Binary Metal Ceramic Tubes for Accident Tolerant LWR Fuel Cladding
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Accident Tolerant Fuel↗
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Systematic Technology Evaluation Program for SiC/SiC Composite-based Accident-Tolerant LWR Fuel Cladding and Core Structures: Revision 2015
Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Weld Development of FeCrAl Thin-Wall Cladding for LWR Accident Tolerant Fuel
The development of new cladding materials to replace Zr-based cladding for light water reactors (LWRs) has been actively pursued with improved high-temperature oxidation resistance to significantly suppress hydrogen production from an accident such as a loss of coolant event. FeCrAl alloy with a typical composition of approximately Fe-15Cr-5Al has been selected as a candidate material because of its improved high-temperature oxidation resistance over the present LWR zircaloy cladding. As a result of higher thermal neutron absorption for FeCrAl alloy, the cladding wall thickness needs to be reduced by approximately half compared to that of Zircaloy cladding to be comparable to the current LWR fuel performance under the normal operating condition. New weld techniques for joining the endplug to thin-walled cladding need to be developed. Both laser beam weld and pressure resistance weld were employed. This paper summarizes the recent progress on the laser weld development for FeCrAl cladding with 350 µm wall thickness. The results of microstructural characterization and mechanical property testing of the weld for the cladding-endplug set are discussed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NEUP Final Report: Multilayer Composite Fuel Cladding and Core Internals for LWR Performance Enhancement and Severe Accident Tolerance
Accident tolerance of LWR fuels and structures is of paramount importance, as highlighted by the accident at the Fukushima Daiichi nuclear power station. The ability of fuel cladding and core internals to resist runaway oxidation during a beyond design basis accident (BDBA), as well as to minimize corrosion during steady state operation and design basis accidents (DBAs), determines its degree of accident tolerance. A gap was identified in the accident tolerant fuel (ATF) concepts under consideration, as most consisted either of coatings which may not hold their integrity under prolonged operation, or different materials altogether which incur significant neutronic penalties. A compromise was therefore sought – by minimizing the amount of coating material while ensuring a strong microstructural bond, it was thought that a Zircaloy-steel layered composite would achieve the best of both approaches.The goal of this NEUP project was to develop a multi-metallic layered composite (MMLC) tailored to addressing accident tolerance of LWR fuel cladding and core internal structures using an innovative fabrication technology. The MMLC developed in this program is expected to enhance the accident tolerance of LWRs, thereby reducing the cost by recovering lost operating margins and/or increasing operating windows of peak cladding temperature, peak linear power, reducing steady-state corrosion, and enhancing severe accident tolerance.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Report on status of execution of SiC step document
Advanced fuel claddings made entirely or mainly of silicon carbide (SiC) ceramics and/or composites are considered very attractive elements of the accident-tolerant fuels for the light water reactors. In order to translate the promise of SiC composite materials into a reliable fuel cladding, a coordinated program of component level design and materials development must be carried out with many key feasibility issues addressed a-priori to inform the process. With the primary objective of developing a draft blueprint of a technical program that addresses the critical feasibility issues; assesses design and performance issues related with manufacturing, operating, and off-normal events; and advances the technological readiness levels in essential technology elements, a draft plan for the Systematic Technology Evaluation Program for SiC/SiC Composite Accident-Tolerant LWR Fuel Cladding and Core Structures was developed in the FY-14 Advanced Fuels Campaign of the U.S. Department of Energy’s Fuel Cycles Research and Development Program. This document summarizes the status of execution of the technical plan within the activities at the Oak Ridge National Laboratory.
36 MATERIALS SCIENCE↗
Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel
The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be. International fabrication options were explored in Europe and Asia, but this proved to be impractical, if not impossible. Consequently, experimental investigation of the Zr-Be binary system was dropped and exploration binary Zr-Y binary system was initiated. The motivation behind the Zr-Y system is the known thermodynamic stability of yttria over zirconia.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis
SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. Here, the results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses.
Accident tolerant fuel↗
Parametric Evaluation of SiC/SiC Composite Cladding with UO 2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod
SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. Here, to ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel–cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.
Accident tolerant fuel↗
Studying silicon carbide for nuclear fuel cladding
Various advanced cladding concepts are currently u
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Material selection for accident tolerant fuel cladding
Alternative cladding materials are being investigated for accident tolerance, which can be defined as >100X improvement (compared to current Zr-based alloys) in oxidation resistance in steam environments at ≥1200°C for short (≤4 h) times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. Therefore, commercial Ti 2 AlC that is not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore may be challenging to use as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation assisted Cr-rich α’ formation. The composition effects and critical limits to retaining protective scale formation at >1400°C are still being evaluated.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Evaluating the irradiation effects on the elastic properties of miniature monolithic SiC tubular specimens
The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. In conclusion, the optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.
