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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Radiolytic evaluation of a new technetium redox control reagent for advanced used nuclear fuel separations

Technetium is a problematic radioisotope for used nuclear fuel (UNF) and subsequent waste management owing to its high environmental mobility and coextraction in reprocessing technologies as the pertechnetate anion (TcO 4 - ). Consequently, several strategies are under development to control the transport of this radioisotope. A proposed approach is to use diaminoguanidine (DAG) for TcO 4 - and transuranic ion redox control. Although the initial DAG molecule is ultimately consumed in the redox process, its susceptibility to radiolysis is currently unknown under envisioned UNF reprocessing conditions, which is a critical knowledge gap for evaluating its overall suitability for this role. To this end, we report the impacts of steady-state gamma irradiation on the rate of DAG radiolysis in water, aqueous 2.0 M nitric acid (HNO 3 ), and in a biphasic solvent system composed of aqueous 2.0 M HNO 3 in contact with 1.5 M N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) dissolved in n-dodecane. In addition, we report chemical kinetics for the reaction of DAG with key transients arising from electron pulse radiolysis, specifically the hydrated electron (e aq - ), hydrogen atom (H ˙ ), and hydroxyl ( ˙ OH) and nitrate (NO 3 ˙ ) radicals. The DAG molecule exhibited significant reactivity with the ˙ OH and NO 3 ˙ radicals, indicating that oxidation would be the predominant degradation pathway in radiation environments. This is consistent with its role as a reducing agent. Steady-state gamma irradiations demonstrated that DAG is readily degraded within a few hundred kilogray, the rate of which was found to increase upon going from water to HNO 3 containing solutions and solvents systems. This was attributed to a thermal reaction between DAG and the predominant HNO 3 radiolysis product, nitrous acid (HNO 2 ), k(DAG + HNO 2 ) = 5480 ± 85 M -1 s -1 . Although no evidence was found for the radiolysis of DAG altering the radiation chemistry of the contacted DEHiBA/n-dodecane phase in the investigated biphasic system, the utility of DAG as a redox control reagent will likely be limited by significant competition with its degradation by HNO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Overview of the distributed image processing infrastructure to produce the Legacy Survey of Space and Time

The Vera C. Rubin Observatory is preparing to execute the most ambitious astronomical survey ever attempted, the Legacy Survey of Space and Time (LSST). Currently the final phase of construction is under way in the Chilean Andes, with the Observatory’s ten-year science mission scheduled to begin in 2025. Rubin’s 8.4-meter telescope will nightly scan the southern hemisphere collecting imagery in the wavelength range 320–1050 nm covering the entire observable sky every 4 nights using a 3.2 gigapixel camera, the largest imaging device ever built for astronomy. Automated detection and classification of celestial objects will be performed by sophisticated algorithms on high-resolution images to progressively produce an astronomical catalog eventually composed of 20 billion galaxies and 17 billion stars and their associated physical properties. In this article we present an overview of the system currently being constructed to perform data distribution as well as the annual campaigns which reprocess the entire image dataset collected since the beginning of the survey. These processing campaigns will utilize computing and storage resources provided by three Rubin data facilities (one in the US and two in Europe). Each year a Data Release will be produced and disseminated to science collaborations for use in studies comprising four main science pillars: probing dark matter and dark energy, taking inventory of solar system objects, exploring the transient optical sky and mapping the Milky Way. Also presented is the method by which we leverage some of the common tools and best practices used for management of large-scale distributed data processing projects in the high energy physics and astronomy communities. We also demonstrate how these tools and practices are utilized within the Rubin project in order to overcome the specific challenges faced by the Observatory.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Integrating the PanDA Workload Management System with the Vera C. Rubin Observatory

The Vera C. Rubin Observatory will produce an unprecedented astronomical data set for studies of the deep and dynamic universe. Its Legacy Survey of Space and Time (LSST) will image the entire southern sky every three to four days and produce tens of petabytes of raw image data and associated calibration data over the course of the experiment’s run. More than 20 terabytes of data must be stored every night, and annual campaigns to reprocess the entire dataset since the beginning of the survey will be conducted over ten years. The Production and Distributed Analysis (PanDA) system was evaluated by the Rubin Observatory Data Management team and selected to serve the Observatory’s needs due to its demonstrated scalability and flexibility over the years, for its Directed Acyclic Graph (DAG) support, its support for multi-site processing, and its highly scalable complex workflows via the intelligent Data Delivery Service (iDDS). PanDA is also being evaluated for prompt processing where data must be processed within 60 seconds after image capture. This paper will briefly describe the Rubin Data Management system and its Data Facilities (DFs). Finally, it will describe in depth the work performed in order to integrate the PanDA system with the Rubin Observatory to be able to run the Rubin Science Pipelines using PanDA.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2021

