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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Vadose Zone Transport of Tritium and Nitrate under Ponded Water Conditions

Vadose zone transport of tritium and nitrate can be important considerations at radioactive waste sites, landfills, or areas with industrial impacts. These contaminants are of particular concern because they typically have a relatively higher mobility in the subsurface compared to other compounds. Here, we describe a semiarid site with tritium and nitrate contamination involving a manmade ponded water source above a thick unsaturated zone at Los Alamos National Laboratory in New Mexico, USA. This study demonstrates the value of vadose zone flow and transport modeling for the development of field investigation plans (i.e., identifying optimal borehole locations and depths for contaminant characterization), and how a combination of modeling with isotope and geochemical measurements can provide insight into how tritium and nitrate transport in the vadose zone in semiarid environments. Modeling results suggest that at this location, tritium transport is well predicted by classical multiphase theory. Our work expands the demonstrated usefulness of a standard tritium conceptual model to sites with ponded surface conditions and agrees with previous results where a standard model was able to explain the evolution of a tritium plume at an arid waste disposal site. In addition, depth-based analyses of δ 18 O and δ 2 H of pore waters helped confirm the extent of pond infiltration into the vadose zone, and the δ 15 N of nitrate showed that the contaminant release history of the site was different than originally assumed.

58 GEOSCIENCES↗
DOE OSTI2025

Electrochemical Recovery of Iron from Spent Pickle Liquor by Chloride-Based Molten Salt Electrolysis

Spent pickle liquor (SPL) is a waste stream generated in the steel industry that presents both a disposal challenge and a potential resource for metal recovery. This study investigates the electrochemical extraction of high-purity iron metal from SPL using chloride-based molten salt electrolysis (CMSE). Electrochemical characterization of diluted SPL confirmed Fe 2+ as the dominant species. Thermal dehydration of SPL under inert atmosphere yielded FeCl 2 -rich solids, which were directly employed as feedstock for electrolysis in LiCl–KCl eutectic melts at 500 °C. Cyclic voltammetry of FeCl 2 in this melt revealed well-defined Fe 2+ /Fe 0 redox behavior within the electrochemical stability window of the supporting electrolyte. High coulombic efficiency (>85%) electrodeposition was achieved demonstrating that iron metal can be produced from dehydrated SPL by CMSE. The electrodeposited iron exhibited >98 wt% purity, which was further enhanced to 99.9 wt% via arc melting. The resulting iron powder was ferromagnetic, and its size distribution was found to be suitable for powder metallurgy applications. This work demonstrates a scalable, energy-efficient pathway for valorizing SPL into high-purity iron metal, advancing circular economy strategies in the steel industry.

Materials science↗
DOE OSTI2023

Understanding the Structure and Dynamics of Conjugated Polymers by Advancing Deuteration Chemistry and Neutron Scattering (Final Report)

The overarching goal of the proposed work was to set up a partnership between the University of Southern Mississippi (USM) and Oak Ridge National Laboratory (ORNL) to develop novel approaches to measure the backbone rigidity of conjugated polymers (CPs) and understand the critical role of sidechains on the backbone conformation and the materials macroscopic property. The backbone rigidity greatly influences the electronic properties of CPs, which ultimately determines the functionality and performance of these materials. Improvements in the electronic properties of CPs would allow for enhanced charge transport in semiconductor devices, improved photovoltaic performance, recycling of waste heat in thermoelectrics, and discovery of new phenomena that will enable the next generation of energy technologies. Although significant progress has been made to optimize the optical and electronic properties of CPs, largely through Edisonian methodologies, it remains a challenge to experimentally characterize conjugated backbone conformation (chain rigidity, torsion, planarity, and short-range order) and relate these to the fundamental optical and electronic properties (electronic coupling, charge transport, etc.). This has left fundamental gaps in our knowledge of the most basic structure/property relationships within these systems, precluded the study of fundamental physical phenomena, and constrained the design and realization of new electronic and device functionalities. Thus, the major goal of this work is to use novel deuteration methodologies via systematic synthetic approaches, and neutron scattering techniques to comprehensively characterize the structural and dynamic properties of CPs in contrast-matching solvents. Our work would, for the first time, elucidate the relationship between backbone rigidity and macroscopic properties. They will also allow a rational formulation of design principles for next-generation CPs that are resilient to disorder through precise control of the delocalized electrons along the polymer backbone. Overall, this project will advance our understanding of the structure, dynamics, and fundamental physics of these materials, which is crucial for enabling the prediction, design, control, and manipulation of current and emerging material electronic properties.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Nitrate-To-Ammonia Electroconversion at Neutral pH on Polycrystalline Vanadium Sulfide Derived from Vanadium Disulfide

