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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Determination of the hydrogen heat of transport in Zircaloy-4

During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. Here, the hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Milestone 1.2.12: The Fate of Oxygen Radical Species in Corroded Aluminum Alloys under Irradiation

The U.S. Department of Energy tasked Idaho National Laboratory to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in heliumbackfilled canisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen (H 2 ) generation from corrosion layers that would be present on the fuel assembly surfaces. However, limited attention has been directed to the oxygen radical species that are concurrently generated in these H 2 producing radiation environments. This report collates and summarizes the available experimental and computational studies on the fate of these reactive oxygen radicals during the irradiation of aluminum oxyhydroxide and hydroxide polymorphs typically formed on ASNF. Based on this review, it is determined that radiation-induced oxygen-centered radicals do not react to give molecular oxygen gas (O 2 ), but rather remain trapped in the aluminum corrosion layers in their original crystal lattice positions. Overall, O 2 generation will not be a concern for extended storage of ASNF in helium-backfilled canisters. That is, no O 2 contribution to pressurization of sealed dry storage systems or creation of a flammable atmosphere is anticipated.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

UZrCN Synthesis via Arc Melting

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO 2 ) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO 2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

UZrCN Synthesis via Arc Melting - A Novel Synthesis Study

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO2) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

36 MATERIALS SCIENCE↗
DOE OSTI2024

UZrCN Formation via Arc Melting – A Novel Synthesis Study

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO2) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: Uranium diboride and uranium carbide

Here, the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breach of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium diboride and uranium monocarbide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UB 2 , and UC. Presented here is the literature for UB 2 and UC degradation modes, thermodynamics, and oxidation performance of the pure compounds and reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation behavior for high uranium density fuels for use in LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Corrosion of novel reactive silver ink and commercial silver-based metallizations in diluted acetic acid

Silver-based metallizations in photovoltaic modules are susceptible to corrosion by acetic acid generated in ethylene vinyl acetate encapsulated modules, resulting in power losses over time. Here, three silver-based metallizations are exposed to diluted acetic acid, in concentrations representative of that found in field-exposed modules. Compositional, morphological, and structural changes of the metallizations are studied over 3000 h of exposure to diluted acetic acid using Raman spectroscopy mapping, X-ray diffraction, and scanning electron microscopy. The three metallizations studied are: 1) a commercial high-temperature fire-through Ag paste, commonly used for Si diffused junction solar cells; 2) a commercial low-temperature paste normally used for silicon heterojunction cells; and 3) a novel low-temperature reactive silver ink shown to be suitable for photovoltaic applications. We find distinct corrosion rates for the high-temperature silver paste and reactive silver ink in the presence of diluted acetic acid. However, the low-temperature silver paste appears to be more corrosion resistant, likely due to a polymer layer that protects the silver particles.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Generalized and Mechanistic PV Module Performance Prediction From Computer Vision and Machine Learning on Electroluminescence Images

Electroluminescence (EL) imaging of photovoltiac (PV) modules offers high-speed, high-resolution information about device performance, affording opportunities for greater insight and efficiency in module characterization across manufacturing, research and development, and power plant operations and management. Predicting module electrical properties from EL image features is a critical step toward these applications. In this article, we demonstrate quantification of both generalized and performance mechanism-specific EL image features, using pixel intensity-based and machine learning classification algorithms. From EL image features, we build predictive models for PV module power and series resistance, using time-series current-voltage (I-V) and EL data obtained stepwise on five brands of modules spanning three Si cell types through two accelerated exposures: damp heat (DH) (85°C/85% RH) and thermal cycling (TC) (IEC 61215). Overall, 195 pairs of EL images and I-V characteristics were analyzed, yielding 11700 individual PV cell images. A convolutional neural network was built to classify cells by the severity of busbar corrosion with high accuracy (95%). Generalized power predictive models estimated the maximum power of PV modules from EL images with high confidence and an adjusted-R 2 of 0.88, across all module brands and cell types in extended DH and TC exposures. Mechanistic degradation prediction was demonstrated by quantification of busbar corrosion in EL images of three module brands in DH, and subsequent modeling of series resistance using these mechanism-specific EL image features. For modules exhibiting busbar corrosion, we demonstrated series resistance predictive models with adjusted-R 2 of up to 0.73.