The Hazards Posed by Mesoscale Lightning Megaflashes

Lightning megaflashes extending over >100-km distances have been observed by the Geostationary Lightning Mappers (GLMs) on NOAA’s R-series Geostationary Operational Environmental Satellites (GOES). The hazards posed by megaflashes are unclear, however, because of limitations in the GLM data. We address these by reprocessing GOES-16 GLM measurements from 1 January 2018 to 15 January 2020 and integrating them with Earth Networks Global Lightning Network (ENGLN) observations. ENGLN verified 194 880 GLM megaflashes as natural lightning. Of these, 127 479 flashes occurred following the October 2018 GLM software update that standardized GLM timing. Reprocessed GLM/ENGLN lightning maps from these postupdate cases provide a comprehensive view of how individual megaflashes evolve. This megaflash dataset is used to generate statistics that describe their hazards. The average megaflash produces 5–7 cloud-to-ground (CG) strokes that are spread across 40%–50% of the flash extent. As flash extent increases beyond 100 km, megaflashes become concentrated in key hot-spot regions in North and South America while the number of CG and intracloud events per flash and the overall peak current increase. CGs in the larger megaflashes occur over 80% of the flash extent measured by GLM, and the majority contain regions where the megaflash is the only lightning activity in the preceding hour. These statistics demonstrate that there is no safe location below an electrified cloud that is producing megaflashes, and current lightning safety guidance is not always sufficient to mitigate megaflash hazards.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Glass structure and crystallization in boro-alumino-silicate glasses containing rare earth and transition metal cations: a US-UK collaborative program

Nuclear wastes generated from reprocessing of used nuclear fuel tend to contain a large fraction of rare earth (RE, e.g., Nd3+), transition (TM, e.g., Mo6+, Zr4+), alkali (A, e.g., Cs+), and alkaline earth cations (AE, e.g., Ba2+, Sr2+). Various strategies have been considered for immobilizing such waste streams, varying from nominally crystal-free glass to glass-ceramic to multi-phase ceramic waste forms. For glass and glass-ceramic waste forms, the added glass-forming system is generally alkali-alkaline earth-aluminoborosilicate (i.e., Na-Ca-Al-B-Si oxide). In a US-UK collaborative project, summarized here, we investigated the glass structure and crystallization dependence on compositional changes in simulated nuclear waste glasses and glass-ceramics. Compositions ranged in complexity from five – to – eight oxides. Specifically, the roles of Mo and rare earths are investigated, since a proposed glass-ceramic waste form contains crystalline phases such as powellite [(AE,A,RE)MoO4] and oxyapatite [(RE,AE,A)10Si6O26], and the precipitation of molybdenum phases is known to be affected by the rare earth concentration in the glass. Additionally, the effects of other chemical additions have been systematically investigated, including Zr, Ru, P, and Ti. A series of studies were also undertaken to ascertain the effect of the RE size on glass structure and on partitioning to crystal phases, investigating similarities and differences in glasses containing single RE oxides of Sc, Y, La, Ce, Nd, Sm, Er, Yb, or Lu. Finally, the effect of charge compensation was investigated by considering not only the commonly assessed peralkaline glass but also metaluminous and peraluminous compositions. Glass structure and crystallization studies were conducted by spectroscopic methods (i.e., Raman, X-ray absorption, nuclear magnetic resonance (NMR), optical absorption, photoluminescence, photoluminescence excitation, X-ray photoelectron spectroscopy), microscopy (i.e., scanning electron microscopy, transmission electron microscopy, electron probe microanalysis), scattering (i.e., X-ray and neutron diffraction, small angle measurements), and physical characterization (i.e., differential thermal analysis, liquidus, viscosity, density). This paper will give an overview of the research program and some example unpublished results on glass-ceramic crystallization kinetics, microstructure, and Raman spectra, as well as some examples of the effects of rare earths on the absorption, luminescence, and NMR spectra of starting glasses. The formal collaboration described here has resulted in the generation of a large number of results, some of which are still in the process of being published as separate studies.

