Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials: Accident Tolerant Fuels and Advanced Characterization
Sponsored by: TMS Structural Materials Division, TMS: Mechanical Behavior of Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Dong Liu, University of Oxford; Peng Xu, Idaho National Laboratory; Simon Middleburgh, Bangor University; Christian Deck, General Atomics; Erofili Kardoulaki, Los Alamos National Laboratory; Robert Ritchie, University of California, Berkeley
Tuesday 8:00 AM
March 1, 2022
Room: 204A
Location: Anaheim Convention Center
Session Chair: Alexander Leide, United Kingdom Atomic Energy Authority; Joshua Kane, Idaho National Laboratory; Takaaki Koyanagi, Oak Ridge National Laboratory; Kumar Sridharan, University of Wisconsin-Madison
8:00 AM
Advanced Characterization and Multiscale Testing for SiC Ceramic Matrix Composite Cladding as Accident Tolerant Fuel Candidate Materials for LWR Applications: Peng Xu1; David Frazer1; Tsvetoslav Pavlov1; Nikolaus Cordes1; Fabiola Cappia1; David Kamerman1; Sean Gonderman2; Christian Deck2; Jack Gazza2; 1Idaho National Laboratory; 2General Atomics
SiC cladding is pursued by the nuclear industry as a candidate material for accident tolerant fuels for the light water reactors (LWRs). It offers superior oxidation resistance and strength retention in high temperature steam during severe accidents. Although the irradiation behavior of SiC has been well studied in the past, the irradiation data is still lacking for SiC ceramic matrix composite (CMC) as fuel cladding material at prototypic LWR conditions. The Idaho National Laboratory is partnering with General Atomics to perform test reactor irradiation at the Advanced Test Reactor (ATR) at the prototypical Pressurized Water Reactor conditions to evaluate its in-pile performance. The SiC composite cladding materials will be examined before and after irradiation. Multiscale characterization and testing will be conducted to correlate radiation induced microstructure changes to its thermophysical and mechanical properties. Cladding integrity and corrosion will be evaluated non-destructively and destructively.
8:20 AM
Fiber/Matrix Debonding of SiC/SiC Composites Evaluated Using the Micropillar Compression: Omer Karakoc1; Takaaki Koyanagi1; Takashi Nozawa2; Yutai Katoh1; 1Oak Ridge National Laboratory; 2National Institutes for Quantum and Radiological Science and Technology
Micro-scale mechanical testing was employed to evaluate carbon interphase properties of silicon carbide fiber-reinforced SiC matrix (SiCf/SiCm) composites that play a key role in the overall mechanical behavior of the composite. To elucidate the optimum composite performance, we measured the interfacial mechanical properties of as-fabricated SiC/SiC composites with different fiber/matrix interface microstructures. Interface properties were determined using the slanted interface micro-pillar compression test, selecting micro-pillar specimens contained inclined carbon interphase interfaces. The interface properties and fracture behavior were explained as dependent on the microstructure of the interfaces as characterized by TEM and Raman spectroscopy. In addition, the micropillar test results were correlated with macroscopic mechanical properties of SiC/SiC composites neutron-irradiated at 300–500°C to ~10 displacements per atom in the High Flux Isotope Reactor. The analysis suggests that radiation effects, rather than interface debonding strength in the as-fabricated condition, are the dominant factor explaining radiation-induced degradation of the macroscopic properties.
8:40 AM
Strain Rate Sensitivity Studies of Commercial FeCrAl Alloy: Hamdy Abouelella1; Chengying Xu1; Korukonda Murty1; 1North Carolina State University
Since Fukushima nuclear accident in Japan 2011, FeCrAl Alloys have been considered as one of the accident tolerant fuel-cladding and structural materials for their superior oxidation resistance. While such oxidation process caused severe degradation in zirconium based clads at high temperatures, protective aluminum oxide layer in FeCrAl prevents further corrosion attack. Next-generation nuclear reactors are projected to work at higher operating temperatures and stresses. Therefore, the structural materials should have adequate mechanical properties that can withstand typical operating conditions expected in the nuclear reactor core. Since FeCrAl Alloy is a potential candidate as structural material to replace Zircaloys, the mechanical behavior of commercial FeCrAl Alloy is investigated at strain rates from 10^-3 to 10^-5 and at four different temperatures 550C,600C,650C,700C . Strain rate sensitivity parameter and the activation are evaluated for an understanding of the basic underlying deformation mechanism(s). This research is supported by DOE/NEUP project DE-NE0008874.
