Microstructural Processes in Irradiated Materials: Fusion Materials and High-Temperature Alloys
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Thak Sang Byun, Pacific Northwest National Laboratory; Chu-Chun Fu, Commissariat Ó l'Únergie atomique et aux Únergies 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 2:00 PM
March 1, 2017
Room: Del Mar
Location: Marriott Marquis Hotel

Session Chair: Gary Was, University of Michigan; Chad Parish, Oak Ridge National Laboratory

2:00 PM  Invited
IOM/Mehl Award Lecture: Microstructure of Irradiated Materials: Steven Zinkle1; 1University of Tennessee; Oak Ridge National Laboratory
    Energetic particle irradiation can induce pronounced microstructural changes and corresponding dramatic property changes in materials. This presentation will provide an overview of radiation-induced microstructural changes, with particular emphasis on similarities and differences between metals and ceramics. There are several key temperature regimes for all irradiated materials (defined by the onset temperatures for migration of interstitials and vacancies, thermal dissolution of in-cascade produced vacancy clusters, and thermal evaporation of cavities). In general, radiation tolerance in one temperature regime does not universally translate to radiation tolerance in other temperature regimes due to different controlling physical parameters. The fluence dependence of defect accumulation also is generally significantly different in the various temperature regimes. The roles of primary knock on atom energy, damage rate, atomic mass, crystal structure, and other material parameters will be briefly discussed.

2:50 PM  
Microstructural Processes in Neutron-irradiated Tungsten: Chad Parish1; Xunxiang Hu1; Lauren Garrison1; Philip Edmondson1; Kun Wang1; Lance Snead2; Yutai Katoh1; 1Oak Ridge National Laboratory; 2Massachusetts Institute of Technology
    We performed mixed spectrum neutron irradiation of single and polycrystalline tungsten over a wide range of irradiation conditions, to determine the microstructural processes that might occur during service as a fusion plasma-facing materials. Irradiation temperatures ranged from ~80░C to ~1000░C. Neutron doses were from a fraction of a dpa to ~5 dpa. Hardness increased and toughness decreased markedly, and many of the high-dpa samples fractured in handling before tensile testing. Positron annihilation indicated significant vacancy populations; isochronal annealing helped determine the vacancy mobility. Microscopy and atom probe found significant levels of the transmutation products Re and Os, which formed complex precipitate microstructures. Grain boundaries and voids gettered Re but not Os. In addition to mapping the microstructure of tungsten as a function of irradiation temperature and dose, this talk will discuss the effect of the microstructural features on the key issue of irradiation embrittlement in tungsten.

3:10 PM  
Evolution of Microstructure of Tungsten under Irradiation with Tungsten Ions: Emmanuel Autissier1; Marie-France Barthe1; Pierre Desagrdin1; CÚcile Genevois1; Brigitte Decamps1; Robin SchaŘblin2; Yves Serruys3; 1CNRS; 2ETH Zurich; 3CEA
    To study and understand the evolution of the microstructure of tungsten under conditions similar to those expected in future fusion reactors such as ITER and DEMO, irradiations with W ions were performed, in well prepared tungsten samples, using various conditions (energy, temperature and damage). Positron annihilation spectroscopy (PAS) and Tranmission Electron Microscopy (TEM) were used to characterize defects induced by irradiation, from the small vacancy clusters to large cavities. The defects distribution changes with damage dose between 0.01 and 0.06 dpa and irradiation temperature between -182 and 700 ░C. The effect of ion energy (1.2  20 MeV) on this distribution will be discussed. Acknowledgements: These studies are supported by the European Commission in the framework of the EUROfusion Consortium. Irradiations were performed at JANNuS (Joint Accelerators for Nanoscience and Nuclear Simulation) Orsay at CSNSM (France) and Saclay at CEA (France) which are part of the EMIR French accelerators network.

3:30 PM  
Understanding the Effects of Helium Implantation Damage in Tungsten: Combining Multi-technique Experiments and Atomistic Modeling: Felix Hofmann1; Duc Nguyen-Manh2; Daniel Mason2; Mark Gilbert2; Sergei Dudarev2; Isaure deBroglie3; Jeffrey Eliason4; Ryan Duncan5; Alexei Maznev5; Keith Nelson5; Christian Beck1; Wenjun Liu6; 1University of Oxford; 2Culham Centre for Fusion Energy; 3╔cole Polytechnique; 4University of Minnesota; 5Massachusetts Institute of Technology; 6Argonne National Laboratory
    Tungsten-based materials are key candidates for fusion armor components. During service they will be exposed to high temperatures, intense neutron flux and ion bombardment. Here we use helium-ion-implantation to study the interaction of injected helium with displacement damage. X-ray micro-diffraction and laser-induced transient grating measurements show significant lattice swelling and changes in elastic and thermal transport properties following ion implantation. These experimental observations are analyzed using atomistic, elasticity and thermal transport calculations. We find that damage is dominated by Frenkel defects with helium-filled vacancies. Interestingly the retained defect population, and its effects on physical properties, are not a trivial functions of the implanted ion dose. Combining such multifaceted experimental observations and multi-scale modeling allows us to begin to establish a joined-up picture of the complex effects helium-implantation-induced damage has on the structure and properties of tungsten.