McCloy, John S.↗
DOE OSTI2018

Testing of an Integrated Iodine Scrubber and Polishing Bed System

In 2016 an initial engineering evaluation and design of the off-gas abatement systems required for a hypothetical 1000 t/yr used nuclear fuel (UNF) reprocessing facility treating 5 yr–cooled, 60 GWd/tIHM UNF was completed. In this study two types of silver-based iodine sorbents were considered. The first was a silver-exchanged mordenite (AgZ), and the second was a silver-functionalized silica aerogel. In the initial design of the dissolver off-gas (DOG) iodine capture beds, the sorbent capacity was assumed to be 40 mg I/g sorbent. The total iodine released from the fuel, used as input for the design of the DOG system, was 1.635 kg/day. Tramp halogens increased this to ~5.1 kg/day on an iodine equivalent basis. The daily consumption of iodine sorbent was ~166 kg (including the mass transfer zones that have lower iodine loading). The iodine capture system was designed to utilize five beds in parallel. In this design each bed was ~1 m in diameter and 1.5 m long. This design allowed the beds to remain on line for 20 days, but at that point all five beds would need to be replaced, resulting in considerable material handling. It was proposed that the bed life could be extended and the number of parallel beds reduced if an aqueous iodine scrubber preceded the solid sorbent iodine capture beds. The testing described in this report was intended to provide an evaluation of an integrated system in which bulk iodine would first be removed by an aqueous NaOH scrubber with a target iodine decontamination factor (DF) of 10 to 50. This would reduce the sorbent usage by 90 to 95% and significantly reduce the associated frequent remote handling of the iodine filter beds. The effluent from the scrubber would then be passed through an AgZ sorbent bed to remove residual iodine. The integration of these unit operations was intended to demonstrate an iodine DF of >1000, which would facilitate regulatory compliance with regards to total volatile radionuclide emission from a reprocessing facility. Initial baseline performance testing of the scrubber system includes the characterization of CO 2 removal by the aqueous NaOH scrubber as a function of temperature, caustic concentration, packing length, liquid-to-gas mass ratio, and liquid flow rate. It was found that when the column was well wetted, very high CO 2 DFs were achieved (>650 and potentially as high as 15,000), preventing the ability to elucidate the effects of each specific variable tested. The L/G ratio in this system was not representative of typical L/G ratios found in industrial-scale scrubber systems. By significantly lowering the liquid recirculation rate to the point that column wetting was marginal, some significant variations in the CO 2 DF or CO 2 effluent concentrations could be observed. The additional tests conducted under these more adverse conditions still showed CO 2 DFs >100. The only operational parameter with significant effects was determined to be column length, with increased column length resulting in higher CO 2 DFs. Following these CO 2 tests, a single iodine scrubber test was performed deliberately at a condition indicated by the CO 2 demonstration to result in poor column performance. This was intended to establish a lower performance bound for iodine removal by the aqueous scrubber. The result showed that even under the adverse operating conditions, an iodine DF of >100 could be achieved. Finally, the aqueous scrubber and a AgZ polishing bed were integrated. This simple caustic scrubber system achieved an iodine DF of >12,000 for just the scrubber portion. Using this iodine DF, the iodine loading on the polishing bed was reduced by 99.993%. The performance of that integrated system was compared to a system including only a AgZ sorbent bed. The system without the scrubber also showed iodine DFs >3000 until breakthrough was observed. The loading on the saturated portion of the bed averaged 87.2 mg I 2 /gm AgZ. Comparing the penetration of iodine into the sorbent beds for these two systems showed a marked difference; the penetration with the scrubber was only 0.65 cm at a maximum loading of 0.11 mg I 2 /gm AgZ, whereas for the case without the caustic scrubber, the bed was saturated for the first 7 cm and breakthrough of the 13.3 cm bed occurred.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Mu*STAR ADSR Fuel Conversion Facility Evaluation and Cost Analysis

