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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Reducing Actinide Production Using Inert Matrix Fuels

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Fundamental Aspects of Zeolite Waste Form Production by Hot Isostatic Pressing

The direct conversion of iodine-bearing sorbents into a stable waste form is a research topic of interest to the US Department of Energy. The removal of volatile radioactive 129I from the off-gas of a nuclear fuel reprocessing facility will be necessary in order to comply with the regulatory requirements that apply to facilities sited within the United States (Jubin et al., 2012a), and any iodine-containing media or solid sorbents generated by this process would contain 129 I and would be destined for eventual geological disposal. While recovery of iodine from some sorbents is possible, a method to directly convert iodineloaded sorbents to a durable waste form with little or no additional waste materials being formed and a potentially reduced volume would be beneficial. To this end, recent studies have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by hot isostatic pressing (HIPing) (Bruffey and Jubin, 2015). Silver mordenite (AgZ), of the zeolite class of minerals, is under consideration for use in adsorbing iodine from nuclear reprocessing off-gas streams. Direct conversion of I-AgZ by HIPing may provide the following benefits: (1) a waste form of high density that is tolerant to high temperatures, (2) a waste form that is not significantly chemically hazardous, and (3) a robust conversion process that requires no pretreatment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Hot Isostatic Pressing of Engineered Forms of I-AgZ

Hot isostatic pressing (HIP) is being considered for direct conversion of 129 I-bearing materials to a radiological waste form. The removal of volatile radioactive 129 I from the off-gas of a nuclear fuel reprocessing facility will be necessary to comply with regulatory requirements regarding reprocessing facilities sited within the United States, and any iodine-containing media or solid sorbents generated by offgas abatement will require disposal. Zeolite minerals such as silver-exchanged mordenite (AgZ) have been studied as potential iodine sorbents and will contain 129 I as chemisorbed AgI. Oak Ridge National Laboratory (ORNL) has conducted several recent studies on the HIP of both iodine-loaded AgZ (I-AgZ) and other iodine-bearing zeolite minerals. The goal of these research efforts is to achieve a stable, highly leach resistant material that is reduced in volume as compared to bulk iodine-loaded I-AgZ. Through the use of HIP, it may be possible to achieve this with the addition of little or no additional materials (waste formers). Other goals for the process include that the waste form will be tolerant to high temperatures and pressures, not chemically hazardous, and that the process will result in minimal secondary waste generation. This document describes the preparation of 27 samples that are distinct from previous efforts in that they are prepared exclusively with an engineered form of AgZ that is manufactured using a binder. Iodine was incorporated solely by chemisorption. This base material is expected to be more representative of an operational system than were samples prepared previously with pure minerals.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Design and Test Plan for an Integrated Iodine Scrubber and Polishing Bed System

The capture and subsequent immobilization of four regulated volatile radionuclides ( 3 H, 14 C, 85 Kr, and 129 I) and relevant semivolatile species from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of significant research interest on the part of the US Department of Energy and other international organizations. Significant research and development has been conducted over the past decade. In 2016 an initial engineering evaluation and design of the off-gas abatement systems required for a hypothetical 1000 t/yr UNF reprocessing facility treating 5 yr–cooled, 60 GWd/tIHM UNF was completed. One of the key findings of that report was that the consumption rate of silver-based iodine sorbents in the dissolver off-gas primary iodine capture bed is very high and may warrant the evaluation of alternative methods to capture the bulk of the iodine that could significantly reduce the associated frequent remote handing of the iodine filter beds. This report is intended to describe the design of an experimental system that can be used to examine the use of aqueous scrubbing to remove the bulk of the iodine from the dissolver off-gas stream prior to a silver-based solid sorbent that would be used to provide the final iodine capture or polishing step. This report also provides a description of the initial series of tests that are proposed for this system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Testing of an Iodine and Tritium Removal System for Advanced Tritium Pretreatment Off-Gas

