Search NASA⌕ Search

SEARCH · Search NASA

Results for “Reprocessing”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,801 records · Page 100
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Safeguards Challenges for Molten Salt Reactors

There are now a wide variety of molten salt reactor (MSR) technologies being proposed or developed that are gaining international interest and momentum toward deployment. There are two sub-categories of this class of reactors - those that are liquid fueled (i.e., the molten salt is both the fuel and the coolant) and those that have solid fuel with the molten salt used as coolant only. The liquid-fueled MSRs present unique challenges to current safeguards approaches. These challenges exist because there are currently no conceptual approaches to applying safeguards to unique, tightly coupled nuclear energy systems with the reactor core, balance of plant, and fuel cycle (everything else including salt processing and material separations) combined in a single facility. The application of safeguards to liquid-fueled MSRs will have to take into consideration several technical factors including: The homogeneous mixture of fuel, coolant, fission products, and actinides (with the attendant extremely high radiation field); Continuous variation in isotopic concentrations in the fuel salt, including removal (passive or active) of fission products, rare earth elements, and noble metals; High operating temperatures of the fuel salt; Online processing where some fraction of the inventory can be removed while the reactor is operational; Unique refueling schemes, including the ability to continuously feed the core with fresh fissile or fertile material; The presence of frozen fuel, potentially requiring a different safeguards process from the liquid fuel; and The presence of fuel outside the vessel. Additionally, If the thorium fuel cycle is employed, there will be additional complications since the resulting radiation signatures will be different from those of the uranium-based fuel cycle. The existing IAEA inspection regimes are based on the uranium-plutonium fuel cycle, item counting for nuclear reactors, and bulk material accountancy for the front and back end of the nuclear fuel cycle. These techniques and associated instrumentation for bulk accountancy have been developed predominantly for enrichment, fuel fabrication, and aqueous reprocessing. However, none of these bulk accountancy measures can be directly applied to liquid-fueled MSRs. This paper will explore some of the possible challenges in applying safeguards to molten salt reactors.

Kovacic, Donald N↗
DOE OSTI2018

Application of a Triple Bubbler Sensor for Determining the Density, Surface Tension, and Depth in Molten Salts

Electrochemical (i.e. pyroprocessing) technology is being widely studied throughout the world as a potential alternative to aqueous reprocessing of spent nuclear fuel (SNF). Central to this process is the electrorefiner (ER), in which the useful uranium in the SNF is electrochemically transported through a molten salt electrolyte to a cathode for later recovery. As part of this process, uranium, plutonium, and other actinides accumulate in the ER salt over time. Material accountancy and safeguards of these special nuclear material buildups in near real-time is a significant challenge due to the elevated processing temperatures, remote operation within a hotcell, and high radiation fields. Analytical techniques outside of the hotcell can be used to provide the SNM concentrations of salt samples from the ER. However, determining the actual mass of these materials in the ER presents a significant challenge because the density and volume within the vessel is largely unknown. The Idaho National Laboratory (INL) has developed a triple bubbler sensor capable of determining the density, surface tension, and depth of a fluid in near real-time. Previous experiments in aqueous media has shown that the sensor is capable of determining the density, surface tension and depth with accuracies on the order of 0.2%. The goal of this current work is to test the sensor in several molten salt media to calibrate and validate the triple bubbler. Experiments have been performed in LiCl-KCl eutectic salt. Calibration factors were 1.0015 and 1.26 for density and surface tension, respectively. With the above calibration factors applied, the depth of the salt in the test vessel was determined to within 0.4%. These results indicate that the triple bubbler sensor has the potential to accurately (less than 1% uncertainty) determine the density, surface, tension, and depth of the ER salt in near real-time. The triple bubbler sensor designed at INL can significantly enhance the material accountancy, safeguards, and process monitoring of the electrochemical processing of SNF.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2018

Thermodynamic Properties of Rare Earth Chlorides in Molten Salts: A Novel Approach to Experimental Electrochemistry of Molten Salts

The thermodynamic and electrochemical properties of rare earth chlorides in molten salt solutions are of interest from the standpoint of both pyrochemical processing of spent nuclear fuel and for the development of Generation IV nuclear reactors. Knowledge of key properties like activity coefficients and phase diagrams will accelerate the development of safe reliable carbon free energy. In this work, three conduits of research have been pursued and reported within the larger theme of minimization of waste in spent fuel reprocessing and associated technology development. The activity coefficient of rare earth chlorides has been investigated using ground breaking theory and experimental approach. Furthermore, an ionic size dependent activity coefficient model is presented. This book also contains research into quaternary phase diagrams for LiCl-KCl-CsCl-RECl3 molten salt systems. This book should prove an excellent introduction to anyone pursuing a career in experimental high temperature electrochemistry and molten salt chemistry.

