Materials Systems for the Future of Fusion Energy: On-Demand Oral Presentations
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Additive Manufacturing Committee, TMS: Computational Materials Science and Engineering Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Jason Trelewicz, Stony Brook University; Kevin Field, University of Michigan; Takaaki Koyanagi, Oak Ridge National Laboratory; Yuanyuan Zhu, University Of Connecticut; Dalong Zhang, Pacific Northwest National Laboratory
Monday 8:00 AM
March 14, 2022
Room: Nuclear Materials
Location: On-Demand Room
Development, Production, and Qualification of Berrylide Neutron Multiplier Parts for the He Cooled Breeding Blanket: Ramil Gaisin1; Michael Duerrschnabel1; Michael Klimenkov1; Pavel Vladimirov1; Sergey Udartsev2; 1Karlsruhe Institute for Technology; 2Ulba Metallurgical Plant
In the helium-cooled blanket concept for the DEMO fusion reactor, solid blocks of titanium beryllide TiBe12 will multiply and moderate neutrons to produce tritium in lithium ceramics. However, the production of large blocks faced many issues due to the extreme hardness and fragility of beryllides. Various casting and powder metallurgy methods were tested before a series of massive titanium beryllide blocks with dimensions up to Ø150mm×200mm was obtained. The resulting blocks have a homogeneous microstructure consisting of single-phase TiBe12 grains with traces of residual beryllium and beryllium oxide. In terms of specific compressive strength, the obtained TiBe12 surpasses all materials, except diamond, in the 700–1000°C temperature range. Long-term thermal cycling tests with rapid heating and cooling, simulating operation in a fusion reactor, showed high resistance of beryllides to thermal shocks. The results obtained are supported by extensive microstructural investigations.
In-situ TEM of Quantum De-trapping and Transport of SIA Clusters in Tungsten: Kazuto Arakawa1; 1Shimane University
To precisely predict the lifetime of nuclear-fusion materials, accurate understanding of the dynamic properties of radiation-produced defects is required. Defects are unavoidably bound to static trapping centres such as impurity atoms, meaning that their diffusion is dominated by de-trapping processes. We successfully monitored the de-trapping and migration of nanoscale SIA clusters in the form of dislocation loops, strongly trapped by impurity atoms in tungsten, by triggering de-trapping out of equilibrium at cryogenic temperatures, using high-energy electron irradiation and in-situ transmission electron microscopy [1]. We reveal the quantum de-trapping of defects below around 1/3 of the Debye temperature, in contrast to a traditional notion that de-trapping occurs only by thermal activation. In the presentation, we also show the impacts of this intrinsically high mobility of the dislocation loops in the microstructural evolution upon irradiation.[1] K. Arakawa et al., Nature Mater. 19 (2020) 508.
Microstructural Examination of Radiation Damage in Tungsten: Michael Klimenkov1; Ute Jäntsch1; Ramil Gaisin1; Steffen Antusch1; Michael Rieth1; Hans-Christian Schneider1; Dmitry Terentyev2; Wouter Van Renterghem2; 1Karlsruhe Institute of Technology; 2Belgian Nuclear Research Centre
Tungsten is the prime candidate material for plasma-facing components in future fusion reactors due to its several advantageous properties such as a high melting point, high sputtering resistance and low coefficient of thermal expansion. The microscopic examination of neutron irradiated tungsten is of high importance for assessing possible limits of operation conditions and life-time of plasma facing components. Transmission electron microscopy examinations of tungsten irradiated at various temperatures and damage doses identify four defect types that adversely affect the material properties: (1) interstitial and vacancy clusters, (2) dislocation loops and (3) voids. In addition, transmutation processes lead to an enrichment of the matrix with rhenium or osmium and their radiation-induced segregation on the structural defects. Under certain conditions, this segregation leads to the formation of σ-W(ReOs)2 and χ-W(ReOs)3 precipitates. The defects lead to swelling, radiation hardening and embrittlement, which can considerably limit the service life of tungsten components.
Neutron Irradiated Tungsten Defect, Surface Chemistry, and Microstructural Characterization: Chase Taylor1; Masashi Shimada1; Yasuhisa Oya2; 1Idaho National Laboratory; 2Shizuoka University
Fusion plasma facing components (PFCs) will be exposed to high fluxes of energetic neutrons, hydrogen isotopes, and helium. In addition, a variety of mixed materials end up on the surfaces of PFCs due to sputtering and redeposition from PFCs of other compositions. The end result will be the generation of new material surfaces with evolved properties with respect to the pristine surface. Tungsten surface and bulk properties significantly influence the behavior of tritium retention. These collective effects impact plasma-surface interactions, safety, and fuel cycle management. Multiple fission campaigns irradiated tungsten samples under thermal and fast neutron fluxes. Samples then were characterized using XPS, PAS, and TEM in order to investigate the surface chemistry, defect structure, and microstructure, respectively. Deuterium irradiation and TDS were performed, which showed marked differences based on the sample surface condition. The present multi-technique analysis elucidates the complex interactions that occur in neutron irradiated tungsten PFCs.
