Development and Characterization of SiC-Clad Control Rods & SiC Guide Tubes for Accident-Tolerant PWR Fuel Assemblies
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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Irradiation of SiC Clad Fuel Rods in the HFIR
During 2009 and- 2010, new test capability for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was developed that allows testing of advanced nuclear fuels and cladding under prototypic light-water-reactor (LWR) operating conditions (i.e., cladding and fuel temperatures, fuel average linear heat generation rates, and cladding fluence). For the initial experiments for this test program, ORNL teamed with commercial fuel/cladding vendors who have developed an advanced composite-wound SiC cladding material for possible use in LWRs. The first experiment, containing SiC-clad UN fuel, was inserted in HFIR in June 2010, and the second experiment, containing SiC-clad UO2 fuel, was inserted in October 2010. Two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in November 2011 at an estimated fuel burnup of approximately 10 GWd/MTHM; and two capsules (one containing UN fuel and the other UO2) were withdrawn from their respective assemblies in February 2013 at an estimated fuel burnup of approximately 20 GWd/MTHM. These capsules are currently awaiting PIE. This paper will describe the experiment, as-run operating conditions for these capsules, and current PIE plans and status.
Irradiation SiC fuel rods HFIR↗
SiC cladding R&D execution plan
To advance the development of accident-tolerant fuel cladding for light-water reactors, the Advanced Fuels Campaign actively addresses the R&D gap between the specialized technological progress in SiC-based composite materials, led by industrial partners, and the fundamental scientific understanding of material behavior under reactor operating conditions. The laboratory activities are systematically organized based on the Development Strategy document, which outlines critical barriers to the use of SiC composites in nuclear energy applications for the coming decade and beyond. To complement the strategy, this execution plan specifies a detailed year-by-year work package and establishes key milestones for assessing progress in technological advancement. This plan has been formulated in collaboration with principal investigators at national laboratories, stakeholders, and industry partners with the goal of closing technology gaps and readying SiC-based cladding technologies for full-scale commercial testing. The research conducted at the national laboratories is categorized into three primary subjects: (1) addressing challenges associated with the performance of SiC cladding under normal operating environments, (2) identifying the failure limits of SiC cladding under accident conditions, and (3) modeling fuel performance for SiC cladding systems. Comprehensive laboratory activities integrate the development of advanced modeling tools for SiC-based cladding systems with experimental validation via integrated testing of multiphysics phenomena. This document lays out the laboratory plan to support the development of SiC-based technologies, which have the potential for significant operational and safety benefits for the U.S. nuclear industry. This plan aligns with industry developer needs, relevant Nuclear Energy University Programs, and Nuclear Energy Advanced Modeling and Simulation Programs to ensure SiC cladding technologies are matured as efficiently as possible and will be maintained and updated to accommodate the dynamically evolving progress of development, capability needs, and resource availability.
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On Development and Wear Behavior of Microwave-Processed Functionally Graded Ni-SiC Clads on SS-304 Substrate
In this study, the functionally graded Ni-SiC claddings with gradual change in SiC contents were deposited on SS-304 substrate using microwave hybrid heating technique. Experimental trials were carried out inside a multimode domestic microwave device at 2.45 GHz and 900 W. The optimal exposure time of 900 W microwave power was varied with compositional gradient, and it was from 300 to 360 s. The developed functionally graded clad (FGC) was characterized by SEM/EDS, XRD and Vicker’s micro-hardness. Microstructural analysis results revealed that the FGC of approximately 2 mm thickness was developed and was free from any type of interfacial cracks and visible porosity. The maximum micro-hardness was at the top FGC layer, and its value was 1025 ± 30 HV. Three types of single-layer claddings were also developed to compare the tribological behavior of FGC. FGC exhibited 32 and 1.2 times more wear resistance than SS-304 substrate and Ni + 30% SiC layer, respectively. In the case of FGC, wear mainly occurs due to debonding of carbide particles from the matrix, while plastic deformation and strong abrasion are responsible for material loss in the case of the SS-304 substrate.
36 MATERIALS SCIENCE↗
Evaluation of the First Generation Dual-purpose Coatings for SiC Cladding
The loss of coolant and station blackout at the Fukushima Daichi nuclear power station highlighted the need to improve the safety margins of current and future Light Water Reactors (LWRs). Accident Tolerant Fuel (ATF) development is designed to find solutions to mitigate or prevent such events through improvement of fuels and fuel cladding. One promising ATF concept is the replacement of zirconium-based alloy cladding with silicon carbide (SiC) fiber-reinforced SiC matrix (SiCf-SiC) composites, due to the inherent accident resistant features of SiC. However, during normal operations, SiC corrodes in reactor coolant potentially at an unacceptable rate. In addition, it is likely for the SiC-based cladding to undergo microcracking under an applied stress, which potentially degrades the hermeticity.
