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

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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Radiation Chemistry Research for Improved Nuclear Fuel Cycles

Nuclear energy is a key component of a global, low-carbon energy future. One of the challenges facing large-scale deployment of nuclear power is the long-term disposition of used nuclear fuel. Due to the political and technical challenges of siting deep geologic storage facilities, reprocessing of used nuclear fuel is desirable to reduce the total volume and decay heat of waste requiring long-term storage, enabling efficient usage of geologic storage repositories. However nuclear fuel reprocessing is complicated by intense multi-component radiation fields that degrade solvent systems that are used for this purpose. Resulting radiolytic degradation can result in reduced performance over time, and generation of additional waste. Thus, an understanding of the radiation-induced chemistry of these reprocessing solvent systems is crucial to for the development of new, radiation-resistant processes for efficiently treating used nuclear fuel. Controlled irradiation experiments combined with high-performance chromatography and mass spectrometry have provided comprehensive knowledge of radiolytic degradation pathways and kinetics, which is needed to predict the effect of radiation-induced chemistry on the performance of the nuclear fuel reprocessing technology. This talk will give an overview of recent progress in research on the radiation chemistry of fuel reprocessing molecules conducted by the Idaho National Laboratory Center for Radiation Chemistry Research (CR2) in conjunction with the Molecular Mass Spectrometry and Chromatography group.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Electrochemical Manipulation and Radiolytic Evaluation of Organic Phase Neptunium

Presentation for Seed LDRD proposal presentation: This Seed LDRD research proposes to test an innovative approach to precisely control the oxidation state distribution of neptunium (Np) under used nuclear fuel (UNF) reprocessing conditions using novel electrochemical methods. Current UNF reprocessing technologies are not financially viable in the US, requiring significant scientific and technological innovation to improve cost efficiency. Reducing the number of process cycles is one avenue for increasing cost efficiency, of which resolving the challenge associated with isolating Np is critical. Under envisioned process conditions, Np is present in a mixture of extractable (Np(IV) and Np(VI)) and inextractable (Np(V)) oxidation states, the distribution of which is dependent on several factors, including ionizing radiation dose, that change throughout a reprocessing scheme. Consequently, Np unintentionally partitions into various UNF reprocessing phases and product streams, ultimately increasing the number of process cycles to isolate UNF components. Here, we propose to employ novel, high surface area, optically transparent, Ligand Modified tin-doped indium oxide Electrodes (LMES). These LMEs bind Np, and therefore enable concurrent generation and spectroscopic characterization of Np oxidation states in organic solutions. These proof-of-concept experiments will: (i) facilitate optimization of the proposed electrochemical system conditions for Np oxidation state manipulation; (ii) identify the accessible electrochemical window for Np complexes in the organic phase; (iii) report characteristic optical spectra for each accessible complexed Np oxidation state; and (iv) determine the lifetime and partitioning of a given atypical Np oxidation state in the organic phase in the presence and absence of ionizing radiation fields. The data gathered by this Seed LDRD will provide support for the design of an electrochemical process concept for the precise manipulation of Np oxidation states in UNF reprocessing solvent systems, with the intention of providing advanced control over Np mass transfer, and thus greater process efficiency and economy.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2013

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

MARIE: A Python-Based Framework for Comprehensive Fuel Recycling Modeling

One of the most pressing challenges to the continued deployment of nuclear energy systems is in the ultimate management and disposition of discharged fuel assemblies. While reprocessing and recovery of valuable materials from UNF assemblies has been considered as part of an overall strategy for minimization of the volume of reactor-based wastes to be managed, the deployment of commercial-scale reprocessing facilities presents an enormous economic challenge. The MARIE software package has been developed as a means of confronting this challenge. Representing components of a generic fuel reprocessing operation as individual physical processes, MARIE is designed as a modular framework intended to allow for analysis and cost-optimization for a hypothetical reprocessing facility while realistically accounting for the physical characteristics of the used fuel source term, such as decay heat, activity, and radiation dose (informing corresponding shielding requirements). Capabilities supported by MARIE include head-end operations such as fuel shearing, voloxidation, and dissolution; generic solvent extraction operations informed by available open-literature data; a suite of unit operations intended to represent electrochemical processing of used fuel assemblies (i.e., oxide reduction, electrorefining, and electrowinning); and finally, accounting for both costs and physical features of discharged waste streams, which can be used to inform follow-on analyses such as the feasibility of deep-borehole disposal of HLW. This paper presents an overview of the MARIE software capabilities, including how individual unit operations are implemented to enable a larger-scale optimization of a hypothetical reprocessing operation on aspects such as cost and recovery of valuable materials.

