Tackling Structural Materials Challenges for Advanced Nuclear Reactors: Advanced Manufacturing
Sponsored by: TMS Corrosion and Environmental Effects Committee, TMS Nuclear Materials Committee, TMS: Advanced Characterization, Testing, and Simulation Committee
Program Organizers: Miaomiao Jin, Pennsylvania State University; Xing Wang, Pennsylvania State University; Karim Ahmed, Texas A&M University; Jeremy Bischoff, Framatome; Adrien Couet, University of Wisconsin-Madison; Kevin Field, University of Michigan; Lingfeng He, North Carolina State University; Raul Rebak, GE Global Research
Wednesday 8:00 AM
October 12, 2022
Room: 330
Location: David L. Lawrence Convention Center
Session Chair: Kevin Field, University of Michigan
8:00 AM
Electrodeposition of Functionally Graded Interlayers for Enhanced Divertor/Heatsink Bonding for Fusion Reactors: Holly Garich1; Katherine Lee1; Brian Skinn1; 1Faraday Technology, Inc.
Energy-positive fusion reactors require the development of divertor modules at the bottom of the vacuum vessel to extract heat and ash produced by the fusion reaction, minimize plasma contamination, and protect the walls from thermal/neutronic loads. Functionally graded interlayers are investigated to provide improved thermomechanical properties to handle the anticipated heat flux from the fusion plasma to heatsinks joined to the tungsten plasma-facing components. Specifically, functionally-graded bonding interlayers, consisting of either tungsten-included copper composite or iron/tungsten alloy are designed to afford high and low tungsten concentrations near the plasma-facing component- and heatsink-facing surfaces, respectively. Pulsed electrodeposition enables these functionally graded interlayers to provide smooth gradients in the coefficient of thermal expansion, facilitating strong brazed joints between plasma-facing and heatsink components. This work presents data on the preparation and performance of functionally graded bonding interlayers by the pulse deposition method.
8:20 AM
ICME and ML Modeling Framework of U-10%wt Mo Fabrication Processes: Ayoub Soulami1; William Frazier1; Yucheng Fu1; Lei Li1; Kyoo Sil Choi1; Zhijie Xu1; Vineet Joshi1; 1Pacific Northwest National Laboratory
Uranium with 10wt% molybdenum (U-10Mo) has been recognized as a promising candidate to replace high-enriched uranium fuel due to its ability to meet the neutron flux demands of U.S. high power research reactors. Manufacturing the U-10Mo alloy involves a complex series of thermomechanical processing steps. As part of this study, several models have been developed for the individual processes. The coupling between individual processes uses the concept of ICME which aims to bridge the information passing between interacting models and investigates the impact of manufacturing processes on material microstructure evolution. To enable fast prediction and optimization of U-10Mo properties, a data-driven surrogate model has been developed using the established simulation data. The simulated microstructures cover a wide range of initial grain size, uranium carbide volume fraction, and rolling reductions. This data-driven surrogate model demonstrates good accuracy in predicting the U-10Mo fuel microstructures and guides the selection of proper processing parameters.
8:40 AM Invited
Heavy Ion Irradiation Response of an Additively Manufactured 316L Stainless Steel: Xinghang Zhang1; Zhongxia Shang1; Cuncai Fan1; Lin Shao2; Thomas Voisin3; Y Wang4; Nick Richter1; 1Purdue University; 2Texas A&M University; 3Lawrence Livermore National Lab; 4University of California, Los Angeles
Additive manufacturing has become an appealing technique to fabricate three-dimensional metallic materials and components for nuclear reactors. However, response of additively manufactured alloys to high-dose heavy ion irradiations at elevated temperatures is still not well understood. Here, we will present the response of an additively manufactured 316L austenitic stainless steel to ex situ and in situ heavy ion irradiations. Microscopy studies show a reduced defect density in the additively manufactured sample compared with its cold worked counterpart after heavy ion irradiation to 200 dpa. In situ irradiation studies captured microstructure evolutions during radiation and reveal potential mechanisms of enhanced radiation tolerance. The present work advances our understanding on the high-temperature irradiation response of additively manufactured steels for nuclear reactor applications.
9:10 AM
Progress Toward Additive Manufacturing of Ferritic-martensitic, In situ Tempered Steels for Nuclear Applications: Calvin Lear1; Todd Steckley1; Mehmet Topsakal2; Simerjeet Gill2; Thomas Lienert3; Stuart Maloy4; 1Los Alamos National Laboratory; 2Brookhaven National Laboratory; 3Optomec; 4Pacific Northwest National Laboratory
Although laser-directed energy deposition (L-DED) offers potentially significant cost savings for metallic reactor components (tooling independence, building to near final shape), L-DED microstructures are not ideal for some promising structural materials. Wrought Gr 91 steel, for example, is both radiation tolerant and resistant to high-temperature creep, while L-DED Gr 91 is mostly fine martensite and extremely brittle. We proposed to remedy this deficiency without a traditional tempering (undesirable due to cost and large components) by adapting principles of multi-pass welding used on boilers and pressure vessels to L-DED. Build parameters were tailored such that prior (lower) layers are tempered by excess heat from later (upper) beads. This process results in an inner Gr 91 component with suitable metallurgical properties. Here, we explore the impact of build pre-heat (~25, 90, 200, and 400 °C) through microstructural characterization using EBSD and through phase analysis using high-energy X-ray diffraction at NSLS-II.
