Microstructural Processes in Irradiated Materials: Austenitic Alloys
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Thak Sang Byun, Pacific Northwest National Laboratory; Chu-Chun Fu, Commissariat ŕ l'énergie atomique et aux énergies alternatives (CEA); Djamel Kaoumi, University of South Carolina; Dane Morgan, University of Wisconsin-Madison; Mahmood Mamivand, University of Wisconsin-Madison; Yasuyoshi Nagai, Tohoku University
Wednesday 8:30 AM
March 1, 2017
Room: Del Mar
Location: Marriott Marquis Hotel
Session Chair: Djamel Kauomi, North Carolina State University; Zhijie Jiao, University of Michigan
8:30 AM Invited
The Role of Deformation in Irradiation Assisted Stress Corrosion Cracking: Gary Was1; Drew Johnson1; Ian Robertson2; Diana Farkas3; 1University of Michigan; 2University of Wisconsin; 3Virginia Tech
The combination of radiation and a chemically aggressive environment gives rise to unique deformation modes and equally unique degradation modes such as irradiation assisted stress corrosion cracking. IASCC occurs in austenitic alloys exposed to irradiation while under stress in high temperature water. The mechanism is not well understood, but recent evidence has pointed to the interaction between dislocation channels and grain boundaries as a key factor driving the degradation. More specifically, the high local elastic stress at dislocation channel-grain boundary intersections is believed to be the key factor in crack nucleation. Yet very few sites result in crack nucleation. This talk will examine the response of irradiated austenitic stainless steels to stress in high temperature water. The nature of the dislocation channels and of the grain boundaries themselves on the cracking behavior will also be discussed in an effort to understand the selectivity of crack nucleation.
Plastic Deformation Mechanisms Accompanying Stress Corrosion Cracking in Highly Irradiated Austenitic Steels: Maxim Gussev1; Kevin Field1; Donovan Leonard1; Gary Was2; Keith Leonard1; 1Oak Ridge National Laboratory; 2University of Michigan
Irradiation-assisted stress corrosion cracking (IASCC) is one of the most severe degradation issues for austenitic stainless steels in the core internal components of light water reactors (LWR). Plastic deformation plays an important role in crack initiation and propagation. However, there is still limited understanding of plastic deformation mechanisms acting in high-irradiated austenitic steels under LWR-relevant conditions. In the present work, EBSD analysis, including Transmission Kikuchi Diffraction (TKD), was employed to investigate IASCC initiation and crack propagation. Constant extension rate testing (CERT) and compact tension (CT) crack growth rate specimens were analyzed following mechanical testing. The investigated materials included commercial-grade steels and model alloys irradiated in the BOR-60 reactor at 320C. The role of grain orientation on crack initiation was investigated using CERT specimens. Crack propagation paths and plastic strain fields in the irradiated CT-specimens were analyzed in detail. Near-crack-tip misorientation fields were investigated using TKD and (S)TEM coupled with FIB.
Study of Microstructural Evolution of 304 Stainless Steels by Atom Probe Tomography: Bertrand Radiguet1; Bertrand Michaut2; Brigitte Décamps3; Faiza Sefta4; Joël Malaplate2; 1GPM UMR CNRS 6634 - Université et INSA de Rouen; 2CEA Saclay, DEN/DANS/DMN/SRMA; 3CSNSM Orsay; 4EDF R&D, département MMC, Groupe Métallurgie
Austenitic stainless steels are widely used as internal structures of pressurized water reactors. In service, these steels are subjected to severe irradiation condition that can induce hardening, loss of ductility and swelling. Understanding of microstructural changes at the origin of these degradations is an important task for safety and life-extension reasons. Two austenitic stainless steels (304 and 304L) were irradiated at 450°C in JANNUS facility in Saclay. 10MeV Fe ions were used. The dose corresponds to a damage of 100dpa at about 750nm below the surface (before implantation peak to avoid the effect of injected self-interstitials atoms). The microstructure of the irradiated steels was characterized by atom probe tomography in order to get information about solute segregation and precipitation at high dose. The comparison between the two steels can provide information about the influence of carbon content on structural evolution. In this talk, experimental results will be presented and discussed.
Post-irradiation Annealing Effect on the Irradiated Microstructure of a BWR-irradiated 304L Stainless Steel: Zhijie Jiao1; Justin Hesterberg1; Gary Was1; 1University of Michigan
Susceptibility of the 304SS to IASCC is largely controlled by the persistent irradiated microstructure. Post-irradiation annealing (PIA) is an effective way to reduce the IASCC susceptibility by modifying the irradiated microstructure. A 304LSS previously irradiated to 5.9 dpa in a BWR was used. PIA was conducted at temperatures 500 and 550C to durations of 1–20 hours. The irradiated microstructure including dislocation loops, Ni/Si clusters and radiation-induced segregation at the grain boundaries was examined using TEM/STEM and atom probe tomography. The population of dislocation loops was found to reduce by 26% after PIA at 500C:1 hr and by 71% at 550C: 1 hr, and become negligible after 550C:20 hrs. The evolution of irradiation-induced Ni/Si clusters was found to follow a similar trend as dislocation loops but at a different rate. The effect of PIA on the overall microstructure will be presented. Its connection with irradiation hardening and will also be discussed.