The purpose of the reported effort is the development of a conceptual process for conversion of used nuclear fuel (UNF) from light water reactors (LWRs) to a fluoride salt form to facilitate its use for power generation in a Mu*STAR Accelerator-driven Subcritical Reactor (ADSR). This document evaluates potential processes for converting UNF into a fluoride salt, evaluates the costs/benefits of using enriched lithium in the carrier salt, and analyzes comparable costs of similar facilities to make a rough order of magnitude estimate for the cost to design, construct, and commission a Mu*STAR ADSR fuel conversion facility. This document also makes a high-level comparison of the Mu*STAR ADSR fuel cycle with other important fuel cycles. The results of the cost analysis show that the Mu*STAR ADSR fuel conversion facility will cost more than $100M, possibly up to around $1B, to design and construct. The cost will depend on the size of the hot cell required, which will depend on the processing options chosen for converting the UNF to a molten salt. Operations and maintenance of the facility will be on the order of $10M per year. The design, fabrication, and installation of the remote process equipment to support the fuel conversion will be an additional cost of more than $50M. Cost surveys of Department of Energy hot cell designs and previous Department of Energy hot cell construction suggest that the cost variance may be large, and the risk of escalating costs may be high in the regulation-mandated atmosphere of nuclear facility design and construction. The molten salt for the Mu*STAR ADSR contains all of the nonvolatile isotopes in the UNF, and there is never a separation of the plutonium from the fission products. This makes this conversion process much more proliferation resistant than the reprocessing options used for other closed fuel cycles. When normalized to the amount of electricity produced, the Mu*STAR ADSR fuel cycle’s expected cost might be less than any reprocessing fuel cycle and possibly less than the once-through cycle now used. Simulations show that once the conversion to fluoride is made, that up to seven times as much energy can be extracted from the UNF than was produced by the LWR while making it [Bowman, 2010]. Using once-only, on-site conversion to fluoride means no mining, no enrichment, no manufacturing of fuel rods, and no transportation costs. Typically, the UNF was produced over a 40-year period in an LWR; it can potentially be used for 280 years in Mu*STAR, producing the same power as the LWR did, ultimately requiring disposal of approximately the same volume of material as the LWR generated. Mu*STAR burns most of the higher actinides, which could reduce both the cost of disposal and significantly reduce the longevity of its radiotoxicity. Mu*STAR ADSR’s ability to demonstrate unique and valuable peripheral benefits are additional arguments for this technology, including (1) turning UNF into a valuable commodity (thus, its disposal will not be such an immediate problem, so continued operation of existing LWRs will be more acceptable); (2) burning fissile or fertile fuel including excess weapons-grade plutonium in subcritical mode; and (3) providing high-temperature process heat to convert natural gas to synthetic diesel fuel, water to hydrogen, or for other process-heat applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Monitoring of Actinide Concentrations in Molten LiCl-KCl Salt using Alpha Spectroscopy: Final Progress Report

This project addresses the need for real time monitoring of actinide concentrations for spent fuel reprocessing utilizing alpha spectroscopy. This project specifically focuses on spent fuel reprocessing with the pyroprocessing technique that uses high temperature molten salt. Pyroprocessing uses molten salt such as lithium chloride and potassium chloride to separate out uranium and plutonium from spent fuel. To prevent any possible diversion of high purity plutonium in this process, monitoring of the actinide concentrations within the refinery are essential for safeguards and non-proliferation efforts. Alpha spectroscopy enables the identification and quantification of actinide isotopic concentrations. However, current assay methods involving intrusive sampling and ex-suit measurement require weeks to provide accurate information. A near real-time and in-situ measurement system is required to quickly quantify actinide isotopic concentrations within the salt. Semiconductor radiation detectors provide a compact, high energy resolution option for completing these measurement needs. Silicon carbide (SiC) has been shown to operate effectively at high temperatures and high chemical stability in the molten salt environment. 4H-SiC Schottky diode alpha particle detectors were fabricated and tested to determine their alpha spectroscopy capabilities. The detectors need to operate at the elevated temperatures, survive the molten salt environment, be radiation hard, and have sufficient energy resolution under all these conditions to determine isotopic concentrations. Along with the detector, a method of producing samples of the actinides in the salt is required. To produce a thin film of the actinides on the detector, which is a necessity for alpha spectroscopy, electrodeposition can be used. Electrodeposition uses an electric potential difference to selectively deposit uranium or other actinides from the salt using an anode and cathode. Deposition of the material directly onto the sensor would produce a thin film of material that would enable materials accountability measurements to be taken in real time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Quantify the Extent of Physisorption on Silver Based Sorbents under VOG Conditions