9:30 AM Invited
Structural Material Design for Power Plants Using Additive Manufacturing: Wei Xiong1; 1University of Pittsburgh
Additive manufacturing (AM) can be used to design complex shape components, which have application in power plants for repairing or component processing. However, it often introduces complex thermal cycles during beam melting causing undesired microstructure-property relationships. Therefore, the process-structure-property correlations are challenging to model, and thus impede the materials and process design for power plant construction applications. Fortunately, the AM technique itself sometimes enables us to accelerate alloy design by acting as a high-throughput design tool. In this talk, we will review some new opportunities regarding AM of structural materials that can potentially be used for power plants. We will highlight some case studies on alloy development for AM through graded alloy processed by directed energy deposition with both powder and wire-based techniques. Through case studies, such as Haynes 282 and P91/740H graded alloy AM, we will identify both challenges and design pathways.
10:00 AM Break
10:20 AM
Phase Field Modeling of Hot Isostatic Pressing for Joining of Dissimilar Metals: Albert Lin1; Yongfeng Zhang1; 1University of Wisconsin - Madison
Powder metallurgy hot isostatic pressing (PM-HIP) is a joining technique proven to manufacture structural and pressure-bearing reactor components with properties matching or exceeding those created by traditional methods. Here, a phase field model is constructed in the open-source MOOSE framework to simulate the interdiffusion between Fe- and Ni-based alloys during PM-HIP. The CALPAD free energy models of the Fe-Cr-Ni ternary system for the ferritic and austenitic phases of steel, in conjunction with mobilities of alloying elements, are used to describe and predict the interdiffusion and phase evolution in the interdiffusion zone. The model is applied to interdiffusion between 316 austenitic/A508 ferritic steels and between 316 steel/Ni-based 600M alloys to demonstrate the capabilities for concurrent diffusion and phase transformation. The generality of the MOOSE framework and the thermodynamic basis leave the model open to additional modeling of carbide phases, thermal cycling, and other critical issues of dissimilar metal joining.
10:40 AM
Understanding of Alloying Additions for Design of Gas Atomization Reaction Synthesis Produced Oxide Dispersoid Strengthened Alloys: Emma Cockburn1; 1Emma Cockburn
Mechanical alloying (MA) has been a focus in developing processing methods for oxide dispersion strengthened (ODS) alloys for ultrahigh temperature and/or high-flux radiation tolerant applications, e.g., nuclear power. While capable of providing “mechano-chemical” mixing of yttria and alloy (Fe- or Ni-based) powders, MA is time-consuming and may introduce contamination and inhomogeneities. Gas atomization reaction synthesis (GARS) for producing powders for ODS alloy processing provides cleaner feedstock powders with Cr-enriched surface oxides and Y-containing intermetallics, creating oxide dispersoids during laser-powder bed fusion additive manufacturing and solid-state friction/stir consolidation by indirect extrusion fabrication. This work will investigate the control/suppression these intermetallics by adjusting the amount of alloying elements available during the solidification reaction in the atomization process. Through thermodynamic calculations, X-ray diffraction data, and microstructural analysis of castings and powders, valuable details about alloying levels can be connected to intermetallic data. Funded by USDOE-ARPA-e program through Ames Lab contract no. DE-AC02-07CH11358.
11:00 AM
Studying Microstructural Evolution in an Oxide Dispersion Strengthened 14YWT Ferritic Steel Tube Manufactured using SolidStirTM Technology: Shubhrodev Bhowmik1; Pranshul Varshney1; Osman El Atwani2; Stuart Maloy3; Kumar Kandasamy4; Nilesh Kumar1; 1University of Alabama, Tuscaloosa; 2Los Alamos National Lab; 3Pacific Northwest National Lab; 4Enabled Engineering
Next generation nuclear reactors are expected to operate in much harsher conditions necessitating use of advanced materials in such environments. In this regard, 14YWT ferritic steel has shown a great potential for being used as fuel cladding tube and other structural components in advanced nuclear reactors. An advanced manufacturing technique, SolidStirTM Technology, patented by Enabled Engineering and based on the principle of friction stir processing, was used in the present work to successfully demonstrate fabrication of a thin-walled tube using ball-milled 14YWT powders. A detailed macrostructural and microstructural characterization using advanced characterization tools have been carried out to understand structural evolution at different length scales in the 14YWT tube processed using SolidStirTM. The preliminary processing and microstructural characterization results indicate that the SolidStirTM is a potential tool for manufacturing fuel cladding tubes at large scale.
11:20 AM Invited
Neutron Irradiation Effects in PM-HIP Nuclear Structural Alloys: Janelle Wharry1; Caleb Clement1; Yangyang Zhao1; Sri Sowmya Panuganti1; Yu Lu2; Yaqiao Wu2; Donna Guillen3; David Gandy4; 1Purdue University; 2Boise State University; 3Idaho National Laboratory; 4Electric Power Research Institute
The objective of this talk is to summarize current understanding of irradiation effects on the microstructure and mechanical properties of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). The nuclear industry has growing interest in replacing castings and forgings with PM-HIP components due to their near-net shape production, microstructural uniformity, and reduced reliance on machining and welding. Here, we directly compare PM-HIP alloys to cast/forged counterparts through an Advanced Test Reactor (ATR) neutron irradiation campaign to 1 and 3 displacements per atom (dpa) at 300-400°C. We focus on Ni-base Alloy 625, Grade 91 steel, and SA508 pressure vessel steel. Uniaxial tensile testing reveals consistent irradiation hardening and embrittlement between PM-HIP and cast/forged alloys. Dislocation loop, nanocluster, and void evolution are also relatively consistent between PM-HIP and cast/forged materials but are dependent upon initial microstructure. Results show promise for code qualification of PM-HIP structural materials.