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

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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Exploring the influence of transition metals on f-element bonding

The United States aims to triple its nuclear energy production by 2050, which will result in increased uranium usage and spent nuclear fuel generation. This highlights the need for a comprehensive understanding of actinide coordination chemistry, which is crucial for the extraction, processing, purification, and fabrication of uranium-based fuels. Moreover, it is vital for the reprocessing or safe disposal of nuclear waste and effective remediation efforts. Achieving this successfully requires an in-depth understanding of f-electron behavior, as “the role of 5f electrons in bond formation remains a fundamental topic in actinide chemistry”. Introducing a second metal into the system can increase structural dimensionality and diversify structural architecture. Heterometallic systems can also alter material properties, such as magnetic and spectroscopic characteristics, luminescence, and actinide mobility. Additionally, secondary transition metals, even when present only in the second coordination sphere and not directly coordinated, can influence the electron density at the actinyl metal center. In this study, uranium heterometallic single crystals were synthesized by incorporating transition metals such as iron(III), iron(II), nickel(II), manganese(II), copper(I), and cobalt(II). The crystals were formed using 2,6-pyridine dicarboxylic acid and other structurally similar ligands with varying functional groups. The synthesized crystals were analyzed using an extensive array of analytical and computational characterization techniques, including single crystal X-ray diffraction, Raman and infrared spectroscopy, and density functional theory calculations.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2024

Silicon Carbide Composite Cladding Qualification

This report examines the process to qualify a new fuel-cladding type focusing on the data requirements and types of irradiation testing necessary to meet the regulatory burden and related guidance applicable to fuel qualification of accident-tolerant fuel. The report considers the use of accelerated fuel qualification techniques and lead-test specimen programs that may shorten the timeline for qualifying fuel for use in a nuclear reactor at the desired burnup. The assessment framework particularly emphasizes the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of evaluation models in the fuel qualification process, and the assessment of the experimental data used to develop and validate the empirical models and empirical safety criteria.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effect of Fluoride Anions on Nd(III) Electrode Processes and Nd Metal Recovery in LiCl–KCl–NdCl 3

Here, this work investigated the impact of fluoride anions on Nd metal recovery in LiCl–KCl–NdCl 3 electrolytes by introducing LiF at a molar F/Nd ratio of 9 at 773 K. The voltammetric measurements confirmed that the fluoride ions facilitated single-step Nd(III)/Nd reduction by promoting the stability of the Nd(III) state compared to two-step reduction (Nd(III)/Nd(II) and Nd(II)/Nd) in all-chloride electrolytes. The stabilized Nd(III) state in the LiF-containing electrolyte effectively suppressed partial Nd(III)/Nd(II) reduction and the back-dissolution of Nd metal via comproportionation. By suppressing these reactions, high round-trip Coulombic efficiencies of 77–86% were achieved in the LiF-containing electrolyte during selective Nd deposition–removal cycles compared to 53–59% in all-chloride electrolytes. The Faradaic yield for Nd metal recovery was estimated at 38.6% in the LiF-containing electrolyte following long-term electrolysis, more than three times higher than 11.3% in all-chloride electrolytes. These results confirm the beneficial effect of fluoride ions on Nd metal recovery in low-melting chloride-based electrolytes, promising efficient Nd recovery at low temperatures.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

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

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

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Metal-oxide cages open up strategy for processing nuclear waste

We report the ability to separate the radioactive element americium from spent nuclear fuel would lower the long-term hazards of nuclear waste. An inorganic molecular cage that selectively binds to americium opens up a separation strategy.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Gas-phase stability of large lanthanide:diglycolamide clusters evaluated using collision-induced dissociation

Efficient utilization of long-term deep geologic storage repositories is critical for the large-scale deployment of nuclear energy. As the minor actinides (americium and curium) are the largest contributor to the decay heat of used nuclear fuel after a few hundred years, their removal from used fuel prior to disposal can significantly increase the amount of waste that can be stored in a given volume. The development of ligands that can effectively separate the minor actinides from other components of used nuclear fuel are crucial to efficient utilization of geological storage repositories. N,N,N’,N’-tetraoctyl diglycolamide (TODGA) is a promising extractant for the separation of lanthanides and the minor actinides from other components of used nuclear fuel. While the use of TODGA in f-element separations has been investigated in process-based formulations, gas-phase metal ion cluster experiments enable the study of covalent interactions in reprocessing systems absent from solvent effects. Exploring fundamental differences in lanthanide-ligand covalent interactions can impact the development and implementation of actinide-lanthanide separation systems in nuclear fuels reprocessing. In this study, lanthanide-TODGA clusters were synthesized in the gas-phase and identified using mass spectrometry fragmentation experiments. Large europium and samarium clusters were identified that contained up to 8 and 10 bound TODGA ligands, respectively; this was surprising due to the size and multidentate binding that is normal for TODGA. In addition, while smaller clusters showed evidence of sequential ligand fragmentation, larger clusters displayed the loss of neutral TODGA with applied collision voltage. Interestingly, the voltage required for this removal decreased as more TODGA ligands were bound to the metal, suggesting that the metal coordination sphere was becoming more saturated and TODGA ligands were more weakly bound as the clusters got larger.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Code of Record: DOE Standard Canister (DOESC)

