7.8 M [Na+]) in an engineering-scale apparatus with two columns, each containing ~0.9 L of resin. Testing of this system used a simulant of the LAW solution, and substituted ReO4 - as a surrogate for TcO4 -. Results were then compared using computer modeling. Bench-scale testing was also performed, and examined an unconstrained resin bed, while engineering-scale tests used both constrained and unconstrained beds in a two-column, lead and lag sequential arrangement.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Characterization of Tank 16 Annulus Samples

The closure of Tank 16H will require removal of material from the annulus of the tank. Samples from Tank 16H annulus were characterized and tested to provide information to evaluate various alternatives for removing the annulus waste. The analysis found all four annulus samples to be composed mainly of Si, Na, and Al and lesser amounts of other elements. The XRD data indicate quartz (SiO 2 ) and sodium aluminum nitrate silicate hydrate (Na 8 (Al 6 Si 6 O 24 )(NO 3 ) 2 .4H 2 O) as the predominant crystalline mineral phases in the samples. The XRD data also indicate the presence of crystalline sodium nitrate, sodium nitrite, gibbsite, hydrated sodium bicarbonate, and muscovite. Based on the weight of solids remaining at the end of the test, the water leaching test results indicate approximately 20-35% of the solids dissolved after three contacts with an approximately 3:1 volume of water at 45 C. The chemical analysis of the leachates and the XRD results of the remaining solids indicate sodium salts of nitrate, nitrite, sulfate, and possibly carbonate/bicarbonate make up the majority of the dissolved material. The majority of these salts were dissolved in the first water contact and simply diluted with each subsequent water contact. The water leaching removed large amounts of the uranium in two of the samples and {approx}1/3 of the {sup 99}Tc from all four samples. Most of the other radionuclides analyzed showed low solubility in the water leaching test. The preliminary data on the oxalic acid leaching test indicate the three acid contacts at 45 C dissolved from {approx}34-47% of the solids. The somewhat higher dissolution found in the oxalic acid leaching test versus the water leaching test might be offset by the tendency of the oxalic acid solutions to take on a gel-like consistency. The filtered solids left behind after three oxalic acid contacts were sticky and formed large clumps after drying. These two observations could indicate potential processing difficulties with solutions and solids from oxalic acid leaching. The gel formation might be avoided by using larger volumes of the acid. Further testing would be recommended before using oxalic acid to dissolve the Tank 16H annulus waste to ensure no processing difficulties are encountered in the full scale process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Waste Stream Treatment and Waste Form Fabrication for Pyroprocessing of Used Nuclear Fuel

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Characterization of Tank 16H Annulus Samples Part II: Leaching Results

The closure of Tank 16H will require removal of material from the annulus of the tank. Samples from Tank 16H annulus were characterized and tested to provide information to evaluate various alternatives for removing the annulus waste. The analysis found all four annulus samples to be composed mainly of Si, Na, and Al and lesser amounts of other elements. The XRD data indicate quartz (SiO 2 ) and sodium aluminum nitrate silicate hydrate (Na 8 (Al 6 Si 6 O 24 )(NO 3 ) 2 .4H 2 O) as the predominant crystalline mineral phases in the samples. The XRD data also indicate the presence of crystalline sodium nitrate (NaNO 3 ), sodium nitrite (NaNO 2 ), gibbsite (Al(OH) 3 ), hydrated sodium bicarbonate (Na 3 H(CO 3 ) 2 .2H 2 O), and muscovite (KAl 2 (AlSi 3 O 10 )(OH) 2 ). Based on the weight of solids remaining at the end of the test, the water leaching test results indicate 20-35% of the solids dissolved after three contacts with an approximately 3:1 volume of water at 45°C. The chemical analysis of the leachates and the XRD results of the remaining solids indicate sodium salts of nitrate, nitrite, sulfate, and possibly carbonate/bicarbonate make up the majority of the dissolved material. The majority of these salts were dissolved in the first water contact and simply diluted with each subsequent water contact. The water leaching removed large amounts of the uranium in two of the samples and approximately 1/3 of the 99 Tc from all four samples. Most of the other radionuclides analyzed showed low solubility in the water leaching test. The oxalic acid leaching test result indicate approximately 34-47% of the solids in the four annulus samples will dissolve after three contacts with an approximately 3:1 volume of acid to solids at 45°C. The same sodium salts found in the water leaching test comprise the majority of dissolved material in the oxalic acid leaching test. However, the oxalic acid was somewhat more effective in dissolving radionuclides than the water leach. In contrast to the water leaching results, most constituents continued to dissolve during subsequent cycles of oxalic acid leaching. The somewhat higher dissolution found in the oxalic acid leaching test versus the water leaching test might be offset by the tendency of the oxalic acid solutions to take on a gel-like consistency. The filtered solids left behind after three oxalic acid contacts were sticky and formed large clumps after drying. These two observations could indicate potential processing difficulties with solutions and solids from oxalic acid leaching. The gel formation might be avoided by using larger volumes of the acid. Further testing would be recommended before using oxalic acid to dissolve the Tank 16H annulus waste to ensure no processing difficulties are encountered in the full scale process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences. Developing models, testing material, characterizing processes, and analyzing performance all have overlapping application regardless of the salt formation of interest.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences. Developing models, testing material, characterizing processes, and analyzing performance all have overlapping application regardless of the salt formation of interest.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Study on a regeneration process of LiCl-KCl eutectic based waste salt generated from the pyrochemical process

