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Comparison of Gas Phase Fragmentation Behaviors of Nuclear Fuel Cycle Ligands in Lanthanide and Americium Metal Ligand Nitrate Clusters

Introduction (120 words): Transport of metal ions across the aqueous-organic phase boundary is an essential step in a hydrometallurgical nuclear fuel reprocessing strategy. The study of transport agents for nuclear fuel elements is imperative to guide the design of ligands that boost the separation efficiency of the recovery process from fission products. However, limited studies have been made on the chemistry of these transport agents when complexing with transuranic elements in gas-phase where all surrounding factors are essentially excluded. This work investigates the reagent ligand complexations to transuranic and other metals and their dissociations in the gas phase. Comparisons are made between 4f and 5f elements and between ligands. Methods (120 words): (N,N-diisobutylcarbamoylmethyl)phenyloctylphosphine oxide (CMPO) and N,N,N',N'-tetraoctyldiglycolamide (TODGA) have been selected to complex with metal nitrates. The actinide americium and lanthanides neodymium, samarium, and europium were investigated as part of this work. The lanthanides were selected to act as size and electron configuration analogues of the minor actinides. Metal complexes with two ligands and two nitrates ([M(NO3)2(CMPO)2]+, for example) are studied in Bruker micrOTOF-Q II mass spectrometer equipped with collision-induced dissociation capability. The comparisons of the mass spectra are made in groups of homogenous ligands and mixed TODGA-CMPO ligands clusters. Comparisons are also made based on the complexed metals (Am and lanthanides). Preliminary data (300 words): Collision-induced dissociation mass spectrometry data are collected on two ligands complexed with metal nitrates where the two ligands are homogenous, with (CMPO)2 or (TODGA)2, or heterogeneous, with (TODGA)(CMPO). Several fragmentation patterns are observed among complexes with the CMPO ligand whereas the TODGA ligand commonly dissociates intact from the complex. Most of the metal complexes exhibit similar fragmentation patterns, but there are a few notable deviations in fragmentation patterns between the Am and Ln-bearing complexes. For the [M(CMPO)2(NO3)2]+ complexes, the initial loss of nitrate in the form of nitric acid is observed in all four complexes. However, [Am(CMPO)2(NO3)2]+ exhibits an additional fragmentation not found in the lanthanide complexes. Also, a significantly different ratio of the second nitric acid loss is found in the Am complex. These deviations may indicate the different interaction behaviors between actinides and lanthanides. The [M(TODGA)2(NO3)2]+ complexes exhibit the fragmentation as the loss of one TODGA ligand as an intact form and the loss of nitrate as nitric acid. The Am complex exhibits an additional fragmentation after losing the TOGDA ligand, which is not observed among the Ln complexes. The heterogeneous [M(TODGA)(CMPO)(NO3)2]+ complexes exhibit both similarities and differences between the Am and Ln complexes. For example, the heterogenous Am complex does not exhibit the loss of an intact TODGA ligand while all three Ln complexes do. This indicates that TODGA may bind more strongly to Am than Ln. Additionally, the intensity of the loss of CMPO ligand (as partially or whole) is found to be significantly larger than that of the loss of TODGA (as partially of whole) indicating that TODGA is bound to the metal significantly stronger than CMPO. Planned computational analysis will help understand the deviation in fragmentation behaviors between americium and lanthanide metal centers, or between TODGA and CMPO ligands. Novel aspect (20 words): Gas-phase actinide and lanthanide complex formation and fragmentation provide insight into the coordination environment differences of f-element metals.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Comparison of Gas Phase Fragmentation Behaviors of Nuclear Fuel Cycle Ligands in Lanthanide and Americium Metal Ligand Nitrate Clusters