The electrochemical nitrate reduction reaction (NO3RR) offers a pathway to produce NH3 for fuel and fertilizer from waste NO3-. In this work, a polycrystalline vanadium sulfide (VSx), which is derived from solvothermally grown and annealed VS2, is shown to exhibit excellent NO3RR activity (2.3 +- 0.6 mg.cm-2 geo..h-1 @ -0.92 VRHE) and Faradaic efficiency to NH4+ (69 +- 6% at -0.69 VRHE) in buffered neutral pH electrolyte containing 0.1 M NO3-. A variety of characterization techniques are leveraged to support the VSx assignment, including X-ray photoelectron spectroscopy, near-edge X-ray absorption fine structure spectroscopy, selected area electron diffraction, and X-ray diffraction measurements. The VS2 annealing step reduces the oxide character and generates VSx, which, based on the improved NO3RR activity, results in the creation of active sites for NO3- binding. To help shed light on NO3RR on VSx, VS2 is used as a model system, and a grand-canonical density functional theory (GC-DFT) investigation of VS2 shows strong evidence that S vacancies are active sites for NO3RR, where NO3- outcompetes H+ for adsorption at the S-vacancy sites. Moreover, GC-DFT results highlight a thermodynamically favorable reaction to generate NH4+ in an aqueous electrolyte at relevant cathodic potentials. As an annealed material, VSx may contain undersaturated V sites, which show an electronic structure similar to the theoretically calculated S-vacancy site of VS2, and these sites may contribute to the observed increase in NO3RR activity and selectivity for NH4+ on VSx versus unannealed VS2. Finally, kinetic isotope effect measurements suggest that the kinetic rate-limiting step of the NO3RR on VSx is not proton-coupled, indicating it may be the first electron transfer to adsorbed NO3*.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

DECOVALEX-2023: Task F2 Salt Final Report

The subject of Task F of DECOVALEX-2023 concerns performance assessment modelling of radioactive waste disposal in deep mined repositories. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. In Task F2- (salt), these objectives have been accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks. Coupled-process submodels and deterministic simulations of the entire PA model for a reference scenario for waste disposal in domal salt have been conducted. The task specification has been updated continuously since the initiation of the project to reflect the staged development of the conceptual repository model and performance metrics. Thermal, hydrological, mechanical, and chemical properties of individual components of the engineered and natural system were chosen for relevance by participating teams. The salt reference case system was characterized using data and measurements collected at relevant underground research laboratories (URLs), field sites, and simulation results from teams with specialized modelling capability. Participating teams made a wide range of model assumptions from compartmentalized networks to full 3D models of the salt formation. No single contributed model includes full-fidelity representation of all the features, events, and processes (FEPs) detailed in the task specification, but almost all features and processes are represented in at least one model. Despite differences in the modelling strategies developed by participating teams, all models indicate that salt compaction and radionuclide diffusion are key processes in the repository, and for the FEPs and model scenario considered, little of the disposed radionuclides will migrate beyond the repository seal over the 100,000 year simulations. In general, the model output quantities have the largest differences over the short term and near the waste. The models tend to be more similar further from waste and at later time. Disparities between the models are believed to be due to differing simplifications from the task specification, some of which are chosen simplifications to reduce complexity, and some are restrictions imposed by the modelling tools. A second round of this task has been accepted for DECOVALEX-2027 in conjunction with Task F1 on crystalline PA modelling. The future round includes waste package heating, improved modelling of salt creep closure, additional comparisons of coupled-process sub-models, and the impact of repository engineering design on radionuclide migration in the repository. Participants will also propose and finalize a set of uncertain inputs for the reference case simulations, propagate these uncertainties in a set of realizations, and conduct sensitivity analyses on the simulation results.

