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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 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In Situ High-Temperature Synchrotron Diffraction Studies of (Fe,Cr,Al) 3 O 4 Spinels

The modeling of a loss-of-coolant-accident scenario involving nuclear fuels with FeCrAl cladding materials in consideration to replace a Zircaloy requires knowledge of the thermodynamics of oxidized structures. At temperatures higher than 1500 degrees C, oxidation of FeCrAl alloys forms (Fe,Cr,Al) 3 O 4 spinels. In situ high-energy X-ray diffraction in a conical nozzle levitator installed at beamline 6-ID-D of the APS was used to study the structural evolution of the oxides as a function of the temperature. Single-phase (spinel) and multiphase (spinel-corundum-FeAlO 3 ) regions are mapped as a function of the temperature for three different compositions of FeCrAl oxidation products. The thermal expansion coefficients and cation distribution in the spinel structure have been refined. The temperature at which complete melting of the fuel cladding is expected has been determined by the liquidus temperatures of the oxidized products to be between 1657 and 1834 degrees C in a 20% O 2 /Ar atmosphere using the cooling trace method. The liquidus temperature increases with increasing Al and Cr content in the spinel phase.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Thermally-Conductive and Mechanically-Robust Graphene Nanoplatelet Reinforced UO 2 Composite Nuclear Fuels

Low thermal transport behavior along the radial direction of nuclear fuel pellets and pellet-cladding mechanical interaction significantly impact fuel performance and the safety of current nuclear energy systems. Here we report a new strategy of advanced fuel design in which highly thermally-conductive and mechanically-robust graphene nanoplatelets are incorporated into UO 2 fuel matrix to improve fuel thermal-mechanical properties. The 2D geometry of the graphene nanoplatelets enables a unique lamellar structure upon fuel consolidation by spark plasma sintering. The thermal conductivity along the radial direction of the sintered fuel pellets at room temperature reaches 12.7 and 19.1 wm –1 K –1 at 1 wt.% and 5 wt.% loadings of the graphene nanoplatelets, respectively, representing at least 74% and 162% enhancements as compared to pure UO 2 fuel pellets. Indentation testing suggests great capability of the 2D graphene nanoplatelets to deflect and pin crack propagation, drastically improving the crack propagation resistance of fuel matrix. Here, the estimated indentation fracture toughness reaches 3.5 MPa·m 1/2 by 1 wt.% loading of graphene nano-platelets, representing a 150% improvement over 1.4 MPa·m 1/2 for pure UO 2 fuel pellets. Isothermal annealing of the composite fuel indicates that the graphene nano-platelet is able to retain its structure and properties against reaction with UO 2 matrix up to 1150 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

MoSi 2 Oxidation in 670-1498 K Water Vapor

Molybdenum disilicide (MoSi 2 ) has well documented oxidation resistance at high temperature (T > 1273 K) in dry O 2 containing atmospheres due to the formation of a passive SiO 2 surface layer. But, its behavior under atmospheres where water vapor is the dominant species has received far less attention. Oxidation testing of MoSi 2 was performed at temperatures ranging from 670–1498 K in both 75% water vapor and synthetic air (Ar-O2, 80%–20%) containing atmospheres. Here the thermogravimetric and microscopy data describing these phenomena are presented. Over the temperature range investigated, MoSi 2 displays more mass gain in water vapor than in air. The oxidation kinetics observed in water vapor differ from that of the air samples. Two volatile oxides, MoO 2 (OH) 2 and Si(OH) 4 , are thought to be the species responsible for the varied kinetics, at 670–877 K and at 1498 K, respectively. Finally, we observed an increase in oxidation (140–300 mg/cm 2 ) from 980–1084 K in water vapor, where passivation is observed in air.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The Impact of Interface Bonding Efficiency on High Burn-up Spent Fuel Vibration Integrity during Normal Transportation

Finite element analysis (FEA) was used to investigate the impacts of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces on spent nuclear fuel (SNF) vibration integrity. The FEA simulation results were validated and benchmarked with ORNL reversal bending fatigue test results on surrogate rods consisting of stainless steel (SS) tubes with alumina-pellet inserts.