Title (20 word): Comparison of Gas Phase Fragmentation Behaviors of Nuclear Fuel Cycle Ligands in Lanthanide and Americium Metal Ligand Nitrate Clusters Introduction (120 words): Transport of metal ions across the aqueous-organic phase boundary is an essential step in a hydrometallurgical nuclear fuel reprocessing strategy. The study of transport agents for nuclear fuel elements is imperative to guide the design of ligands that boost the separation efficiency of the recovery process from fission products. However, limited studies have been made on the chemistry of these transport agents when complexing with transuranic elements in gas-phase where all surrounding factors are essentially excluded. This work investigates the reagent ligand complexations to transuranic and other metals and their dissociations in the gas phase. Comparisons are made between 4f and 5f elements and between ligands. Methods (120 words): (N,N-diisobutylcarbamoylmethyl)phenyloctylphosphine oxide (CMPO) and N,N,N',N'-tetraoctyldiglycolamide (TODGA) have been selected to complex with metal nitrates. The actinide americium and lanthanides neodymium, samarium, and europium were investigated as part of this work. The lanthanides were selected to act as size and electron configuration analogues of the minor actinides. Metal complexes with two ligands and two nitrates ([M(NO3)2(CMPO)2]+, for example) are studied in Bruker micrOTOF-Q II mass spectrometer equipped with collision-induced dissociation capability. The comparisons of the mass spectra are made in groups of homogenous ligands and mixed TODGA-CMPO ligands clusters. Comparisons are also made based on the complexed metals (Am and lanthanides). Preliminary data (300 words): Collision-induced dissociation mass spectrometry data are collected on two ligands complexed with metal nitrates where the two ligands are homogenous, with (CMPO)2 or (TODGA)2, or heterogeneous, with (TODGA)(CMPO). Several fragmentation patterns are observed among complexes with the CMPO ligand whereas the TODGA ligand commonly dissociates intact from the complex. Most of the metal complexes exhibit similar fragmentation patterns, but there are a few notable deviations in fragmentation patterns between the Am and Ln-bearing complexes. For the [M(CMPO)2(NO3)2]+ complexes, the initial loss of nitrate in the form of nitric acid is observed in all four complexes. However, [Am(CMPO)2(NO3)2]+ exhibits an additional fragmentation not found in the lanthanide complexes. Also, a significantly different ratio of the second nitric acid loss is found in the Am complex. These deviations may indicate the different interaction behaviors between actinides and lanthanides. The [M(TODGA)2(NO3)2]+ complexes exhibit the fragmentation as the loss of one TODGA ligand as an intact form and the loss of nitrate as nitric acid. The Am complex exhibits an additional fragmentation after losing the TOGDA ligand, which is not observed among the Ln complexes. The heterogeneous [M(TODGA)(CMPO)(NO3)2]+ complexes exhibit both similarities and differences between the Am and Ln complexes. For example, the heterogenous Am complex does not exhibit the loss of an intact TODGA ligand while all three Ln complexes do. This indicates that TODGA may bind more strongly to Am than Ln. Additionally, the intensity of the loss of CMPO ligand (as partially or whole) is found to be significantly larger than that of the loss of TODGA (as partially of whole) indicating that TODGA is bound to the metal significantly stronger than CMPO. Planned computational analysis will help understand the deviation in fragmentation behaviors between americium and lanthanide metal centers, or between TODGA and CMPO ligands. Novel aspect (20 words): Gas-phase actinide and lanthanide complex formation and fragmentation provide insight into the coordination environment differences of f-element metals.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

SFDS: The criticality assembly for the experimental research of the nuclear spent fuel dissolver in CIAE - 14401

SFDS is the criticality assembly for the experimental research of the spent fuel dissolver in China Institute of Atomic Energy. The spent fuel dissolver, as the bottleneck of spent fuel reprocessing, has so many characteristics such as the dynamic changes, heterogeneous distribution, complex compositions that it is hard to analysis by theoretical method. Without enough experiments, clear bias and uncertainty of the analysis tools the handling limits for criticality safety in dissolver can't be determined. The SFDS, physical designed by MONK 9A code, uses two kinds of nuclear fuels, 3% enrichment solid UO{sub 2} rods and 5.5% enrichment UO{sub 2}(NO{sub 3}){sub 2} solution, will meet the demand of the critical experiments about the dissolvers. Seven series of critical experiments will be present on it, those are on pitches of rods, uranium concentrations of solutions, nitric acid concentrations of solutions, fission products, void fractions, temperatures and dissolve phases. More kinds of experiments could be introduced with the flexible structures of SFDS in the future. The current work on it are testing and debugging with full systems by water and it will be first critical later in 2015. It will become the first benchmark assembly for critical experiments research in the nuclear fuel cycle system in China. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Application of Principal Component Analysis to Electrochemical Reprocessing PM and NMAC

