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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Effects of sample bias on adhesion of magnetron sputtered Cr coatings on SiC

Swelling of SiC at 300 C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of –50V, –75V, and –100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. Additionally, at the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Comparison of Thermal Diffusivity Degradation of Irradiated Zirlo and SiC-SiC Composite

Zirlo (Zircaloy cladding material from Westinghouse) has been widely used as the fuel cladding material in Light Water Reactors because of its high melting point, low susceptibility to corrosion, good thermal and mechanical shock resistance, and low degradation of mechanical properties under irradiation. But as Zirlo reacts with water at high temperature, which can trigger serious accidents, efforts have been made to explore better candidates to replace Zirlo in the water-cooled reactors. SiC-SiC composite is considered as one of the most viable options due to its comparable mechanical performances and corrosion resistance. However, one of the most important properties, thermal diffusivities of both Zirlo and SiC-SiC composite, and their degradation after irradiation, are not well known yet. Ion-irradiated samples are used to study the thermal property degradation caused by irradiation instead of the neutron-irradiated ones due to the significantly lower radioactivity. The microstructural damage is estimated from the software simulation SRIM/TRIM based on the energy and fluence of the ion, and the target material, and represented by DPA (displacement by atom). Different from neutron-irradiated samples, the damaged layer of ion-irradiated samples is only a few to a few tens of micrometers thick, and the damage is depth dependent. It is impossible to directly measure the bulk property change, and a microscale measurement technique is necessary. Laser-based photothermal reflectance technique is one of the ideal methods to measure local thermal conductivity and diffusivity with high spatial resolution. The main principle of this technique is to heat the solid sample using an intensity-modulated laser, and scan the sample with a constant intensity continuous-wave (CW) laser to detect the thermal wave propagation. On bulk, isotropic material, if the heating laser intensity is modulated as a sine wave, thermal diffusivity of the sample can be derived from the phase lag of the thermal wave, scan distance and modulation frequency in a straight forward manner. By using more specific analytical models, thermal conductivity and diffusivity in complicated cases, such as layered structure materials, composite materials, and/or anisotropic materials, are measurable as well. In this work, thermal diffusivity of the light damage layer of the post-irradiation Zirlo and SiC-SiC composite are measured and compared. The Zirlo sample is irradiated by Ni{sup 9+} ion, and SiC-SiC composite material is irradiated by Si{sup 7+} ion. The peak DPA of both sample is similar as ∼10. For ion-irradiated materials, DPA is known to increase with depth before it reaches the peak, and then sharply drop to zero. In term of light damage layer in this manuscript, we mean by the layer with a DPA smaller than 5% maximum value. In this layer DPA increases slowly and smoothly with depth, and the structure and thus thermal properties are more spatially uniform. For the SiC-SiC composite, the measurements are separately conducted on fibers along radial or axial direction, and on the matrix. Our measurement results show that with similar microstructural damage, thermal diffusivity of matrix SiC reduces more significantly than that of Zirlo. For fiber SiC that has a much lower thermal diffusivity than the matrix SiC before irradiation, the post-irradiation effect on thermal diffusivity is not obvious.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Mechanisms of stored energy release in silicon carbide materials neutron-irradiated at elevated temperatures

Understanding the stored energy release behavior of SiC is crucial for the design of SiC-based fuel cladding for light water reactors because stored energy release affects response during accident conditions. Differential scanning calorimetry was used to evaluate the stored energy of monolithic SiC and SiC fiber–reinforced SiC matrix composite variants following neutron irradiation under light water reactor–relevant dose and temperature conditions. The tests were performed at heating rates of 0.5, 3, and 20 K/min up to 1,273 K. The onset temperature of the energy release roughly corresponded to the irradiation temperature, and energy release was delayed as the heating rate increased. The monolithic and composite specimens exhibited similar stored energy release behavior up to ~1,000 K, and above that temperature, the energy release depended more on material type. The energy release measured in this study is below the material-specific heat capacity at temperatures below ~1,000 K but exceeds it under certain material and annealing conditions at temperatures above ~1,000 K. The amount of irradiation-induced volumetric swelling of the specimens was an indication of the total stored energy. Based on this observation, the energy release behavior was described by a swelling recovery model.

