Search NASA⌕ Search

SEARCH · Search NASA

Results for “SiC cladding”

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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

A Radial Heat Flow Apparatus for Thermal Conductivity Characterisation of Cylindrical Samples

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. In this frame, the thermal conductivity (TC) of cladding prototype sections are measured both before and after neutron irradiation with the purpose of better understanding the effects of neutron irradiation on the pyrolytic carbon interphase linking the SiC fibres to the SiC matrix. To this end, an apparatus has been designed and built to measure the TC of short cylindrical samples, whose lengths range from 5 to 20 cm, by the radial heat flow method. To avoid oxidation issues, the experiments are carried out in vacuum. Today the most common measurement method used to characterise the TC of materials is the so-called laser-flash technique, whereby a short high intensity laser pulse is shot on one surface of a thin flat sample and the energy transfer through the sample is analysed on the other side. Though much faster and easier to implement than the apparatus presented in this work, laser-flash equipment does have issues when dealing with highly anisotropic materials or capturing shape effects such as those present in large aspect ratio samples such as cladding tube sections. The two aforementioned effects are exactly what the developed equipment has been built to capture. The radial heat flow method is based on the principle of measuring the temperature difference building up through a piece of material as this material acts as a thermal resistance against a known heat flow crossing it. This paper presents a new implementation of one of the classical thermal conductivity measurement techniques. Since the development of the laser-flash method in the sixties, steady-state measurements such as the radial heat flow have fallen in disuse. In this discussion, it is shown that a direct measurement method can be used and even be preferable in the case of highly anisotropic materials such as SiC/SiC composites. The advantage of this experimental setup is that the actual transverse conductivity is obtained, whereas laser-flash measurements yield a response in which both the transverse and axial conductivities are mixed. Further steps of this study will include the measurements of samples irradiated with neutrons in the MITR test-reactor. The aim will be to evidence the role played by the pyrolytic carbon interphase by linking the TC measurements with energy filtered electron microscopy. We would like to acknowledge General Atomics and Westinghouse for their support in the frame of the CARAT research program. Providing us with actual prototype clad sections is a unique opportunity. Furthermore the Swiss CCEM.ch project MeAWaT is acknowledged for partially financing the development of the measurement device. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents mechanical testing capability demonstration activities performed in fiscal year (FY)2025 at Oak Ridge National Laboratory(ORNL)on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~13GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide(SiC)thermometry (TM) was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target. Mechanical testing at ambient temperature demonstrated the expected irradiation-induced hardening, with yield strength(YS)and ultimate tensile strength(UTS)values increasing substantially relative to unirradiated counterparts. However, this strengthening was accompanied by a reduction in ductility, as indicated by lower uniform and total elongations(UE and TE). Both coated and uncoated claddings exhibited similar mechanical response, though Cr-coated specimens showed surface cracking perpendicular to the loading direction, attributable to the hardness mismatch between the coating and substrate. Fracture in all cases remained ductile, and no coating spallation was observed following HFIR irradiation. Complementary efforts were directed toward the fabrication of test specimens from commercially irradiated cladding (rod 47I, ~31.1 GWd/t average burnup). Axial sectioning and computer numerical control(CNC)machining successfully produced ATT geometries suitable for benchmarking against HFIR-irradiated specimens. This capability enables direct comparison of cladding behavior between test reactor and commercial reactor environments, thereby supporting the validation of HFIR as a surrogate irradiation platform for accident tolerant fuel (ATF) development. Once HFIR irradiations are completed in FY26,the relevant comparison tests will be completed. Collectively, the FY 2025 PIE campaign has provided mechanical performance data for irradiated advanced claddings The demonstrated capabilities support the framework for mechanical testing and further evaluations in subsequent years. These efforts will represent an important contribution toward the licensing and deployment of Cr-coated zirconium alloy cladding as a near-term ATF solution.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Integral Heat Transfer Test Method for Determining Thermal Resistance of Silicon Carbide Cladding and Liquid Tin System at Elevated Temperatures - Master's Thesis Presentation University of Idaho

Presentation for master's work at University of Idaho discussing Integral Heat Transfer Test Method for Determining Thermal Resistance of Silicon Carbide Cladding and Liquid Tin System at Elevated Temperatures. Presentation covers slight literature review results/previous work, discussions on system design and approach for thermal property measurement of SiC composite and Tin system, modeling efforts, and results.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