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

Use of the Manuela 3D/Radiation Scanning Tool at Nuclear Facilities - 20437

Licensees, facility operators and decommissioning contractors have a need for comprehensive, accurate and precise radiological and topographical data to plan and execute safe and efficient Decontamination and Decommissioning (D and D) projects. Traditional approaches rely greatly upon labor and administratively intensive data collection methods that require considerable time and effort to glean the relevant data to plan and perform work in a radiological work area. Orano Decommissioning Services has developed and utilized at French nuclear facilities a lightweight portable scanning instrument called MANUELA (Mobile Apparatus for Nuclear Expertise Localization Assistance). Recently Orano conducted demonstrations at two United States nuclear facilities using MANUELA that may prove useful in areas of the fuel cycle including: - ALARA (As Low As Reasonably Achievable) planning during facility operations; - Scoping and site characterization during the D and D planning phase; - Remedial action support during active decommissioning work; - Radiation and modeling planning for plant modifications; - Support Final Status Survey and demonstration of 'end state' compliance for license termination. MANUELA is currently being evaluated for use during decommissioning of nuclear facilities and could aide in demonstrating compliance with end state criteria after decommissioning. Of particular interest are decommissioning projects planned for various sites across the United States. Most of these sites present unique challenges and opportunities to safely and efficiently decommission in a regulatory compliant manner. The two demonstration site's current conditions of relatively low radiation levels inside containment proved to be ideal platforms to deploy the MANUELA instrument for evaluation of the complex geometries typically found in a nuclear facility. The MANUELA instrument was able to precisely identify local hot spots of the reactor system's primary equipment and was able to simultaneously overlay the dose information onto the scanned 3D images. Using data like this, a licensee would be able to make informed technical and personnel decisions on methods to remediate, shield or properly dispose of the radioactive source term. MANUELA can be used throughout the decommissioning process to validate progress and continue to refine the work activities to optimize the efforts of the crew which reduces downtime waiting for survey results, whereas traditional survey methodology relies on a physical survey to be performed and the data transcribed onto survey maps that are only two dimensional. The traditional format tends to be inaccurate and difficult to interpret the precise location(s) of hot spots or generalized dose rates. The resultant scan also allows the user to know the exact extent of the survey versus transposing information onto a map with only a portion of the map area scanned. MANUELA also compiles all of the relevant survey data into a downloadable table format which includes the measurement identification number, time stamp, point space location on the XYZ plane, and the dose rate value. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Unveiling Feedstock Variability: Insights into Corn Stover Conversion - Part I: Physicochemical Properties and Self-Degradation

Transforming agricultural waste into biofuels and bioproducts is crucial to advancing a low-carbon bioeconomy. However, the inherent variability in the composition and quality introduces uncertainties in the conversion efficiency and poses challenges in process development. Through integrating a high-throughput conversion system, material characterization techniques, and advanced data analysis tools, this study investigates the variability of corn stover and its subsequent impacts on carbohydrate conversion. The findings reveal that indoor storage substantially reduces the moisture and ash content and soil contamination, while other properties remain largely unchanged. Self-degradation due to microbial activity during storage decreases the carbohydrate content of corn stover but enhances glucose and xylose yields. A negative correlation is observed between sugar yields and lignin content across samples with varying ash and moisture content. The inhibitory effect of lignin diminishes in self-degraded samples likely due to the disrupted cell wall structure. Although self-degradation slightly increases cellulose crystallinity, no strong correlation was observed between the crystallinity and sugar yield. Hot water pretreatment under mild conditions effectively mitigates inherent variability, consistently improving the sugar yield from corn stover by up to 50%. By elucidating the feedstock variability and its impact on convertibility, these findings offer valuable insights into appropriate feedstock handling and management, highlighting potential strategies to address variability challenges.

09 BIOMASS FUELS↗
DOE OSTI2023

NNSA Funded 03-0016, Ion Beam Facility Characterization Compilation (Characterization Funded and Completed FY2021)

This document is the compilation of the sampling activities associated with the FY2021 NNSA funded characterization of the 03-0016, Ion Beam facility at Los Alamos National Laboratory. Included within this document are the radiological and non-radiological sampling plans, results from the executed sampling plans, areas not sampled, summary of findings for each area sampled, and anticipated waste streams.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Evaluation of Radiography for TRISO Buffer Layer Density Measurement

