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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

A new non-diffusional gas bubble production route in used nuclear fuel: Implications for fission gas release, cladding corrosion, and next generation fuel design

The relationship between noble metal phase particles and fission gas bubble production in used nuclear fuel is described. The majority of Te atoms within noble metal phase undergo radioactive decay to form stable Xe within a few hours after particle formation. This results in the production of clusters of Xe atoms contained within the solid metal matrix exhibiting an equivalent gas bubble pressure approaching 1 GPa. These high pressure bubbles are stabilized by the UO2 within the bulk of the fuel. However, when these bubbles form near the fuel / cladding interface, in combination with local and temporal damage caused by fission recoil, they are capable of overcoming the fracture strength of the UO2 and rupturing catastrophically. The force of the resulting bubble rupture is sufficient to eject noble metal phase particles several microns into the cladding. This proposed mechanism explains the observance of noble metal phase in cladding and is consistent with a host of morphological features found near the fuel / cladding interface.

Schwantes, Jon M.↗
DOE OSTI2021

Produce ZrH Moderator Material

This reports complete Milestone M2.2.3 Produce zirconium hydride moderator material. In this milestone, we report the successful fabrication of FeCrAl alloys cladded zirconium hydride moderator. Different types of crucible designs were presented, including the basic design with bar cladding material, basic design with tube cladding material, and crucible design with position holding function. Mo and FeCrAl were used as the trial cladding materials. The welding techniques and procedures were reported and discussed. Totally three welding techniques were adopted in the development of moderator cladding, and they are electron beam welding (EBW), laser welding (LW), and gas tungsten arc welding (GTAW). After welding, all cladded crucibles were evaluated with two kinds of leak testing, the helium leak test for minor leakage and the bubble test for major leakage. Due to the concern of the poor neutronics performance, Mo was not ideal for future moderator cladding application. Therefore, detailed characterization of the Mo cladded zirconium hydride was not pursued. Instead, the characterization of FeCrAl cladded zirconium hydride was performed. We discussed two material conditions (i.e., as machined and pre-oxidized conditions) prior to the cladding process, evaluated thermal stability of cladded zirconium hydride moderator through directly measuring hydrogen release, and characterized the zirconium hydride following the thermal desorption measurement. The results showed pre-oxidized FeCrAl is capable of efficiently preventing hydrogen release from the moderator assembly.

36 MATERIALS SCIENCE↗
DOE OSTI2026

High-Temperature SiC Cladding End Plug Irradiation Design and HFIR Readiness

This report documents the design of a High Flux Isotope Reactor (HFIR) irradiation experiment intended to evaluate irradiation effects on the hermeticity of silicon carbide (SiC) end plug specimens under a radial fast neutron flux gradient at representative light-water reactor (LWR) temperatures of approximately 300 °C. The overarching goal of this work is to statistically evaluate SiC end plug hermeticity and mechanical properties following irradiation using the high-throughput irradiation capability discussed here. Each specimen consists of a short section of SiC fiber–reinforced SiC (SiC/SiC) tube with a single monolithic SiC end plug joined to one end. The experiment allows up to 66 specimens to be irradiated in six different stacks within a dry, sealed irradiation capsule derived from the previously developed high-temperature SiC/SiC cladding bowing experiment. The neutronics basis, thermal analysis, and HFIR readiness of the experiment are discussed in this report for two possible design cases. The first design case is based on existing approval documentation and components that are on hand and approved for use, so the experiment insertion would require only specimen receipt, specimen pre-irradiation characterization, experiment assembly, and final fabrication package approval. The second design case provides improved thermal robustness and the preferred end plug geometry but requires fabrication of a modified holder and revisions to the HFIR approval documentation, in addition to the other activities required for the first design case, before insertion.