36 MATERIALS SCIENCE↗
Elastic moduli reduction in SiC-SiC tubular specimen after high heat flux neutron irradiation measured by resonant ultrasound spectroscopy
The initial results of a post-irradiation examination study conducted on a SiC-SiC tubular specimen irradiated under a high radial heat flux are presented herein. The elastic properties of the specimen were evaluated before and after the irradiation using the resonant ultrasound spectroscopy (RUS) technique. The composite tubular specimen was considered as an orthotropic elastic with nine elastic constants (Young's moduli, shear moduli and Poisson's ratios—three components of each) for representing its full elastic deformation behavior. All the elastic moduli decreased after irradiation; the reduction was as high as 35% in one of the moduli. The significant decrease in the moduli indicates the presence of microcracks. The results from a computational study show significant stress development in the specimen due to irradiation, primarily caused by differential swelling across the thickness of the specimen. The evaluated stresses exceed the proportional limit stress of the material, indicating the likelihood of matrix microcracking, and thus corroborating the results obtained from RUS. X-ray Computed Tomography (XCT) study confirmed the presence of cracks in the irradiated specimen. These cracks occurred at the inner region of the specimen and propagated in axial and hoop directions. Finally, these XCT results are in agreement with the RUS results and stress distribution results from the computational study.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Chromium-Based Mitigation Coatings on SiC Materials for Fuel Cladding
SiC composites are undergoing assessment under the Advanced Fuels Campaign for Light Water Reactors (LWR) as an Accident Tolerant Fuel (ATF) cladding. As noted, the main concern has been the hydrothermal corrosion of SiC under all LWR chemistries, resulting in cladding oxidation and recession unacceptable for the duration of burn-up. Improvements in future fuel burn-up further increase recession mitigation requirements. Hermeticity of ceramic cladding may also be a concern in terms of quality control. Coatings are being investigated as a practical solution to address concerns regarding SiC cladding for both accident conditions and normal operations. SiC materials were coated by various technologies to produce Cr-based mitigation coatings as a corrosion barrier in LWR water chemistry. In all cases, deposition of Cr compounds requires that SiC behave as a cathode during processing, which requires surface electrical conductivity. Morphology and phase analysis of the coatings are still undergoing investigation, mainly for predicting differential swelling, thermal expansion and microcracking. Autoclave corrosion and in-pile irradiation in a PWR water loop will follow in mid-2016. (authors)
36 MATERIALS SCIENCE↗
Cold spray technology in nuclear energy applications: A review of recent advances
Cold spray technology is a promising solid-state, powder-based deposition methods for fabrication of coatings, near-net-shape manufacturing, and component repair. In recent years, this technology has been investigated for many applications in the nuclear energy sector. For example, it has been explored for the deposition of coatings of corrosion and oxidation materials on zirconium-alloys fuel claddings in light water reactors (LWR) for achieving more accident-tolerant fuel (ATF) cladding. Other examples of its use for nuclear energy applications include coatings for enhancements in wear resistance and heat transfer, near-net shape manufacturing of next-generation fuel claddings of oxide dispersion-strengthened (ODS) steels, and for mitigation and repair of potential chloride induced stress corrosion cracking in used fuel storage canister systems. This work reviews published literature on the application of cold spray process for nuclear energy applications, while pointing to gaps that have yet to be addressed.
36 MATERIALS SCIENCE↗
Ion irradiation testing of Improved Accident Tolerant Cladding Materials
This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
BISON Fuel Performance Analysis of FeCrAl cladding with updated properties
In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Hydrothermal Corrosion of SiC Materials for Accident Tolerant Fuel Cladding with and Without Mitigation Coatings
As a candidate material for accident-tolerant fuel cladding for light water reactors (LWR), SiCf–SiC composite materials possess many attractive properties. However, prior work has shown that SiC is susceptible to aqueous dissolution in LWR coolant environments. To address this issue, candidate coatings have been developed to inhibit dissolution. For this study, CVD SiC samples were prepared with Cr, CrN, TiN, ZrN, NiCr, and Ni coatings. Uncoated SiC and SiCf–SiC samples were also prepared. The samples were exposed for 400 h in 288 °C water with 2 wppm DO in a constantly-refreshing autoclave to simulate BWR–NWC. Cr and Ni coated samples lost less mass than the uncoated SiC sample, indicating an improvement in performance. The CrN coating resisted oxidation, but some of the coating was lost due to poor adhesion. The TiN coated sample gained significant mass due to oxidation of the coating. ZrN and NiCr coatings showed significant corrosion attack. SiCf–SiC ceramic matrix composite materials dissolved much faster than the CVD SiC sample, demonstrating the need for mitigation coatings if CMCs are to be used in LWRs. This work demonstrates the promise of Cr, Ni and CrN coatings for corrosion mitigation in LWRs, and shows that NiCr and ZrN are not promising coating materials.
Raiman, Stephen S.↗
Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment
Accident Tolerant Fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady state reactor operational performance. A potential ATF concept is SiC-SiC composite claddings. Fuel performance simulations were conducted on a SiC-SiC based cladding concept utilizing a multilayered approach for improved performance. This cladding concept referred to as the Duplex concept is a duplex structure composed of a monolithic SiC layer placed on the outside of a SiC-SiC composite. A liquid metal is added to fuel-cladding gap for improved heat dissipation from the fuel. The monolithic SiC layer is used to improve the coolant corrosion characteristics and protect the SiC-SiC composite layer from exposure to the coolant. The fuel performance code BISON was used to conduct fuel performance simulations on the cladding concepts. Comparisons are made with a current prototypic fuel rod design (UO 2 fuel enclosed in Zircaloy-4 cladding). Representative steady-state cases were considered for normal power and two cycle power histories. Additionally, a PCI ramp case was simulated to analyze potential anticipated operational occurrences. Transient response during a Loss of Coolant Accident and a Reactivity Initiated Accident were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard power cases evaluated. The ramping evaluations indicate potential fracturing of the SiC-SiC composite of the composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In conclusion, the rod fails early at low enthalpy for RIA but survives a LOCA with minimal material loss due to high temperature steam corrosion.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