36 MATERIALS SCIENCE↗
DOE OSTI2013

SiC MODIFICATIONS TO MELCOR FOR SEVERE ACCIDENT ANALYSIS APPLICATIONS

The Department of Energy (DOE) Office of Nuclear Energy (NE) Light Water Reactor (LWR) Sustainability Program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. The Fuels Pathway within this program focuses on fuel system components outside of the fuel pellet, allowing for alteration of the existing zirconium-based clad system through coatings, addition of ceramic sleeves, or complete replacement (e.g. fully ceramic cladding). The DOE-NE Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) is also conducting research on materials for advanced, accident tolerant fuels and cladding for application in operating LWRs. To aide in this assessment, a silicon carbide (SiC) version of the MELCOR code was developed by substituting SiC in place of Zircaloy in MELCOR’s reactor core oxidation and material property routines. The purpose of this development effort is to provide a numerical capability for estimating the safety advantages of replacing Zr-alloy components in LWRs with SiC components. This modified version of the MELCOR code was applied to the Three Mile Island (TMI-2) plant accident. While the results are considered preliminary, SiC cladding showed a dramatic safety advantage over Zircaloy cladding during this accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Assessment of SiC-CMC compatibility with components of actinide fuel systems

Silicon carbide (SiC) fiber reinforced SiC ceramic matrix composites (CMC) represent an alternative cladding material for accident tolerant fuel light water reactor (LWR) applications, however, data and technology gaps exist that need to be understood before SiC-CMCs can be considered. One technology gap is the understanding of fuel-clad chemical interactions (FCCI) in SiC-CMC clad LWR systems. A review of SiC interactions with relevant elements to the LWR fuel system is discussed. Insights are drawn from research focusing on the use of SiC for high temperature electronics applications and high temperature structural materials applications as well as insights from tristructural-isotropic (TRISO) coated particle fuel development, where SiC is a critical component of the TRISO design. The interactions discussed in this review are framed relative to the expected conditions for SiC-CMC cladding application. Furthermore, the general observation is that limited risk of FCCI exists under normal operating conditions for SiC-CMC cladding systems in LWR applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multiscale Modeling of the Mechanical Response of Silicon Carbide Composite Within the Accelerated Fuel Qualification Framework

The accelerated fuel qualification (AFQ) framework has been used for the initial development of multiscale modeling of silicon carbide (SiC) fiber reinforced composite (SiC-SiC). The AFQ framework provides a methodology to leverage physics-informed multiscale modeling along with a reduced set of empirical test data to reduce the time and cost of licensing and qualification of new nuclear fuel systems while maintaining the overall nuclear power plant safety case. SiC-SiC is being proposed for in-core applications, most notably fuel cladding, for current and next-generation nuclear reactors because of its high temperature stability, irradiation tolerance, and ability to withstand many accident conditions. As these composites exhibit multiscale architectures and complex microstructure-based fracture mechanics, it is an appealing use case for the AFQ methodology. While the end goal of this work is a single multiscale model that can be used for predictive in-core performance, current focus is on the individual various length scale models. Four individual models have been initially developed from microscale to engineering system level to capture key physics-based effects across different length scales. These models include a microscale homogenized tow model, a mesoscale fast Fourier transform–based weave model that integrates the homogenized tow model, a mesoscale finite element–based weave model, and a system-level BISON fuel performance model. Results of these models have undergone an initial comparison with separate-effects test data showing a good match to experimental results. By using the AFQ framework during model development, several near-term benefits have been secured including a reduction in development time for the SiC-SiC cladding, more targeted irradiation testing, and a better understanding of uncertainty.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Assessment of Possible Cycle Lengths for Fully Encapsulated Microstructure fueled light water reactor Concepts

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel “as-is” into existing LWRs while retaining their thermal–hydraulic characteristics. Another mandatory constraint is use of low enriched uranium (at or below 20 w/o). The feasibility of using this fuel is assessed by looking at two factors: cycle lengths and fuel material failure rates. Other considerations (e.g., safety parameters such as reactivity coefficients, feedback, etc.) were not considered at this stage of the study. The study includes the examination of increases in the TRISO kernel sizes without changing the thickness of any of the coating layers. In addition, cases where the buffer layer thickness is allowed to vary are also considered. The study shows that a naïve use of UO2 (even up to 20 w/o enrichment) results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. The increase of fissile inventory can be accomplished through multiple means, including higher particle packing fraction, higher enrichment, larger fuel kernel sizes, and the use of higher density fuels (that contain a higher number of U atoms per unit volume). In this study, starting with the recognized highest packing fraction practically achievable (44%), combinations of the other means have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Assessment of possible cycle lengths for fully-ceramic micro-encapsulated fuel-based light water reactor concepts

