Transmutation Effects in Fusion Reactor Materials: Critical Challenges & Path Forward: Structural, Plasma-facing & Functional Materials
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Arunodaya Bhattacharya, Oak Ridge National Laboratory; Steven Zinkle, University of Tennessee; Philip Edmondson, The University Of Manchester; Aurelie Gentils, Université Paris-Saclay; David Sprouster, Stony Brook University; Takashi Nozawa, National Institutes for Quantum and Radiological Science and Technology (QST); Martin Freer, University of Birmingham
Wednesday 8:30 AM
March 22, 2023
Room: 27B
Location: SDCC
Session Chair: Jaime Marian, University of California; Estelle Meslin, CEA-Saclay, University of Paris-Saclay
8:30 AM Invited
Discrete and Continuum Models for the Sources of Nonlinear Strain for Macrscopic Simulations of Reactor Components: Sergei Dudarev1; Max Boleininger1; Peter Derlet2; Pui-Wai Ma1; Daniel Mason1; Luca Reali1; 1UK Atomic Energy Authority; 2Paul Scherrer Institut
Multiscale modelling has been attempting to find macroscopic solutions by focusing on energies of defects and their transition energy pathways. This physics-based evolution-centred approach appeared detached from notions of macroscopic elasticity and their FEM implementations, requiring computing internal and external forces acting on components to evaluate stresses and deformations. Energies and lattice distortions are not directly related; the divergence of elastic self-energy density shows that it is the electronic structure at the core that control the structure of a defect and its dynamics. Macroscopic phenomenological models suffer from lack of transferability in unknown operating environments, including nuclear fusion. A new multiscale treatment for radiation effects in nuclear reactor components, where defects, helium impurities, and dislocations act as microscopic distributed sources of macroscopic elastic strain and stress fields, provides a basis for a self-consistent, not hierarchical, approach to the simulation of the effect of irradiation on fusion reactor components.
9:10 AM
Stability of a Li2TiO3 Candidate Solid-breeder Material Following Li Transmutation: German Samolyuk1; Philip Edmondson2; Yuri Osetskiy1; 1Oak Ridge National Laboratory; 2The University of Manchester
Generation of tritium fuel in future nuclear fusion power reactors is likely to occur through transmutation driven by interactions of Li atoms with energetic neutrons. In solid-state breeder blanket concepts based around Li ceramics such as Li2TiO3, the designed transmutation of Li into tritium will result in the formation of vacancies that may degrade the material properties. Here, we have applied first-principles density functional theory (DFT) based approaches to investigate the stability of the structure with increasing vacancy concentration as an approximation to the transmutation-induced transition from Li to tritium. The criterion for the loss of structural stability within the lattice is the presence of negative phonon frequency. This presentation will focus on the stability of lithium metatitanate (Li2TiO3) and the results will be discussed in terms of the upper limit of 6Li enrichment of the initial breeder, and the limit on exposure/lifetime.
9:30 AM
Grain-boundary Effects on the Irradiated Damages in W-Re Alloys: Sanghyuk Yoo1; Younggak Shin2; Seunghwan Oh1; Hyoungrul Park1; Younghyun Kim2; Anseung Yoo2; Ohkyoung Kwon3; Keonwook Kang1; Byeongchan Lee2; 1Yonsei University; 2Kyung Hee University; 3Korea Institute of Science and Technology Information
Under irradiated damages lies a mobility discrepancy between vacancies and self-interstitial atoms (SIAs). In W-Re alloys, Re atoms in W, either as solutes or transmutated elements, reduce the mobility of SIAs, buying time for Frenkel pairs to recombine, retarding damages. The presence of grain boundaries changes the irradiated damages working as a sink of interstitials. In particular, ultrafine grains found in a thin-film specimen or a sintered sample have a larger net grain-boundary area, and at the same time, a shorter distance for SIAs to travel to a grain boundary, necessarily changing the defect kinetics. Here, we report the irradiated damages in W-Re alloys from atomistic calculations. In particular, a virtual specimen is prepared to directly be compared with thin-film experiments. The virtual specimen measures 370 nm in thickness with 6 billion atoms. We will try to explain a governing relation of the grain boundary effects on irradiate damages.