US regulations govern release limits for volatile radionuclides contained in the gaseous effluents of used nuclear fuel reprocessing facilities. Of the four volatile radionuclides that are most restricted by the release limits ( 3 H, 14 C, 85 Kr, and 129 I), 129 I will require the greatest degree of abatement. Within a reprocessing facility, the vessel off-gas stream will likely contain 1%–5% of the total iodine at parts per billion (ppb) concentrations. Studies conducted by the authors have now investigated both I 2 and methyl iodide (CH 3 I) adsorption by two silver-bearing sorbents, hydrogen-reduced silver-exchanged mordenite (Ag 0 Z) and silver functionalized aerogel (AgAerogel) at a range of concentrations up to 1,000 ppb. Throughout this suite of tests, and particularly in the case of CH3I adsorption tests, there was a significant discrepancy in the amount of iodine recovered on sorbent beds compared to the amount of iodine that was believed to be delivered to the test system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Mid-Atlantic U.S. Offshore Carbon Storage Resource Assessment Project (Final Technical Report)

The greatest potential for carbon storage in the northeastern United States lies in the offshore geologic formations comprising the outer continental shelf (Monteverde et al., 2011). Offshore storage can be linked to large point-sources of carbon dioxide (CO 2 ) while avoiding many of the logistical difficulties and potential risks encountered when siting onshore projects, especially in densely populated areas of the East Coast. Offshore CO 2 storage assessment and research in the United States is underway to address uncertainty in potential storage resources, particularly in the mid- and north-Atlantic offshore area. Given the current knowledge base and access to publicly available data, the objectives of the Mid-Atlantic U.S. Offshore Carbon Storage Resource Assessment Project (MAOCSRAP) were fourfold: (1) complete a systematic carbon storage resource assessment of the mid-Atlantic offshore coastal region from the Georges Bank Basin (GBB) through the Long Island Platform (LIP) to the southern Baltimore Canyon Trough (BCT); (2) define key input parameters to reduce uncertainty for offshore storage resource and efficiency estimates; (3) perform a preliminary assessment of risk factors, uncertainties, and data gaps; and (4) engage industry and regulatory stakeholders through development of a Road Map to assist future project planning and implementation. The MAOCSRAP project has compiled, inventoried, and assimilated various publicly available data sets to provide a strong technical basis on which future carbon storage studies and applications can be built. The knowledge infrastructure necessary to support the development of full-scale offshore carbon storage include: Defining the geologic characteristics of candidate storage sites (Chapter 2). Using existing seismic data to better define the continuity of the storage zones and seals (Chapter 3). Cataloguing the hydrogeologic properties of mid-Atlantic offshore storage sites (Chapter 4). Calculating prospective CO 2 storage resources using net effective pore volumes and fluid displacement properties specific to offshore lithologies (Chapter 5). Examining risk factors related to offshore storage (Chapter 6). Communicating with industry and other stakeholders about the future prospects for offshore storage (Chapter 7). Ensuring technology transfer to industry and other stakeholders (Chapter 8). Led by Battelle, this project was conducted by public and private entities with expertise in offshore geology and resources for the study region, including state geological surveys of Delaware, Maryland, and Pennsylvania; United States Geological Survey-Woods Hole Coastal and Marine Science Center and Haifa University; Rutgers University; Harvard University; and Lamont-Doherty Earth Observatory (LDEO) at Columbia University. The storage resource assessment was completed for a broad region offshore of the U.S. East Coast, from Massachusetts to Virginia. The results include high-level storage resource estimates for areas not previously characterized and improved storage resource estimates for geographically expansive portions of offshore geologic units. Results of the project provide a foundation for developing CO 2 storage along offshore areas of the mid-Atlantic United States. A combination of information was used to provide confidence in project results. Key findings of the major project tasks are summarized as follows. Subsurface Data Analysis: Legacy seismic, well-log, core, and biostratigraphic data were digitized, reprocessed, and analyzed using modern techniques, augmenting previous characterization efforts. The sequence stratigraphic framework for the Cretaceous and Jurassic strata was developed by correlating well-log depositional sequences to seismic reflectors identified in the BCT, LIP, and GBB. The integration of well-log stratigraphy with seismic stratigraphy and petrophysical analysis helped to identify and map potential storage zones within sequences of the Logan Canyon (LC), Missisauga, and Mohawk formations, as well as the regionally extensive caprock comprised of the Dawson Canyon shale. Approximately 4,000 kilometers (km) of seismic data was reprocessed to create a structural framework that was used to tie together all available well data, constrain sequence stratigraphic interpretations, and make regional assessments of offshore carbon storage resources. Prospective storage resource estimates were 150 to 1136 megatons (Mt) for the combined storage zones. This suggests mid-Atlantic U.S. offshore formations can store decades of CO 2 from industrial sources in the region. Preliminary local reservoir simulations suggest that injection rates of 1 Mt CO 2 /year may be sustained for 30 years in single injection wells in the BCT. Advanced geologic modeling and new data acquisition are needed to address data gaps and advance carbon capture and storage (CCS) in key offshore areas selected for further investigation. Offshore Risk Factors: Offshore geologic risk factors include soft sediment deformation, unit continuity, sedimentological and structural features, seismicity and hydrates. CO 2 storage risks include inadequate seals, migration/leakage, and chemical interactions leading to decreased storage. Sensitive habitats, environmental impacts, disturbance to seafloor, and other risks need to be identified in advance of project activities and integrated into detailed mitigation plans for all project phases. Stakeholder Engagement: Input and participation from government, industry, and environmental groups provided input into the Road Map and address next steps needed for project deployment. Early engagement and ongoing communication, as well as policy framework development, is key to large-scale deployment of CCS. Next Steps: This project represents an important first step by completing a high-level CO 2 storage resource assessment and building the knowledge infrastructure necessary to improve quantitative storage resource estimates. The data sets that have been curated under this project provide an opportunity to conduct research and development (R&D) needed to address data gaps and reduce risk and uncertainty. Offshore characterization and validation strategies that are systematically designed to provide data and infrastructure that can be upscaled to meet commercial requirements could be developed off of this foundation.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2020

