Search NASA⌕ Search

SEARCH · Search NASA

Results for “Metallic Fuel Database”

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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Validation Gap Assessment of a MAP Package Containing Fresh, Metallic Sodium-Cooled Fuel

This report describes an assessment of sodium fast reactor assemblies loaded within an AREVA MAP package. The assessment was performed to determine the state of the validation basis for sodium fast reactor fuel within a transportation package originally intended for fresh, light-water reactor fuel assemblies. A similar study is being simultaneously released (Cumberland, 2025), and substantial parts of the explanatory text are identical to that parallel work, which follows the same workflow. Both studies employed the TSUNAMI-3D and TSUNAMI-IP sequences from the SCALE code system to determine correlation coefficients between various configurations and databases of validation assessments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1997

Metal-Air Batteries: (Latest citations from the Aerospace Database)

The bibliography contains citations concerning applications of metal-air batteries. Topics include systems that possess different practical energy densities at specific powers. Coverage includes the operation of air electrodes at different densities and performance results. The systems are used in electric vehicles as a cost-effective method to achieve reliability and efficiency. Zinc-air batteries are covered more thoroughly in a separate bibliography. (Contains 50-250 citations and includes a subject term index and title list.)

FUEL CELLS↗
DOE OSTI2024

Generation of Enrichment-Dependent Thermal Neutron Scattering Data

This work details the generation of enrichment-dependent thermal neutron scattering cross sections for several crucial uranium fuel compounds. The evaluations of the thermal scattering law (TSL) and associated cross sections for uranium dioxide (UO 2 ), uranium carbide (UC), and uranium nitride (UN) were performed using standard ab initio lattice dynamics (AILD) methods. The data for uranium metal was produced using a novel hybrid approach of molecular dynamics combined with lattice dynamics methods. 235 U enrichments of 5%, 10% (LEU+), 19.75% (HALEU), 93% (HEU), and 100% were considered, in addition to natural uranium. The enrichment-dependent masses and free atom cross sections were used in the generation of elastic and inelastic thermal neutron scattering cross sections, while the calculation of the phonon density of states (DOS) and resulting TSL considered only the natural isotopic composition of uranium. The use of an identical DOS for all enrichments is expected to have minimal impact on the final data, as the small change in uranium mass should not significantly affect lattice vibrations. The cross sections are shown to exhibit significant dependence on 235 U enrichment. The submission of this data to the National Nuclear Data Center (NNDC) for release in the ENDF/B-VIII.1 database should support the design of advanced reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Assessment of thermodynamic stability of sapphire in eutectic molten chloride environment

The continued development of molten salt reactors requires highly efficient and stable instruments that can efficiently track monitor conditions of the molten salt during long-term application operation in both the fuel and coolant/heat transfer fluid loops. Sapphire (Al 2 O 3 ) fibers have shown tremendous potential by enabling radiation resistance and a broader operational range of temperature . Thermodynamic modeling using the ThermoCalc and the SGTE Molten Salts (SALT1) and Pure Substances (Pure5) database is done performed to understand the compatibility of Al 2 O 3 fibers with NaCl-MgCl 2 eutectic molten salt in the temperature range of 600 to 1000°C. The thermodynamic calculations show that sapphire fibers are not expected to be compatible with long term exposure to molten chloride salts within this temperature range. Furthermore, various pure metallic elements were evaluated as potential cladding materials for part of the Al 2 O 3 fibers. Based on the thermodynamic analysis, molybdenum (Mo) and nickel (Ni) could be effective cladding materials to enhance the stability of Al 2 O 3 in NaCl and MgCl 2 chloride salt molten bath in the desired temperature range. The presence of a metallic cladding leads to the suppression of aluminum dissolution from Al 2 O 3 , thus improving its stability in chloride salts. Additionally, it is also shown that Al 2 O 3 remains stable up to 2300 K in the presence of pre-existing Al 2 MgO 4 and Al 2 NiO 4 in the eutectic molten chloride bath environment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Characterization of Long-Term Service Coal Combustion Power Plant Extreme Environment Materials

