Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Structural Materials IV
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory; Raul Rebak, GE Global Research; Clarissa Yablinsky, Los Alamos National Laboratory
Thursday 8:30 AM
March 2, 2017
Room: Point Loma
Location: Marriott Marquis Hotel
Session Chair: Raul Rebak, GE Global Research; Thak Sang Byun, Pacific Northwest National Laboratory
Grain Boundary Damage Precursors Leading to Intergranular SCC Initiation of Cold-Worked Alloy 600 and Alloy 690 in PWR Primary Water: Ziqing Zhai1; Mychailo Toloczko1; Stephen Bruemmer1; 1Pacific Northwest National Laboratory
Stress corrosion crack (SCC) initiation of cold-worked (CW) alloy 600 and alloy 690 materials was assessed in 360°C simulated PWR primary water using constant load tensile tests instrumented for in-situ detection of initiation by direct current potential drop (DCPD) technique. Fundamental differences in grain boundary damage evolution were observed between the two Ni-base alloys. Intergranular (IG) attack and coalescence of short cracks occurred in short exposure times for the four CW alloy 600 heats promoting detection of cracking by DCPD in 200-2000 hours. In contrast, SCC initiation was not detected by DCPD in any of the alloy 690 specimens (6 heats) after ~10000 hours of exposure. However, small creep cavities were discovered at grain boundaries in certain highly CW heats and led to the formation of shallow IG surface cracks. The mechanisms controlling grain boundary damage evolution and the transition to SCC initiation will be discussed.
Mechanical Property Measurements of a New Metal Matrix Material for Nuclear Reactor Applications: Donna Guillen1; Mychailo Toloczko2; Anthony Guzman2; Ramprashad Prabhakaran2; Jesse Willett2; 1Idaho National Laboratory; 2Pacific Northwest National Laboratory
Microhardness indentation and tensile tests were performed on unirradiated and irradiated samples of a new metal matrix composite material developed to provide fast neutron flux test capability in the Advanced Test Reactor. Hafnium aluminide (Al3Hf) intermetallic particles were mixed with aluminum powder and hot pressed to produce material with 20.0, 28.4 and 36.5% Al3Hf by volume. Specimens were irradiated for up to four reactor cycles where they experienced average temperatures up to 348 K and doses up to ~4 dpa. Material hardness increases with dose and plateaus at around 2 dpa. Samples with a higher Al3Hf vol% exhibit higher hardness since the intermetallic particles contribute to an increased strength of the material. The tensile tests show that these materials exhibit ductile fracture when tested at 200°C.
Microstructural Evolution of Thermal Recovery in Ti3AlC2-Ti5Al2C3 and Ti3SiC2: Caen Ang1; Chad Parish1; Chunghao Shih2; Steven Zinkle3; Yutai Katoh1; 1Oak Ridge National Laboratory; 2Oak Ridge National Laboratory; General Atomics; 3University of Tennessee
Ti3AlC2 -Ti5Al2C3 and Ti3SiC2 were irradiated with neutrons a dose of ~2 dpa at temperatures of 400 and 700 °C. Typical of HCP ceramics, anisotropic c-axis expansions of 3.1% and 1.5% were recorded at ~400°C. Fracture surfaces indicate the inability to form bridging ligaments and delamination morphologies. However, at ~700°C, the dimensional stability, mechanical and thermal properties – appear to be maintained due to thermal mobility of defects. Thermophysical data from annealing show the recovery of electrical conductivity and thermal diffusivity from ~500°C, indicating the stability of the Al and Si layers. No cavities were found at ~700°C. The consistent percolation of defects with a Burger’s vector x<0001> requires further investigation due to the implications of radiation-induced TiC(1-x) crystallography (suggesting reduced radiation resistance). Defect growth may also indicate anisotropic grain coarsening. It is shown that early stages of defect accommodation occurs via both interstitial and antisite planar defects, the latter being implied by HAADF images. At least three types of planar defects are proposed to explain these observations.
Microstructural Characterization of AA6061-AA6061 HIP Bonded Cladding Interface: Abhishek Mehta1; Le Zhou1; Dennis Keiser2; James Cole2; Yongho Sohn1; 1University of Central Florida; 2Idaho National Laboratory
In Materials Management and Minimization Reactor Conversion Program, AA6061 is employed as a cladding to encapsulate the Zr-laminated monolithic U-10Mo fuel with hot isostatic pressing (HIP), carried out at 560ºC for 90 minutes with a pressure of 104 MPa. While the fuel/Zr and Zr/AA6061 interfaces have been characterized, the interface between AA6061 and AA6061 (e.g., cladding bond line) has drawn attention, because it was documented to be susceptible to cracking during reactor application. Microstructural characterization of AA6061-AA6061 HIP bonded cladding interface was conducted using SEM equipped with XEDS and TEM capable of EELS. The interface consisted of discontinuous layer of Mg2Si precipitates and traces of oxide-dispersion. In addition, various fine precipitates, mainly rich in Al, Fe and Si, were also observed. Quantitative microscopy was carried out to measure the linear density of Mg2Si at the interface, and correlated to variation in starting AA6061 characteristics and HIP processing parameters.