Skutnik, Steve [ORNL] (ORCID:000000016441135X)↗
DOE OSTI2017

Novel Glass Formulations for Post Operational Clean Out of Highly Active Storage Tanks - 17596

The UK has a successful history of reprocessing nuclear fuel from both UK nuclear reactors and for overseas customers. Highly Active Liquor (HAL), a waste product from reprocessing, is stored in stainless steel tanks before being incorporated into a stable glass matrix in one of three vitrification lines at Sellafield. Over the next 4 years Sellafield Ltd will cease reprocessing nuclear fuel, and the reprocessing plants and storage tanks will enter a Post Operational Clean Out (POCO) phase to remove the radioactive inventory and prepare for future decommissioning. POCO feeds are expected to differ greatly from current HAL, and will contain high levels of molybdenum solids. This paper, and subsequent poster presentation, introduces the vitrification challenges expected with processing POCO waste-streams. The main challenge is the transfer and processing of high molybdenum containing solids. Details of the application of new glasses onto the active vitrification plant will also be included. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

The German product quality control of the compacted metallic nuclear waste - 15311

The treatment and safe disposal of spent nuclear fuel is a key challenge of nuclear energy utilization and has to be accomplished regardless the future nuclear of electricity production or phase-out policies. Until June 2005 approx. 6662 tons of spent fuel from German nuclear power plants was sent to the La Hague (France) and Sellafield (UK) plants for reprocessing. An equivalent amount of nuclear waste compounds has contractually to be returned to Germany, which is currently in progress. The returnable waste streams comprise mainly highly-active vitrified waste containing a major part of fission products and actinides, and intermediate-active super-compacted metallic waste containing reprocessed fuel element's structural materials and some technological waste from the reprocessing process itself. Shearing and dissolution of spent fuel leaves scraps of fuel metal cladding, the end-pieces and spacers which are contaminated, activated and irradiating from their life in a reactor core. In fact, this metallic intermediate level radioactive waste makes up a bit more than half of the total volume of the reprocessed waste. At La Hague the metallic waste is super-compacted. Several, typically eight, compacted discs are stacked in a stainless steel container, named CSD-C as for 'Colis Standard de Dechets Compactes'. 4104 containers of such CSD-Cs in 152 transport/storages casks have to be returned to Germany. Each one of them has to be checked and quality certified for its conformity with German national product specifications and acceptance criteria by independent national product quality control inspectors (PKS). The residue package documentation has to be complete and the characterization methods have to be comprehensive. Due to the heterogeneity of metallic waste the characterization and control of its properties is very complex. The nuclear characterization is performed by means of non-destructive measurements, where only a few key nuclides are measured directly. In fact, the proper calculation of the whole nuclear inventory relies on numerous supplementary simulations. The measurement and calculation methodology is scrutinized by PKS in detail and key points are benchmarked with our own simulations for the sake of an independent nuclear inventory declaration check. With respect to the vast amount of documentation and parameters, which have to be checked and verified, PKS had to optimize their checking procedures. For this purpose dedicated proof tools have been developed that shall be applied for selected indicative cross-checks of the relevant parameters. This contribution is dedicated to the procedure of quality control by PKS and examples of actually performed computational verifications of declared nuclear inventory. (authors)

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

Development of a Regulatory Strategy for Post Operational Clean Out Activities at Sellafield Limited Nuclear Licensed Site - 20273

The Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear licensed site carried out its final commercial shear of spent nuclear fuel in November 2018, since when the facility has undergone rundown prior to cessation of operations. The Magnox Reprocessing Facility (MRF) is also due to cease operating, around the end of 2020. Following the cessation of spent fuel reprocessing, the facilities and supporting infrastructure will transition into Post Operational Clean Out (POCO), which will become an increasingly significant portion of Sellafield Limited's (SL) activities. POCO is defined as the set of activities undertaken directly after a nuclear facility comes to the end of its operational life, in order to place the plant in a suitable state for the subsequent decommissioning steps. The decommissioning strategy at SL is for deferred dismantling of these facilities, which will be subject to a period of Surveillance and Maintenance (S and M) following completion of POCO. This paper outlines the development and implementation of a regulatory strategy for POCO activities at the Sellafield site by the Office for Nuclear Regulation (ONR), the independent nuclear safety, transport and security regulator in the United Kingdom. The paper also describes how this regulatory strategy is applied to nuclear installations undergoing transition into POCO and subsequent decommissioning. ONR works jointly with the Environment Agency, as the environmental regulators in England, to ensure that matters relevant to them have been duly considered. In developing a regulatory strategy, ONR has taken a site-wide approach, moving away from regulation on a facility-by-facility basis, to promote consistency, as well as maximising efficiency and effectiveness. The aim of this integrated approach was to prevent complications during future decommissioning and dismantling that have been experienced previously at similar facilities in France. The strategy recognised that once reprocessing operations are concluded, the hazard present will be reduced significantly, which provides the opportunity for ONR to consider how best to deliver a proportionate approach to regulatory interventions. Once POCO has been completed and the facility has transitioned into a period of S and M, the scale and frequency of ONR's interventions will reduce, commensurate with the safety case covering the remaining hazard and activities associated with maintaining the facility prior to final decommissioning. This approach allows ONR to consider risks holistically, considering the full life cycle of the facility and by taking due account of the interdependencies between facilities across the site. The strategy also aligns with existing regulation of high hazard and risk reduction activities, which remain ONR's top priority. As SL progresses with POCO activities in THORP and across the rest of the site, ONR will review its regulatory strategy to ensure that the intended safety benefits are realised. (authors)

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

Updates to a Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 20420

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The U.S. Department of Energy (DOE) was required to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980 including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the West Valley Site that evaluates potential future consequences for human health and the environment. This forms the core of the spatial domain of the West Valley PPA Model. The PPA Model, developed using the GoldSim system modeling software, is a tool intended to provide support for decision making that evaluates uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-grouped contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of high level radioactive waste resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The PPA Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife could be exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the PPA Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. PPA Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes. With this information, the decision makers can drive informed, defensible decisions regarding decommissioning of the Site. This paper includes an update of the information presented at WM2019 [1]. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

New Thermoplastic Vulcanizate Based on Acetoacetoxy Functionalized Natural Rubber/Polyamide12 Blend Filled with Carbon Black

Thermoplastic vulcanizates (TPVs) based on blends of natural rubber (NR) grafted with poly(acetoacetoxyethyl methacrylate), NR-g-PAAEM, and polyamide12 (PA12) were prepared by the dynamic vulcanization technique. NR-g-PAAEM was first prepared and its surface free energy was evaluated by means of contact angles. Blends of NR-g-PAAEM and PA12 were then prepared at a 60/40 blend ratio (wt%) using a dynamic vulcanization technique. The results revealed that TPVs exhibited a dispersed-phase morphology. However, the TPVs filled with carbon black (CB) showed much smaller dispersed rubber particles when compared to the unfilled TPV. TEM analysis also revealed that the CB appeared to be mainly located within the NR-g-PAAEM phase. The presence of CB generally raised the melt viscosity and increased shear during melt-mixing, resulting in finer dispersion of the rubber particles. Additionally, a significant enhancement in the tensile properties of the TPV was also achieved by the incorporation of CB. The optimum tensile properties were attained by the addition 30 phr of CB. The results also showed that the incorporation of CB into the NR-g-PAAEM/PA12 TPVs improved their stress relaxation properties. Moreover, the effects of reprocessing on the mechanical properties of the TPV filled with CB were also investigated. Although the TPV filled with CB exhibited reprocessing capability, reprocessing at high temperature and shear forces was found to negatively affect its tensile properties, especially after the third reprocessing cycle.

36 MATERIALS SCIENCE↗
DOE OSTI2015

AUTOMATED TRANSIENT IDENTIFICATION IN THE DARK ENERGY SURVEY

We describe an algorithm for identifying point-source transients and moving objects on reference-subtracted optical images containing artifacts of processing and instrumentation. The algorithm makes use of the supervised machine learning technique known as Random Forest. We present results from its use in the Dark Energy Survey Supernova program (DES-SN), where it was trained using a sample of 898,963 signal and background events generated by the transient detection pipeline. After reprocessing the data collected during the first DES-SN observing season (2013 September through 2014 February) using the algorithm, the number of transient candidates eligible for human scanning decreased by a factor of 13.4, while only 1.0% of the artificial Type Ia supernovae (SNe) injected into search images to monitor survey efficiency were lost, most of which were very faint events. Here we characterize the algorithm’s performance in detail, and we discuss how it can inform pipeline design decisions for future time-domain imaging surveys, such as the Large Synoptic Survey Telescope and the Zwicky Transient Facility. An implementation of the algorithm and the training data used in this paper are available at at http://portal.nersc.gov/project/dessn/autoscan.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2011