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Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs
An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO{sub 2} volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)
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Mechanical Properties and Characterization of Coated SiC Fuel Cladding
SiC fiber-reinforced SiC matrix (SiC{sub f}-SiC) composites continue to undergo process modifications to address remaining concerns for Light Water Reactor (LWR) fuel cladding. The critical concerns include the radiolysis-assisted hydrothermal corrosion and the loss of hermeticity during normal operation. Coating technologies are being developed to provide a solution to both problems with no or minimum modification to the SiC{sub f}-SiC clad structures. Several barrier coating options have been considered and examined for these purposes. A comprehensive screening program is being developed to determine their feasibility in LWR environments before the coated fuel claddings are further advanced toward the ideal 'drop-in' replacements for the aging LWR fleet. There have been several coating technologies proposed. All are based on chromium and zirconium interacting with the coolant because they have a proven performance history. However, it is difficult to deposit these elements in reduced form (Zr{sup 0}, Cr{sup 0}) due to the high enthalpy associated with the stable oxide. The Cr chemical deposition techniques use electrochemical 'hard chrome' reduction of Cr{sup 3+/6+} protected by chelating ligands in solution. The collaborative partner for this endeavor is NEO Knoxville, LLC, and two types of hard chrome systems (trivalent and hard chrome) are being investigated. There are also two feasible vapor techniques previously demonstrated: commercial physical vapor deposition (PVD) methods and an experimental, proprietary vacuum plasma spray (VPS) by Plasma Processes, LLC. Both of these use Zr{sup 0} and Cr{sup 0} in high vacuum systems. In total, eight coatings and four bond coats are being evaluated. In mid-2016, these were inserted into a reactor and a characterization program is underway. (authors)
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Early Implementation of SiC Cladding Fuel Performance Models in BISON
SiC-based ceramic matrix composites (CMCs) [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability (LWRS) Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites (CMCs) as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.
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U.S. Department of Energy Accident Resistant SiC Clad Nuclear Fuel Development
A significant effort is being placed on silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding by Light Water Reactor Sustainability (LWRS) Advanced Light Water Reactor Nuclear Fuels Pathway. The intent of this work is to invest in a high-risk, high-reward technology that can be introduced in a relatively short time. The LWRS goal is to demonstrate successful advanced fuels technology that suitable for commercial development to support nuclear relicensing. Ceramic matrix composites are an established non-nuclear technology that utilizes ceramic fibers embedded in a ceramic matrix. A thin interfacial layer between the fibers and the matrix allows for ductile behavior. The SiC CMC has relatively high strength at high reactor accident temperatures when compared to metallic cladding. SiC also has a very low chemical reactivity and doesn't react exothermically with the reactor cooling water. The radiation behavior of SiC has also been studied extensively as structural fusion system components. The SiC CMC technology is in the early stages of development and will need to mature before confidence in the developed designs can created. The advanced SiC CMC materials do offer the potential for greatly improved safety because of their high temperature strength, chemical stability and reduced hydrogen generation.
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Understanding the interface interaction between U3Si2 fuel and SiC cladding
Abstract Triuranium disilicide (U 3 Si 2 ) fuel with silicon carbide (SiC) composite cladding is being considered as an advanced concept/accident tolerant fuel for light water reactors thus, understanding their chemical compatibility under operational and accident conditions is paramount. Here we provide a comprehensive view of the interaction between U 3 Si 2 and SiC by utilizing density functional theory calculations supported by diffusion couple experiments. From the calculated reaction energies, we demonstrate that triuranium pentasilicide (U 3 Si 5 ), uranium carbide (UC), U 20 Si 16 C 3 , and uranium silicide (USi) phases can form at the interface. A detailed study of U 3 Si 2 and SiC defect formation energies of the equilibrated materials yielding the interfacial phases U 20 Si 16 C 3 , U 3 Si 5 and UC reveal a thermodynamic driving force for generating defects in both fuel and cladding. The absence of either the U 3 Si 2 or SiC phase, however, causes the defect formation energies in the other phase to be positive, removing the driving force for additional interfacial reactions. The diffusion couple experiments confirm the conclusion with demonstrated restricted formation of U 3 Si 5 , UC, and U 20 Si 16 C 3 /USi phases at the interface. The resulting lack of continuous interaction between the U 3 Si 2 and SiC, reflects the diminishing driving force for defect formation, demonstrating the substantial stability of this fuel-cladding system.
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A Scoping Analysis Of The Impact Of SiC Cladding On Late-Phase Accident Progression Involving Core–Concrete Interaction
The overall objective of the current work is to carry out a scoping analysis to determine the impact of ATF on late phase accident progression; in particular, the molten core-concrete interaction portion of the sequence that occurs after the core debris fails the reactor vessel and relocates into containment. This additional study augments previous work by including kinetic effects that govern chemical reaction rates during core-concrete interaction. The specific ATF considered as part of this study is SiC-clad UO 2 .