14 SOLAR ENERGY↗
DOE OSTI2021

Electroanalytical Sensors for Liquid Fueled Fluoride Molten Salt Reactor

ThorCon is developing a thermal thorium / uranium molten salt fueled reactor that uses a fuel salt consisting of NaF - BeF 2 - ThF 4 - UF 4 - UF 3 , which uses 19.75% enriched uranium. It is not a breeder reactor but requires regular additions of fissile material. On-site fuel processing is limited to adding make-up fuel, and adding beryllium to maintain redox balance in the fuel salt. The fluoride salt is an excellent flux that will strip the protective oxygen away from all metal surfaces it contacts. In order to protect the metal surfaces in the primary loop, no free fluorine is allowed. Maintaining a ratio of UF 4 to UF 3 provides a chemical buffer that can absorb any free fluorine released during the fission process before it can harm the structure. Periodic additions of metallic beryllium to the fuel salt will react with UF 4 to form BeF 2 and UF 3 , which will maintain the desired UF 4 /UF 3 ratio. A means to measure the UF 4 /UF 3 ratio is important to maintain the fuel salt in a non-corrosive state and could also provide additional insight into the corrosion behavior and chemical constitution of the system. A promising measurement technique is based on electrochemistry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY22 Status Update)

High-level purpose of this work: This report summarizes work carried out by Sandia National Laboratories (SNL) in the fiscal year 2022 (FY22) to evaluate the potential occurrence of stress corrosion cracking (SCC) on spent nuclear fuel (SNF) dry storage canisters. The U.S. currently lacks a repository for permanent disposal of SNF; thus, dry storage systems will be in use for much longer time periods than originally intended. Gap analyses by the US Department of Energy (DOE), the Nuclear Regulatory Commission (NRC), the Nuclear Waste Technical Review Board (NWTRB), and the Electric Power Research Institute (EPRI) have all determined that an improved understanding of the occurrence and risk of canister SCC is critical to demonstrating the safety of long-term dry storage. Should canister penetration by SCC occur, the containment boundary represented by the canister would be breached. A loss of the inert environment (helium) within the canister could occur and intrusion of air and moisture could react with and damage the fuel within the canister. For this reason, the DOE is funding an effort to evaluate the potential occurrence and consequences of dry storage canister SCC and to develop prevention, mitigation, and repair technologies for this degradation mechanism.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Degradation Mechanism Due to Water Ingress Effect on the Top Contact of Cu(In,Ga)Se2 Solar Cells

The impact of moisture ingress on the surface of copper indium gallium diselenide (CIGS) solar cells was studied. While industry-scale modules are encapsulated in specialized polymers and glass, over time, the glass can break and the encapsulant can degrade. During such conditions, water can potentially degrade the interior layers and decrease performance. The first layer the water will come in contact with is the transparent conductive oxide (TCO) layer. To simulate the impact of this moisture ingress, complete devices were immersed in deionized water. To identify the potential sources of degradation, a common window layer for CIGS devices—a bilayer of intrinsic zinc oxide (i-ZnO) and conductive indium tin oxide (ITO)—was deposited. The thin films were then analyzed both pre and post water soaking. To determine the extent of ingress, dynamic secondary ion mass spectroscopy (SIMS) was performed on completed devices to analyze impurity diffusion (predominantly sodium and potassium) in the devices. The results were compared to device measurements, and indicated a degradation of device efficiency (mostly fill factor, contrary to previous studies), potentially due to a modification of the alkali profile.