The United States Department of Energy (DOE) Spent Nuclear Fuel (SNF) Packaging Demonstration seeks to develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE SNF for road-ready dry storage (RRDS); and establish the processes that will be used in a future production facility. The Packaging Demonstration will utilize the DOE Standard Canister (DOESC) for packaging select DOE-managed SNF types for interim storage, transportation and disposal as part of RRDS. Placing DOE-managed SNF into RRDS is part of the strategic framework for SNF on the Idaho National Laboratory (INL) site. To comply with DOE, INL and Nuclear Regulatory Commission (NRC) requirements, this Code of Record establishes the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 3 as the Code that will govern DOESC and internal support structure constructioni, with certain clarifications. This Code of Record establishes a rationale for proceeding without construction certification of the DOESC (i.e., “N-stamping) and the extent to which a Registered Professional Engineer is required for the DOE Spent Fuel Packaging Demonstration. Given the (i) standard industry practice to pursue independent licensure of commercial storage casks and transportation packages by the NRC in lieu of ASME certification and (ii) guidance provided by 10 CFR 830, DOESC construction activities need not be certified (i.e., “stamped”) to the ASME BPVC. However, additional quality assurance requirements will apply as outlined in this Code of Record.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

NDMAS: Data storage, visualization, analysis, delivery, and more

Slides representing Nuclear Data Management and Analysis System (NDMAS) Data for properly preserving our publicly funded data, including DOE Public Access Plan and the users of NDMAS. From Capture, Storage, Delivery & Analysis, Archival, and Qualification of Fuel fabrication, PIE, ART Experiment monitoring and operations, and processes for recording work and for looking ahead.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Increased Peak Temperatures for Spent Fuel Cladding Performance during Dry Storage (Rev.1)

NUREG-2214, “Managing Aging Processes in Storage (MAPS) Report,” (2019) documents the technical basis for age-related degradation mechanisms on spent nuclear fuel assemblies during dry storage operations. The technical bases have been reviewed, and hoop cladding stress calculations were conducted using the Fuel Analysis under Steady-state and Transients (FAST) code at elevated normal condition temperatures of 425°C and 450°C, and at an off-normal and accident conditions temperature of 600°C. The NUREG-2214 conclusions are substantiated by the referenced literature. The calculations demonstrate that the aging mechanisms primarily driven by hoop stress such as hydride reorientation, delayed hydride cracking, thermal and athermal (low-temperature) creep, and localized mechanical overload are either not credible or do not compromise the fuel assembly’s performance for increased peak normal condition temperatures of 425°C or 450°C during the up to 60-year dry storage period. Other aging mechanisms considered are also not credible for compromising the fuel assembly’s performance during 60-year dry storage with a peak temperature of 400ºC as specified in ISG – 11 Rev. 3, and the same is indicated for the increased peak normal condition temperatures of 425°C or 450°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

SR19039 - All Metal Vacuum Scroll Pump Development for Tritium

Currently the Savannah River Site tritium gas processing mission relies on a pumping train combination of a Normetex scroll pump backed by a Metal Bellows MB-601 diaphragm pump. Both of these pumps have wetted components that are "all-metal" so that no polymers or elastomers that can degrade due to tritium gas exposure are exposed to the process fluid. Normetex, the original all-metal vacuum scroll pump manufacturer, has gone out of business. Another French based company, Eumeca, has "spun off" from the Normetex company and is now producing the original, 9 cfm all metal Normetex vacuum scroll pump. It should be noted that Normetex, now Eumeca, is currently the only supplier of a functional all metal vacuum scroll pump to meet the SRS tritium gas processing requirement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

DOE SNF Packaging Demonstration Project: Worktable and Canister Loading Sleeve Design [Slides]

This presentation is a briefing for summer interns and others outside the project, to give them a summary of ongoing design activities. The DOE SNF Packaging Demonstration Project is an effort to develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed spent nuclear fuel (SNF) for road ready dry storage (RRDS). The project is to remotely load and seal two types of DOE-managed SNF from the CPP-603 facility into DOE Standard Canisters, place the sealed DOE Standard Canisters into commercial vendor over-canisters (equivalent to commercial MPCs), seal and place the over-canisters into commercial vendor storage overpacks, and place the storage system (over-canister and storage overpack) onto the CPP-2707 cask pad.