In this report, data from an electrorefiner (ER) for nuclear fuel reprocessing is evaluated for process monitoring (PM) conclusions. This data comes from tests performed at the Idaho National Laboratory in 2022. Multivariate approaches utilizing methods of Principal Component Analysis (PCA) is applied. This is based off established work in process monitoring for fault detection in industrial facilities. This report will discuss the background, methods, and results of the application and some of the conclusions and applications that can be drawn from them. PCA is applied to two different electrorefiner (ER) operations that occurred at Idaho National Laboratory between August and October 2022. The first operation occurred with little incident while the second had several noted faults in the equipment in operational logs. The data from the first run was used to train the data for “normal” operations and applied to both sets of data to determine when operations were in an “off-normal” condition and identify where the fault occurs through PCA. PCA was able to identify off-normal events and identify the cause for off-normal operations. These identified off-normal events matched with the events and their causes in the operational logs. However, small amounts of variance in the data led to false detection of “off-normal” events. Thus, careful selection of training data and a-posteriori conclusions based off operator assessments will both be required for application of PCA to PM applications. This work demonstrated that multivariate approaches and latent variables are applicable to pyroprocessing PM applications and can be further expanded in future work as quality variables such as salt concentration from sensors and sampling become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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↗

Expression of Interest for off-gas system design build

Idaho National Laboratory (INL) and Pacific Northwest National Laboratory (PNNL) are studying a used nuclear fuel reprocessing system that could be used to recycle used nuclear fuel by separating waste fission products from re-useable, valuable uranium that can be used in making new nuclear fuel. This process includes dissolving the mixture of uranium and fission products in a nitric and hydrofluoric acids, then using an organic extraction process to remove and recover the uranium. The waste fission products remain in the acidic aqueous “raffinate” stream.” The fission products in this raffinate waste stream are expected to require stabilization in a vitrified glass waste form for safe disposal. INL and PNNL plan to construct and operate a facility in which to test the performance of the electric melter process that would receive a non-radioactive but chemically similar simulant of this raffinate. This test system will be located at INL in or near Idaho Falls, ID. In the melter, which operates at a temperature of about 1,050 to 1,150oC, the water, nitric acid, and other volatile and semi-volatile materials evolve into an off-gas stream, while the fission products (in this case, non-radioactive simulants of fission products) are melted with glass forming chemicals into a molten glass. Upon cooling, the molten glass provides a durable waste form for long-term disposal. Expressions of interest are requested from companies who would like to receive a request for proposal to design and build this off-gas system. We expect to request proposals and desire to award a contract prior to the end of September 2019.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Strategies for Decommissioning the Fuel Storage and Reprocessing Complex at Chalk River Laboratories - 16183

For over 60 years, Canadian Nuclear Laboratories (CNL), previously Atomic Energy of Canada Limited (AECL), has been a world leader in developing peaceful and innovative applications from nuclear technology. Resulting from these years of nuclear research there are legacy liabilities consisting of outdated and unused research facilities and buildings, a wide variety of buried and stored wastes, and affected lands. In 2006, the Government of Canada established the Nuclear Legacy Liabilities Program (NLLP) to safely and cost-effectively eliminate, to the extent practicable, the nuclear legacy liabilities. The National Research Experimental (NRX) fuel rod storage and handling facility, as well as two dedicated fuel reprocessing facilities at the Chalk River Laboratories (CRL) known as the Building 200 (B200) Complex, are examples of these legacy liabilities. It is planned to remove these facilities as part of a logical progression of activities to reduce CNL's liabilities under the NLLP and to prepare the property for the physical redevelopment of the CRL site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Comparison of Industrialized Late 20th Century Flowsheets for Reprocessing Used Nuclear Fuel