Tristructural isotropic (TRISO) fuel particles consist of a central uranium-bearing kernel and a series of coating layers designed to retain fission products and to ensure fuel performance. Several parameters such as thickness and density must be measured for these coating layers to show that they conform with fuel specifications. Current methods for measuring the density of pyrolytic carbon and silicon carbide layers (liquid gradient density column) and the buffer layer (mercury porosimetry) generate Resource Conservation and Recovery Act (RCRA) radiological-mixed waste. In addition, measurement of buffer and inner pyrolytic carbon layer densities require hot sampling or interrupted coating runs and the mercury porosimetry method used for buffer density measurement only measures the mean buffer density, not the interparticle distribution. A new approach has been evaluated to measure the density of coating layers in TRISO particles based on the dependence of x-ray attenuation in radiographs on material density. This method does not generate RCRA mixed waste, measures density on a particle-by-particle basis, and in principle is capable of measuring the density of all coating layers in a single process. Initial results using thinned TRISO particle sections to evaluate radiography measurement of density as a quality control characterization method are reported herein. In this work, the primary focus is on measurement of the density of the buffer layer; however, with appropriate calibration the method should be applicable to other coating layers. Improvements to the initial method and a full demonstration of the method on the remaining coating layers may be pursued as a future effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Characterization results for the 2020 tank farm 2H and 3H evaporator overhead samples

The 2020 annual 2H and 3H Evaporator overhead samples were received at Savannah River National Laboratory (SRNL) on December 22, 2020. The delivered samples were analyzed for concentration of cesium-137 ( 137 Cs), strontium-90 (90 Sr) and iodine-129 ( 129 I) to validate compliance with the Effluent Treatment Project Waste Acceptance Criteria (ETP WAC, Rev. 7). The concentration of all three radionuclides, 137 Cs, 90 Sr, and 129 I, in the 2H Evaporator overhead sample were found to be less than the corresponding ETP WAC limits. Radiochemical analysis of the 3H Evaporator overhead sample indicated a 137 Cs content about an order of magnitude higher than the acceptable threshold value, and 90 Sr and 129 I concentrations within the acceptance limits.

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

Summary of Analytical Results from Samples Supporting Tank Closure Cesium Removal (TCCR) Batch 3 and Modeling Results for Cs Loading on CST

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 3 for processing through the TCCR unit. Some results from analysis of those samples have been previously reported. This report documents the remaining analyses of the in-tank batch contact samples as well as the results of ZAM (Zheng, Anthony, Miller) isotherm modeling performed for comparison to the measured results. Results of the additional analyses include analysis of the loading of other radionuclides besides 137 Cs on the crystalline silicotitanate (CST) contained within the in-tank batch contact test samples. Results from those analyses revealed the next highest contributor to the activity on the CST was 90 Sr with an average loading of 2.76E+08 dpm/g CST compared to 3.56E+10 dpm/g CST for the 137Cs. Isotopes of plutonium were also detected on the samples. ZAM modeling was performed using the measured composition of the Tank 10H Batch 3 qualification samples. The modeling predicted a maximum Cs loading approximately 2.2x higher than the measured result. This is a slightly lower ratio (expected/measured) compared to what was observed for the prior TCCR in-tank batch contact testing performed for Batches 1A and 2 where the ZAM results were 2.7-2.8x higher than the measured values.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Time-dependent thermal degradation of lost circulation materials in geothermal systems

Treatment of lost circulation can represent anywhere from 5 to 25 % of the cost in drilling geothermal wells. The cost of the materials used for lost circulation treatment is less important than their effectiveness at reducing fluid losses. In geothermal systems, the high temperatures (>90 °C) are expected to degrade many commonly used lost circulation materials over time. This degradation could compromise different materials ability to mitigate fluid loss, creating more non-productive time as multiple treatments are needed, but may result in recovering desired permeability zones within the reservoir section over time. This research aimed to study how thermal degradation of eight different lost circulation materials affected their properties relevant to sealing loss zones in geothermal wells. Here, mass loss experiments were conducted with each material at temperatures of 90–250 °C for 1–42 days to measure the breakdown of the material at geothermal conditions, collecting gases during several experiments to determine the waste produced during degradation. Compaction experiments were conducted with the degraded materials to show how temperatures reduced the rigidity and increased packing of the materials. Viscosity tests were conducted to show the impact of different materials on drilling fluid rheology. Microscope observations were conducted to characterize the alterations to each material due to thermal degradation. Organic materials tend to degrade more than inorganic materials, with organics like microcellulose, cotton seed hulls and sawdust losing 30–50 % of their mass after 1 day of heating at 200 °C, while inorganics like magma fiber only lose ~5–10 % of its mass after one day of heating at 200 °C. Granular materials are the strongest when compacted despite any mass loss, while fibrous and flaky materials are fairly weak and breakdown easily under stress. The materials do not generally affect fluid rheology unless they have a viscosifying agent as part of the mixture. Microscopic analysis showed that more rigid materials like microcellulose and cedar fiber degrade in brittle manners with splitting and fracturing, while others like cotton seed hulls degrade in more ductile manners forming meshes or clumps of material. The thermal breakdown of lost circulation materials tested suggests that each material should also be classified by its degree of thermal degradability, as at certain temperatures the materials can lose the capability to bridge loss zones around the wellbore.