The objective of this DOE-sponsored project was to develop a comprehensive database of mechanical properties, alloy microstructures, and to a lesser extent, the oxidation/corrosion behaviors of coal-fired power plant components, such as boiler tubing, steam headers, and steam piping, which had been in service for at least 100,000 operating hours (preferably more than 200,000 operating hours) under the operating conditions of high temperatures and high mechanical stresses where creep, fatigue, steam-side oxidation, and fireside corrosion were life-limiting factors. The components included in this database consisted of ferritic steels, creep strength enhanced ferritic (CSEF) steels, and 300-series H-grade stainless steels, as well as dissimilar metal welds (DMWs) among these types of materials. As a result of extensive metallurgical characterization and mechanical testing performed in this project, a comprehensive database on mechanical properties and detailed quantitative microstructural information was successfully developed for several long-term serviced EEM components. Such a database can be used by material research communities to develop, calibrate, refine, and validate mechanical behaviors, models, and other assessment tools for accurate prediction of remaining life of major components under similar EEM operating conditions.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2017

Liquefaction of Forest Biomass to "Drop-in" Hydrocarbon Biofuels

The overall goal of the project was to demonstrate solvent liquefaction as a viable path to stable intermediates for subsequent upgrading to fuel blendstocks. Iowa State University (ISU) utilized technology developed by Chevron to produce “green crude” via continuous biomass liquefaction to bio-oil without catalyst or reducing gas, then upgraded the crude to drop-in hydrocarbon liquid transportation fuels. Original benchmarks for the project included conversion of biomass to bio-oil at a yield of 50% or more, produce bio-oil with oxygen content below 20 wt%, and recycle a portion of the bio-oil product for use as solvent, displacing initial hydrocarbon solvent. Initial testing and technology development was completed by Chevron using a custom Small Continuous Liquefaction Unit (SCLU) operated in “once-through” mode. In this system, liquid product and solid residual/char was collected together and required both off-line solids removal and product separation. This unit was disassembled and major components incorporated into a redesigned solvent liquefaction system at ISU capable of processing biomass at rates up to 1 kg/h. Online, continuous solids separation and bio-oil recovery/separation systems were successfully implemented. Online solids removal using an acetone injection to aid in separation achieved over 99% solids removal efficiency, and acetone recovery of greater than 97% was achieved. The gas stripping system developed for the separation of medium wood oil (MWO) for the recycle stream from the heavy wood oil (HWO) product stream was also successful in recovering approximately 93 wt% of the initial solvent from the biomass-derived products. Due to the generation of fine particulate and the high molecular weight of the HWO which both negatively affect the ability to separate the product oil, solvent recycle was not fully implemented in the system. The biomass used in this project is loblolly pine forest residue. The solvent was a blend of two hydrocarbon liquids. The majority of the solvent was comprised of commercially available naphthalene-depleted heavy aromatic solvent. The hydrogen donor solvent was a proprietary cut of light cycle oil (LCO) that was hydrotreated. As much as 25 wt% of the solvent mixture was comprised of the hydrotreated LCO. Higher blend ratios were limited by the economics of providing hydrogen to the process in this manner. Two cases were evaluated for technical feasibility on the SCLU. These are 1. Simulated phenolic solvent recycle and 2. Hydrocarbon-rich solvent processing. The simulated phenolic solvent recycle case was based on an idealistic solvent mixture where only 5% makeup or hydrogen donor solvent entered the system. The resulting solvent mixture at steady state was estimated to be 75 wt% depleted hydrogen donor, 20 wt% biomass derived phenolic monomers, and 5 wt% HLCO. This case resulted in the highest bio-oil organic yield observed in the project of 66.7 wt% on a solvent-less, dry biomass basis and relatively low oxygen content of 12%. Additionally, these conditions deconstructed the biomass so well, that extremely fine particulate was generated such that 5 micron barrier filters were unable to remove the material. Offline addition of acetone, centrifugation and acetone removal were required in order to remove the particulate prior to fractionation with the stripping column. The resulting HWO had a high molecular weight and was therefore very viscous in nature, remaining solid at room temperature. Additionally, the column was unable to sustainably strip enough MWO to close the recycle loop balance while maintaining flowability of product oil exiting the column. It was determined that additional HLCO would be required at the front of the process to decrease product viscosity both by dilution and additional hydrogen transfer into the HWO product oil. The hydrocarbon rich mixture was the preferred mixture for a majority of the project and featured a 25% donor or HLCO content and 75% depleted hydrogen donor mix. A series of three tests were conducted using this mixture and serve as the basis for a paper detailing the system. While the online acetone injection system for high pressure solids removal was demonstrated early on using this mixture, difficulty controlling process flows led to the use of an atmospheric stirred tank filtration system. At these high donor solvent levels, the product stream easily partitioned with sufficient MWO to close the recycle loop balance. Additionally, this mixture produced HWO product oil at just over 55% yield, <5% oxygen content, and a reasonable viscosity at room temperature. A 56 hour campaign was conducted with the ISU pilot to produce a large enough batch of HWO product oil to conduct upgrading experiments at Chevron’s Richmond, CA facility. Benchtop hydrotreating units were used to conduct long-duration hydrogenation/hydrogenolysis reactions in order to provide yield data and help validate assumptions made in the TEA model. Resulting products featured an oxygen content of <1%, near the detectability limit. Hydrocracking simulations were also performed, and a distribution of heavy wood oil to refinery products was determined. These distributions are based on a database of commercial and pilot plant operation used by Chevron to estimate product yield from novel inputs based on the input physical and chemical characteristics. While multi-week benchtop hydrotreating tests were successfully conducted without fouling, analysis and testing of the heavy wood oil suggest it may be better suited for hydrocracking. Metals content in the heavy wood oil would likely create catalyst fouling over time and reduce lifetime of hydrotreating catalysts. Yields and product partitioning were also more favorable from hydrocracking compared to those from hydrotreating.

