Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Structural Materials III
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

Wednesday 2:00 PM
March 1, 2017
Room: Cardiff
Location: Marriott Marquis Hotel

Session Chair: Kumar Sridharan, University of Wisconsin; Walter Luscher, Pacific Northwest National Laboratory

2:00 PM  
Seamless Thin-wall Tube Production of ATF Wrought FeCrAl Alloys: Yukinori Yamamoto1; Sun Zhiqian1; Maxim Gussev1; Kevin Field1; Bruce Pint1; Lance Snead2; Stuart Maloy3; Kurt Terrani1; 1Oak Ridge National Laboratory; 2Massachusetts Institute of Technology; 3Los Alamos National Laboratory
    Nuclear grade wrought FeCrAl alloy is currently under development as one of the candidate materials for accident tolerant fuel cladding in light water reactors. In addition to the alloy design efforts of nuclear grade FeCrAl alloys incorporating the improved oxidation and corrosion resistance, mechanical properties, irradiation characteristics, and weldability, a parallel effort has been devoted to produce seamless thin-wall tubes of the FeCrAl alloys through commercially available tube production technology such as tube drawing process. Because of limited deformability nature of FeCrAl alloys with strong dependence on minor alloying additions, several key factors have been discussed; proper alloy composition ranges including minor alloying additions, microstructure control prior to tube drawing process, selection of area reduction per pass, inter-pass annealing conditions, and so on. The process and compositional windows of wrought FeCrAl alloy tube production together with the quality evaluation results of the drawn tubes will be presented.

2:20 PM  
Charged Particle Irradiation Studies of High Dose Precipitation in Reactor Pressure Vessel Steels: Nathan Almirall1; Takuya Yamamoto1; Peter Wells1; G. R. Odette1; Nicholas Cunningham1; Soupitak Pal1; Scott Tumey2; Keith Williams3; Tim Williams3; 1University of California Santa Barbara; 2Lawrence Livermore National Laboratory; 3Rolls Royce
    Nuclear reactor pressure vessel service life may be limited by nano-scale precipitates and other hardening features which increase the yield strength of steels resulting embrittlement, characterized by ductile to brittle transition temperature increases. Great strides have been made in developing of predictive embrittlement models that are applicable up to intermediate neutron fluence. However, even higher fluence are experienced over extended life. Neutron irradiations are costly and time consuming, often requiring many years to complete. While they do not directly simulate neutrons, charged particle irradiations (CPI) produce 80-year RPV dpa in hours, with a high degree of control over the other irradiation variables. Building on earlier studies this presentation focuses on APT and nanohardness characterization of 70MeV ion irradiations induced precipitation of nm-scale MnNiSi(Cu) phases in a RPV steel matrix that covers a very wide range of compositions. The accelerated CPI data are compared to lower dose rate neutron irradiation results.

2:40 PM  
Effect of Different Processing Routes on the Microstructure and Texture of 14YWT Alloy: Soupitak Pal1; Ershadul Alam1; G Odette1; Stuart Maloy1; David Hoelzar1; John Lewandowski1; 1University of California Santa Barbara
    Thermo-mechanical processing of an extruded and cross rolled plate of a 14YWT nanostructured ferritic alloy produces severe microcracking due to its highly anisotropic microstructure and strong <110>-fiber texture. However, subsequent hot hydrostatic tube extrusion heals the microcracks and alters the texture. The mechanisms leading first to microcracking and subsequent microcrack healing are described in terms of texture evolution, dislocation density, and mobility, and formation of special dislocation boundaries. Our results show that a texture component perpendicular to the deformation direction is key to defect-free processing of 14YWT tubes. The role of the applied stress state on the development of favorable shear texture component of {112}<110> and {111}<110> is discussed based on direct observation of the dislocation substructure based on both TEM and self-consistent viscoplastic modeling.