15 GEOTHERMAL ENERGY↗
DOE OSTI2020

Technology Development for Dry Storage of Aluminum-Clad Spent Nuclear Fuel - 20490

A candidate disposition pathway for the > 13 MTHM of aluminum-clad spent nuclear fuel (ASNF), owned and managed by the U.S. Department of Energy, is the drying and placement of the SNF into sealed-canister dry storage, with the ASNF-in- canisters 'road-ready' for transportation to and final direct disposal in a repository waste package. Technical information gaps in fuel drying, and fuel dry storage behavior, have previously challenged the declaration of technology readiness for drying and placement of this fuel into the DOE Standard Canister design for > 50 years of safe dry storage. The principal technical information gaps included: i) characterization and thermal dehydration behavior of aluminum (oxy)hydroxide films attendant on the cladding due to film formation during reactor operation and during post-discharge up to long-term wet storage histories; and ii) G-value data to enable estimation of the radiolytic gas generation from the cladding with its (oxy)hydroxide films. Thus, the oxide films on the ASNF challenged the safety of a sealed storage canister with thermal and radiolytic decomposition of the waters on the films that can lead to corrosion, pressurization, and flammability issues. These gaps in the technical information base have largely been closed. This paper discusses the investigations at the Idaho National Laboratory (INL) and the Savannah River National laboratory (SRNL), and outlines the pending technology development work for input to an engineering design to enable a road-ready dry storage system for ASNF. The ASNF inventory considered for road-ready dry storage is stored at the Savannah River Site (SRS) and at the Idaho Nuclear Technology and Engineering Center (INTEC) at the INL. The ASNF inventory in the SRS L Basin is from foreign and domestic research reactors (FRR and DRR), and is diverse in terms of design, irradiation, and post-reactor-discharge storage conditions; these factors yield a range of characteristics of cladding oxide films on ASNF. Mixed aluminum (oxy)hydroxide (boehmite and bayerite/gibbsite) films, non-uniform in thicknesses up to a maximum local thickness bounded by 25 μm, were observed on ASNF materials removed from wet storage in the L Basin and in non-sealed dry storage at the SRS following reactor service and a long-term (up to 40+ years) interim storage history. The ASNF inventory at INTEC, in both wet and vented dry storage, is predominantly from the Advanced Test Reactor (ATR), but it also includes DRR and FRR fuel. To address a profound behavior of these films, radiolysis testing of aluminum specimens with mixed type boehmite/bayerite oxide films was performed to develop basic data on G-values for production of radiolytic hydrogen under dry storage conditions with nominal relative humidity, temperature, and cover gas. Modeling and simulation of canister internal environments with postulated inventories of oxide films provides estimation of the evolution of the conditions of the canister loaded with ASNF. Simultaneous Thermal Analysis (TGA/DSC) of hydrated oxide powders, and laboratory-scale aluminum specimens with a bayerite film (∼10 μm) using TGA methods, inform drying time/temperature conditions to be used for the ASNF. A demonstration project is recommended for Verification and Validation of the drying and storage of the ASNF. Remaining major tasks leading up to the hot demonstration include scale-up radiolysis testing and scale-up drying testing. Engineering design with the information from the technology program will establish the safety basis and enable long-term (> 50 years) dry storage compatibility with ASNF in the DOE Standard Canister pending its transportation to and disposal in a repository. This full capability would show the ASNF-in-canister storage system to be road-ready. (authors)

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

Summary of Expedited Results from Samples Supporting Tank Closure Cesium Removal (TCCR) Batch 3