This study identified and compared three flowsheets for reprocessing used nuclear fuel (UNF) industrialized in plants in the United Kingdom (the Thermal Oxide Reprocessing Plant), France (UP2-800/UP3) and Japan (the Rokkasho Reprocessing Plant). The study also identified the major implications for a plant in the United States if it were initiated. All flowsheets employed the established Plutonium Uranium Reduction Extraction (PUREX) solvent extraction technology to separate uranium and plutonium from UNF dissolved in nitric acid. However, differences in the approaches to managing iodine-129, tritium and technetium were identified in the flowsheets. A US plant would also need to separate krypton-85 as well as iodine-129 and tritium for immobilization and disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluation of high resolution X-ray monochromatic wavelength dispersive X-ray fluorescence instrument for determining Pu and U in nuclear reprocessing streams

In this study, an alpha-prototype benchtop instrument was developed and tested for measuring Pu and U in spent fuel dissolver tanks in a nuclear fuel reprocessing plant. The high resolution X-ray (hiRX) instrument was designed with the aim of providing a new capability to measure samples with high radiation background with selectivity, accuracy, and precision using ultra-low volumes (7 μL) of liquid. The instrument was tested with specimens from the operational reprocessing facility at Savannah River National Laboratory. The results showed that bias was <6% for Pu nitrate process solution and < 18% for Pu in U/HNO 3 matrix. Instrumental precision was <1% RSD while between microcell repeatability was <6% RSD. The limit of detection for Pu in HNO 3 matrix was 3 × 10 -5 g/L (100 s, 35 W). Finally, this study demonstrates that hiRX technology has the potential to provide rapid and simultaneous quantification of Pu and U without sample preparation, making it suitable for process monitoring and materials control and accountability applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Assessments and Options for Removal and Immobilization of Volatile Radionuclides from the Processing of Used Nuclear Fuel

Nuclear fuel reprocessing is being considered in the US as part of an effort to help meet the nation’s energy needs. Various options are available to meet these needs (DOE 2010). The recent Fuel Options Study indicated that fuel cycles that utilize continuous recycling are among the “most promising” (Wigeland et al. 2014). After the valuable components have been extracted from the irradiated UO 2 during reprocessing, the remaining waste must be immobilized and sent to a repository. Some of the radionuclides in the fuel are short-lived isotopes that decay to innocuous levels soon after reactor discharge, e.g., 127 Xe (t 1/2 = 36.4 d). However, the preponderance of radionuclides requires long-term, geological storage. In this report, we assess options for capturing and immobilizing radionuclides expected to volatilize during aqueous reprocessing of used nuclear fuel. This document is an update to a previous report prepared in 2011 that was only released as a draft (FCR&D-SWF-2011-000305).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Unique Facilities, Innovative Treatment Processes, and a Talented Workforce Allow Fluor Idaho to Successfully Disposition Transuranic Waste - 18388