Phase Equilibria and Thermochemistry of Advanced Fuels: Modeling Burnup Behavior (Final Report)

Achieving the goal of developing advanced fuel concepts that meet the DOE objectives of being robust, demonstrating high performance, and are more tolerant of accident conditions than current fuel systems will require a thorough understanding of the thermophysical and thermochemical properties of the constituent materials. Non-oxide fuel systems are being explored under the Advanced Fuels Program that hold significant promise for improved performance and accident tolerance, including the uranium silicide-based system considered in the current work. Prospective cladding materials currently considered that contribute to improved accident tolerance include silicon carbide composites and ferritic alloys (Fe-Cr-Al base compositions). Thus, the effort developed thermochemical models and values, supported with targeted experiments, to evaluate the ferritic alloy and silicon carbide composite cladding systems in contrast to current zirconium alloy cladding. The developed detailed understanding will serve to aid in relatively early screening of candidate systems to avoid wasted effort, guide development of new fuel forms, and to provide a basis for predicting and modeling fuel performance. Major deliverables for the project included: Thermochemical assessment and models of phases in the U-Si and U-Si-N systems; thermochemical evaluation supported by experimental measurements of fuel-cladding interactions of silicide fuel with baseline zirconium, ceramic composite, and ferritic alloy cladding; thermochemical assessment and models of phases supported by experimental measurements for silicide fuel with key fission products provided in a dataset and reported in refereed publications. Within the project a significantly refined U-Si phase diagram was developed and reported that now includes homogeneity ranges for key phases, such at the U 3 Si 2 proposed fuel phase, and settles issues with regard to uncertainty in the stability of some phases. Computational efforts together with key experiments has determined phase formation in interactions between U 3 Si 2 and Zircaloy-4 cladding material, a ferritic FeCrAlY alloy of interest as an advanced cladding material, and silicon carbide, also of interest as a fiber-reinforced composite cladding. As expected, very significant reactions occur between U3Si2 and Zircaloy-4, with much less interaction at higher temperatures for the ferritic alloy, and finally interactions with SiC only in the region of contact. A potentially major issue is the stability of U 3 Si 2 fuel that has undergone significant burnup. The result is the loss of the uranium metal, liberating silicon, and the formation of concomitant fission product elements that either dissolve in the U 3 Si 2 phase or form independent, and possibly silicide phases. A combination of experimental determinations of phase formation of U 3 Si 2 reacted with representative fission products yttrium, gadolinium, cerium, zirconium, and molybdenum and first principles calculations has helped understand the fuel chemistry. The behavior of the U 3 Si 2 phase and the partitioning of silicon to possible fission product phases with burnup was thus determined, with significant dissolution in U 3 Si 2 of cerium, gadolinium, zirconium, and plutonium predicted along with independent phase formation of a U-Mo-Si ternary phase, yttrium silicide, and elemental selenium. It can be concluded that at significant burnup there will be a very minor amount of the U 3 Si 2 fuel phase that will decompose to a lower silicide or a uranium alloy as silicon preferentially forms a secondary phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Enhanced Thermal Conductivity UO{sub 2}-BeO and UO{sub 2}-SiC Fuels Behavior

Commercial light-water reactors currently use UO{sub 2} as a nuclear fuel, which has demonstrated several desirable characteristics. However, UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet. This temperature gradient contributes to limits in the operational performance of the reactor due to effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases as demonstrated. These high fuel temperatures can be decreased and reactor performance improved by developing an enhanced thermal conductivity nuclear fuel. A high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling as described. Additionally, fission gas release would be decreased allowing for higher fuel burn-up, and the safety of the reactor would be improved with a faster thermal response and less stored energy in the fuel pins. A ceramic-ceramic composite nuclear fuel has an increased effective thermal conductivity due to adding a second compatible and high thermal conductivity ceramic to UO{sub 2}. SiC and BeO are two materials that demonstrate compatibility with UO{sub 2} and have a high thermal conductivity. In summary, the enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuel performance has been modelled in a light water reactor, and a more complete and detailed understanding was achieved for enhanced thermal conductivity UO{sub 2}-BeO and UO{sub 2}-SiC fuels. BeO and SiC have high melting point, low neutron absorption, low thermal expansion coefficient and chemical compatibility with current UO{sub 2} fuels which makes them a suitable material for thermal conductivity enhancement. In our modeling and simulation work, the enhanced thermal conductivity nuclear fuel showed the capability of decreasing the fuel temperatures, increasing the gap total conductance, reducing the fission gas release, decreasing the gap/plenum pressure, lessening oxygen redistribution, and facilitating a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor operation would be then improved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloy reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