Safeguarding Reactors and Spent Nuclear Fuel [Slides]

Highly enriched fuel contains large quantities of ²³⁵U; Burned fuel may contain large quantities of ²³⁹Pu; Reprocessing is the major threat (discussed at length later), but we need to ensure the amount of plutonium going to the reprocessing plant is what we believe it to be; All nuclear material needs to be accounted for (material accountancy); There may be a loss of continuity of knowledge in which case fuel needs to be measured to determine how much plutonium and uranium it contains.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Progress and Technology Assessment of Melt-Cast Ceramic Waste Forms

The Savannah River National Laboratory (SRNL) conducted research from FY2011 – FY2020 to develop ceramic waste forms to treat High-Level Waste (HLW) resulting from reprocessing of commercial Used Nuclear Fuel (UNF) as part of the Department of Energy (DOE) office of Nuclear Energy (NE) Fuel Cycle Technologies (FCT) research program. The objective of the research was to develop a reference ceramic waste form and process technology needed to immobilize the combined HLW raffinate stream, including undissolved solids and soluble technetium, and potentially the TRU waste stream resulting from reprocessing processes. This report compiles the relevant literature produced under, and pertinent to, the program objectives and provides a perspective on the technology development as well as prospects for future.

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

Screening IMS Seismic Detections using Dynamic Correlation Processor: Progress report for 2020

LLNL completed the FY20 workplan entailing application of the Dynamic Correlation Processor (DCP) to International Monitoring System (IMS) primary seismic data in 2019 and 2020. The effective analysis period was curtailed to January 1 through November 12, 2019 due to a data compression change that rendered later data unusable. Fixing the data compression error is a priority for our data source, and we plan to extend the period of data analysis when the fix is implemented. We found that the percentage overlap between detections generated by the International Data Centre (IDC) DFX software and detections generated by DCP varied widely by station. Overlap is 30-40% at some stations, suggesting that a large percentage of DFX detections could be screened (removed) from the detection-event association algorithm. However, detection overlap was much lower, a few percent, at many stations and the percentage overlap was low at stations known to have many mines and other local sources of seismicity. Unexpectedly low detection overlap at many stations prompted us to examine results for the ARCES station in detail. There is considerable mine activity near ARCES, but our initial FY20 analysis found only ~2% DCP/DFX detection overlap. Differences in the pre-processing (e.g. beam recipes) used by DCP and DFX appear to be the cause of low detection overlap at ARCES. DCP uses wideband filters to improve detector robustness and the IDC uses narrow-band pre-filtering to improve sensitivity. After reprocessing ARCES using the IDC beam recipe, DCP/DFX overlap increased to ~74%. This result shows that we must reprocess the IMS network using the IDC station-specific beam recipes if we are to effectively screen DFX detections and improve automatic event building performance at the IDC.