Hott, Daniel [Oak Ridge National Laboratory (ORNL)↗
DOE OSTI2025

Improved damage tolerance of SiC-based nuclear fuel cladding with novel multi-layered SiC coating design at 1200 °C

Continuous SiC fibre reinforced SiC matrix composites (SiC f -SiC m ) with monolithic SiC outer coatings are considered as a damage-tolerant cladding design for loss of coolant accident (LOCA) conditions in light water reactors. However, monolithic SiC coatings are brittle and prone to catastrophic failure. In this study, a SiC f -SiC m cladding with a novel multi-layer SiC outer coating (11 sub-layers, ∼260 µm in total thickness) was investigated under C-ring compression at room temperature and 1200 °C in argon environment. Real-time synchrotron X-ray computed tomography (XCT) was employed to capture crack initiation and propagation processes. Compared to conventional monolithic SiC outer coatings, the multi-layer coating structure facilitated crack deflection and bifurcation enhancing its damage tolerance at both temperatures. Despite pre-existing surface cracks, claddings exhibited stable mechanical-performance at both temperatures. These initial cracks did not critically affect the failure processes as they were not aligned with the maximum stress direction. Furthermore, the microstructure, distribution of residual stresses, and local properties of individual components in the material were thoroughly characterized, and compared with open literature on conventional claddings with monolithic outer coatings. These results provide new insights into the failure mechanisms of multi-layer SiC coatings and offer guidance for the future design of accident-tolerant nuclear fuel claddings.

36 - MATERIALS SCIENCE↗
DOE OSTI2017

Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

The Nuclear Regulatory Commission (NRC) has considered revision of 10 CFR 50.46C rule to account for the effects of burn-up rate in future analysis of reactor accident scenarios so that safety margins may evolve with the operation and reloading of the reactor as dynamic limits. In order to find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. It is computationally intensive to run a plant model through a long duration operational transient to the time of fuel reload, and this must be repeated for each reload until the time of the accident scenario. Moreover, fuel reload patterns may differ in many ways so that calculations with many different core configurations must be considered. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multi-physics tools by combining existing codes and adding new capabilities. The Parallel Highly Innovative Simulation INL Code System (PHISICS) toolkit for neutronic and reactor physics is coupled with the Reactor Excursion and Leak Analysis Program (RELAP5-3D) for the Loss of Coolant Accident (LOCA) analysis and RAVEN for the Probabilistic Risk Assessment (PRA) and margin characterization analysis. Expansion of code capability was also necessary. For RELAP5-3D to process a single sequence of cores in a continuous run required modification of RELAP5-3D to restart a sequence of input decks, each with a slightly modified neutronics model or thermal-hydraulic flow region and culminating in an accident scenario. A new multi-deck input processing capability was developed and verified for this analysis. In conclusion, the combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical Pressurized Water Reactor (PWR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1974

Development of a metal-clad advanced composite shear web design concept

An advanced composite web concept was developed for potential application to the Space Shuttle Orbiter main engine thrust structure. The program consisted of design synthesis, analysis, detail design, element testing, and large scale component testing. A concept was sought that offered significant weight saving by the use of Boron/Epoxy (B/E) reinforced titanium plate structure. The desired concept was one that was practical and that utilized metal to efficiently improve structural reliability. The resulting development of a unique titanium-clad B/E shear web design concept is described. Three large scale components were fabricated and tested to demonstrate the performance of the concept: a titanium-clad plus or minus 45 deg B/E web laminate stiffened with vertical B/E reinforced aluminum stiffeners.

Laakso, J. H.↗
DOE OSTI2018

Applicability of HBU Fuel Cask R and D Project to Al Clad Fuel Storage - 18558

Aluminum clad spent nuclear fuel (SNF) not destined to be reprocessed at Savannah River's H Canyon will eventually be moved from existing wet and dry storage facilities into standardized dry storage systems (e.g., Universal Canisters), ideally suited for disposal. With the current lack of disposal capacity in the U.S., placement of this SNF into dry storage requires plans for demonstrating to the regulator the safety of both the extended storage and the subsequent transportation to a future disposal facility of this SNF. For commercial SNF, concerns related to extended dry storage and subsequent transportation have been, and continue to be, addressed through several demonstration projects including, but not limited to, one for the lower burnup SNF at Idaho National Laboratory (INL) and one for the higher burnup SNF at Dominion's North Anna Power Station. Lessons learned from the more recent Orano led higher burnup demonstration project (HBU Demo Project) provides relevant information for advancing the concept of a 'smart cask' system for dry interim storage, which will allow for collection of data to support the safety demonstration for extended dry storage and subsequent transportation of the aluminum clad SNF. This 'smart cask' would eliminate the need for multiple demonstration projects that may have to be performed for different types of aluminum clad SNF designs and/or for potentially age distinguishing parameters, such as burnup or initial enrichment. This 'smart cask' would be developed and deployed by leveraging the HBU Demo Project's experience associated with modifying and deploying a TN-32B bolted carbon steel vertical storage cask system to monitor key parameters related to the criticality, structural, confinement, and thermal safety of the SNF and the canister system containing this SNF. This paper will focus on how the extensive design, licensing, and fabrication experience accumulated by the HBU Demo Project's cask system can be leveraged to support the design, licensing, and fabrication of a 'smart cask' system intended to support/confirm the interim dry storage, potential extended storage, and subsequent transportation of aluminum clad SNF, such as the Advanced Test Reactor (ATR) SNF. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Polymer large core optical splitter 1 × 2 Y for high-temperature operation