Bagri, Prashant↗
DOE OSTI2016

IODINE ADSORPTION IN FIXED BED ADSORBER TESTS

I-129 emissions control is needed for used fuel aqueous reprocessing. I-129 capture has been studied for 40 years; Two sorbents tested in recent years – Silver Zeolite (AgZ) and silver-functionalized Aerogel (AGAero). Results show needed iodine adsorption efficiencies are possible, DFs up to10,000. Mass transfer zone (MTZ) depths ~2 in. for I2 and 8 in. for CH3I, but increase over time; deeper sorbent beds needed for efficient CH3I capture. Ag utilization between 10-70% for most test durations; lower for higher NOx. Iodine sorption capacities: ~10 g/100 AgZ and ~20 g/100 g AgAero; lower for high NOx. Nearly all adsorbed I was physisorbed iodine for I2 and CH3I.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A Triple Bubbler System to Measure Density, Surface Tension, and Depth of Electrorefiner Molten Salt

In pyroprocessing of used nuclear fuel (UNF), uranium from the fuel is electrochemically dissolved at the anode and transported through a molten salt electrolyte to the cathode—where it is later recovered for reuse. As part of this process, special nuclear materials (SNM) dissolve into the molten salt electrolyte. Monitoring the mass of these materials, particularly plutonium, is important for material accountability and safeguards of the electrorefiner (ER) used in pyroprocessing. The mass of plutonium in the ER vessel is a function of the volume and density of the salt as well as the plutonium concentration. Measuring these parameters presents challenges due to the high temperature, radiation, and accessibility of the ER within a hotcell. Here, the main motivation for this work was to develop a method capable of measuring the density and depth (volume) of salt under the extreme operating conditions within the ER to enhance the SNM accountability and safeguards of pyroprocessing. In addition, these online measurements would be useful for monitoring level changes associated with removal of electrolyte from the ER and different unit operations associated with electrochemical reprocessing of used nuclear fuel.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

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↗
DOE OSTI2019

A Self-Irradiation Correction for the Hybrid K-Edge Densitometer

Hybrid x-ray fluorescence and K-edge densitometry is used at nuclear fuel reprocessing facilities to determine the concentrations of uranium and plutonium in the input accountability vessel. The uranium concentration is determined by K-edge densitometry, that is, from the step-difference in transmission on either side of the K-absorption edge measured using a continuous x-ray spectrum on a vial of the solution defining a well-known geometry. Historically this was done using a basic region of interest analysis but recently to accommodate more challenging solution full spectral analysis has been developed with considerable benefits. In parallel with the transmission measurement, the relative plutonium-to-uranium concentration is obtained from the relative strength of the x-ray–induced K-shell x-ray fluorescence production measured at a backward angle in energy-dispersive mode. However, gamma-ray emissions from fission products present in the solution also cause fluorescence of the characteristic actinide x-rays. Uncorrected, these self-irradiation–induced x-rays can introduce a bias into the reported U:Pu result. The commonly implemented method to correct this interference requires acquisition of a long (~8 hour) passive spectra for the sample. Although effective, it is often impractical to perform a passive measurement on every sample such that a representative passive spectrum is used. We have developed an alternative correction methodology using only the active x-ray fluorescence spectrum that eliminates the requirement for the time-consuming passive count. The active self-interrogation methodology will be discussed and initial performance results presented.

Mcelroy Jr, Robert↗
DOE OSTI2019

Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using GeoMelt ICV - WM2020 Conference Paper

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt® In-Container Vitrification (ICV)™ technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV™ melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt® technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt®, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt® ICV™ process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass-formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt® ICV™ at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering-scale. Since 2018, a 10-metric ton full-scale GeoMelt unit (GeoMelt® Richland) was designed, installed, and commissioned at Perma-Fix Northwest in Richland, Washington for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt ® Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt® technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt® technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt® can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper.