58 GEOSCIENCES↗
DOE OSTI2020

Mechanistic Understanding of Silver Sorbent Aging Processes in Off-Gas Treatment

The effect of aging in various off-gas streams including dry air, humid air, 1% NO/N 2 , 2% NO 2 /air, and a gas mixture consisting of 99.25% humid air (dew point: -15 °C), 0.25% NO/N 2 , and 0.5% NO 2 /air in terms of concentration on reduced silver exchanged mordenite (Ag 0 Z) and silver functionalized silica aerogel (Ag 0 -aerogel) were studied in this project. Utilizing simple and sophisticated experimental information combined with molecular and macroscopic mechanistic modeling, this project has significantly advanced the current understanding of aging for silver bearing adsorbents in nuclear-fuel-reprocessing off-gas streams. In addition, this project has produced models that can be utilized to predict the influence of adsorbent aging on the uptake kinetics and capacity of aged adsorbents for iodine and iodine-containing organic compounds. The results of the study have been published in the open literature, and can be used to help DOE better predict the removal of radioactive gases from nuclear-fuel-reprocessing off-gas streams.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Domestic Safeguards Material Control and Accountancy Considerations for Molten Salt Reactors

Molten salt reactors (MSRs) are a class of nuclear reactor designs with features and operational characteristics that vary significantly more than the class of light water reactors (LWRs). MSR design concepts can be broadly categorized as solid-fueled reactors with molten salt as the coolant, liquid-fueled reactors with fuel dissolved in molten salt coolant and liquid-fueled reactors with fuel dissolved in molten salt that is contained in distinct fuel tubes with a nonfissile coolant. Proposed MSR concepts include features that are specific to each design. MSR concepts vary widely across these design features, including the physical, chemical, and isotopic composition of fresh and irradiated fuel. Operational neutron energy spectrums and breeding ratios also vary significantly across design concepts. Some concepts are burner reactors designed to transmute the spent nuclear fuel from LWRs or pressurized heavy water reactors (PHWRs), while others are breeder reactors designed to breed fissile 233 U from naturally occurring fertile 232 Th. Some MSR concepts are designed to be a part of a once-through fuel cycle, while others involve chemical separation. Those that include plans to recycle the fuel differ by whether the chemical processing would be done onsite, as a process connected to the fuel salt itself, or offsite at a reprocessing facility similar to how LWR or PHWR spent fuel is reprocessed in some countries. Each overall design concept contains various aspects of each of these features to produce a unique facility

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Initial Assessment of TeF 6 Adsorption

Research and development that supports the management of off-gases from nuclear fuel reprocessing has historically been focused on the off-gas streams that arise from aqueous reprocessing technology. However, as Gen-IV reactor development pathways move toward deployment, alternative spent nuclear fuel (SNF) processing and disposition pathways have been considered more actively. This work is focused upon aspects of fluoride volatility (FV) processing. The versatility of this method for deployment against multiple types of spent fuel encourages the continued advancement of both the primary separations processes and the secondary processes, including waste treatment, material control and accountability, and engineering designs. Recent work noted that TeF 6 , the most highly volatile fluorinated compound produced during FV processing, did not have a clear abatement technology recommended in the literature. Thus, this work performed scoping tests on activated alumina and copper shot to assess whether they could be used to remove TeF 6 from gas streams that bear F 2 . Previous work on this topic was not well described in the literature and was not performed with excess F 2 in the stream as would be typical of spent fuel processing via FV. To support an understanding of the concentration of TeF 6 contacting the adsorbent beds, a series of preliminary testing identified the TeF 6 production rate and the equilibrium concentration of TeF 6 in the gas stream contacting the adsorbent. Excess F 2 was determined to not affect the ability of activated alumina to remove TeF 6 from the gas stream quickly and completely. A determination of whether excess F 2 affected the distribution depth of Te in the sorbent bed is still pending analysis of the used sorbent. Literature suggests that when activated alumina is near saturation, TeF 6 could migrate from the sorbent bed. Future testing should investigate this possibility. Copper metal did not adsorb TeF 6 in the presence of F 2 across the sorbent temperature range of 50 to 335°C. F 2 was fully removed by the copper bed. Although preliminary thermodynamics would indicate adsorption to be energetically favorable, other factors that impact adsorption (e.g., slow kinetics, excess fluorine on the copper surface, an unfavorable transition state) are likely preventing the adsorption of TeF 6 at an easily measurable rate. The work also investigated TeF 6 production from multiple forms of Te in the temperature range of 100 to 250°C. No previous study had assessed the initial reaction rates for this process. A carefully designed study allowed determination of the activation energy for the production of TeF 6 from Te metal. The substantial amount of data collected during this study merits analysis beyond what is described here. A more in-depth kinetic analysis will be pursued. Additional analytical results will provide the ability to benchmark adsorption coefficients for TeF 6 , understand the distribution of TeF 6 within the alumina bed, and better understand the effect of F 2 partial pressure on Te fluorination. The data from this report, as supplemented by these additional analyses, will be submitted to a peer-reviewed journal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Iodine Retention of Long-chain Organic Iodides on Silver-based Sorbents under DOG and VOG Conditions