10:00 AM Break
10:15 AM Invited
Role of Grain Boundary Phenomena on Stress Corrosion Cracking in LWR Environments: Daniel Schreiber1; Matthew Olszta1; Stephen Bruemmer1; 1Pacific Northwest National Laboratory
Stress corrosion cracking (SCC) is historically problematic in the high-temperature, corrosive, and high stress environment of light water reactors (LWR). Despite decades of research, we have yet to achieve a complete mechanistic understanding of SCC susceptibility and crack propagation. More recently, advancements in high resolution microscopy, in tandem with site-specific specimen preparation, have enabled unique insights into the role of grain boundary chemistry and its evolution in LWR environments on SCC. This work will present a combination of high-resolution analytical transmission electron microscopy and atom probe tomography to explicate the nature of grain boundary oxidation leading to intergranular SCC, and in particular the related changes in grain boundary chemistry ahead of the crack front. Comparisons will be made between model high-purity alloys and commercial alloys to bridge from fundamental degradation mechanisms to the complex heterogeneous degradation that is often observed in service alloys and environments.
Mechanical Characterization of In Service Inconel X-750 Annulus Spacers: Cameron Howard1; Peter Hosemann1; Scott Parker1; Malcolm Griffiths2; Colin Judge2; David Poff2; 1UC Berkeley; 2Canadian Nuclear Laboratories
The observation of helium bubbles segregated along grain boundaries and the intergranular failure of Inconel X-750 springs call for the necessity of further investigations using small scale mechanical testing. Novel techniques to perform in-situ, micro-sized, lift-out experiments are developed to assess the yield strengths and grain boundary strengths of Inconel X-750 in-service material, removed from the core of a CANDU reactor after 14 and 19 full power service years. Lift-out three-point bend test results indicate profound temperature effects in matrix yield strength as well as cold-working effects produced by the spring manufacturing process. Lift-out push-to-pull tensile tests provide an effective method for measuring grain boundary strengths by isolating individual grain boundaries and testing specimens until failure. In addition, tensile tests on helium implanted Inconel X-750 material performed at the Australian National Science and Technology Organization were done as a comparative study.
Microstructural Evolution and Mechanical and Fracture Behavior of CASS under Accelerated Thermal Aging: Timothy Lach1; Thak Byun1; 1Pacific Northwest National Laboratory
Cast austenitic stainless steels (CASSs) are used for many large components of reactor coolant systems, such as primary coolant piping and pump casings. The thermal aging-induced embrittlement is one of the most serious concerns in pursuit of extended operations of the massive components. In this ongoing long-term study, we aim to provide a comprehensive knowledge base for the integrity assessment of CASS components for extended operations, with a focus on thermal degradation mechanisms in the cast CF3, CF3M, CF8, and CF8M alloys at reactor-relevant accelerated aging temperatures of 290-400°C. Effects on mechanical and fracture properties and microstructure have been evaluated using uniaxial tensile tests, Charpy impact tests, analytical SEM and STEM, and atom probe tomography. The microstructural evolution, fracture mechanisms, and degree of thermal aging embrittlement in these alloys, including spinodal decomposition of ferrite, segregation of solute, and precipitation of second phases, will be discussed.
Irradiation-induced Nanoclusters in Cu-Nb and Cu-Nb-Si: Jaeyel Lee1; Pascal Bellon1; Robert Averback1; 1University of Illinois at Urbana-Champaign
Inducing finely dispersed nanoprecipitation to alloys is a promising mean to develop irradiation-tolerant materials for nuclear application. The key for the alloys to be effective is their stability at very high temperatures and irradiation. There have been reports on remarkably enhanced thermal stability of compound precipitates in Cu-Mo-Si by taking advantage of stable silicide formation. Here, we extended the study to Cu-Nb-Si systems, and report Nb-Si nanocluster formation under irradiation and their thermal stability. Using X-ray diffraction (XRD) and transmission electron microscopy (TEM), we showed that Nb-Si clusters are formed in a Cu90Nb4Si6 alloy during Kr ions irradiation at room temperature with high number density (~1023 m-3) and narrow size distribution. Those nanoclusters showed coarsening resistance during annealing at 500C for 1hr. Nanoclustering of Nb in Cu-Nb and their thermal coarsening behavior are studied for comparative analysis and the origins of coarsening resistance of Nb-Si clusters will be discussed.
11:45 AM Introductory Comments DOE-BES Program/Mechanical Behavior & Radiation Effects