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

Exploiting Knowledge of Technetium in the Subsurface to Develop Remediation Strategies – 18223

Technetium-99 (Tc) contamination remains a major environmental problem at legacy nuclear reprocessing sites, such as the Hanford Site in south-central Washington state, where the total inventory of released Tc is estimated at 700 Ci. The environmental risk associated with Tc is due to its long half-life (213,000 years) and the high mobility of the oxidized anionic pertechnetate species (Tc(VII)O4-). Thus, there is a critical need to resolve scientific and technical issues underpinning the ability to safely, cost-effectively, and efficiently remediate Tc-contaminated vadose zone sediments and groundwater. Developing enhanced attenuation strategies for released Tc that reduce contaminant migration, while mitigating risk to human health and the environment, requires characterization of any retardation processes and mass flux. The current conceptual model for Tc transport in the vadose zone assumes that Tc is predominantly present in its oxidized form as TcO4-, which is highly soluble and mobile. However, under certain conditions, Tc can become entrained in solid phases. Characterization of Tc-containing mineral phases from sediment cores extracted from various Hanford waste disposal sites were analyzed to determine Tc present in vadose zone sediments as (i) TcO4- that remains in water films associated with fine-grained sediments; (ii) Tc co-precipitated with other lanthanides in phosphate mineral phases; and (iii) Tc present as TcO2•nH2O in localized reducing zones. These results will be used to develop improved long-term Tc remediation strategies that are most likely to succeed during field application, through stimulation of conditions that enhance Tc attenuation.

Pearce, Carolyn I.↗
DOE OSTI2019

Status PIE Report on Legacy EBR-II and FFTF Metallic Fuel Experiments

A long term research focus of the US Department of Energy (DOE) Advanced Fuels Campaign (AFC) is to investigate technologies that can allow for increased actinide utilization in nuclear fuel for fast neutron spectrum reactors. In order to reach these goals, many experimental campaigns have been tested in an approximated fast reactor neutron spectra at the Idaho National Laboratory Advanced Test Reactor (ATR) in order to study the performance of metallic, nitride, ceramic and advance fuel forms. In addition and complementary to ATR testing, systematic characterization of experiments irradiated in true fast reactors are performed in order to assess ATR in qualifying fast reactor fuel. For this purpose, complementary postirradiation examinations (PIE) are performed on several metallic fuel pins previously irradiated in the EBR II. Advanced PIE techniques developed since the initial PIE of EBR II samples in the 1980’s have been applied in new examinations of irradiated samples during past EBR-II operations. The EBR-II metallic fuel pins analyzed in this report are X501 G591, an early minor actinide-bearing metallic fuel alloy; X430-T653, a U-19Pu-10Zr metallic fuel alloy with a larger diameter compared to EBR-II driver fuel, and more prototypic of future commercial fast reactors; X521-G594, a unique experiment to test the feasibility to burn weapon grade Pu without reprocessing; X496 CL47&CL39 from a long-life metallic fuel experiment with the goal to reach very high burnup; and X486-J555 a U-10Zr Mk-IIIA qualification fuel test that will serve as first of a kind “modelling driven PIE.” These legacy pins were evaluated against the historical fuel performance of previously irradiated fuel from literature. Fuel performance was investigated through a suite of PIE techniques including neutron radiography, gamma spectrometry, dimensional inspection, fission gas release measurements, chemical burnup analysis, optical microscopy, and advance microscopy. So far, the PIE campaign revealed fuel performance and characteristics that are generally in line with literature or EBR-II experience. Some exceptions are present, and they are directly discussed in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Radiological Degradation of Tetraoctyl Diglycolamide in n-Alkyl Solvents: Influence of Solvent Ionization Potential

Reprocessing of used nuclear fuel reduces the requirements for long-term storage repositories. Cost-efficient processes depend on knowledge of radiation effects, but gaps occur in understanding how radiation results in the degradation of fuel separation molecules. This project investigates the influence of different n-alkyl solvents on the radiolytic fragmentation behavior of N,N,N',N'-Tetraoctyl-3-oxapentane-1,5-diamide (TODGA), a promising ligand for extraction of the lanthanides and minor actinides from dissolved nuclear fuel. Findings from this work suggest a linkage between the ionization potential of the solvent and the degradation products observed in the resulting mass spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Crystalline Aluminosilicates for Radionuclide Remediation