The aqueous reprocessing of used nuclear fuel releases four key volatile radionuclides ( 3 H, 14 C, 85 Kr, and 129 I) from the used fuel into the off-gas streams of a reprocessing facility. Some semi-volatile radionuclides such as 106 Ru may also be released into the gas streams. Compliance with US regulations may require removal of these radionuclides from the off-gas streams before their discharge to the environment (Jubin et al. 2017). Iodine is released during multiple unit operations within a plant and can be found in gas streams such as the shear off-gas, dissolver off-gas, vessel off-gas (VOG), and waste solidification off-gas. To achieve DFs >1,000, treatment of the plant off-gas streams must include, at minimum, removal of both the dissolver off-gas and VOG. A 2015 analysis showing that the speciation of iodine in the VOG is primarily composed of organic alkyl iodides and that abatement technology for organic iodides is not well developed has produced increased interest in removing iodine from VOG. The lack of knowledge surrounding iodine removal from the VOG prompted the Office of Nuclear Energy within the US Department of Energy to initiate experimental efforts targeted at understanding organic iodide removal from prototypic VOG streams. In 2018 a joint test plan was developed and issued by Oak Ridge National Laboratory and Idaho National Laboratory. This test plan is intended to guide a multiyear experimental effort that will provide substantial amounts of information about the removal of organic iodides from VOG streams. This document reports on the progress of Idaho National Laboratory and Oak Ridge National Laboratory in the execution of the joint test plan as of the end of fiscal year 2019. Silver-based sorbents are the primary removal technology being characterized, and this report includes data on iodine adsorption by reduced silver-exchanged mordenite (also called silver-exchanged zeolite [AgZ]) (as a baseline iodine sorbent) and silver-functionalized silica aerogel (AgAero or Ag Aerogel) (as an advanced iodine sorbent). Conclusions are preliminary in nature and reflect the ongoing execution of the test plan through close collaboration between the two laboratories.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

An Enhanced Safeguards Approach for Accelerator Driven Systems utilized to Close the Fuel Cycle

Further to our previous safeguards approach for Accelerator Driven Systems, which focused on estimates of fissile material production using relevant proton accelerator systems and corresponding safeguards needs for fuel storage, the subcritical reactor, and spent fuel storage material balances areas, this report is more expansive and considers utilization of ADS for either burning of transuranics or breeding of fissile materials. We find that the recycled fuels likely intended for ADS will be thermally and radioactively hot to such a degree that it is likely reprocessing and fuel fabrication will have to be co - located with the ADS reactor facility to avoid impractical hot fuel transportation issues. As such, we consider in detail the full ADS system to include material balance areas for spent fuel receiving, reprocessing, storage & cooling, fuel fabrication, subcritical reactor area, and waste storage & handling. Furthermore, aqueous - based separation methods like PUREX cannot tolerate the intense heat of the ADS fuels, so pyroprocessing will likely be required. With these considerations, we developed an Enhanced Safeguards Approach for ADS beyond the work done in our first report, and conclude that significant diagnostic development is needed , a nd provide safeguards recommendations. We have also included an appendix regarding some country programs, in particular the Chinese ADANES burner/breeder program a nd the Indian thorium - based breeder program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