9:00 AM
Impact of Lithium Accommodation on Defect Chemistry in ZrO2: Gareth Stephens1; Yan Ren Than2; 1Nuclear Futures Institute Bangor; 2National University of Singapore
Identifying the mechanism by which Li accelerates zirconium alloy corrosion will allow new alloying additions to be considered and new water chemistry regimes to be investigated, improving the efficiency and performance of future nuclear power reactors, reducing the cost of operation and design.Density functional theory (DFT) was used to identify the most stable accommodation mechanisms for Li in ZrO2 at an atomic scale through the binding properties of valance electrons. A Brouwer diagram has been developed using Fermi-Dirac statistics that predicts the nature of the defect structures and their competing species concentrations. This was then combined with experimental data to corroborate the most stable accommodation mechanisms of Li in ZrO2. The solubility of Li in bulk ZrO2 is predicted to be low indicating that accelerated corrosion due to bulk Li accommodation is unlikely.
9:20 AM
In Situ Study of High Temperature Mechanical Behavior of Irradiated FeCrAl Alloys: Tianyi Sun1; Tongjun Niu1; Dongyue Xie2; Adam Gabriel3; Lin Shao3; Jian Wang2; Haiyan Wang1; Xinghang Zhang1; 1Purdue University; 2University of Nebraska-Lincoln; 3Texas A&M University
FeCrAl alloy is one of the near-term cladding materials in accident tolerant fuel systems. The improved high-temperature steam oxidation resistance of FeCrAl alloy compared with Zr alloy leads to a longer responding time and less hydrogen generation under accident conditions. With growing interest in FeCrAl alloy for nuclear applications, the mechanical behavior of FeCrAl alloy at high temperatures and irradiation states is of great importance. In this study, we used in situ SEM micropillar compression technique to investigate the mechanical behavior of both annealed and irradiated FeCrAl alloy at various temperatures, ranging from 200 – 500 °C. The temperature-dependent deformation behavior is quantified and discussed.
9:40 AM Break
10:00 AM
Accelerating Advanced Fuel Development and Analysis by Combining Modelling and Experiment: Simon Middleburgh1; Phylis Makurunje1; Fabio Martini1; Mustafa Bolukbasi1; Lee Evitts1; Dave Goddard2; Antoine Claisse3; William Lee1; Nicholas Barron2; 1Bangor University; 2National Nuclear Laboratory (NNL); 3Westinghouse Electric Sweden AB
A range of advanced technology fuels (ATFs) are now being considered that aim to improve operational characteristics, such as residence times and reduction in center-line temperatures, with accident tolerance providing an overall economic and viable fuel candidate for both light water reactors and reactors currently under development, including lead-cooled fast reactors. Here we present a range of developments being undertaken to support ATF development with a focus on the manner in which modelling is guiding experiment and how experiment dictates the modelling required to develop a new fuel candidate - here the focus shall be on a UO2-based composite fuel system. Atomic scale simulations as well as neutronic modelling has been performed on the UO2-based composite, whilst manufacture and design of the fuel has been progressing based on what we have predicted - including the use of novel manufacturing methods based on spray drying.
10:20 AM
NOW ON-DEMAND ONLY - Assessment of Local Deformation Behavior in Mesoscale Tensile Specimens via Digital-image Correlation: Yachun Wang1; David Frazer1; Daniel Murray1; Geoffrey Beausoleil1; Mahmut Cinbiz1; 1Idaho National Laboratory
Advanced nuclear reactor development and demonstration goals challenge the qualification of in-core fuel cladding and structural materials due to material availability and dose limitations of nuclear facilities. Therefore, focused-ion beam (FIB) milled subsize specimens may offer a pathway to assess the advanced nuclear reactor materials’ mechanical performance. One challenge with mesoscale tests is to map local strain distribution on the sample. Surface metrology on using scanning electron microscopy (SEM) while testing and subsequent strain calculations via digital image correlation (DIC) offer an accurate local strain examination on the FIB-milled samples. In this study, computer-generated speckle patterns were applied on the FIB-milled surrogate aluminum and Zircaloy-4 samples using focused electron beam (FEB) deposition process resulting high-contrast nano-scale speckle patterns. In-situ SEM tensile testing, and DIC analysis were performed to investigate local deformation and its effect on the failure behavior. The effect of the randomness of the speckle patterns were also investigated.