14 SOLAR ENERGY↗
NASA NTRS2014

A Cubesat Asteroid Mission: Propulsion Trade-offs

A conceptual design was performed for a 6-U cubesat for a technology demonstration to be launched on the NASA Space Launch System (SLS) test launch EM-1, to be launched into a free-return translunar trajectory. The mission purpose was to demonstrate use of electric propulsion systems on a small satellite platform. The candidate objective chosen was a mission to visit a Near-Earth asteroid. Both asteroid fly-by and asteroid rendezvous missions were analyzed. Propulsion systems analyzed included cold-gas thruster systems, Hall and ion thrusters, incorporating either Xenon or Iodine propellant, and an electrospray thruster. The mission takes advantage of the ability of the SLS launch to place it into an initial trajectory of C3=0.

corrosion↗
NASA NTRSDate not supplied

Hot hydrogen testing of carbides developed for nuclear thermal propulsion

Carbides are candidate materials for use in nuclear thermal propulsion, either as the primary material in nuclear fuel elements or as protective coatings. During operation, the fuel elements are exposed to flowing hydrogen gas at high temperature and it is important that fuel mass loss resulting from hydrogen corrosion be kept to a minimum. In this study, ZrC and NbC were fully consolidated by spark plasma sintering at 2100 °C, and relative densities greater than 98% were achieved. The samples were characterized by scanning electron microscopy and x-ray diffraction before and after hydrogen testing. The samples were exposed to hot flowing hydrogen gas at a pressure of 1 atm., at a temperature of 2500 K, and for durations up to 3 hours. It was found that the mass loss rate decreased with increasing exposure, before reaching a steady state value of about 1 mg/m2s. Oxygen in the samples increased the mass loss rate. Samples for which oxygen could be detected by x-ray diffraction had their mass loss rate increase by a factor of 10. Overall, the mass loss rates compare quite favorably to those previously reported.

M. P. Volz↗
DOE OSTI2024

Molten Salt Reactors: What about Materials? An Overview and Future Needs

ORNL Molten Salt Reactor Experiment (MSRE) – 60’s- early 70’s Developed INOR-8 (aka Hastelloy N); MSRE operated over the period Dec 1965 to 1969. (not continuous); Operated for >13,000 hr at T from ~550 to ~650°C using U, Li-7, Be and Zr fluoride salt fuel; Post-Irradiation Material Evaluation (corrosion + neutron irradiation) of surveillance specimens; Noted ”beneficial effect of Nb” in alloy (on intergranular Te embrittlement) More recent (post-1980) Materials R&D for Gen IV …. Alloys 316L, 316H, 304, Alloy 800H, Alloy 617 (!!), Hastelloy N (mod), plus numerous experimental alloys developed

36 - MATERIALS SCIENCE↗
DOE OSTI2020

Preliminary Deposition Modeling: For Determining the Deposition of Corrosive Contaminants on SNF Canisters

This report fulfills the M3 milestone M3SF-20PN010207045. During fiscal year (FY) 2019, Pacific Northwest National Laboratory (PNNL) worked to develop initial deposition and particle tracking models which could be used to better understand how corrosive contaminants deposit on the surfaces of spent nuclear fuel (SNF) canisters during dry storage. These models may be useful in understanding the likelihood of chloride induced stress corrosion cracking (CISCC) on SNF canisters

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1978

EPRI-NASA Cooperative Project on Stress Corrosion Cracking of Zircaloys

Examinations of the inside surface of irradiated fuel cladding from two reactors show the Zircaloy cladding is exposed to a number of aggressive substances, among them iodine, cadmium, and iron-contaminated cesium. Iodine-induced stress corrosion cracking (SCC) of well characterized samples of Zircaloy sheet and tubing was studied. Results indicate that a threshold stress must be exceeded for iodine SCC to occur. The existence of a threshold stress indicates that crack formation probably is the key step in iodine SCC. Investigation of the crack formation process showed that the cracks responsible for SCC failure nucleated at locations in the metal surface that contained higher than average concentrations of alloying elements and impurities. A four-stage model of iodine SCC is proposed based on the experimental results and the relevance of the observations to pellet cladding interaction failures is discussed.

Cubicciotti, D.↗