Fluor Idaho is utilizing state-of-the-art equipment, treatment technologies, and employee expertise at the Department of Energy's (DOE) Idaho Site to treat contact-handled transuranic (CH-TRU) waste, as well as extremely challenging remote-handled transuranic (RH-TRU) waste forms, to prepare the material for shipment and permanent disposal at the Department's Waste Isolation Pilot Plant (WIPP). Nowhere else in the DOE Complex - perhaps the world - will one find the compilation of tools and an experienced workforce with an unparalleled history of safely retrieving, characterizing, treating, repackaging, certifying, and shipping TRU waste. Since 2003, the Advanced Mixed Waste Treatment Project (AMWTP) has used both conventional and unique retrieval concepts, high-tech characterization equipment - including real-time radiography and radioassay - robotics, and automated treatment processes to retrieve and treat 65,000 cubic meters of above-ground CH-TRU waste. Additionally, engineered controls and specialized personal protective equipment have allowed employees to make thousands of entries into high-radiation areas without a skin or clothing contamination incident - evidence of the employees' commitment to safety. AMWTP has encountered and found solutions for many challenging waste forms, including polychlorinated biphenyl-contaminated wastes, TRU waste containers with up to 1,000 fissile gram equivalent plutonium-239, oversized TRU boxes, and partially absorbed organic sludge waste. By utilizing additional facilities and equipment at the Accelerated Retrieval Project V and VII, Fluor Idaho has developed new processing lines and capabilities to ensure worker safety and milestone completion. These two soft-sided buildings, located at a facility adjacent to AMWTP, were originally constructed to exhume Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) TRU wastes from a section of a 97-acre landfill. After a total of 1,150 cubic meters of targeted waste were removed, personnel, working in collaboration with state regulators, successfully implemented a unique regulatory strategy that allowed these CERCLA facilities to be modified and permitted to treat Resource Conservation and Recovery Act (RCRA) wastes from the AMWTP - a savings of tens of millions of dollars over the cost of constructing new buildings. Adding to the AMWTP's technological uniqueness is its ability to treat wastes from other facilities in the DOE Complex. Two prior decision documents and an agreement with the state of Idaho have allowed AMWTP to treat wastes from 15 other DOE sites. It is the only project in the complex to have done so. At another project located next to AMWTP, targeted waste exhumation within the Subsurface Disposal Area is approximately two years ahead of schedule. Crews have removed more than 8,600 cubic meters of CH-TRU and hazardous wastes from eight exhumation areas that equal a combined footprint of 4.75 acres. Less than one acre remains to be exhumed to satisfy the 2008 Record of Decision waste exhumation end state between the DOE, Environmental Protection Agency, and state of Idaho. Construction of the final building where targeted waste exhumation will continue was completed during the summer of 2017. Exhumation will be completed within the next few years. Following exhumation completion, an evapotranspiration barrier (or cap) will be constructed over the entire landfill. Fluor Idaho continues to make progress on a more challenging project involving an inventory of RH-TRU waste that also contains the reactive metal sodium. Crews are utilizing hot cells, once used for spent nuclear fuel reprocessing and high-level waste treatment at the Idaho Nuclear Technology and Engineering Center, to render this challenging waste form safe for disposal in a national geologic waste repository. The treatment methods being employed on this RH-TRU waste are unique in the DOE Complex. Minute concentrations of water are sprayed onto the sodium-contaminated material in a process called 'spritzing' to chemically change the sodium from a reactive metal to a non-reactive salt. In other cases, sodium-contaminated waste is dipped in water to initiate a reaction in a process called 'immersion.' Lastly, for more challenging waste shapes that could not be successfully treated with the two prior methods, crews lower the sodium-contaminated RH-TRU waste into a one-of-a-kind distillation unit that uses high temperature and vacuum to volatilize the sodium and recapture it in a condenser unit for disposal. The RH-TRU waste treated with this technology is then safe for permanent disposal. Fluor Idaho is proud to operate DOE's CH- and RH-TRU waste treatment facilities, as well as use innovative processes and its talented workforce to complete this important work-scope. Such capabilities allow the contractor to meet the commitments made by the DOE to the state of Idaho. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Safeguarding Commercial Breed and Burn Sodium-Cooled Fast Reactors