3:00 PM  
Impact of the Neutron Irradiation on the Structure and Properties of the 6061 Al Alloy Produced by Ultrasonic Additive Manufacturing: Maxim Gussev1; Kurt Terrani1; Chad Parish1; Aaron Selby1; Niyanth Sridharan1; Dana McClurg1; Zachary Thompson1; Mark Norfolk2; Sudarsanam Babu3; 1Oak Ridge National Laboratory; 2Fabrisonic LLC; 3University of Tennessee
    Ultrasonic Additive Manufacturing (UAM) is a promising technique to produce complex shaped parts including in-reactor components; however, the radiation tolerance of the UAM-produced parts has to be verified. In the present work, UAM-produced specimens of 6061 aluminum alloy were irradiated in the High Flux Isotope Reactor (HFIR, ORNL). Prior irradiation, part of the specimens were subjected to post-welding anneal attempting to improve the properties. The post-irradiation investigation was carried out using tensile testing with digital image correlation (DIC), FIB-TEM, SEM, and EBSD. The impact of radiation on the mechanical properties and structure was discussed. Portevin-Le-Chatelier phenomenon (serrated flow in the tensile curves and deformation band appearance) was also explored in detail using DIC. The main focus was made on the performance of the welding interfaces, their strength, and local structure variations. UAM-specimens demonstrated pronounced radiation hardening, but ductility changes were complex due to high property anisotropy of the UAM-material.

3:20 PM  
Creep Fatigue Crack Growth of T91: Test Design and Data Analysis: Marta Serrano1; Rebeca Hernandez Pascual1; Mercedes Hernández-Mayoral1; 1CIEMAT
    Structural material candidates for Gen IV reactors operating at high temperature are susceptible to creep damage effects in quasi-static loadings and to creep-fatigue damage effects in cyclic loadings. Following a fracture mechanics approach, their life assessment is based on knowledge of operating condition, accurate stress analysis and relevant high temperature material data. In this paper creep-fatigue crack growth tests of two different T91 heats at 625ºC and different hold time are presented. Test design is described and the application of the ASTM E-2760-10 is discussed. Some of the data were part of the ASTM/EPRI RR in Support of Standard Test Method for Creep-fatigue Crack Growth Testing. The research leading to these results is partly funded by the Euratom FP7/2007-2013 under grant agreement No. 604862 (MatISSE project) ) and in the framework of the EERA (European Energy Research Alliance) Joint Programme on Nuclear Materials

3:40 PM Break

4:00 PM  
Property Evolution Due to Thermal Aging of Cast Duplex Stainless Steels As Measured by Multi-Scale Mechanical Methods: Samuel Schwarm1; Sarah Mburu1; R. Prakash Kolli1; Carl Cady2; Stuart Maloy2; Sreeramamurthy Ankem1; 1University of Maryland, College Park; 2Los Alamos National Laboratory
    Extending the operational life of cast duplex stainless steel piping in light water reactors to 80 years requires an improved understanding of microstructural evolution and corresponding changes in mechanical behavior during service. We are investigating the effects of thermal aging on the mechanical properties of cast CF–3 and CF–8 stainless steels. Bulk mechanical tests have been performed to measure properties such as tensile strength, impact energy, and ductility during aging embrittlement. The results show an increase in strength and decrease in ductility and impact energy after aging to 12,900 hours. Smaller length scale tests, such as instrumented nanoindentation, reveal the effects of aging on local properties of the constituent ferrite and austenite phases. The measured values are utilized to evaluate the influence of microstructural changes on thermal aging embrittlement of the steels. This work is supported by the DOE Nuclear Energy University Programs (NEUP), contract number DE-NE0000724.

4:20 PM  
Microstructural Heterogeneity of Deformed and Annealed FeCrAl Alloys with Nb Addition: Zhiqian Sun1; Philip Edmondson1; Yukinori Yamamoto1; 1Oak Ridge National Laboratory
    Nuclear-grade wrought FeCrAl alloys are currently under development as promising accident-tolerant fuel cladding materials for light water reactors due to their superior oxidation and corrosion resistance in high-temperature steam environments when compared with commercial Zr-based alloys. Small addition of Nb in FeCrAl alloys is known to promote the Fe2Nb-type Laves phase precipitation in bcc-Fe matrix, which improves the high-temperature performance of FeCrAl alloys. In the study, the microstructures and mechanical properties of deformed Nb-containing FeCrAl alloys before and after annealing at 600-900°C were investigated. Deformation heterogeneity as well as certain texture formation was found after applying warm-rolling at 300°C, which resulted in non-uniform, various types of microstructural evolutions (recovered/recrystallized grains and Lave phase particle distribution) during annealing. Based on the results including mechanical property evaluation, process optimization of tube production of Nb-containing FeCrAl alloys was discussed in conjunction with various microstructural factors to balance their high strength with tube fabricability.