09 BIOMASS FUELS↗
DOE OSTI2020

Preliminary Plan for Evaluation of Reactor Pressure Vessel Surveillance Materials from Palisades Nuclear Generating Station

The Palisades Nuclear Generating Station (PNGS), located in Michigan, is owned and operated by Entergy. It is a Combustion Engineering 2-loop pressurized water reactor (PWR) producing 805 MWe (2,565 MWth). The PNGS was built between 1967 and 1970, with approval to operate at full power in 1973; the plant’s original license was due to expire on March 24, 2011. An application for 20-year extension was filed in 2005 with the Nuclear Regulatory Commission and was granted on January 18, 2007. Although the plant was scheduled for decommissioning by 2031, Entergy currently plans to close the PNGS in 2022. The PNGS included in its reactor pressure vessel (RPV) surveillance program a capsule, designated A-60, containing specimens of one of the vessel plates and a weld metal with nickel content of about 1.36 wt% and copper content of about 0.20 wt%. This capsule was irradiated to a fluence of 1.8 x 10 20 n/cm 2 . The capsule was removed from its surveillance position in 1995 and has been resident in the spent fuel pool since that time. The material is also of special interest because of its very high nickel content and because of the potential for development of NiMnSi (nickel-manganese-silicon) precipitates, dubbed “late blooming phases.” The surveillance program also includes a capsule, designated T-150, dedicated specifically for thermal aging, which would provide results for at least 33 effective full power years, which is beyond the current thermal-aging database for such materials. Given that license extensions to 60 years of operation have been approved by the US Nuclear Regulatory Commission for most of the currently operating light water reactors in the United States, and that the first extension to 80 years was recently approved, there exists the probability that some RPVs will reach and possibly exceed a fast neutron fluence (> 1 MeV) of 1 x 10 20 n/cm 2 . This is a fluence regime with no US surveillance data and very little test reactor data, except for the Light Water Reactor Sustainability Program–sponsored University of California Santa Barbara Advanced Test Reactor (ATR) ATR-2 project. Thus, the materials in the A-60 capsule represent a valuable resource for directly exploring the effects of commercial surveillance irradiation on a typical plate and a high-nickel weld with similar materials irradiated in the ATR-2 test reactor project. This report provides background information for the surveillance program, previous results of surveillance materials testing, the preliminary plan for capsule retrieval and disassembly, and the plan for testing and microstructural examination of the mechanical test specimens of these unique materials to assess the features induced by very high irradiation fluence or very long thermal aging time