The design, fabrication and measurement of the properties of the large core 1 × 2 Y optical planar splitters for high-temperature operation are demonstrated. The splitters were designed by the beam propagation method optimized for operation wavelengths 650 nm and for assembling with optical fibre having core diameter 600 µm and cladding 630 µm. The splitters consisted of optical glue core and cladding layers made of Cyclo Olefin Polymer. The applied materials allowed for higher operating temperatures (up to 120 °C) than standard splitters made of acrylate plastic (up to 80 °C). The splitters show optical losses lower than 4.0 dB at 532 nm/650 nm, 2.6 dB at 850 nm, 3.4 dB at 1300 nm and 5.1 dB at 1550 nm at room temperature. The optical losses measured in the thermal shock test chamber at temperatures up to 120 °C showed that the best sample had losses lower than 4.2 dB at 850/1300 nm, 4.6 dB at 532 nm and 7.6 dB at 650 nm. Such splitters would be very useful in places with elevated temperatures, for instance in engine parts where the expected temperature can reach as high as 120 °C.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2024

Metallic Coating of Cerium Oxide Microspheres

The ability to remove heat is paramount to nuclear fuel performance and longevity. Retaining fission product and separating fuel from reactor coolant and the environment is also necessary to prevent radiological contamination. Conventional nuclear fuel for commercial light water reactors and radioisotope power systems (RPS) is composed of oxide powders pressed into a pellet (cm-scale) and then sealed into a metal cladding to confine the fuel. What typical fuels lack is a method to surround each particle of nuclear fuel in metal, thus providing a more intimate protection layer for accident tolerance and boosting the thermal extraction from the fuel element. In such a way, metal-coated fuel particles increase heat extraction efficiency over clad-pellet designs while increasing the accident tolerance of the fuel. Metal oxide microspheres have wide-ranging applications, including the realm of fuels for nuclear reactors and RPS. Microspheres of uranium oxide/uranium carbide, mixed uranium/plutonium oxides, transuranics, and thorium fuels have been extensively studied. Pacific Northwest National Laboratory has also demonstrated the production of 238 PuO 2 microspheres for RPS applications. Metal-coated oxide microsphere fuels may also be attractive for other applications such as nuclear thermal rockets, future nuclear reactor designs, and catalysts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Design rules for 316H nuclear components cladded with nickel or tungsten

The existing Class A metallic materials qualified in the ASME Boiler & Pressure Vessel Code Section III, Division 5 rules for high temperature nuclear reactors are not ideally suited for long term molten salt corrosion resistance in high temperature molten salt reactors. Potentially, corrosion or corrosion driven environmentally assisted cracking could limit the design life of structural components in molten salt reactors. A solution to this limitation is the use of cladded components – overlay the Class A material with a thin layer of some non-qualified, corrosion-resistant material. However, this necessitates the development of design methods for cladded components that do not require long-term testing of clad materials in order to support the near-term deployment of molten salt reactors. This report develops such a methodology along with a complete set of rules presented in a format compatible with an ASME nuclear Code Case. These design rules are for 316H Class A components cladded with either nickel or tungsten. The report also discusses the development of a set of general criteria for selecting clad materials beyond the specific 316H/tungsten and 316H/nickel systems and general acceptance tests for checking the mechanical integrity of the clad/base metal interface. Finally, the report includes a set of fully-documented sample problems detailing the application of the rules to high temperature cladded components