Crystalline aluminosilicates have a wide range of applications such as media for the capture and/or immobilization of radionuclides. Several different types of metal-exchanged zeolites have been successfully demonstrated as effective sorbents for gaseous, aqueous, or solid-state capture and immobilization of radionuclides including iodine [e.g., I2(g), CH3I(g)], xenon, krypton, radium (e.g., 226Ra and 228Ra), selenium (e.g., 79Se6+), technetium (e.g., 99TcO4-), strontium (e.g., 90Sr2+), cesium (e.g., 135Cs+, 137Cs+), mixed alkali halide streams (i.e., iodide and chloride salts), and actinides (e.g., U6+, Th4+). These zeolites include sodalite (SOD), mordenite (MOR or Z-type), faujasite (FAU), Linde type-A (LTA), cancrinite (CAN), gismondine (GIS), and heulandite (or clinoptilolite) (HEU). The structural building blocks of these seven materials vary from cages (e.g., a-cages, ß-cages, e-cages) to channels linking various cage types and these are mostly comprised of aluminosilicates. This chapter will delve into the crystallographic differences between these framework materials and provide a brief overview of literature data where these seven crystalline aluminosilicates have been utilized for the capture and/or immobilization of radionuclides. Metal-exchanged zeolites (e.g., MOR, FAU, LTA) have often been the source of study for gaseous capture of iodine where the more promising compositions include Ag-exchanged sorbents; these types of materials can be made in a variety of robust forms including berl saddles and extrudates. Sodalite can be used as an immobilization strategy for halide waste streams (e.g., Cl, I), including those from electrochemical reprocessing of used nuclear fuel, as well as TcO4–. Cancrinite has been demonstrated to enhance sorption capacity of sediments for several problematic ionic radionuclides including Cs+, SeO42–, and TcO4–. Actinides such as U6+ and Th4+ have been successfully captured from solutions using MOR, FAU, LTA, and HEU-type aluminosilicates. Finally, mixtures of aluminosilicates have been included in a set of zeolite materials that show promise as sorbents for 226Ra and 228Ra removal from mine waters.

zeolites, aluminosilicates, radionuclide remediati↗
DOE OSTI2021

Multielectrode sensor for concentration and depth measurements in molten salt

A multi electrode sensor that provides in-situ, real time measurements for molten salts and other process fluids such as real-time concentration and salt level measurements for nuclear systems such as molten salt reactors, nuclear reprocessing facilities utilizing molten salts and concentrated solar power systems. The sensor has multiple electrodes with unique lengths which are connected to a potentiostat. Measurements are taken when the electrodes are immersed in the process fluid.

Hoyt, Nathaniel C.↗
DOE OSTI2022

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↗
DOE OSTI2022

DEVELOPMENT OF AN APPROACH FOR DOWNBLENDING AND DISPOSITION OF IRRADIATED GRAPHITE FUEL

The IGR reactor, a pulse-type test experimental reactor, reached full power in 1961. The reactor contains a center axial experimental ampoule that is subjected to a large pulse of neutrons (neutron flux maximum density of 7×1016 n/cm2·s, peak power 10GW) when the reactor is operated. The fuel form for the IGR reactor is graphite cubes and rods impregnated with highly enriched (90%) uranyl-dinitrate (UO2(NO3)2·6H2O). This fuel is eligible for return to Russia for reprocessing under the Russian Research Reactor Fuel Return program. However due to some technical and licensing issues it was decided to immobilize this fuel in Kazakhstan. The disposition approach thus selected for the spent HEU fuel involves down-blending of the fuel followed by cementation. Development of the process for down-blending and subsequent cementation requires a process that satisfies many criteria, including fissile content, compressive strength, permeability, and containment of radioactivity, among others. One key goal of the final product is to meet the criteria for termination of safeguards. This paper will describe the overall challenge associated with disposition of irradiated HEU graphite spent fuel, the identification of the objectives of the disposition process, and the constraints associated with development of the process and the waste form.

Gaines, Kris↗

Refine your search to explore more results.