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2014

NASA Tech Briefs, June 2014

Topics include: Real-Time Minimization of Tracking Error for Aircraft Systems; Detecting an Extreme Minority Class in Hyperspectral Data Using Machine Learning; KSC Spaceport Weather Data Archive; Visualizing Acquisition, Processing, and Network Statistics Through Database Queries; Simulating Data Flow via Multiple Secure Connections; Systems and Services for Near-Real-Time Web Access to NPP Data; CCSDS Telemetry Decoder VHDL Core; Thermal Response of a High-Power Switch to Short Pulses; Solar Panel and System Design to Reduce Heating and Optimize Corridors for Lower-Risk Planetary Aerobraking; Low-Cost, Very Large Diamond-Turned Metal Mirror; Very-High-Load-Capacity Air Bearing Spindle for Large Diamond Turning Machines; Elevated-Temperature, Highly Emissive Coating for Energy Dissipation of Large Surfaces; Catalyst for Treatment and Control of Post-Combustion Emissions; Thermally Activated Crack Healing Mechanism for Metallic Materials; Subsurface Imaging of Nanocomposites; Self-Healing Glass Sealants for Solid Oxide Fuel Cells and Electrolyzer Cells; Micromachined Thermopile Arrays with Novel Thermo - electric Materials; Low-Cost, High-Performance MMOD Shielding; Head-Mounted Display Latency Measurement Rig; Workspace-Safe Operation of a Force- or Impedance-Controlled Robot; Cryogenic Mixing Pump with No Moving Parts; Seal Design Feature for Redundancy Verification; Dexterous Humanoid Robot; Tethered Vehicle Control and Tracking System; Lunar Organic Waste Reformer; Digital Laser Frequency Stabilization via Cavity Locking Employing Low-Frequency Direct Modulation; Deep UV Discharge Lamps in Capillary Quartz Tubes with Light Output Coupled to an Optical Fiber; Speech Acquisition and Automatic Speech Recognition for Integrated Spacesuit Audio Systems, Version II; Advanced Sensor Technology for Algal Biotechnology; High-Speed Spectral Mapper; "Ascent - Commemorating Shuttle" - A NASA Film and Multimedia Project DVD; High-Pressure, Reduced-Kinetics Mechanism for N-Hexadecane Oxidation; Method of Error Floor Mitigation in Low-Density Parity-Check Codes; X-Ray Flaw Size Parameter for POD Studies; Large Eddy Simulation Composition Equations for Two-Phase Fully Multicomponent Turbulent Flows; Scheduling Targeted and Mapping Observations with State, Resource, and Timing Constraints;

Source record↗
DOE OSTI2024

Life-Cycle Assessment Integration into Scalable Open-Source Numerical Models (LiAISON) for Analyzing Emerging Low-Carbon Technologies