HFIR SiC Bowing Test Ready to Insert

This report describes the successful assembly of a High Flux Isotope Reactor (HFIR) irradiation experiment designed to assess radiation-induced lateral bowing of silicon carbide fiber–reinforced, silicon carbide matrix composite (SiC/SiC) components under a radial fast neutron flux gradient. Excessive bowing of a SiC/SiC channel box in a boiling water reactor could potentially interfere with control blade movements. Similar concerns exist for SiC/SiC fuel cladding in light water reactors. The experiment described herein will provide experimental validation of the structural response of a miniature SiC/SiC channel box and tube specimens with pressurized water reactor diameters during irradiation. The significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were characterized using detailed three-dimensional neutronic calculations. The three-dimensional displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses that predicted the deformation and stresses in the channel box specimen. The specimens were thoroughly characterized prior to irradiation using traditional dimensional inspection and surface profilometry so that these measurements can later be compared with similar measurements that will be made post-irradiation to determine radiation-induced deformations. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped using a digital microscope and a three-dimensional stage. This technique allowed for accurate measurements of the marker spacings, which can be compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment was successfully assembled and is scheduled for insertion during HFIR cycle 492, which is currently scheduled to run from May 25, 2021 to June 18, 2021.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

High Hydrogen Content Graphene Hydride Compounds & High Cross-­ Section Cladding Coatings for Fast Neutron Detection

The objective is to develop and implement a superior low-cost, large area (potentially >32in), easily deployable, close proximity, harsh environment innovative neutron sensor needed for next generation fuel cycle monitoring. We will exploit recent breakthroughs at the PI’s lab on the electrochemistry of epitaxial graphene (EG) formed on commercial SiC wafers, a transformative nanomaterial system with superior radiation detection and durability properties to develop a new paradigm in detection for fast neutrons, a by-product of fission reactors. There are currently few effective detection/monitoring schemes, especially solid-state ones at present. This is essential for monitoring and control of future fuel cycles to make them more efficient and reliable. By exploiting these novel materials, as well as innovative hybrid SiC/EG/Cladding device architectures conceived by the team, will develop low-cost, high performance solutions to fast-neutron detection. Finally, we will also explore 3-terminal device implementations for neutron detectors with built-in electronic gain to further shrink these devices and improve their sensitivity.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Mach Conference, 2022.

There has been considerable interest in chemical vapor infiltration (CVI) manufactured silicon carbide fiber and silicon carbide matrix (SiC/SiC) composite tubes due to their superior mechanical properties. The SiC/SiC tubes are known to remain stable even after prolonged exposure to radiation. However, there are many variables in the manufacturing of SiC/SiC composite tubes such as elastic constants of constituent SiC fiber and SiC matrix, braiding angle, porosity, and others whose values vary over a range. The intention is to identify the distribution of elastic constants of SiC/SiC tubes which would help model the performance of the novel cladding material. To do so, a sensitivity analysis will be carried out to determine the dominant variables that influence the elastic constants of SiC/SiC tubes. A chosen number of combinations of dominant variables will be sampled through the Latin hypercube sampling (LHS) method. The elastic constants of composite tube namely Young's moduli in the circumferential and longitudinal direction, Poisson's ratio, and shear modulus will be calculated at sampled points through finite element (FE) analysis. A polynomial response surface (PRS) will be built for each of the elastic constants to be used as a surrogate to FE analysis. Once the predictive accuracy of PRS is verified at the validation set of data points, the PRS will be invoked 105 times in Monte Carlo simulations (MCS). The uncertainty can be quantified by calculating the coefficient of variation (CV) based on MCS-generated data which will be presented at the conference.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Development and Implementation of Digital Image Correlation for Out-of-Cell Burst Testing