The Economic, repository and proliferation implications of advanced nuclear fuel cycle

The goal of this project was to compare the effects of recycling actinides using fast burner reactors, with recycle that would be done using inert matrix fuel burned in conventional light water reactors. In the fast reactor option, actinides from both spent light water and fast reactor fuel would be recycled. In the inert matrix fuel option, actinides from spent light water fuel would be recycled, but the spent inert matrix fuel would not be reprocessed. The comparison was done over a limited 100-year time horizon. The economic, repository and proliferation implications of these options all hinge on the composition of isotopic byproducts of power production. We took the perspective that back-end economics would be affected by the cost of spent fuel reprocessing (whether conventional uranium dioxide fuel, or fast reactor fuel), fuel manufacture, and ultimate disposal of high level waste in a Yucca Mountain like geological repository. Central to understanding these costs was determining the overall amount of reprocessing needed to implement a fast burner, or inert matrix fuel, recycle program. The total quantity of high level waste requiring geological disposal (along with its thermal output), and the cost of reprocessing were also analyzed. A major advantage of the inert matrix fuel option is that it could in principle be implemented using the existing fleet of commercial power reactors. A central finding of this project was that recycling actinides using an inert matrix fuel could achieve reductions in overall actinide production that are nearly very close to those that could be achieved by recycling the actinides using a fast burner reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Dismantling of the 904 Cell at the HAO/Sud Facility - 13466

La Hague facility, in France, is the spent fuel recycling plant wherein a part of the fuel coming from some of the French, German, Belgian, Swiss, Dutch and Japanese nuclear reactors is reprocessed before being recycled in order to separate certain radioactive elements. The facility has been successively handled by the CEA (1962-1978), Cogema (1978-2006), and AREVA NC (since 2006). La Hague facility is composed of 3 production units: The UP2-400 production unit started to be operated in 1966 for the reprocessing of UNGG metal fuel. In 1976, following the dropout of the graphite-gas technology by EDF, an HAO workshop to reprocess the fuel from the light water reactors is affiliated and then stopped in 2003. - UP2-400 is partially stopped in 2002 and then definitely the 1 January 2004 and is being dismantled - UP2-800, with the same capacity than UP3, started to be operated in 1994 and is still in operation. And UP3 - UP3 was implemented in 1990 with an annual reprocessing capacity of 800 tons of fuel and is still in operation The combined licensed capacity of UP2-800 and UP3 is 1,700 tons of used fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The benefits of a fast reactor closed fuel cycle in the UK

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nuclear Material Monitoring in Molten Salt Systems using Cyclic Voltammetry

As designs for the next generation of nuclear power plants and reprocessing facilities develop, molten salt based systems are receiving increasing attention as fuel containment, cooling and reprocessing media. Traditionally, aqueous systems for power production and fuel reprocessing are associated with well-developed methods for the monitoring and accountancy of nuclear material. Many of these techniques are proving insufficient in providing the same information in molten salt systems. The International Atomic Energy Agency (IAEA) has called for the continued evaluation of advanced measurement techniques in systems such as molten salt as part of the Generation IV International Forum (GIF) and International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). Cyclic voltammetry (CV) is an electrochemical technique that offers promising applications in molten salt systems for the in situ monitoring of nuclear material. A technique such as CV offers a method for monitoring nuclear material in a system relying on molten salt for power production, cooling or reprocessing. Cyclic voltammetry has been shown to provide information regarding the diffusion of electroactive species in molten salt. Once a diffusion coefficient data base is assembled, CV can be performed on molten salt systems to determine concentrations of species in situ.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Self-Adjusting Metal–Organic Framework for Efficient Capture of Trace Xenon and Krypton

The capture of the xenon and krypton from nuclear reprocessing off-gas is essential to the treatment of radioactive waste. Although various porous materials have been employed to capture Xe and Kr, the development of high-performance adsorbents capable of trapping Xe/Kr at very low partial pressure as in the nuclear reprocessing off-gas conditions remains challenging. Herein, we report a self-adjusting metal-organic framework based on multiple weak binding interactions to capture trace Xe and Kr from the nuclear reprocessing off-gas. The self-adjusting behavior of ATC-Cu and its mechanism have been visualized by the in-situ single-crystal X-ray diffraction studies and theoretical calculations. The self-adjusting behavior endows ATC-Cu unprecedented uptake capacities of 2.65 and 0.52 mmol g -1 for Xe and Kr respectively at 0.1 bar and 298 K, as well as the record Xe capture capability from the nuclear reprocessing off-gas. Further, our work not only provides a benchmark Xe adsorbent but proposes a new route to construct smart materials for efficient separations.