Decarbonizing the industrial sector is a significant challenge in achieving a net-zero greenhouse gas (GHG) emissions economy by 2050 and the Paris Agreement, i.e., a global climate change mitigation target of achieving a maximum average temperature change potential of 1.5 degrees C or less by 2100 with respect to pre-industrial levels. In the United States (US), the industrial sector accounts for 23% of total GHG emissions and is home to a number of hard-to-electrify activities. The chemicals subsector has the single largest subsector emissions profile after direct emissions from fossil fuel combustion and leakage from fossil fuel distribution systems. Within the chemicals subsector, many processes depend on hydrogen or ammonia precursors. Decarbonizing these two commodities would contribute significantly to decarbonizing the industrial sector as hydrogen could also be used for low carbon steel production (e.g., hydrogen-based direct reduction of iron) and other industrial applications. Emerging technologies require the application of prospective life cycle assessment (LCA), which can account for technology (foreground) scaling and process improvements via learning-by-doing, among others. In many cases, the future system context (background) in which the technologies are assumed to operate in is equally relevant. Background scenarios generated by integrated assessment models (IAM) can coherently incorporate potential future dynamics of the energy-climate-human-land system. Further, IAM scenarios are harmonized across socioeconomic and climate change mitigation pathways, which facilitates the comparability of prospective LCAs using different IAMs. We introduce an open source prospective LCA framework, the Life-cycle Assessment Integration into Scalable Open-source Numerical models (LiAISON), to analyze the non-linear relationships between technology foreground and the future energy system background across a series of midpoint and resource use metrics The integration of LCA and IAM data is achieved using prospective environmental Impact assessment (PREMISE). We showcase it by assessing two Power-to-Hydrogen (PtH2) processes, namely Solid Oxide Electrolysis (SOE) and Polymer Electrolyte Membrane Electrolysis (PEME). We compare the technologies to a baseline of hydrogen production via natural gas-based Steam Methane Reforming (SMR) in a US context of multiple energy system and climate change mitigation futures. Besides providing an analysis that specifies the LCA results ranges with temporal and geospatial explicitness across the two technologies, metrics, and impact assessment methods, this research also aims to establish a base framework that can be expanded to use other IAM generated scenarios and US open-source life cycle inventory (LCI) databases. We find that the temporal environmental performance of either technology or their difference to SMR is directly influenced by the underlying background dynamics. Under baseline projections (i.e., no decarbonization goals), neither process reaches parity with the incumbent technology across several environmental metrics. Under the decarbonization scenarios, the underlying sectoral shifts result in declining impacts over time, compared to 2020 levels, except for metal depletion levels, which increase. The background shifts postulate a heavily decarbonized economy and energy system, which help technologies reach parity with SMR between 2040-2050 (RCP2.6) and 2030-2040 (RCP1.9) for global warming. Despite declines across several other metrics over time, neither PtH2 technology break even with SMR by 2100 besides for global warming.

decarbonizing↗
DOE OSTI2018

Tritium permeability measurement in hydrogen-tritium system

Fusion power promises to provide electricity generation with outstanding safety and environmental performance. Safety plays a crucial role in fusion material selection since tritium behavior in materials determines two key safety evaluation source terms: in-vessel inventory source term and ex-vessel release term, which are used in reactor safety assessments for licensing fusion facilities. Tritium permeation through materials at elevated temperature during long pulse operation expected in a burning plasma environment will become a major safety issue and development of tritium permeation barrier materials is required to minimize ex-vessel release to the environment [1]. Extensive work on hydrogen and deuterium behavior in fusion materials has been conducted, but very little database is available for tritium, the radioactive fuel for future reactors, due to the cost and difficulty associated with handling tritium [2]. A majority of the experiments were carried out at high hydrogen/deuterium partial pressure (> 100 Pa), but the expected tritium partial pressure in the blanket system is at low tritium partial pressure (< 100 Pa), where the tritium behavior deviates from diffusion-limited to surface-limited permeation. At these pressures the behavior of tritium is suspected to deviate from diffusion-limited to surface-limited permeation or isotope-limited [3]. The use of tritium enables us to investigate hydrogen isotope behavior in low partial pressure range with beta radiation detection. The physics of two-component hydrogen (H+T) permeation through metal in the surface-limited regime and through liquid breeding material are still too poorly understood to confidently design environmentally benign and safe fusion systems. In this paper, we discuss development of tritium permeation system for low tritium partial pressure [4], and discuss the results from two-component hydrogen (H+T) permeation in 20 mm disc samples of pure alpha-iron at H2 partial pressure of 100 Pa, wide T2 partial pressure range from 10-12 to 10 Pa, and sample temperature up to 600 °C to uncover the basic scientific understanding of tritium permeation behavior at realistic blanket conditions for DEMO. [1] FESAC report on “Opportunities for Fusion Materials Science and Technology Research Now and During the ITER Era” (2012) DOE/SC-0149 [2] R.A. Causey, R.A. Karnesky, and C. San Marchi, “4.16 Tritium Barriers and Tritium Diffusion in Fusion Reactors”, Comprehensive Nuclear Materials (2012) Chapter 4.16. [3] P. Humrickhouse “Permeation of Tritium and Hydrogen in the Diffusion-Limited and Surface–Limited Regimes”, submitted to Journal of Nuclear Materials. [4] R.J. Pawelko, M. Shimada, K. Katayama, S. Fukada, P.W. Humrickhouse, and T. Terai, “Low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic”, Fusion Engineering and Design 102 (2016) 8