3:50 PM Break

4:05 PM  
Microstructure and Mechanical Properties of Neutron-irradiated Tungsten Foil for Laminate Composites: Lauren Garrison1; Chad Parish1; Xunxiang Hu1; Taehyun Hwang1; Takaaki Koyanagi1; Jens Reiser2; Lance Snead3; Yutai Katoh1; 1Oak Ridge National Laboratory; 2Karlsruhe Institute of Technology; 3Massachusetts Institute of Technology
    Fusion reactors need a robust and ductile tungsten-based material. Tungsten foil composites are one promising option, but the post-irradiation behavior of the foil needs to be understood. Rolled foil, annealed and recrystallized foil, and single crystal tungsten were irradiated in HFIR at temperatures up to 800°C and doses up to ~5 dpa. After ~1 dpa, all tungsten materials developed a network of small, needle-shaped Re and Os rich precipitates. In addition, the foil materials developed large precipitates at the grain boundaries. The rolled foil had the highest hardness of the tungsten materials both before and after irradiation up to 0.6 dpa. The single crystal tungsten and annealed tungsten foils had essentially the same hardness level before irradiation, but after only 0.006 dpa the single crystal tungsten hardened significantly more than the annealed foil. The relation of microstructure to mechanical properties and consequences for tungsten-foil-based composites will be discussed.

4:25 PM  Invited
Mechanism of Reduced Radiation Damage Identified in Equiatomic Multicomponent Single Phase Alloys: Flyura Djurabekova1; Fredric Granberg1; Kai Nordlund1; William J. Weber2; Yanwen Zhang2; 1University of Helsinki; 2Oak Ridge National Laboratory
    Recently a new class of metal alloys, of single-phase multicomponent composition at equal atomic concentrations showed promising mechanical, magnetic, and corrosion resistance properties, in particular, at high temperatures. These features make them potential candidates for components of next-generation nuclear reactors that will involve high temperatures combined with corrosive environments and extreme radiation exposure. Until recently, however, the radiation tolerance of these alloys at high doses remained unexplored. In this work, a combination of experimental and modeling efforts reveals a substantial reduction of damage accumulation under prolonged irradiation in single-phase NiFe and NiCoCr alloys compared to elemental Ni. This effect is explained by reduced dislocation mobility, which leads to slower growth of large dislocation structures. Moreover, there is no observable phase separation, ordering, or amorphization, pointing to a high phase stability of this class of alloys. [1] F. Granberg et al., PRL 116 (2016) 135504.

4:55 PM  
Comparison of Neutron and Ion Irradiation Effects on Microstructure of MA957: Jing Wang1; Nathan Bailey2; Mychailo Toloczko1; Daniel Schreiber1; Frank Garner3; Y. Kupriianova4; A. Kalchenko4; V. Voyevodin4; Lin Shao5; 1Pacific Northwest National Laboratory; 2University of California at Berkeley; 3Radiation Effects Consulting; 4Kharkov Institute of Physics and Technology; 5Texas A&M University
    Heavy-ion irradiations have regained popularity for investigating irradiation effects on the microstructure of materials for reactor core internals. Because of differences between ion and neutron irradiation environments, there is an expectation that there will be some differences in microstructural evolution induced by the two methods. Studies conducted many years ago attempted to quantify the magnitude of these differences for selected alloys, and there are ongoing efforts by various groups to quantify differences in microstructural evolution of ferritic-martensitic alloys between the two methods using improved ion irradiation techniques and microscopy capabilities. Reported here is a detailed comparison of the microstructure of MA957 for chromium ion irradiation and neutron irradiation, both to ~100 dpa at temperatures ~400-550░C. Atom probe tomography and transmission electron microscopy were used to quantify void swelling, alpha-prime, and YTiO populations, revealing some differences, but with generally consistent trends with dose and temperature between the two irradiation methods.

5:15 PM  
Neutron Irradiation Damage in Ferritic ODS Steel MA957: Xiang Liu1; Yinbin Miao2; Wei-Ying Chen2; Yaqiao Wu3; James Stubbins1; 1University of Illinois at Urbana Champaign; 2Argonne National Laboratory; 3Center for Advanced Energy Studies
    Oxide dispersion strengthened (ODS) steels are known for their excellent irradiation resistance due to the introduction of Y-Ti-O nanoclusters into the matrix. The Y-Ti-O/matrix interface can act as sinks for both point defects and helium atoms under neutron irradiation. In this study, neutron irradiation damage in a ferritic ODS steel MA957 was characterized by transmission electron microscopy and atom probe tomography. The MA957 samples were irradiated at ATR at 300C and 450C up to 5 dpa. Irradiation effects on the cluster stability, dislocation structure, alpha prime precipitates, and voids were studied.

5:35 PM  
Impact of He Concentration on the Microstructure of W Using TEM with In Situ Ion Irradiation: Robert Harrison1; Matheus Tunes1; Graeme Greaves1; Jonathan Hinks1; Stephen Donnelly1; 1University of Huddersfield
    As a plasma-facing material, W will be exposed to high-heat flux, high-energy neutrons (>14MeV) and He production from (n,α) reactions as well as injection from the plasma. W has been irradiated with 15, 30 or 60keV He+ ions using in-situ TEM between 500 and 1000░C to a dose of ~3 DPA at the MIAMI facility. A high density of small dislocation loops with sizes around 5–20nm were observed at temperatures <1000░C. Loops were characterised as b = ▒1/2<111> type with no <100> type observed so far, unlike in previous work on self-ion irradiations, this microstructure is attributed to the presence of He. Ordered arrays of He bubbles were observed for all He-appm/DPA ratios at 500░C. Bubble-lattice spacing increased with increasing He+ ion energy (i.e. decreasing He concentration), showing a trend between He concentration and bubble lattice parameters, shedding new light on the formation and development of bubble-lattices.