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SiC cladding corrosion and mitigation
Silicon carbide (SiC) is a leading candidate for accident tolerant fuel. Testing of SiC in autoclaves at 360 C. degrees in simulated coolant resulted in little corrosion. However, testing of SiC tubes in the Massachusetts Institute of Technology (MIT) reactor manufactured in the same way had unacceptable corrosion levels. This in-reactor SiC corrosion issue is now being addressed by Westinghouse in tests being carried out at our Churchill, PA facility and in tests in the MIT reactor. The suspected driver for the excessive corrosion is the generation of oxidative water radiolysis products in the core. The MIT reactor was identified as running at H{sub 2} levels of ∼20 to 22 cc/kg coolant (cubic centimeters standard temperature and pressure per kilogram coolant) which were below those normally specified in operating pressurized water reactors (PWRs) of 25 to 50 cc/kg. Hydrogen is used to react with the radiolysis products to reduce them to water. The current MIT reactor tests will be run at H{sub 2} concentration levels of ∼50 cc/kg. The applicability of this approach has been supported by recent papers in the literature. Parallel testing at Churchill is being carried out to verify the efficacy of this approach using autoclaves operating at 345 C.degrees with H{sub 2}O{sub 2} injection to simulate the water radiolysis products. Various concentrations of H{sub 2} are being tested to determine the effect on SiC corrosion rate. Further studies are underway to investigate the effects of different SiC fabrication and structural parameters and to optimize for improved corrosion performance. This paper compares the results of the corrosion tests in the MIT reactor and the autoclave tests with H{sub 2}O{sub 2} injection and the effect of H{sub 2} injection on the SiC corrosion rate. The results of the autoclave tests with H{sub 2}O{sub 2} addition indicate that corrosion rates for SiC can be achieved that are much higher than those achieved in a reactor. It appears that the use of H{sub 2}O{sub 2} addition in flow through loops to simulate the corrosion rate of SiC samples in reactor is feasible. The use of H{sub 2} addition to reduce the corrosion rate in-reactor of SiC appears to be feasible.
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Developing a Macro-scale SiC-cladding Behavior Model Based on Localized Mechanical and Thermal Property Evaluation on Pre- and Post-Irradiation SiC-SiC Composites (Final Report)
Silicon carbide is being investigated for accident tolerant fuel cladding applications due to its good neutronic performance, high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. The development and investigation of these materials is particularly important in the light of the Fukushima event and subsequent emphasis in DOE on accident tolerant fuel (ATF) concepts. In this work, we propose to develop and improve upon existing small-scale mechanical and thermal characterization methods, including micro-cantilever bend tests and fiber push-out tests, and time-domain thermo-reflectance measurements. These techniques will be applied to evaluate micro-scale properties of SiC-SiC composite constituents (matrix, fibers, interphase). The results will be coupled with meso-scale fiber and void structural information and be used as input to develop a comprehensive, finite element-based model based on constituent properties. The goal of this multi-scale model is to be used to predict anisotropic mechanical and thermal cladding behaviors. The model will be benchmarked against measurements of bulk properties and validated for radiation-damaged materials by utilizing irradiated SiC-SiC composites available from General Atomics. The characterization approach developed under this program will have particular application in the property prediction of reactor irradiated materials, as only small volumes need to be tested, minimizing associated hazards and costs. The micro-scale test methods developed here will be used along with meso-scale structural data to inform a finite element model of the full SiC-based cladding system. This model will capture fabrication and irradiation effects and importantly include for the first-time separate material data for each system component. The incorporation of these micro-scale effects will lead to a more accurate model of SiC-SiC composite behavior, enabling further advancements in the development and design of SiC-based materials for improved fuel cladding performance. The aim of this work is to develop advanced localized material characterization techniques to directly measure mechanical and thermal properties of the individual constituents of SiC-based claddings at the relevant micro-scale. SiC-SiC composites will be evaluated before and after irradiation and these results will be coupled with macro-scale properties and microstructural information in order to provide the input parameters for a comprehensive finite element model, which will be developed in this program.