4:40 PM  
Complex SiC-SiC Composite Structures for Nuclear Applications: Ekaterina Novitskaya1; Hesham Khalifa2; Alexander Kritsuk1; Olivia Graeve1; 1University of California, San Diego; 2General Atomics, Corp.
    SiC fiber-reinforced SiC matrix (SiC-SiC) composites are attractive materials for advanced high-temperature reactor components due to their ability to retain essential thermo-mechanical properties in the presence of high neutron fluence, molten salts, and high temperatures. In this study, we report a detailed study of complex SiC-SiC composites, produced by a transient eutectic phase (TEP) hot-press consolidation process. Samples with varying amounts of reinforcing fiber plies and geometries were analyzed by microhardness measurements to evaluate the local mechanical properties of the composites with respect to structural and geometrical differences. The microhardness of both matrix and fibers of the samples was found to be independent of the structural and geometrical features. SEM characterization verified that samples with three fiber plies and two holes 20 mm apart, demonstrate the most uniform internal microstructure as well as sintering aid distribution. These findings are of special interest to the future production of SiC-SiC complex composites.

5:00 PM  
Effects of Ion-irradiation Damage on Mechanical Behavior in Silicon Carbide: David Armstrong1; Helen Pratt1; Steve Roberts1; Yevhen Zayachuk1; 1University of Oxford
     Silicon Carbide and Silicon Carbide composites are being studied as potential materials in multiple nuclear applications ranging from accident tolerant fuel cladding to structural materials in fusion breeder blankets; however, data on the effects of irradiation on their mechanical behaviour is sparse. To simulate neutron damage, silicon ion implantation was performed on direct sintered silicon carbide to a peak dose of ~3.8 dpa with damage depth of ~1.4μm. Nanoindentation, Raman spectroscopy and TEM were used to study the effects of irradiation.The irradiation damage increases the hardness from 38GPa to 43GPa. The elastic modulus was seen to decrease, from 471GPa 430GPa; this is thought to be due to lattice swelling. Raman spectroscopy and TEM showed that the damaged layer had become amorphous during irradiation. The implications of this for the use of SiC for nuclear applications in both fusion and fission reactors will be discussed.

5:20 PM  
Study on the Microstructure and Mechanical Behavior of the New Type SA508-IV Reactor Pressure Vessel (RPV) Steel by Different Methods: Xue Bai1; Sujun Wu1; Peter K. Liaw2; Lin Shao3; 1Beihang University; 2The University of Tennessee, Knoxville; 3Texas A&M University
    A new type SA508-IV reactor pressure vessel (RPV) steel, containing high content of Ni and just 0.041 wt.% Cu, underwent the thermo-mechanical embrittlement processing (TMEP) deterioration process. The as-received state microstructure reveals that Cr and Mn atoms replace Fe and form alloying cementites, (Fe,Cr)3C and (Fe,Mn)3C, through in-situ nucleation. Due to the slower diffusion coefficient, Cr precipitates in the outer layer of the Mn clusters. In the subsequent embrittlement process, needle-shaped Mo2C, fine copper-rich clusters (CRCs) and P-rich precipitates are formed. The mechanical test results showed that the DBTT fits a linear function of the square root of embrittlement time at 520 ºC (t 1/2) from 10h to 90h degradation and the degree of embrittlement reaches saturation after 90h. Furthermore, this SA 508-IV RPV steel was irradiated at 290ºC to 3 and 30 dpa with 3.5 MeV Fe2+ ions to study the cascade effect similar to that caused by neutrons.