This report details the progress made towards utilizing in-situ methods to observe and quantify unirradiated Zircaloy-4 (Zr4) fuel cladding deformation during out-of-cell design basis accident conditions, specifically during loss-of-coolant accident (LOCA) burst testing in the Severe Accident Test Station at Oak Ridge National Laboratory. Digital image correlation (DIC) was implemented to calculate cladding strain during burst and infrared (IR) thermography was used to attempt a correlation of strain with temperature and to investigate thermal gradients. For both techniques, early experimentation revealed experimental modifications were necessary for implementation. For DIC, this manifested as shortened cladding lengths which could be centered in front of the viewing chamber bored through the IR furnace side, and for IR thermography this meant conducting burst tests in air without a reaction tube and the incorporation of a SiC shell to eliminate direct IR reflections from the tungsten lamps. In the following, the technical issues and subsequent origin of these modifications are discussed, laying the framework for a proof-of-concept DIC/IR characterization of a burst test in the final section.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

High Temperature Steam Corrosion of Cladding for Nuclear Applications: Experimental

Stability of cladding materials under off-normal conditions is an important issue for the safe operation of light water nuclear reactors. Metals, ceramics, and metal/ceramic composites are being investigated as substitutes for traditional zirconium-based cladding. To support down-selection of these advanced materials and designs, a test apparatus was constructed to study the onset and evolution of cladding oxidation, and deformation behavior of cladding materials, under loss-of-coolant accident scenarios. Preliminary oxidation tests were conducted in dry oxygen and in saturated steam/air environments at 1000OC. Tube samples of Zr-702, Zr-702 reinforced with 1 ply of a ß-SiC CMC overbraid, and sintered a-SiC were tested. Samples were induction heated by coupling to a molybdenum susceptor inside the tubes. The deformation behavior of He-pressurized tubes of Zr-702 and SiC CMC-reinforced Zr-702, heated to rupture, was also examined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

All SiC/SiC cladding irradiation in LWR environments

The Muroran Institute of Technology has been in charge of developing SiC/SiC claddings by the DEMO-NITE method, and to evaluate stability in reactor water environment, in high temperature steam environment and in neutron exposure environment. Those issues have been studied during five years and carried out in 3 experiments (Scarlet-1, Scarlet-2 and Inspire-1) irradiated in the Halden Boiling Water Reactor (HBWR) in Norway. The main conclusions follow. The corrosion of SiC/SiC is accelerated by the neutron irradiation. The corrosion of SiC/SiC significantly depends on the condition of water chemistry, and the corrosion of SiC/SiC is drastically decreased under hydrogen water chemistry. Under boiling water condition, the corrosion rate of SiC/SiC is slightly increased. The significant degradation at a joint was confirmed by the galvanic corrosion caused by the joining of dissimilar materials. As expected, there is no degradation of full SiC/SiC segment prepared by diffusion bonding. Microstructural change of CVD-SiC coat on SiC/SiC tubes was not observed. The effectiveness of CVD (Chemical Vapor Deposition) SiC coat as environmental barrier coating is confirmed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

A novel fiber-fretting test for tribological characterization of the fiber/matrix interface

Ceramic matrix composites exhibit excellent high temperature properties and are candidate materials for structural applications from jet engines to nuclear fuel cladding. The strength and deformation behavior of these composites are intrinsically tied to the properties of the fiber/matrix interface. This research introduces a novel technique to evaluate the fundamental friction and wear characteristics of these interfaces. A case study on SiC f /PyC/SiC m composites applies in situ fiber fretting to investigate the kinetic friction coefficient and cyclic wear characteristics of the interphase. The experimental methodology is presented in explicit detail to encourage other researchers to explore the capabilities of this technique. Additionally, testing was carried out as a function of fiber roughness, PyC thickness (≈10, 500, 1200 nm), cycle count (10, 100, 1000), and frequency (1, 5, and 10 Hz). The friction coefficient and wear behavior changed with cycle length due to a transition from adhesive to abrasive mechanisms. SEM fractography provided insights to tribological evolution from adhesive sliding to three-body abrasive wear. The structural disorder at the tribo-surface relating to graphitic content and dangling bond density was investigated using Raman spectroscopy. Properties are contextualized with comparison to uniaxial tensile behavior of the parent composite.

36 MATERIALS SCIENCE↗