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Electrorefiner Speciation and Phase Model for Prediction of Operation Lifetime

The Mk IV electrorefiner (ER) at Idaho National Laboratory has recovered uranium metal from used nuclear fuels for 27 years. Thus, large quantities of fission product chlorides more electropositive than uranium have accumulated in the ER salt, increasing the complexity of phase equilibria and salt speciation. Importantly, the melt temperature of the salt increases with ER operation and will ultimately exceed design limits without intervention. Methods to remove fission products from molten chloride salt are known, but their implementation at scale is not yet demonstrated. In this work, historical ER composition and differential scanning calorimetry (DSC) are utilized to develop coupled speciation and chloride salt solution thermochemistry models, which enable accurate present-day analysis of ER salt as well as forecast its future disposition. The developed thermochemical models are applicable to chloride salts generally and have been made available in the Molten Salt Thermal Properties Database – Thermochemical.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project

The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Time-Resolved PIV/PTV Measurements on Interior Subchannels of a Wire-Wrapped 61-pin Hexagonal Fuel Bundle

Liquid metal fast reactor fuel bundles using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. Often, the fuel pins are helically wrapped with a wire spacer. The primary function of the wire spacers is to maintain a gap between adjacent fuel pins. They also enhance sub-channel mixing to increase convective heat transfer and mitigate vortex-induced vibration. Texas A and M University has conducted isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). These tests produced high spatial and temporal resolution measurements of the velocity field and pressure at different locations in the fuel bundle. The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques such as particle image velocimetry (PIV), particle tracking velocimetry (PTV), and laser Doppler velocimetry (LDV). The result of this experimental activity will be an experimental database of pressure and velocity field measurements that will be suitable for validating computational fluid dynamics (CFD) methods, such as Reynolds-averaged Navier-Stokes (RANS) and highly resolved large eddy simulation (LES). PIV results on the set of edge subchannels near the hexagonal duct have been presented in another paper. This paper presents PIV results on two sets of interior subchannels, defined as Interior- 1 and Center-2, at a Reynolds number (Re) of 19,000 in the turbulent regime.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Electrochemical and Thermodynamic Properties of U in LiCl-KCl-UCl{sub 3} salt system

Pyro-processing technology has been considered and examined by several countries as a promising method for the recovery and recycle of uranium and actinide elements from used nuclear fuel. One of the key processes in pyro-processing is an electro-refiner (ER) where the metallic fission products are loaded into an anode basket and dissolved into lithium-potassium chloride (LiCl-KCl) eutectic salt at 773 K. Many researchers have been exploring the fundamental characterizations of U in LiCl-KCl system by using various electrochemical methods, resulting in several reported values of diffusion coefficient, apparent standard potential, and activity coefficient of U in the salt system. In addition, several researchers measured kinetic parameter of U (exchange current density, i{sub 0}) by using Tafel and linear polarization (LP) methods. Despite available data sets on this ternary salt system, there are discrepancies due to different experimental conditions, design, size of the cell, and measurement techniques. In addition, the i{sub 0} for U/U{sup 3+} reaction has only been measured since 2009; therefore, only few data sets are available to date although the i{sub 0} is critical for the development of dynamic ER model. The value of i{sub 0} is a kinetic characteristic in an ideal equilibrium state; therefore, it is very challenging to be measured by experiments. The authors have estimated the values of i{sub 0} via Tafel and linear polarization methods (LP); however, these methods are lack of maintaining electrode surface area during the measurements. In addition, some data are not following the trends from other literature studies due to its difficulties of the measurement. Therefore, further experimental determinations have to be accomplished to understand the i{sub 0} kinetics and the database for the i{sub 0} of U/U{sup 3+} couple in LiCl-KCl eutectic salt needs to be established systematically. In the present study, an experimental setup was designed and used in a consistent manner to build a meaningful database for U in LiCl-KCl system. LiCl-KCl-UCl{sub 3} salt was prepared at different concentrations and temperatures ranging from 0.5 wt% to 4 wt% and 723 K to 798 K, respectively. Then, CV, open circuit potential (OCP), and electrochemical impedance spectroscopy (EIS) were performed to understand the electrochemical, thermodynamic, and kinetic properties of U. Specially, i{sub 0} values have not been measured via EIS method; therefore, this study provides a good chance to confirm the i{sub 0} values with the new technique.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2007