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with accident-tolerant fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rate of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Role of Defects in Swelling and Creep of Irradiated SiC

Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and modeling effort to discover the previously invisible defects in irradiated SiC and to determine the contributions of these defects to radiation swelling. Knowledge of the most stable defect structures and the rate controlling processes during defect evolution is essential for development of predictive models for swelling and creep as a function of temperature and radiation dose. This research has been enabled by state-of-the-art imaging techniques, such as the aberration corrected scanning transmission electron microscopy (STEM) (FEI TITAN) closely coupled with multi-scale models of stable defect clusters and their evolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called the endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Failure Characterization of SiC/SiC Woven Tubular Composites

In the nuclear power industry, fuel-pellet swelling and pellet-cladding interactions induce complex stress states in the cladding material. SiCf/SiC composite tubes (SiC fibers embedded in SiC matrix) have been proposed as a candidate material for the cladding material. SiCf/SiC has high resistance to radiation-damage, low neutron crosssection and good damage tolerance in extreme radiation environments. The drawback of using SiC is the brittle nature of the material. However, using SiC in a woven ceramic matrix composite introduces a pseudoductility effect without catastrophic failure. In the current work woven, tubular cladding with continuous silicon carbide fibers and a high-purity stoichiometric silicon-carbide matrix was compared with monolithic silicon carbide samples.

Nance, James↗
DOE OSTI2016

Preliminary analysis of core heat-up for ATFs under accident conditions

The accident-tolerant fuel (ATF) system is aim at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the conventional UO{sub 2}-Zr fuel system when the reactor is in normal operation. In this paper, the improved performance of UO{sub 2}-SiC, UO{sub 2}-FeCrAl fuel systems, and of fully ceramic micro-encapsulated (FCM) fuels: FCM-SiC and FCM-FeCrAl, on the core heat-up progress in a pressurized water reactor (PWR) is analyzed by using the best-estimate system code RELAP5 and severe accident analysis program. For this purpose two accident scenarios are selected: a large break LOCA as a design base accident and an extended station blackout (SBO) severe accident. For the LBLOCA with just low-pressure safety injection, larger volumetric heat capacity and thermal conductivity of ATF materials contribute to a larger safety margin in terms of the peak cladding temperature compared with UO{sub 2}-Zr. Thanks to higher melting temperatures, FCM-SiC is failed by melting almost 1200 s later than UO{sub 2}-Zr. The accumulated production of hydrogen for ATF materials is around 10 times smaller than that for UO{sub 2}-Zr. However, high temperature oxidation of SiC cladding generates carbon monoxide which is combustible, while the melting point of FeCrAl is not high enough. In the SBO scenario with mitigation measure, the advantage of ATF is more evident. If the mitigation measure could not be taken, core heatup is inevitable for all types of fuel system, but the temperature rising of ATF core is much slower. In general, improved high-temperature properties and resistance to oxidation play important roles for ATF cores.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, together with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10 -12 atm-cc/s for helium and 2.83 x 10 -12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as is manifested by the un-detectable deuterium permeation flux at various feeding gas pressures. A large helium leakage rate was detected for the uncoated SiC/SiC composite tube exposed to atmosphere, indicating it is inherently not hermetic. The hermeticity of coated SiC/SiC composite tubes is strongly dependent on the coating materials and the preparation of the substrate SiC/SiC composite samples. To simulate the practical application environment, monolithic CVD SiC tubes were exposed to neutron irradiation at the High Flux Isotope Reactor under high heat flux from the internal surface to the external surface. Although finite element analysis and resonant ultrasound spectroscopy measurement indicated that the combined neutron irradiation and high heat flux gave rise to a high probability of cracking within the sample, the hermeticity evaluation of the tested sample still exhibited gas tightness, emphasizing that SiC cracking is inherently a statistical phenomenon. The developed permeation testing station is capable of measuring the gas permeation flux in the range of interest with full confidence based on the presented results. It is considered a critical pre- /post-irradiation examination technique to characterize SiC-based cladding materials in asreceived and irradiated states to aid the research and development of ATF.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