The reprocessing of used nuclear fuel would release volatile radionuclides into the off-gas of a processing plant, including 3 H, 14 C, 85 Kr, and 129 I. For a reprocessing plant sited within the United States and governed by US regulations, the abatement of some or all of these radionuclides from the plant off-gas would be required before discharge of that gas to the environment. A potential simplification of the off-gas management system can be achieved through an efficient advanced tritium pretreatment (ATPT) step in which the UO 2 fuel is oxidized by NO 2 prior to dissolution. Oxidation of the UO 2 fuel matrix results in the release of tritium contained within. Up-front removal of tritium from the fuel in a pretreatment step, followed by the abatement of tritium from the advanced tritium pretreatment off-gas (ATPTOG) stream, can minimize or eliminate the distribution of tritium throughout the plant, also decreasing or eliminating the need for tritium capture on multiple off-gas streams. The use of NO 2 as the oxidant allows tritium pretreatment to be performed at lower temperatures and may result in the quantitative release of iodine from the fuel. A total of five tests examined AgA as a potential iodine sorbent at an operating temperature of 150°C and both 3A MS and silica gel as potential tritium sorbents with an operating temperature of 40°C. 3A MS and silica gel are intended for use as regenerative sorbents, in which the adsorbed tritium (as tritiated water or tritiated nitric acid) is loaded onto the sorbent at operating temperature and then subsequently desorbed at high temperature into a small-volume gas stream for recovery and conversion to a waste form. This testing leaves many fundamental questions regarding the ATPTOG iodine and tritium abatement systems unanswered. These include questions regarding sorbent capacity and sorbate desorption. Overall, this testing indicates that tritium and iodine abatement from an ATPTOG stream by solid sorbents is currently at a low stage of technological development and that deployment of such an abatement system would require substantial additional research and development.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Performance of Silver-Exchanged Mordenite and Silver-Functionalized Silica-Aerogel Under Vessel Off-gas Conditions

US regulations 40 CFR 61, 40 CFR 190, and 10 CFR 20 govern release limits for volatile radionuclides contained in the gaseous effluents of used nuclear fuel (UNF) reprocessing facilities. Of the four volatile radionuclides that are restricted by the release limits ( 3 H, 14 C, 85 Kr, and 129 I) 129 I will require the greatest degree of abatement. Previous studies have shown that overall plant decontamination factor (DF) for 129 I must, at minimum, exceed 1,000 to meet regulatory requirements. Iodine-containing off-gas will be present in the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas, and the shear off-gas. The VOG will most likely contain 1%–5% of the total iodine at part per billion (ppb) concentrations. A number of studies have examined iodine abatement from the dissolver off-gas, which contains greater than 95% of the iodine inventory of the plant, but very few have examined the recovery of iodine from the VOG stream. Reduced silver-exchanged mordenite (Ag 0 Z) has been considered a promising iodine sorbent for use in UNF reprocessing plants, but recent efforts have identified silver-functionalized silica-aerogel (AgAerogel) as a potential alternative. The testing reported here reflects a substantial increase in the body of work available on iodine adsorption by both silver sorbents under prototypical VOG conditions. Both I2 and methyl iodide (CH 3 I) adsorption tests have been conducted with Ag 0 Z and AgAerogel at a range of concentrations up to 1,000 ppb. These experiments were performed with two specific aims: (1) to assess the effect of iodine concentration on its adsorption by the sorbent and (2) to compare the performance of Ag 0 Z and AgAerogel in the removal of iodine from a prototypical VOG stream.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Joint Test Plan for the Evaluation of Iodine Retention for Long-Chain Organic Iodides

A number of reports have identified four volatile radionuclides that arise from the reprocessing of nuclear fuel ( 3 H, 14 C, 85 Kr, and 129 I) that require controls limiting their environmental release from reprocessing facilities in order to meet US regulatory requirements. Of these, 129 I has the longest half-life and highest potential biological impact.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