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal (TCCR) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 3 for processing through the TCCR unit. SRNL received and characterized a set of dip samples collected from Tank 10H (one surface and one variable depth sample). No solids were observed in either sample and therefore the samples were combined, and a suite of analyses were performed on the composite sample. The density of the combined sample was 1.174 g/mL (0.18 %RSD), which was consistent with the measured sodium concentration of 3.54 M (0.08 %RSD). The 137Cs activity in the sample was 4.75E+07 dpm/mL (4.28 %RSD) and the total Cs concentration was calculated to be 1.54 mg/L (1.16E-05 M). The in-tank batch contact samples consist of 0.1 g of crystalline silicotitanate (CST) contained within a teabag device. Duplicate samples were submerged in Tank 10H supernate for a period of about 13 days, after which time they were retrieved and transferred to SRNL for analysis. At SRNL the CST was rinsed to remove excess salt solution and the CST was then air dried before being digested for analysis. Results of the analysis indicated a 137 Cs loading of 3.66E+10 ± 1.40E+09 dpm/g or 16.5 ± 0.63 Ci/kg CST . This value represents a bounding upper limit as it includes the addition of two sigma uncertainty from replicate analysis of the individual teabag samples as well as the addition of the small amount of 137 Cs activity measured in the rinse solutions. The above values are based on the air-dried mass of CST. Correcting to the true dry mass using a F-factor of 0.8191 results in a maximum loading of 20.2 ± 0.77 Ci/kg CST .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

High volume packing fraction TRISO-based fuel in light water reactors

We report that for a decade, fully ceramic microencapsulated (FCM) fuel, containing tri-structural isotropic (TRISO) fuel particles in a silicon carbide (SiC) matrix, has been investigated as an accident-tolerant fuel for light water reactors (LWRs). Other examples exist of TRISO-based concepts for LWR fuels with different matrix materials. Previous studies assumed TRISO particle volume packing of approximately 0.44 in SiC (or another) matrix, the highest realistic packing fractions possible with conventional manufacturing. Recent advances in advanced manufacturing have yielded the development and demonstration of a fuel form that consists of conventionally manufactured TRISO particles in a 3D-printed SiC matrix with significantly higher possible TRISO packing fractions (0.5–0.7). This increased uranium loading enhances the viability of using TRISO-based particle fuel forms in LWRs. The viability of high-packing-fraction TRISO-based particle fuel forms in LWRs is assessed from the perspective of fuel cycle length, achievable fuel burnup, reactivity coefficients, and fuel cycle performance. Higher-packing-fraction TRISO-based fuel enables either longer cycle lengths (by ~25% at a packing fraction of 0.55 relative to 0.44) at a constant enrichment or decreased enrichments (by ~25% at a packing fraction of 0.55 relative to 0.44) at a constant cycle length. Studies of different fuel kernel types (uranium nitride, uranium oxycarbide, and uranium carbide) yield similar results, although the cycle length of uranium oxycarbide is shorter than for uranium nitride or uranium carbide (due to the lower density of uranium oxide). This work also characterized the production of 14 C resulting from neutron absorption in 14 N during operation for uranium mononitride fuel kernels; the ratio of 14 C/N was 1–2 at. % at discharge. For the fuel cycle evaluation, the activity of spent nuclear fuel and high-level waste at 100 and 100,000 years was lower for high-packing-fraction fuels than for conventional LWR fuel. Environmental impact metrics were similar overall, but higher on the front end of the fuel cycle and lower on the back end of the fuel cycle. Reactivity coefficients of higher-packing-fraction TRISO-based fuel were reasonable compared with those of conventional fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AFC Metal Fuel Workshop Presentations

Compilation of PowerPoint slide decks from presentations given at the 3rd Annual AFC Metal Fuels Workshop. The purpose of this workshop is to bring the national metallic fuel stakeholders together to discuss status and needs of metallic fuel and review the DOE AFC Metal Fuel R&D. Desired Outcomes - Advance Metallic Fuel Technology: 1. Stakeholder Community Building 2. National Program R&D Accountability, Planning Support, and Increased Stakeholder Advocation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Environmentally Sound Processing Technology: JANNAF Safety and Environmental Protection Subcommittee and Propellant Development and Characterization Subcommittee Joint Workshop

Topics covered include: Risk assessment of hazardous materials, Automated systems for pollution prevention and hazardous materials elimination, Study design for the toxicity evaluation of ammonium perchlorate, Plasma sprayed bondable stainless surface coatings, Development of CFC-free cleaning processes, New fluorinated solvent alternatives to ozone depleting solvents, Cleaning with highly fluorinated liquids, Biotreatment of propyleneglycol nitrate by anoxic denitrification, Treatment of hazardous waste with white rot fungus, Hydrothermal oxidation as an environmentally benign treatment technology, Treatment of solid propellant manufacturing wastes by base hydrolysis, Design considerations for cleaning using supercritical fluid technology, and Centrifugal shear carbon dioxide cleaning.

Pickett, Lorri A.↗