Analytical Approach for Estimating Preliminary Mass of ARES I Crew Launch Vehicle Upper Stage Structural Components

In January 2004, President Bush gave the National Aeronautics and Space Administration (NASA) a vision for Space Exploration by setting our sight on a bold new path to go back to the Moon, then to Mars and beyond. In response to this vision, NASA started the Constellation Program, which is a new exploration launch vehicle program. The primary mission for the Constellation Program is to carry out a series of human expeditions ranging from Low Earth Orbit to the surface of Mars and beyond for the purposes of conducting human exploration of space, as specified by the Vision for Space Exploration (VSE). The intent is that the information and technology developed by this program will provide the foundation for broader exploration activities as our operational experience grows. The ARES I Crew Launch Vehicle (CLV) has been designated as the launch vehicle that will be developed as a "first step" to facilitate the aforementioned human expeditions. The CLV Project is broken into four major elements: First Stage, Upper Stage Engine, Upper Stage (US), and the Crew Exploration Vehicle (CEV). NASA's Marshall Space Flight Center (MSFC) is responsible for the design of the CLV and has the prime responsibility to design the upper stage of the vehicle. The US is the second propulsive stage of the CLV and provides CEV insertion into low Earth orbit (LEO) after separation from the First Stage of the Crew Launch Vehicle. The fully integrated Upper Stage is a mix of modified existing heritage hardware (J-2X Engine) and new development (primary structure, subsystems, and avionics). The Upper Stage assembly is a structurally stabilized cylindrical structure, which is powered by a single J-2X engine which is developed as a separate Element of the CLV. The primary structure includes the load bearing liquid hydrogen (LH2) and liquid oxygen (LOX) propellant tanks, a Forward Skirt, the Intertank structure, the Aft Skirt and the Thrust Structure. A Systems Tunnel, which carries fluid and electrical power functions to other Elements of the CLV, is included as secondary structure. The MSFC has an overall responsibility for the integrated US element as well as structural design an thermal control of the fuel tanks, intertank, interstage, avionics, main propulsion system, Reaction Control System (RCS) for both the Upper Stage and the First Stage. MSFC's Spacecraft and Vehicle Department, Structural and Analysis Design Division is developing a set of predicted mass of these elements. This paper details the methodology, criterion and tools used for the preliminary mass predictions of the upper stage structural assembly components. In general, weight of the cylindrical barrel sections are estimated using the commercial code Hypersizer, whereas, weight of the domes are developed using classical solutions. HyperSizer is software that performs automated structural analysis and sizing optimization based on aerospace methods for strength, stability, and stiffness. Analysis methods range from closed form, traditional hand calculations repeated every day in industry to more advanced panel buckling algorithms. Margin-of-safety reporting for every potential failure provides the engineer with a powerful insight into the structural problem. Optimization capabilities include finding minimum weight panel or beam concepts, material selections, cross sectional dimensions, thicknesses, and lay-ups from a library of 40 different stiffened and sandwich designs and a database of composite, metallic, honeycomb, and foam materials. Multiple different concepts (orthogrid, isogrid, and skin stiffener) were run for multiple loading combinations of ascent design load with and with out tank pressure as well as proof pressure condition. Subsequently, selected optimized concept obtained from Hypersizer runs was translated into a computer aid design (CAD) model to account for the wall thickness tolerance, weld land etc for developing the most probable weight of the components. The flow diram summarizes the analysis steps used in developing these predicted mass.