A new class of reactors is under development in the United States. These new reactor designs use fast neutron spectra and rely on fuel technology advancements to reach fleet-wide operating conditions where neither enrichment nor reprocessing fuel are required to sustain or grow the fleet. Achieving this level of operation requires nuclear fuel that can remain in the reactor environment for much longer than current limits. The transfer of irradiated fuel is also required as the initial fissile source for new and replacement reactors. This class of reactors is known as breed and burn reactors and can be deployed as a fleet whereby the transfer of fuel is used to power subsequent generations of reactors. The use of breed and burn reactors reduces, and potentially eliminates, the need for enrichment and reprocessing to sustain a reactor fleet. Currently, enrichment and reprocessing plants require extremely complicated safeguards implementation and offer States diversion pathways for indirect use material. By eliminating the need for these facilities, a nuclear power fleet based on the BREED AND BURN reactor concept has the potential to indirectly reduce the burden on IAEA safeguards. However, this complex dynamic system does pose new safeguards challenges that are explored further in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Gaseous fuel reactors for power systems

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Kendall, J. S.↗

Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

Beatty, Randy L.↗

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↗

Next Generation System Analysis Model: Recently Added Features and Future Plans

To better enable informed decision-making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management System (IWMS) program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the development and application of system analysis tools capable of analyzing various system options for the management of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).With these tools, IWMS architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool expressly designed to be capable of modeling features within various IWMS architectures. NGSAM imports data from Oak Ridge National Laboratory (ORNL)’s unified database (e.g., historic assembly information, thermal profiles for assembly heat, and at-reactor dry storage loadings) to ensure that each simulation initializes with a realistic representation of the state of commercial SNF in the U.S. Recent major enhancements implemented into NGSAM in the period since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort car acquisition and buffer car acquisition • The addition of heavy haul truck (HHT) and barge routes for some sites, as well as support for user-defined inter-modal routes • Updates to the logic that checks the transportation cask thermal limit maps prior to package transport • An allocation method that predicts when reactor sites would pack assemblies from their spent fuel pools for dry storage, and prioritizes shipments directly from the pools of those reactor sites in the preceding periods (before the predicted loadings to dry storage), thus reducing the number of casks loaded into dry storage at reactor sites • The addition of “reactor site family” operational limits to restrict the number of loads of SNF taken from the pool and from dry storage at a given reactor site each year • Added support for multiple canister loading map options and packages with multiple compatible transportation overpacks • Updates to the handling of non-commercial SNF, including a new database that contains data to support the updates • Updates to allow analysis of hypothetical scenarios which include repackaging at reactor sites, e.g., for possible comparative analysis with other scenarios • Implementation of additional output reports, or modification of existing ones • The ability to generate and implement user edits via the NGSAM website • The ability to model loading SNF from pool storage at an interim storage facility (ISF) into dry storage at the ISF • The ability to model consolidating SNF from different existing storage containers at a DOE site into the same DOE standard canister • The ability to model transferring SNF casks from one transportation mode to another, e.g., from HHT to rail, referred to as transloading. These new features have improved NGSAM capabilities and users’ experience with the model. Preliminary NGSAM requirements for modeling advanced reactor fuels, reprocessing, treatment, and conditioning were considered, and this paper describes them at a high level. Other nuclear fuel cycle system analysis tools developed under sponsorship of DOE-NE, like the VISION code developed at Idaho National Laboratory, might be better suited for initial high-level analysis of those technologies and advanced fuel cycles. As technologies are developed and system concepts evolve, NGSAM could provide value by providing more detailed modeling of transport, storage, and disposal of spent fuel and wastes from advanced reactors and advanced fuel cycles at the fuel element and waste container level.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Brief User's Guide to the Excel ® -Based DF Calculator

To understand the importance of capturing penetrating forms of iodine as well as the other volatile radionuclides, a calculation tool was developed in the form of an Excel ® spreadsheet to estimate the overall plant decontamination factor (DF). The tool requires the user to estimate splits of the volatile radionuclides within the major portions of the reprocessing plant, speciation of iodine and individual DFs for each off-gas stream within the Used Nuclear Fuel reprocessing plant. The Impact to the overall plant DF for each volatile radionuclide is then calculated by the tool based on the specific user choices. The Excel ® spreadsheet tracks both elemental and penetrating forms of iodine separately and allows changes in the speciation of iodine at each processing step. It also tracks 3 H, 14 C and 85 Kr. This document provides a basic user's guide to the manipulation of this tool.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