A regeneration process of LiCl-KCl eutectic waste salt generated from the pyrochemical process of spent nuclear fuel has been studied. This regeneration process is composed of a chemical conversion process and a vacuum distillation process. Through the regeneration process, a high efficiency of renewable salt recovery can be obtained from the waste salt and rare earth nuclides in the waste salt can be separated as oxide or phosphate forms. Thus, the regeneration process can contribute greatly to a reduction of the waste volume and a creation of durable final waste forms. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Advances in Geologic Disposal System Modeling and Application to Crystalline Rock

The Used Fuel Disposition Campaign (UFDC) of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of used nuclear fuel (UNF) and high-level nuclear waste (HLW). Two of the high priorities for UFDC disposal R&D are design concept development and disposal system modeling (DOE 2011). These priorities are directly addressed in the UFDC Generic Disposal Systems Analysis (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media (e.g., salt, granite, clay, and deep borehole disposal). This report describes specific GDSA activities in fiscal year 2016 (FY 2016) toward the development of the enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. The GDSA framework employs the PFLOTRAN thermal-hydrologic-chemical multi-physics code and the Dakota uncertainty sampling and propagation code. Each code is designed for massively-parallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

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DOE OSTI2015

OE-WIPP Event Presentation

Information is given on waste generation at TA-55 and remediation needed to meet WIPP acceptance criteria, including the role of nitrate salts. Breaching of a particular waste-filled drum is reviewed, along with an accident analysis and steps for corrective actions and improved process management.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Analytical results from routine DSSHT and SEHT monthly samples

Strip Effluent Hold Tank (SEHT) and Decontaminated Salt Solution Hold Tank (DSSHT) samples from several of the “microbatches” of Integrated Salt Disposition Project (ISDP) Salt Batch (“Macrobatch”) 8B have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (Inductively Coupled Plasma Emission Spectroscopy - ICPES), and anions (Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatch samples. The Cs removal continues to be excellent, with decontamination factors (DF) averaging 22,100 (114% RSD). The bulk chemistry of the DSSHT and SEHT samples does not show any signs of unusual behavior, other than lacking the anticipated degree of dilution that is calculated to occur during Modular Caustic-Side Solvent Extraction Unit (MCU) processing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Summary of FY11 Sulfate Retention Studies for Defense Waste Processing Facility Glass

This report describes the results of studies related to the incorporation of sulfate in high level waste (HLW) borosilicate glass produced at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). A group of simulated HLW glasses produced for earlier sulfate retention studies was selected for full chemical composition measurements to determine whether there is any clear link between composition and sulfate retention over the compositional region evaluated. In addition, the viscosity of several glasses was measured to support future efforts in modeling sulfate solubility as a function of predicted viscosity. The intent of these studies was to develop a better understanding of sulfate retention in borosilicate HLW glass to allow for higher loadings of sulfate containing waste. Based on the results of these and other studies, the ability to improve sulfate solubility in DWPF borosilicate glasses lies in reducing the connectivity of the glass network structure. This can be achieved, as an example, by increasing the concentration of alkali species in the glass. However, this must be balanced with other effects of reduced network connectivity, such as reduced viscosity, potentially lower chemical durability, and in the case of higher sodium and aluminum concentrations, the propensity for nepheline crystallization. Future DWPF processing is likely to target higher waste loadings and higher sludge sodium concentrations, meaning that alkali concentrations in the glass will already be relatively high. It is therefore unlikely that there will be the ability to target significantly higher total alkali concentrations in the glass solely to support increased sulfate solubility without the increased alkali concentration causing failure of other Product Composition Control System (PCCS) constraints, such as low viscosity and durability. No individual components were found to provide a significant improvement in sulfate retention (i.e., an increase of the magnitude necessary to have a dramatic impact on blending, washing, or waste loading strategies for DWPF) for the glasses studied here. In general, the concentrations of those species that significantly improve sulfate solubility in a borosilicate glass must be added in relatively large concentrations (e.g., 13 to 38 wt % or more of the frit) in order to have a substantial impact. For DWPF, these concentrations would constitute too large of a portion of the frit to be practical. Therefore, it is unlikely that specific additives may be introduced into the DWPF glass via the frit to significantly improve sulfate solubility. The results presented here continue to show that sulfate solubility or retention is a function of individual glass compositions, rather than a property of a broad glass composition region. It would therefore be inappropriate to set a single sulfate concentration limit for a range of DWPF glass compositions. Sulfate concentration limits should continue to be identified and implemented for each sludge batch. The current PCCS limit is 0.4 wt % SO 4 2- in glass, although frit development efforts have led to an increased limit of 0.6 wt % for recent sludge batches. Slightly higher limits (perhaps 0.7-0.8 wt %) may be possible for future sludge batches. An opportunity for allowing a higher sulfate concentration limit at DWPF may lay lie in improving the laboratory experiments used to set this limit. That is, there are several differences between the crucible-scale testing currently used to define a limit for DWPF operation and the actual conditions within the DWPF melter. In particular, no allowance is currently made for sulfur partitioning (volatility versus retention) during melter processing as the sulfate limit is set for a specific sludge batch. A better understanding of the partitioning of sulfur in a bubbled melter operating with a cold cap as well as the impacts of sulfur on the off-gas system may allow a higher sulfate concentration limit to be established for the melter feed. This approach would have to be taken carefully to ensure that a sulfur salt layer is not formed on top of the melt pool while allowing higher sulfur based feeds to be processed through DWPF.