Aggarwal, Pravin↗
DOE OSTI2019

The Impact of Coordination Environment on the Thermodynamic Stability of Uranium Oxides

Amorphous uranium oxides are acknowledged to arise via industrial processes relevant to the nuclear fuel cycle yet evade rigorous structural characterization. An encouraging approach is to develop statistical relationships between uranium–oxygen coordination environments and thermodynamic stability from which general statements about the likelihood of observing particular U–O arrangements can be made. The number of known crystalline uranium oxides is insufficient to build statistical relationships. We have developed a database of theoretical compounds using genetic algorithms with the density functional theory as a foundation to analyze coordination geometries in the uranium–oxygen phase space. We draw fundamental insights into the nature of uranium–oxygen interactions by correlating total energy with the coordination environment. The most stable configuration of U cations with O anions is in an environment with coordination numbers 5–8 in a cubic configuration. Higher and lower coordination numbers are observed only in metastable phases. General trends are observable in the relationships between the coordination number, density, and uranium fraction in each structure. The new insight into uranium coordination enabled by these analyses is foundational for further studies into the characteristic properties of individual uranium oxide materials and for elucidation of potential oxidation pathways for uranium metal.

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

Non-destructive Preirradiation Assessment of UN / U-Si “LANL1” ATF formulation

The goal of the Advanced Non-destructive Fuel Examination (ANDE) work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels, ultimately also to irradiated fuels. The results of these characterizations provide complete pre- and post-irradiation on length scales ranging from mm to nm, guide destructive examination, and inform modelling efforts. Besides technique development and application to samples to be irradiated, the ANDE work package also examines possible technologies to provide these characterization techniques pool-side, e.g. at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) using laser-driven intense pulsed neutron and gamma sources. Neutron tomography and neutron diffraction characterizations were performed on nine pellets; four UN/ U-Si composite formulations (two enrichment levels), three pure U 3 Si 5 reference formulations (two enrichment levels), and two reject pellets with visible flaws (to qualify the technique). The 235 U enrichments ranged from 0.2 to 8.8 wt. %. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U 3 Si 5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. We have also proposed a data format to build a database for characterization results of individual pellets. Neutron data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. This report provides the results for the characterized samples and discussion in the context of ANDE and APIE. We quantified the gamma spectra of several samples in their received state as well as after neutron irradiation to ensure that the neutron irradiation does not add significant activation that would complicate shipment and handling. We demonstrated synchrotron-based 3D X-ray microscopy on the composite fuel materials, providing unparalleled level of detail on the 3D microstructure. Furthermore, we initiated development of shielding containers allowing the characterizations presented herein while allowing handling of irradiated samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Results from the last DD and DT JET campaigns in the framework of the EUROfusion Tokamak Exploitation Work Package activity

JET, the only tokamak capable of operating with deuterium–tritium (D–T) fuel (since TFTR was shutdown in 1999), has provided essential experimental data to support ITER and DEMO design and operation. Within the EUROfusion Tokamak Exploitation Work Package, JET completed its final campaigns (2022–2023), culminating in the third D–T campaign (DTE3). These experiments addressed key challenges in plasma scenarios, exhaust control, and tritium management under reactor-relevant conditions. Significant progress was achieved in demonstrating ITER-like integrated scenarios with impurity seeding, achieving partial divertor detachment and high confinement ($H_{98}(y,2)$ ≈ 0.85) at 3 MA in D–T plasmas. Advanced exhaust regimes such as quasi-continuous exhaust (QCE) and X-point radiator (XPR) were successfully achieved first in D–D and then extended to D–T operation, confirming their relevance for mixed isotope operation. Operational milestones included a new world record of 69 MJ fusion energy in tritium-rich hybrid plasmas and long-pulse H-mode operation up to 60 s, contributing with unique data to the CICLOP database. Physics studies focused on peeling-limited pedestals in support of ITER and improved understanding of edge stability and impurity screening in metallic environments. Extensive usage of the shattered pellet injector (SPI) on JET provided critical information for the design of the ITER disruption mitigation system (DMS). Real-time control systems for D/T ratio control and plasma exhaust were deployed and demonstrated in D–D and D–T, while energetic particle physics investigations unfolded the role of fast ions in turbulence suppression mechanisms. Comprehensive tritium retention studies using gas balance method, post-mortem analysis, and ITER-relevant laser induced desorption spectroscopy (LIDS) diagnostics provided essential input for tritium accountancy strategies. These results are validating the ITER operational concepts, inform DEMO design, and deliver critical experience in nuclear operation and scenario integration.

disruptions↗