Mechanical Behavior of Nuclear Reactor Components: Processing Effects
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

Monday 8:30 AM
March 15, 2021
Room: RM 50
Location: TMS2021 Virtual


8:30 AM  Invited
Development of Modified 3Cr-3WVTa Base Bainitic Steels for Fusion Structural Applications: Yukinori Yamamoto1; Roger Miller1; Arthur Rowcliffe1; 1Oak Ridge National Laboratory
    Development of a modified 3Cr-3WVTa base bainitic steel is currently in progress, targeting lifetime structural component applications in next generation fusion devices such as the helium cooled vacuum vessels operating up to 450°C and the blanket support structures up to 550°C. Major goal is to achieve high-temperature mechanical properties comparable or superior to existing commercial bainitic or ferritic-martensitic (F-M) steels together with no requirement of a post-weld heat treatment (PWHT), to allow a significant cost/time reduction of large-scale component construction and eliminate the design limitation. Alloy modification focused on reducing inhomogeneous cross-weld mechanical properties without PWHT, which was successfully achieved through a design strategy to increase the hardenability and reduce the hardness in as-normalized condition, by tailoring minor alloying additions. The alloy selection and property evaluation will be discussed. Research sponsored by the U.S. Department of Energy, Office of Fusion Energy Sciences, under contract DE-AC05-00OR22725 with UT-Battelle, LLC.

9:00 AM  
Low Temperature Neutron Irradiation and Mechanical Properties of Welded AISI 347: Lauren Garrison1; Nathan Reid1; John Echols1; Kaustubh Bawane1; 1Oak Ridge National Laboratory
    AISI 347 stainless steel has many beneficial properties including general corrosion resistance and weldability, so it is widely used in industrial applications including as pressure vessels and for chemical processing equipment. It has been used for some nuclear applications, but there is limited experience with low temperature (<100°C) neutron irradiation behavior. Alloy 347 from three suppliers was compared as base metal, tungsten inert gas welded, and flux cored arc welded. The three suppliers’ materials had different grain sizes and impurity content. For the welded materials, samples were cut to either focus on the fusion zone or the heat affected region. Base metal and welded samples were irradiated in the High Flux Isotope Reactor to fluences of approximately 0.08 and 0.8 dpa at the ambient coolant temperature of approximately 60°C. The effects of the different supplier starting microstructure, welding, and neutron irradiation were evaluated through microstructure analysis, tensile, and Charpy tests.

9:20 AM  
Neutron Irradiation Response of SA508 Pressure Vessel Steel Prepared by Powder Metallurgy and Hot Isostatic Pressing: Yangyang Zhao1; Caleb Clement1; Shujuan Wang2; Yaqiao Wu2; Katelyn Wheeler3; Donna Guillen3; David Gandy4; Janelle Wharry1; 1Purdue University; 2Boise State University, Center for Advanced Energy Studies; 3Idaho National Laboratory; 4Electric Power Research Institute
    The objective of this study is to investigate neutron irradiation effect on microstructure and mechanical properties of SA508 pressure vessel steel fabricated by powder metallurgy and hot isostatic pressing (PM-HIP). Alloys produced by PM-HIP are proposed for structural and pressure-retaining applications in the nuclear power industry. However, the radiation response of PM-HIP alloys must be understood before they can be deployed. In this work, neutron irradiation to 1 dpa at 300 and 400 °C was performed on PM-HIP SA508 steel and a conventional forged counterpart. It was found that the PM-HIP sample showed higher irradiation embrittlement than the forged counterpart. Microstructural evolution was examined using scanning and transmission electron microscopy (SEM, TEM) and atom probe tomography (APT). The irradiation hardening effect was correlated with the microstructural evolution using the Orowan model. The present study provides insight into assessing the viability of using alloys manufactured by PM-HIP for nuclear reactor internals.