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

Review of on-line and near real-time spectroscopic monitoring of processes relevant to nuclear material management

Spectroscopic chemometric based on-line monitoring of used nuclear fuel (UNF) reprocessing solutions and characterization of legacy nuclear waste (LNW) stored at Hanford is discussed in this manuscript. Using on-line and near real-time monitoring with spectroscopy, as opposed to traditional off-line monitoring, can significantly reduce the cost, risk and improve the efficiency of characterizing UNF reprocessing streams and LNW since the data can be collected non-destructively and analyzed nearly instantly. It can accurately quantify and qualify nuclear material in near-real time which can help to know the immediate UNF reprocessing stream and LNW conditions and detect the diversion of materials. The ability to be applied in a real reprocessing plant and in an actual Hanford waste tank/transfer pipe has been demonstrated by applying this technique to accurately quantify analytes in real UNF streams and LNW samples. Furthermore, the future development of spectroscopy based on-line monitoring is also discussed in this manuscript.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Structure and Thermochemistry of Perrhenate Sodalite and Mixed Guest Perrhenate/Pertechnetate Sodalite

Treatment and immobilization of technetium-99 ( 99 Tc) contained in reprocessed nuclear waste and present in contaminated subsurface systems represents a major environmental challenge. One potential approach to managing this highly mobile and long-lived radionuclide is immobilization into micro- and meso-porous crystalline solids, specifically sodalite. We synthesized and characterized the structure of perrhenate sodalite, Na 8 [AlSiO 4 ]6(ReO 4 ) 2 , and the structure of a mixed guest perrhenate/pertechnetate sodalite, Na 8 [AlSiO 4 ] 6 (ReO 4 ) 2-x (TcO 4 ) x . Perrhenate was used as a chemical analogue for pertechnetate. Bulk analyses of each solid confirm a cubic sodalite-type structure (P$\overline{43}$n, No. 218 space group) with rhenium and technetium in the 7+ oxidation state. High-resolution nanometer scale characterization measurements provide first-of-a-kind evidence that the ReO 4 – anions are distributed in a periodic array in the sample, nanoscale clustering is not observed, and the ReO 4 – anion occupies the center of the sodalite β-cage in Na 8 [AlSiO 4 ] 6 (ReO 4 ) 2 . We also demonstrate, for the first time, that the TcO 4 – anion can be incorporated into the sodalite structure. Lastly, thermochemistry measurements for the perrhenate sodalite were used to estimate the thermochemistry of pertechnetate sodalite based on a relationship between ionic potential and the enthalpy and Gibbs free energy of formation for previously measured oxyanion-bearing feldspathoid phases. The results collected in this study suggest that micro- and mesoporous crystalline solids maybe viable candidates for the treatment and immobilization of 99 Tc present in reprocessed nuclear waste streams and contaminated subsurface environments.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

The corona of the broad-line radio galaxy 3C 390.3

We present the results from a joint Suzaku/NuSTAR broadband spectral analysis of 3C 390.3. The high quality data enables us to clearly separate the primary continuum from the reprocessed components allowing us to detect a high energy spectral cut-off (E{sub cut}=117{sub −14}{sup +18} keV), and to place constraints on the Comptonization parameters of the primary continuum for the first time. The hard over soft compactness is 69{sub −24}{sup +124} and the optical depth is 4.1{sub −3.6}{sup +0.5}, this leads to an electron temperature of 30{sub −8}{sup +32} keV. Expanding our study of the Comptonization spectrum to the optical/UV by studying the simultaneous Swift-UVOT data, we find indications that the compactness of the corona allows only a small fraction of the total UV/optical flux to be Comptonized. Our analysis of the reprocessed emission show that 3C 390.3 only has a small amount of reflection (R ∼ 0.3), and of that the vast majority is from distant neutral matter. However, we also discover a soft-X-ray excess in the source, which can be described by a weak ionized reflection component from the inner parts of the accretion disk. In addition to the backscattered emission, we also detect the highly ionized iron emission lines Fe xxv and Fe xxvi.

79 ASTRONOMY AND ASTROPHYSICS↗