interface bonding↗
DOE OSTI2017

Synergies Between ' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

Candidate cladding materials for advanced nuclear power reactors including fast reactor designs require materials capable of withstanding high dose neutron irradiation at elevated temperatures. One candidate material, HT-9, through various research programs have demonstrated the ability to withstand significant swelling and other radiation-induced degradation mechanisms in the high dose regime (>50 displacements per atom, dpa) at elevated temperatures (>300 C). Here, high efficiency multi-dimensional scanning transmission electron microscopy (STEM) acquisition with the aid of a three-dimensional (3D) reconstruction and modeling technique is used to probe the microstructural features that contribute to the exceptional swelling resistance of HT-9. In particular, the synergies between ' and fine-scale and moderate-scale cavity formation is investigated.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Design of Laboratory System to Recover Tritium from Zirconium Recycle and Plan to Test System

The objective of the tritium separations task is to develop a viable process to separate tritium from the zirconium recycle off-gas stream. In support of this long-term objective, the goal of the present work is to design a test system and prepare a test plan for an alternate tritium capture method suitable for, and compatible with, recycle of the reactive gas (i.e., chlorine) in the zirconium recycle process. The zirconium recycle process is under development to recover zirconium from the Zircaloy® cladding component of used nuclear fuel. Small quantities of all isotopes of hydrogen (all H) are trapped in the Zircaloy as zirconium hydride. Processing releases and converts this hydrogen to hydrogen chloride, including tritium-substitu ted hydrogen chloride, 3 HCl (a.k.a., TCl). This compound of tritium must be removed from the zirconium recycle process off-gas and sequestered to satisfy regulations governing emissions of tritium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Measurement of radiolytic hydrogen generation from surrogate zr-based materials in a "mini-canister"

This report describes a test campaign to investigate radiolytic hydrogen generation from surrogate materials resembling the (Zr-based) cladding of commercial spent nuclear fuel (SNF) with an inventory of residual water post-dryout. The radiolytic hydrogen generation rates from residual waters in the SNF-in-canister system are important for predicting the in-canister environment for sealed SNF dry storage canisters over time and the impact to continued safe dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Radiolytic H 2 Yield from Non-Oxidized Zr Surrogates Under Humid Helium Backfill

This report presents updates from a test campaign to investigate radiolytic hydrogen generation from surrogate materials resembling the (Zr-based) cladding of commercial spent nuclear fuel (SNF) with an inventory of residual water post-dryout. The radiolytic hydrogen generation rates from residual waters in the SNF-in-canister system are important for predicting the in-canister environment for sealed SNF dry storage canisters over time and its impact on continued safe dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Recovery of Zr from ZrCl 4 by Electrolysis in Fused Salts and Ionic Liquids: Literature Review and Test Plan

This report reviewed the current status on the research and development of Zr electrodeposition in high temperature molten salts and room temperature ionic liquids from ZrCl 4 , which is the product of the chlorination process developed for recycling the Zr claddings of the spent nuclear fuels. Due to the pyrophoricity issue, the Zr metal product from the electrodeposition process is desired to be of coarse, dense crystalline or plates (instead of powder or loose, fine crystals) that are not to be easily oxidized. The findings based on the literature review are included in the report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1982

Overview of high-temperature materials for high-energy space power systems

The current state of technology and some of the more pressing research needs and challenges associated with the possible use of high temperature materials in future high energy space power systems are discussed. Particularly, emphasis is on the need to improve and quantify the fundamental understanding of the effects of the following: (1) fast neutron radiation on the properties and behavior of nuclear reactor fuels and claddings; and (2) long term, high temperature, space (vacuum) exposure on the properties of refractory metals considered for use as structural materials in various power conversion systems.

Saunders, N. T.↗
DOE OSTI2019

Resistance pressure weld for nuclear reactor fuel rod tube end plug

A fuel rod for a nuclear reactor, including a cladding tube having a first end with an annular end face, a second end with an annular end face, and a cylindrical body portion extending therebetween, and a first tube end plug including a front portion, an annular lip with an annular end face, and a substantially straight cylindrical body portion extending therebetween, wherein the surface area of the annular end face of the first end of the cladding tube and the annular end face of the annular lip of the first tube end plug are substantially equal, and the annular end face of the first end of the cladding tube and the annular end face of the annular lip of the first tube end plug are connected by a resistance pressure weld.

Barger, Earl Brian↗
DOE OSTI2019

Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces

To understand the performance of carbides under irradiation, the behavior of SiC and ZrC as well as water slurries of these carbides have been systematically studied using gamma-rays and 5.5 MeV He ions. Radiolytic damage to the carbides was determined using a range of analytical techniques including thermogravimetric analysis (TGA), elemental analysis, inductively-coupled plasma optical emission spectroscopy (ICP-OES) and scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDS).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