42 ENGINEERING↗
DOE OSTI2020

Advanced Test Reactor Experiment Safety Analysis

This presentation contains general information about the Advanced Test Reactor experiment safety analysis program. Accidents and dose consequence are discussed in general terms. Pictures of ATR facilities and experiments are included. In cases where experiments are shown, identification numbers have been removed. Department of Energy limits on dose consequence are provided. These limits are found in many non-CUI documents and a derived from publicly available sources. A summary of the presentation content is as follows: -Define Experiment Safety Analysis -Discuss what we are protecting and the applicable limits -Show how safety analysis fits into the experiment process -Explore the scope of experiment activities and associated accidents -Present how we categorize accidents and establish accident frequencies -Identify how controls are implemented and process improvements

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

In-Can Incineration and Vitrification Process for the Treatment of Transuranic Wastes - 20035

Technological alpha contaminated wastes are produced by the nuclear fuel cycle management industry, essentially from MOX fuel production units. These Plutonium Contaminated Materials (PCM) contain a mixture of organic, metal and mineral materials, contaminated with uranium, plutonium and americium. One treatment scenario under study is to incinerate the organic materials, melt the metallic fraction and incorporate the mineral fraction (including ashes from organics incineration) into a glass matrix without sorting or pretreatment (grinding for instance). A scale 1 pilot of this PIVIC process (Process for Incineration and Vitrification In Can) has been commissioned in 2018 at the CEA R and D vitrification facility in Marcoule site. A R and D program aiming to demonstrate the feasibility of this process is currently on going. This paper starts with a description of PIVIC process with explanation of the technologies involved. Then, main results of the first year of operation with the technological demonstration goal are presented. (authors)

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

Finding ways to reduce nuclear waste: searching for the unknown one step at a time

In my home country, Venezuela, research has been stagnant. Due to the political turmoil and the crisis, many educated people have left the country in search for a better life. This has caused a deficit in any technological and scientific advances, making Venezuela one of the first South American countries to have its rate of publications decline by 29% in 2013. Currently, Venezuela lacks the infrastructure and the means to keep up with the research progress as compared to other countries in South America, such as Brazil. Since coming to the United States (US), and currently working for a national laboratory, the active research environment endorses a wide range of careers and engineering programs that allow researchers to thrive at any given field. Researchers have access to funds and tools to succeed in developing materials for the future. There are 17 national laboratories in the US, and all of these have a different research focus/objective. As examples, Los Alamos National Laboratory and Sandia National Laboratory focus is on national homeland security, weapon science, radiation effects, among others. Argonne National Laboratory focuses on nuclear energy, energy storage, high performance computing, etc. At Idaho National Laboratory (INL) the research focuses on innovating nuclear energy and clean energy resources, critical infrastructure materials, along with fuel cycle solutions to manage, dispose and find ways to recycle current and future radiological waste. Compared to other national laboratories, INL focuses slightly more on applied processes and how nuclear energy can be innovated to next reactor design and technologies. The research being conducted at INL made me apply for a Seaborg distinguished postdoctoral position. For the position itself, the researcher must submit a proposal related to actinide chemistry on a research field area. In this position, 50% of my time will be focused on my own proposal. The proposal that I am working on is focused on the innovation of nuclear energy and fuel cycle recycling, which is why I was mainly interested on working at this national laboratory. To give a bit more context of what my proposal is about, a little bit of background is necessary: After the nuclear fuel (UO2) is used in a reactor, the fuel matrix is then characterized by various fission products (FP). Among these FP (including rare earth elements, alkali/alkaline earths, and actinides), many can potentially be recovered through nuclear reprocessing technologies. In pyroprocessing, the used nuclear fuel undergoes electrochemical dissolution into a molten chloride salt mixture in an electrorefiner. Initially, uranium is reduced onto an inert cathode by applied potentials. However, numerous remaining FPs accumulate in the melt and pose challenges for recovery by an inert electrode, particularly the rare earth elements (e.g., Nd, Gd, Pr, Sm) due to their multivalent oxidation states and tendencies toward side reactions, leading to their dissolution in the electrolyte. These recovery challenges result in inefficiencies and necessitate the continual discarding of the molten chloride salt, thereby generating additional waste. Furthermore, the presence of rare earth elements and other fission products in the molten salt electrolyte alters its physical and chemical properties, affecting both uranium recovery efficiency and the longevity of the molten chloride salt. To improve the recovery efficiency of the FP, specifically rare earth elements, I am investigating the fundamental interactions between rare earth elements in the molten chloride salt and their metallic form. The kinetic pathways and the chemical reactions of these elements will give insights on how the recovery efficiency can be improved. The interactions and speciation of these elements are being studied by spectro-electrochemistry at high temperature environments in quartz and other ceramic materials (e.g., alumina crucibles). Some of the challenges I am facing specifically relates the reactivity of some of these elements with different glass and crucible materials. Although my research focuses on fundamental science, it will benefit the applied process by generating new scientific knowledge and closing the gap for an efficient recycling of the waste: one step at a time.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Identifying microstructures susceptible to pulverization in commercially irradiated high burnup UO2 under LOCA conditions