36 MATERIALS SCIENCE↗
DOE OSTI2016

Design of p-type cladding layers for tunnel-injected UV-A light emitting diodes

Here, we discuss the engineering of p-AlGaN cladding layers for achieving efficient tunnel-injected III-Nitride ultraviolet light emitting diodes (UV LEDs) in the UV-A spectral range. We show that the capacitance-voltage measurements can be used to estimate the compensation and doping in the p-AlGaN layers located between the multi-quantum well region and the tunnel junction layer. By increasing the p-type doping concentration to overcome the background compensation, on-wafer external quantum efficiency and wall-plug efficiency of 3.37% and 1.62%, respectively, were achieved for the tunnel-injected UV LEDs emitting at 325 nm. We also show that interband tunneling hole injection can be used to realize UV LEDs without any acceptor doping. The work discussed here provides new understanding of hole doping and transport in AlGaN-based UV LEDs and demonstrates the excellent performance of tunnel-injected LEDs for the UV-A wavelength range.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Computational Design of Improved Fast Reactor Cladding

HT9 ferritic-martensitic (FM) steel has served as a leading candidate for sodium-cooled fast reactor (SFR) cladding due to its favorable resistance to irradiation-induced swelling and good thermal and chemical properties. However, its limited creep strength at temperatures above 600 °C and susceptibility to α′ phase embrittlement under specific conditions could limit its application in next-generation SFRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

The advanced characterization, post-irradiation examination, and materials informatics for the development of ultra high-burnup annular U-10Zr metallic fuel

U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.

Yao, Tiankai↗
DOE OSTI2018

An advanced experimental design for modified burst testing of nuclear fuel cladding materials during transient loading

Here an advanced tube-burst test system was designed to test samples of nuclear fuel cladding under conditions relevant to a postulated design-basis reactivity insertion accident (RIA) in light-water reactors. The system is based on the “driver tube” concept and allows for high-speed testing with internal pressure impulses of 10–1000 ms. To measure strain, the system was equipped with an ultra-high-speed video camera with a telecentric lens providing high focal depth, to compensate for specimen shift. A mirror system was developed to provide 360° view of the tube specimen into a single camera. Images taken during the test allow the use of a digital image correlation approach in strain evaluation. Several experiments were conducted to assess the performance of the experimental setup, and results are reported.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design

The purpose of this document is to provide the technical specifications for the Engineering Scale Design for the drying tests of aluminum-clad spent nuclear fuel (ASNF). The engineering scale design is to accommodate both forced helium drying (FHD) as well as vacuum drying. The experiments will employ mock ASNF assemblies with a limited number of chemically treated aluminum plates to simulate the aluminum hydroxides found on ASNF. The goal here is to understand the transformation that may occur in plates including the loss of chemisorbed water and changes in morphology. Physiosorbed and bulk water found in experiments will be dried in the process and the effect on drying operations noted. Decay heat will be simulated in one assembly using a custom designed resistance heating element. A typical radial arrangement of assemblies will be employed using a modified Type 1A Basket design. For the chamber/canister, a single assembly layer with full radial dimension will be employed. Modifications to the standard canister design and configuration include the addition of a siphon tube to enable recirculation of hot gas for forced helium drying. A ring of ports/flanges are added near the top of the chamber to accommodate the feedthroughs for instrumentation and power needed for the experiment and data collection. Four viewports are added to the main section of the chamber with two in line with one of the fuel assembly and siphon tube. The other two viewports are on the opposite side of the chamber. The purpose here is to provide the option of thermal imaging in addition to the nearly four dozen thermocouples available inside the chamber. Modeling of the chamber, basket, and mock ASNF assemblies was conducted to estimate temperature distributions throughout the chamber during forced helium drying and over the anticipated drying period. This modeling was used as input to decisions on the Engineering Scale Design. Work on modeling continues in the period during setup and testing to develop preliminary drying models. Analysis of experimental results will later be used to validate these and subsequent models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Surface coupling of plasma optical emission spectra with bent metal-clad waveguide

We present the design of compact bent metal-clad waveguides to collect the vertically emitted light and to transmit to the horizontal direction. The intention of the design is to collect the plasma optical emission spectra inside the semiconductor etch process chamber to monitor plasma gas phase chemistry resides in the process chamber. Coupler A has been designed for visible spectra range from 748 to 753 nm at which the maximum coupling efficiency is over 52%, while Coupler B operates at the wavelength range from 775 to 780 nm with coupling strength of 62% and Coupler C covers the visible spectrum from 808 to 813 nm with its maximum efficiency of 60%. The surface couplers reveal good spectral characteristics, considerable alignment tolerance and minimal size of 9 × 5 × 1.6 µm{sup 3} each. Relatively simple geometrical structure over conventional grating coupler is advantageous to lower the fabrication cost as well.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