Mechanical Behavior of Nuclear Reactor Components: Creep, Fatigue, and Fracture
Sponsored by: TMS Materials Processing and Manufacturing Division, TMS Structural Materials Division, TMS: Nanomechanical Materials Behavior Committee, TMS: Nuclear Materials Committee
Program Organizers: Clarissa Yablinsky, Los Alamos National Laboratory; Assel Aitkaliyeva, University of Florida; Eda Aydogan, Middle East Technical University; Laurent Capolungo, Los Alamos National Laboratory; Khalid Hattar, University of Tennessee Knoxville; Kayla Yano, Pacific Northwest National Laboratory; Caleb Massey, Oak Ridge National Laboratory
Wednesday 2:00 PM
March 17, 2021
Room: RM 50
Location: TMS2021 Virtual
2:00 PM Invited
Multiscale Modeling of Creep and Transient Conditions in Steels: Application to HT9 Steel Alloy: Arul Kumar Mariyappan1; Aaron Tallman1; Christopher Matthews1; Laurent Capolungo1; 1Los Alamos National Laboratory
Structural design and certification of metallic components subjected to extreme environments (e.g. high stresses, temperatures, irradiation) relies on the predictions of the evolution of stresses, elastic and inelastic strains during service. Data scarcity creates a need for mechanistic constitutive models which can be used in regimes outside the calibration domains. Direct use of such mechanistic models for engineering scale simulations is computationally expensive. So, in this work, as an alternative, we developed a multiscale framework in three steps. First a mechanistic crystal-plasticity model is developed and validated to predict mechanical responses of materials under extreme environments. Second, a computationally efficient surrogate model (SM) is derived from crystal-plasticity model predictions for a wide range of loading conditions. Finally, the SM is integrated into a finite-element solver to simulate engineering scale components. This framework is then applied to capture transients in pressurized HT9-tubes subjected to stress cycling, thermal cycling, and thermal ramping.
2:30 PM
Creep Crack Growth Behaviour of Austenitic Stainless Steels Alloy 709 and 316H: Suyang Yu1; Jin Yan1; Hangyue Li1; Afsaneh Rabiei2; Paul Bowen1; 1University of Birmingham; 2North Carolina State University
Alloy 709, a new austenitic stainless steel, is considered as the structural material for the next-generation sodium-cooled fast neutron reactor due to its superior creep resistance. Knowledge of crack growth in this new material under creep is essential to the life prediction with the consideration of the possible fracture originating from existing flaws. To address this, creep crack growth behaviour (rates and failure mechanisms) in Alloy 709 are tested using compact tension testpieces at 550 to 750°C. 316H is also tested at 650°C as a baseline. Crack growth rates are characterized using elastic-plastic C* and linear elastic stress intensity factor. Their validity and applications in creep crack growth are discussed. Consistent with their creep properties, Alloy 709 shows a much better creep crack growth resistance than 316H. Failure mechanisms in Alloy 709 are found to depend on temperature and load, and appears to be different from 316H.
2:50 PM
Stress Corrosion Cracking Resistance of FeCrAl Alloys in Light Water Reactor Environments: Raul Rebak1; Liang Yin1; Andrew Hoffman1; 1GE Global Research
In the last decade extensive research has been conducted worldwide to determine the suitability of iron-chrome-aluminum (FeCrAl) alloys as ATF cladding of uranium dioxide nuclear fuel in light waterpower reactors. Since FeCrAl alloys have not been used in nuclear reactors, it is important to characterize their behavior in the entire fuel cycle. Stress corrosion cracking (SCC) studies were conducted for two FeCrAl alloys (APMT and C26M) using compact tension specimens in typical simulated boiling water reactor conditions at 288°C and containing either dissolved hydrogen or oxygen. Crack propagation studies showed that both ferritic FeCrAl alloys were highly resistant to SCC at stress intensities below 40 MPa√m. The current work confirms that ferritic stainless alloys are orders of magnitude more resistant to SCC than austenitic type of materials such as type 304/316 stainless steels.
3:10 PM
Enabling In-situ Crack Growth Testing and Monitoring in VTR Cartridge Loop Environments: Samuel Briggs1; Peter Beck1; Dustin Mangus1; Jake Quincey1; Andrew Brittan1; George Young1; Guillaume Mignot1; Julie Tucker1; 1Oregon State University
The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. While the conceptual design is a 300 MWth sodium-cooled fast reactor, it also incorporates self-contained cartridge loops, enabling experimentation in other advanced reactor coolant environments, including molten salts, gases, and other liquid metals. Oregon State University is supporting this program by developing techniques enabling instrumented, in-situ environmentally-assisted cracking (EAC) studies in various cartridge loop environments. To date, test facilities capable of fracture mechanics-based EAC testing in liquid sodium, molten salt, and supercritical CO2 environments have been developed. Additionally, commercially available non-destructive testing techniques, including potential drop and acoustic emission monitoring, are being adapted for use in the extreme coolant and radiation environments typified by VTR cartridge loops. A general overview of these efforts and initial results will be presented.