High burnup fuel fragmentation (HBFF) has been a concern in the nuclear industry for many years. When UO2 reaches and exceeds pellet average burnups around 55 GWd/tU, microstructural changes in the fuel pellet begin to occur that render the pellet susceptible to fine fragmentation under loss-of-coolant accident (LOCA) conditions. During a LOCA event, the cladding balloons and bursts, potentially releasing part of these fine fuel fragments into the reactor pressure vessel. This process is known as fuel fragmentation, relocation, and dispersal (FFRD). Presently, the US nuclear industry is developing a safety basis for FFRD with the goal of increasing burnups and pressurized water reactor cycle lengths. Increased cycle lengths would push the burnup limits of the fuel rods past the known threshold for HBFF susceptibility. In this work, the microstructural impact on HBFF was investigated by comparing the microstructures of as-irradiated rod segments to their post-LOCA tested counterparts. This investigation found that varied operational histories result in different microstructural evolutions across the pellet radius, which impacts the fragmentation behavior and radial location of the fragmentation. It is noted that regions with a high bubble density and a high density of grain boundaries fragment under LOCA-relevant conditions. In addition to the fragmentation that has been previously reported at the periphery of the fuel, fragmentation was also noted in the dark zone toward the fuel center. Analysis of the power histories of the fuel samples suggests that the dark zone forms in the central regions of the pellet between temperatures of approximately 740 and 960 °C.

McKinney, Casey [ORNL] (ORCID:0000000335383614)↗
DOE OSTI2021

Update on the Simulation of Commercial Drying of Spent Nuclear Fuel

The purpose of this report is to document improvements in the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multiscale Evaluation of Acetohydroxamic Acid (AHA) Radiolysis Under Used Nuclear Fuel Reprocessing Solvent System Conditions

Acetohydroxamic acid (AHA) has been proposed as an alternative agent for the selective separation of plutonium and neptunium from co-extracted uranium during the reprocessing of used nuclear fuel. However, the fundamental radiolytic behavior of this molecule under envisioned process conditions – i.e., acidic biphasic solvent systems – is not sufficiently understood to support process applications. Here we present a systematic irradiation study (steady-state gamma and time-resolved pulsed electron) into the radiolytic integrity of AHA and formation of degradation products in aqueous nitric acid (HNO3) solutions (0.2 M) in presence and absence of an organic phase, comprising current (tri-butyl phosphate - TBP) and future (N,N-di-(2-ethylhexyl)butyramide - DEHBA and di-2-ethylhexylisobutyramide - DEHiBA) reprocessing ligands dissolved in n-dodecane diluent. Experimental data are complimented by predictive multiscale model calculations for the elucidation of underpinning mechanisms.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Nuclear Material Accounting and Control (NMAC) for a Commercial Fuel Debundling Facility in the United States

The concept of fuel debundling proposes that all non-fuel components of a fuel assembly be removed prior to beginning the recycling process. It was introduced to reduce the quantity of and cost of disposing high level radioactive waste during recycling in the United States. However, nuclear material accountancy and control programs could also benefit from fuel debundling due to a reduction in self-shielding material for individual fuel rods compared to a fuel assembly. Current NRC regulations and guidance point to an independent fuel debundling facility with a material-possession-only license as the best option for implementing this concept into a used fuel recycling scheme in the United States. This allows the facility to comply only with simple NMAC requirements under Subpart B of 10 CFR Part 74 without the need to submit an NMAC plan to the NRC, exempting it from more tedious and complex NMAC requirements. However, a recycling facility may desire a Pu mass estimate from measurements of each individual fuel rod to improve accountancy into the dissolver. Many technologies and methodologies were considered for performing NMAC in an independent fuel debundling facility in the United States. Few were able to meet anticipated throughput demands of up to 3,300 rods/day and achieve measurement uncertainties suitable for input accountancy for a recycling facility. The best option was a combination of passive gamma-ray spectroscopy and neutron coincidence counting to meet all requirements for the proposed Fuel Debundling Detector (FUDD). Follow-on work will include a full uncertainty analysis to inform what assumptions can be made to meet input accountancy objectives before beginning experimental activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