9:50 AM Break
10:10 AM Cancelled
Embrittlement and Hardening of Beryllium Under Irradiation at Low Temperatures: Viacheslav Kuksenko1; Ed Darnbrough2; Artem Lunev1; 1UK Atomic Energy Authority; 2University of Oxford
A set of micromechanical and microstructural studies was carried out on beryllium samples irradiated by high energy protons and on beryllium samples implanted with He ions. Irradiation at 50°C to a dose of 0.04 dpa (160 appm of He) led to fracture mode change from transgranular cleavage to grain-boundary cracking. After irradiation, microcantilevers tests showed significant increase of fracture load with almost complete loss of ductility. Nanoindentation tests demonstrated that irradiation induced hardening after 50°C implantation is about 2 times higher than after 200°C (0.1dpa, 2000appm of He). Analysis suggests that the “back dots” should lead to about half of the measured hardness increase, while the rest of the hardening should originate from helium bubbles with the size below the TEM resolution. The possible mechanisms of the described behaviours are analysed in combination with the observed microstructural changes obtained by SEM, APT and Raman spectroscopy.
10:30 AM Cancelled
Melting Behavior of He-implanted Tungsten Visualized by MeV-ultrafast Electron Diffraction: Ling Wang1; Thies Albert2; Zhijiang Chen1; Leora Dresselhaus-Marais3; Samuel Murphy4; Nicholas Hartley1; Laurenz Kremeyer2; Matthias Kling1; Emma McBride1; Benjamin Ofori-Okai1; Alexander Reid1; Adam Summers1; Klaus Sokolowski-Titen2; Xiaozhe Shen1; Xueli Zheng3; Yongqiang Wang5; Siegfried Glenzer1; Mianzhen Mo1; 1SLAC National Accelerator Laboratory; 2University of Duisburg-Essen; 3Stanford University; 4Lancaster University; 5Los Alamos National Laboratory
Tungsten (W) has been perceived as a leading candidate plasma facing material for future fusion reactors. Its behavior under high fluence of helium (He) ion bombardment and other kinds of irradiation has been extensively studied. However, some key knowledge gaps still exist. Here, we use MeV-ultrafast electron diffraction (MeV-UED) to study the melting behavior of He-implanted W. Tungsten thin films were pre-implanted with 25 keV He ions to fluences of 1×1016 ~ 1×1017 ion/cm2 at room temperature. Melting of the samples was driven by intense ultrafast laser excitation, and the ensuing structural evolution was probed by MeV-UED with atomic-level resolutions. By monitoring the decay of Laue diffraction peaks and the appearance of liquid scattering signal, we captured the full solid-liquid phase transition in He-implanted W under different heating rate conditions. These results are important to benchmark large-scale molecular dynamics simulations for materials under extreme conditions.
10:50 AM
Analysis of Irradiation Damage Accumulation in Bi-phase Tungsten Heavy Alloy Microstructures: James Haag1; Weilin Jiang1; Matthew Olszta1; Wahyu Setyawan1; 1PNNL
Tungsten heavy alloys present a combination of highly desirable material properties and favorable initial results for consideration as plasma facing material components in nuclear fusion reactors. Yet in these promising probe studies, there exists a lack of information on irradiation damage accumulation in these multi-component systems and its subsequent bearing upon material properties and thereby service lifetimes. To provide this essential data for material adoption, this study observes the effects of the simulated fusion environment on two identical composition tungsten heavy alloy specimens both before and after thermomechanical processing to observe the effects of microstructure on damage accumulation. These irradiations simulate the five-year expected damage to divertor tiles by subjecting specimens to sequential high temperature Ni and He ion irradiations to mimic displacement damage and He production incurred during extended operation. This data will then be used to greatly improve predictive capacity in the extrapolation of material service lifetimes.