3:30 PM Invited
In-situ Scanning Electron Microscopic Observation of Creep and Creep-fatigue of Alloy 709: Amrita Lall1; Rengen Ding2; Paul Bowen2; Afsaneh Rabiei1; 1North Carolina State University; 2University of Birmingham
Alloy 709 is a 20Cr-25Ni advanced austenitic stainless steel developed as an improvement over the existing advanced austenitic stainless steels. The alloy’s high Ni content provides increased austenite stability, while its high Cr content improves its corrosion resistance at extreme environments of nuclear structures. In this study, in-situ scanning electron microscope (SEM) tensile, creep and creep-fatigue tests at various temperatures from room temperature to 1000 °C will be reported. Electron backscatter diffraction (EBSD) and Energy Dispersive X-ray spectrometry (EDS) were used to observe the microstructural evolution and phase change during the in-situ heating and loading at different temperatures and strain rates and identify the dominant deformation mechanisms in each environmental condition.
4:00 PM
Mechanical Characterization of Neutron Irradiated HT-9 Heats (ORNL, LANL and EBR II) at LWR and Fast Reactor Relevant Temperatures: Ramprashad Prabhakaran1; Mychailo Toloczko1; Kumar Sridharan2; 1Pacific Northwest National Laboratory; 2University of Wisconsin-Madison
9-12 weight% ferritic-martensitic steels such as HT-9 are being considered as candidate structural materials for fast, advanced LWR, and fusion reactors due to their excellent resistance to radiation-induced void swelling, good irradiation creep properties, microstructural stability, and thermal conductivity. However, the extreme hardening and low fracture toughness that occur at irradiation temperatures below 430°C is a serious issue. To address this concern, systematic investigations on the mechanical behavior and microstructure of HT-9 with slight variations in chemical composition and heat treatment are needed over a wide range of doses and temperatures to better understand the source of hardening. Mechanical characterization was performed on three HT-9 heats with different processing conditions irradiated to 3 dpa and 6 dpa (Advanced Test Reactor, Idaho) at 291°C, 360°C and 431°C to understand the effects of radiation damage, and to provide insight on desirable HT-9 processing conditions.
4:20 PM
Burst Behavior of Accident Tolerant Fuel Cladding Concepts under Simulated Loss-of-coolant Conditions: Samuel Bell1; Bruce Pint1; Ken Kane1; 1Oak Ridge National Laboratory
Mitigating the dangers of a loss of coolant accident (LOCA) in a nuclear reactor has been a focus of the nuclear community since the 2011 Fukushima Daiichi disaster. Several accident tolerant alternatives have been proposed to address the oxidation related risks of the incumbent zirconium alloys, including replacement by ferritic iron-chromium-aluminum “FeCrAl” alloys or the application of a metallic oxidation resistant coating. Burst testing was conducted on first (T35Y2) and second (C26M) generation FeCrAl alloys, Zircaloy-2, Zirlo, and Cr coated zirconium claddings in a simulated LOCA environment. Mechanical characterization and general oxide analysis of all cladding concepts was conducted. Second generation FeCrAl showed tremendous improvements in burst strength compared to the other claddings. Cr coatings offered no improvement in burst strength of the underlying zirconium-based alloys but did significantly decrease oxide formation. This work was funded by the Advanced Fuels Campaign, Office of Nuclear Energy, U.S. Department of Energy.
4:40 PM
C-ring Compression of SiC-SiC Cladding at 1200°C with In-situ X-ray Computed Micro-tomography: Dong Liu1; Jon Ell2; Guanjie Yuan1; Peng Xu3; Roger Lu4; Edward Lahoda4; Harold Barnard2; Dula Parkinson2; Robert Ritchie2; 1University of Bristol; 2Lawrence Berkeley National Laboratory; 3Idaho National Laboratory; 4Westinghouse Electric Company
In this work, SiCf-SiC accident tolerant cladding materials were examined using a unique hot cell to perform C-ring compression tests coupled with real-time X-ray computed micro-tomography (XCT) to image cracks formation and propagation through the surface coating and composite substrate at ambient and 1200°C. Two types of cladding with different coating thicknesses were examined. Their mechanical behavior was found to be very different which indicated a need for optimizing the thickness of the surface coating and composite substrate for both strength and damage-tolerance. During the experiments, an XCT scan was collected at each loading step until final failure, with digital volume correlation methods utilized to identify 3D strain concentrations. Incipient cracks were segmented for both types of samples, with their 3D distribution and interactions with the composite substrate characterized. Differences in the maximum local failure strains and their distributions in the two types of samples will be discussed.