Wear Study Comparison of Accident Tolerant FeCrAl Cladding, Zircaloy-2 and SS304 against X750: Raghunath Kanakala1; Christian Williams1; Sobhan Patnaik1; Raul Rebak2; 1University of Idaho; 2GE Global Research
Fuel rods in a nuclear reactor are kept aligned by grids made of X-750 Alloy. The vibration of the rods against the X-750 grids or against debris caught in the grids cause fretting/wear damage on the cladding material. Current work studies the wear behavior of the proposed new accident tolerant cladding material Iron-chromium-aluminum (FeCrAl). Samples with three different cladding materials are tested against the standard sample (for this testing) X-750 alloy which is the material for making grids, like FeCrAl, Zircaloy-2, and SS304. From our current study FeCrAl shows significantly superior wear resistance compared to Zircaloy-2 and SS304 at elevated temperatures of 300oC.
10:10 AM Break
Creep-Fatigue Deformation of 9Cr-1MoV Steel and Weldments: Harrison Whitt1; Tyler Payton1; Wei Zhang1; Michael Mills1; 1The Ohio State University
Creep resistant properties of 9Cr-1MoV steel and weldments make them excellent candidate materials for nuclear reactor components. In order to create more accurate life prediction models, a better understanding of the creep-fatigue deformation behavior is required for 9Cr-1MoV base metal and welded components. Using a force-controlled, dwell-fatigue testing apparatus the effects of dwell time, temperature, and applied load on the deformation properties of 9Cr-1MoV materials are studied. 9Cr-1MoV specimens were found to exhibit inelastic recovery during load-controlled creep-fatigue testing. Scanning electron microscopy (SEM) and electron-backscattered diffraction (EBSD) were used to study the microstructure of 9Cr-1MoV weldments after accumulating creep-fatigue damage with special emphasis on the fine-grained heat-affected zone (FGHAZ). Advanced characterization techniques including scanning transmission electron microscopy (STEM) and transmission kikuchi diffraction (TKD) were used to examine the effects of creep-fatigue damage on the microstructure of 9Cr-1MoV specimens including substructure coarsening, changes in dislocation density and development of microstructural inhomogeneity.
Atomic Scale Behavior of Beryllium in Zirconium: Abhinav Jain1; Dallas Trinkle1; 1University of Illinois
High-temperature oxidation of zircalloys poses a serious safety risk for nuclear fuel cladding applications, thus driving the search for oxidation resistant alloys. Ellingham diagrams suggest preferential oxidation of beryllium over zirconium, but the atomic-scale behavior of Be in the bulk Zr and near the surfaces is unknown. We perform first principle calculations using density functional theory to investigate the stability of Be at possible sites in Zr bulk, as well as the basal and prismatic surfaces. We also use nudged elastic band to compute energy barriers for atomic transport and kinetic Monte Carlo to compute Onsager transport coefficients for Be in Zr. We find that Be favors substitutional sites in the bulk, and preferentially segregates to surface substitutional sites due to charge redistribution. This data can be used with mesoscale modeling to quantify the kinetics of segregation from solid solution or precipitate dissolution.
The Role of Stoichiometry on Ordering Phase Transformations in Ni-Cr Alloys for Nuclear Applications: Fei Teng1; Julie Tucker1; Benjamin Spencer2; Larry Aagesen2; Yongfeng Zhang2; Pritam Chakraborty2; Octav Ciuca3; Grace Bruke3; Emmanuelle Marquis4; Mukesh Bachhav4; 1Oregon State University; 2Idaho National Laboratory; 3University of Manchester; 4University of Michigan – Ann Arbor
Mechanical property degradation due to isothermal ageing is of potential concern for alloys based on the Ni-Cr binary system (e.g., Alloys 625 and 690), particularly in nuclear power applications. The disorder-order phase transformation, which is the primary source of embrittlement, has been studied in Ni-Cr model alloys by a combined experimental and computational approach. Kinetic and Grand Canonical Monte Carlo simulations, based upon density functional theory calculations, are used to study thermodynamic and kinetic aspects of the phase transformation. A corresponding model is being built in Grizzly code, which is based on Multiphysics Object-Oriented Simulation Environment (MOOSE), for lifetime predictions. Model alloys with different stoichiometries have been isothermally aged up to 10,000 hours and characterized via nanoindentation, atom probe tomography, neutron diffraction and transmission electron microscopy. Experimental measurements in hardness and phase fraction as a function of aging time and temperature are discussed in order to assess the model accuracy.
Peuget: How Ion Beam Irradiations Simulate the Radiation Aging of Nuclear Glass: Sylvain Peuget1; 1CEA
In order to ensure the long term structural integrity of the glass used for the conditioning of the high level nuclear wastes, it is important to find the best methods to simulate their irradiation aging. Ion beam irradiation is a very convenient way to accelerate the aging of a material, but the optimization of the irradiation conditions needs to compare the behavior of ion beam irradiated materials with the one of radioactive material that are self-irradiated by radionuclide’s decays. In this study we compare the modification of nuclear glass structure generated by ion beam irradiations with different ions, in either mono or double beam mode, to the one induced by 244 Cm alpha-decays. It is shown that double beam irradiations (simultaneous or sequential) better simulate the aging under alpha decay irradiation, because both components of the alpha decay (recoil nuclei and alpha particle) have a specific effect on the glass structure.