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

Stabilization of Uranium Metal by Chloride Conversion-Economical, Safe, Scalable [Factsheet]

Researchers at Los Alamos National Laboratory have developed SUCCESS, a process to convert uranium metal to uranium tetrachloride enabling production of these valuable compounds for actinide science, molten salt nuclear reactors, and energy research. Conventional methods are problematic due to the oxide reduction processes use rare reagents and generate significant amounts of waste and hydride conversion process using chlorine at high temperature is hazardous. This new method using inexpensive reagents chemically transform uranium metal into uranium halide compounds under mild conditions. The facile conversion to stable uranium tetrachloride could reduce the need to mine uranium for uses other than nuclear, by recycling the uranium without oxidation. The Laboratory is interested in supporting further scale-up efforts using a Cooperative Research and Development Agreement (CRADA) or by granting a license to a qualified entity

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Simplifying Ceramic Waste Form Processing

The glass-bonded sodalite ceramic waste form (CWF) was developed in the 1990s to immobilize chloride salt-bearing waste streams generated during treatment of spent sodium-bonded nuclear fuel. This was necessary because chloride salts are essentially insoluble in borosilicate glasses (on the order of 1%). In the CWF, synthetic sodalite (Na 8 [Al 6 Si 6 O 24 ]Cl 2 ) is generated by the reaction of zeolite 4A and NaCl during processing to sequester the chlorine. The resulting sodalite crystals become microencapsulated in borosilicate glass. Based on the current understanding of the reactions involved and advancements in pyroprocessing operations in which the waste salt is generated, the processes developed for producing fullsize CWF materials with waste salt may be unnecessarily complicated. Aspects of processing that could be simplified and initial studies to justify further engineering development are recommended.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Tank 48H Phase 1 Initial Testing Using Sodium Permanganate to Decompose Tetraphenylborate: Simulant Studies with Shaker Oven and 2-L Vessel

Tank 48H currently holds legacy material including organic tetraphenylborate (TPB) compounds from the operation of the In-Tank Precipitation process. The large quantity of TPB is not compatible with the waste treatment facilities at SRS and must be removed or undergo treatment to oxidize the organic compounds before the tank can be returned to routine Tank Farm service. Tank 48H currently holds approximately 270,000 gallons of legacy material comprised of decontaminated salt solution, approximately 20,000 kilograms of TPB solids, 3,400 kilograms of sludge solids, and 1,800 kilograms of monosodium titanate (MST).

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

Possible impacts of molten salt reactors on the International Monitoring System

Molten salt reactors (MSRs) are gaining support as many countries look for ways to increase power generation and replace aging nuclear energy production facilities. MSRs have inherently safe designs, are scalable in size, can burn transuranic wastes from traditional solid fuel nuclear reactors, can store excess heat in thermal reservoirs for water desalination, and can be used to produce medical isotopes as part of the real-time liquid-fuel recycling process. The ability to remove 135 Xe in real time from the fuel improves the power production in an MSR because 135 Xe is the most significant neutron-absorbing isotope generated by nuclear fission. Xenon-135, and other radioactive gases, are removed by sparging the fuel with an inert gas while the liquid fuel is recirculated from the reactor inner core through the heat exchangers. Without effective abatement technologies, large amounts of radioactive gas could be released during the sparging process. This work examines the potential impact of radioxenon releases on samplers used by the International Monitoring System (IMS) to detect nuclear explosions. Here, atmospheric transport simulations from seven hypothetical MSRs on different continents were used to evaluate the holdup time needed before release of radioxenon so IMS samplers would register few detections. Abatement technologies that retain radioxenon isotopes for at least 120 d before their release will be needed to mitigate the impacts from a molten salt breeder reactor used to replace a nuclear power plant. A holdup time of about 150 d is needed to reduce emissions to the average level of current nuclear power plants.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