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Assessment of different fuel design options with SiC cladding for light water reactors
It is very important to avoid high tensile stresses in SiC if it is to be used safely as the cladding material in LWRs. To achieve this goal, three fuel design options are examined here: fuel pellets with a 10% vol central void, adding 10% vol BeO in UO{sub 2} fuel and replacing the gas gap with Lead-Bismuth Eutectic (LBE) bond. For each case, a neutronic analysis is first applied using CASMO-4E to obtain the required enrichment to achieve a fuel average burnup of 50 MWd/kgU. The fuel performance code FRAPCON, specifically FRAPCON-3.3 and FRAPCON-EP, is modified and then used to simulate the effects of steady-state irradiation in each case for typical PWR peak rod conditions. Key parameters like fuel average temperature and plenum pressure are examined for the relative performance of each approach. The results show that all three options can improve the fuel performance by lowering the fuel temperature as well as the plenum pressure. Fuel pellets with a central void design offers the most advantages since the larger void volume in the fuel rod results in the lowest plenum pressure, which is the limiting condition for safe operation of the SiC cladding. (authors)
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Preliminary characterization and projections of PVD coatings on SiC cladding for Light Water Reactors
Physical Vapor Deposition (PVD) based CrN, Cr and TiN are currently the leading mitigation coatings for SiC-based Light Water Reactor (LWR) fuel cladding. This manuscript describes the morphology, phase analysis and elemental composition of coatings deposited by commercial vendors on SiC clad materials. The coatings showed continuous crack-free morphology with few microstructural flaws. X-ray analysis and spectroscopy indicated high purity >99.9%. Mechanical tests of coating debonding (by ASTM D4541) and scratch-indentation (by ASTM E2546) were conducted. Strong interface bonding was observed and higher resistance to normal and shear stresses was observed from thicker coatings. Preliminary developments in scratch testing, phase analysis, future needs and expected behavior under irradiation are discussed.
Ang, Caen K.↗
Radiation-induced Swelling and Microcracking in SiC Cladding for LWRs
After the Fukushima accident in 2011, SiC has attracted considerable attention as a potential material for fuel cladding that can offer increased accident tolerance in light water reactors (LWRs) compared to zircaloy cladding. Existing measurements on nuclear grade SiC/SiC composites made of high-purity, stoichiometric SiC have already shown very good radiation tolerance in various harsh environments. However, there are a number of challenges that need to be addressed before SiC composites can be successfully used in LWR technologies. One of them is irradiation-induced swelling, which can lead to development of significant stresses in the matrix, followed by microcracking and subsequent release of fission products. Models of irradiation-induced swelling had been largely empirical and therefore they cannot be used to predict the microstructural dependence of microcracking. Existing models are also limited by the lack of understanding of defects that form due to irradiation in the range of temperatures relevant to fuel cladding in LWRs (<1000°C). Many of the defects in this regime of temperature are too small to be detected with traditional transmission electron microscopy (TEM) techniques. In fact the defects observed in traditional TEM account only for 10-45% of the swelling measured in irradiated SiC. Here, we developed a multi-scale simulation methodology combined with state of the art experimental imaging techniques based on aberration corrected scanning transmission electron microscopy (STEM) to provide a scientific basis for predictive models of radiation swelling and microcracking in SiC. Specifically, we have developed a phase field fracture model that predicts microcrack initiation and growth in polycrystalline SiC. The model accounts for single crystal anisotropy, including anisotropy in the elastic constants and anisotropic surface energy. It also considers the reduced fracture strength of grain boundaries. The model was implemented using the MARMOT mesoscale nuclear materials tool. We have parameterized the model for SiC using molecular dynamics simulation results. The model has been verified using single crystal, bicrystal, and polycrystal examples. Swelling model has also been developed based on cluster dynamics simulations of defect evolution in SiC. We found that point defects and small defect clusters have a significant contribution to swelling. This finding is important because point defects and very small defect clusters are not visible in experiments (e.g., in TEM). Combined modeling and experiments can bring critical insights into how swelling depends on irradiation conditions and the microstructure. In situ nanoindentation tests and ex situ nanoindentation tests were performed on assynthesized 3C-SiC and 3C-SiC irradiated under different radiation conditions. Without radiation, bend contour movement and residual bend contours after indentation demonstrate some room temperature plastic flow and residual plastic deformation. Plastic deformation recovery is also observed on unirradiated samples during unloading. Radiation-induced embrittlement is demonstrated by differences in crack geometry in in situ tests, and is consistent with ex situ test results. Failure by single, straight, cracks indicates decreased fracture toughness due to irradiation in 3C-SiC. Future work includes coupling of the swelling and fracture models and development of a master equation that can be used in the BISON code.
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Advancing SiC clad fuel performance model: bridging micro- and macro-scale models and experimental validation
This report presents a workflow for advancing fuel performance modeling of SiC composite cladding for light-water reactors by linking microscale, experimental data-informed finite element analysis with rod-scale fuel performance codes such as BISON. The workflow uses X-ray computed tomography (XCT) to capture the actual geometry and processing-induced defects of as-fabricated SiC composite tube specimens, particularly porosity and wall-thickness variations, and converts the segmented XCT volumes into image-based finite element meshes for high fidelity structural analysis.
Koyanagi, Takaaki [Oak Ridge National Laboratory (↗
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)↗