9:40 AM  
Dose and Temperature Dependence of Microstructure and Mechanical Properties in Ion-Irradiated PM-HIP Inconel 625: Caleb Clement1; Janelle Wharry1; Yangyang Zhao1; David Gandy2; Shujuan Wang3; Yaqiao Wu3; 1Purdue University; 2Electric Power Research Institute; 3Boise State University, Center for Advanced Energy Studies
    The objective of this talk is to understand the role of irradiation dose and temperature on microstructure evolution and mechanical properties of a Ni-based alloy produced by powder metallurgy with hot isostatic pressing (PM-HIP). PM-HIP is an attractive alternative to castings and forgings in the nuclear industry because components exhibit chemical homogeneity and are produced near-net shape, but the integrity of PM-HIP components under irradiation must first be understood. Inconel 625( IN625) manufactured with PM-HIP is irradiated with 4.5 MeV Fe+ ions to 50 and 100 dpa at 400°C and 500°C; conventional forged IN625 is also studied as a control specimen. Nanoindentation is used to evaluate yield strength of the irradiated layer. Microstructure is characterized using scanning and transmission electron microscopy(SEM, TEM). Irradiation hardening is related to the microstructural evolution through the Orowan model. This study forms the basis for assessing the viability of PM-HIP Inconel for nuclear settings.

10:00 AM  Invited
Mechanical Behavior and Radiation Effect in Additively Manufactured 316L Stainless Steel: Meimei Li1; Xuan Zhang1; Wei-Ying Chen1; 1Argonne National Laboratory
    Additive manufacturing (AM) as a disruptive manufacturing technology enables smart designs of nuclear reactor components with complex geometries and provide unprecedent design freedom. It also opens up new opportunities for designing advanced materials with controlled microstructure and properties. This paper summaries the recent effort on investigating the high-temperature mechanical behavior and radiation response of AM 316L SS for its core structural applications in the Transformational Challenge Reactor, a gas-cooled microreactor being developed to demonstrate revolutionary technologies including additive manufacturing. Creep and fatigue behavior of AM 316L stainless steel and the effect of post-build heat treatment was evaluated. Deformation mechanisms in AM 316L stainless steel was studied by synchrotron high-energy X-ray techniques. The evolution of irradiation-induced defects in AM 316L SS was investigated by in situ ion irradiation with transmission electron microscopy.

10:30 AM  
Mechanical Properties of Additively Manufactured 316L Stainless Steel before and after Neutron Irradiation: Thak Sang Byun1; Timothy Lach1; Maxim Gussev1; Kurt Terrani1; 1Oak Ridge National Laboratory
    The Transformational Challenge Reactor (TCR) program plans to build most core components through additive manufacturing (AM) processes, including the laser powder bed fusion (LPBF) for the metallic (316L) components. Mechanical testing and microstructural characterization before and after irradiation have been performed to build a database and assess the performance of AM components in reactor conditions. This presentation is to discuss the outcome of those activities focusing on the mechanical performance of AM 316L stainless steel. Baseline and post-irradiation tensile testing over a wide temperature range of room temperature–600°C were performed for the AM 316L alloy in as-built, stress-relieved, and solution-annealed conditions. The AM 316L stainless steel, regardless of its post-build processing, showed higher strength and comparable ductility compared to those of wrought 316L stainless steel. Mechanical property data indicate that the AM 316L stainless steel demonstrates a high resistance to the neutron irradiation at 300°C and 600°C.

10:50 AM  
Effects of Low-temperature Neutron Irradiation and Post-weld Heat Treatment on Tensile Properties of Welded Zircaloy-4: John Echols1; Nate Reid1; Sara Wonner1; Lauren Garrison1; 1Oak Ridge National Laboratory
    Zircaloy-4 possesses good radiation stability, corrosion resistance, mechanical properties, and low hydrogen uptake. Applications, however, are limited mostly to fuel cladding at operating temperatures above 300°C. Expanding the usage of zircaloy-4 has been proposed for projects including pressure vessels, test reactors, and reprocessing plants. Postweld mechanical properties, however, need to be qualified for these new operating regimes. This work investigates unwelded and TIG welded zircaloy-4 - with and without a post-weld heat treatment (PWHT) - neutron irradiated in the High Flux Isotope Reactor (target doses 1021 and 1022 n/cm2) at Oak Ridge National Laboratory. Low-temperature irradiation (60-90°C) and room-temperature tensile testing were performed. Emphasis is placed on understanding ductility as a function of weld, PWHT, and neutron dose through tensile testing and fractography. For unirradiated materials, peak total elongation (16.5%) is observed at 800°C, 1h PWHT, which also don’t exhibit undesirable ‘blocky alpha’ microstructure observed at longer (24h, 48h) PWHTs.