Self-passivating SMART Alloys for a Fusion Power Plant: Andrey Litnovsky1; Felix Klein1; Xiaoyue Tan1; Jan W. Coenen1; Gerald Pintsuk1; Christian Linsmeier1; Jesus Gonzalez-Julian1; Martin Bram1; Ivan Povstugar1; Thomas W. Morgan2; Yury M. Gasparyan3; Alexey Suchkov3; Diana Bachurina3; Duc Nguyen-Manh4; Mark R. Gilbert4; Damian Sobieraj5; Jan Wrobel5; Joven Lim4; Pawel Bittner1; Anicha Reuban1; 1Forschungszentrum Juelich; 2DIFFER Dutch Institute for Fundamental Energy Research; 3National Research Nuclear University MEPhI; 4CCFE, United Kingdom Atomic Energy Authority; 5Warsaw University of Technology
Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development as plasma-facing materials for DEMOnstration fusion power plant (DEMO). These alloys containing tungsten, chromium and yttrium must suppress oxidation of tungsten under accident conditions providing an acceptable plasma performance during the regular operation. These requirements are vital elements of the structured R&D on SMART materials. A 104-fold reduction of oxidation and more than 40-fold suppression of sublimation of tungsten oxide is attained during a reproduced 10-day accident event. The sputtering resistance of SMART materials and pure tungsten is identical for plasma fluence corresponding to 20 days of continuous DEMO operation. Crucial role of yttrium in resistance to oxidation is revealed via modeling and experiments.Significant efforts are focused on industrial up-scale. Mechanical alloying by an industrial partner and the production of a bulk sample with dimensions of 100mm×100mm×7mm are accomplished toward the realization of a first wall element of DEMO
Development of Tungsten Heavy Alloy Composites for Fusion Applications: Wahyu Setyawan1; Ba Nguyen1; Weilin Jiang1; Md Alam2; James Haag IV3; Jing Wang1; Laila El-Guebaly4; Dalong Zhang1; Ramprashad Prabhakaran1; Charles Henager Jr.1; G. Odette2; Mitsu Murayama3; 1Pacific Northwest National Laboratory; 2University of California at Santa Barbara; 3Virginia Tech; 4University of Wisconsin
W/NiFe tungsten heavy alloy (WHA) composites exhibit superior toughening than W, making them prospective structural materials for divertors. The composites consist of spheroidal W particles embedded in a Ni-based ductile phase. Tensile and bend specimens are tested as a function of temperature. General principles of ductile-phase-toughening are observed in the deformation of these composites. Finite-element models are developed to explore artificially designed lamellar brick-and-mortar microstructures. The strength and ductility of brick-and-mortar microstructures can be tailored by adjusting the brick aspect ratio. Ni+ and He+ ion irradiations are performed to investigate irradiation effects. Helium cavities are located preferentially at interphase boundaries. Atomistic simulations are performed to estimate the He effect on the cohesion of interphase boundaries. In addition, neutron activation calculations are performed to estimate the allowable limit of Ni for disposal and recycling options, He/H gas production, and damage level in the WHAs based on the ARIES-ACT2 power plant concept.
Neutron Radiation Enhanced Grain Growth in Tungsten and Tungsten Alloys under Mixed Spectrum Neutron Irradiation: Hanns Gietl1; Takaaki Koyanagi1; Xunxiang Hu1; Yutai Katoh1; 1Oak Ridge National Laboratory
Tungsten (W) is the main plasma facing material under consideration for fusion devices. However, there are many open questions and challenges regarding the use of W. One major question is the response of W to neutron irradiation.In this study, pure W and several W alloys were subjected to mixed spectrum neutron irradiation in HFIR at a best-estimate temperature of 850 ˚C and above. Microstructural analysis using SEM-EBSD found significant grain growth after irradiation at temperatures remarkably lower than reported recrystallization temperatures for unirradiated materials in a similar duration. Recrystallisation fraction was quantified by a combination of grain size, type of grain boundary and internal grain misorientation. Hardness measurements found softening consistent with grain growth in most materials. Recrystallisation kinetics under irradiation is described by calculation of a radiation enhanced self-diffusion. The research is sponsored by the U.S. Department of Energy, Office of Fusion Energy Sciences.
Characterization of Atomic-scale Defects in Neutron Irradiated Silicon Carbide: Takaaki Koyanagi1; David Sprouster2; Xunxiang Hu1; Yutai Katoh1; 1Oak Ridge National Laboratory; 2Stony Brook University
The characterization of atomic-scale defects in silicon carbide (SiC), including Frenkel pairs, antisites, and small defect clusters, is the key to understanding the dynamic self-stabilizing mechanisms in certain ceramic compounds under irradiation. Although the configurations and energetics of such defects have been studied in theoretical simulations, experimental validation of the results has been limited. The objective of this study is to provide experimental information on atomic-scale defects in SiC to allow better modeling of the microstructural response in a fusion nuclear environment. We used three different techniques for the characterization of microstructures: Raman spectroscopy, positron annihilation spectroscopy, and high-energy x-ray diffraction. Based on the systematic investigation, we present experimental evidence showing how homonuclear bonds (Si-Si and C-C) and vacancy clusters develop under high-dose neutron irradiation. The research is sponsored by the U.S. Department of Energy, Office of Fusion Energy Sciences.
Novel Transitional Layer Structure between Reduced Activation Ferritic Martensitic Steels and Tungsten for Fusion Reactors: Tim Graening1; Ishtiaque Robin2; Ying Yang1; Lizhen Tan1; Yutai Kato1; 1Oak Ridge National Laboratory; 2University of Tennessee Knoxville
Plasma-facing components (PFCs) are among the most critical gaps for fusion energy to establish technical and economic feasibility. Here, a novel transitional multilayer structure joining tungsten and RAFM steels using five layers, i.e., RAFM steel/ FeCrAl/VCrAl/VCrTi/W was investigated. The composition of each interlayer was selected based on computational thermodynamics and diffusion kinetics to prevent the formation of a brittle intermetallic phase region in the temperature range of 620~1150ºC. As a first step, the transitional structure was bonded using spark plasma sintering (SPS). Microstructural analysis using scanning electron microscopy on SPS-bonded and heat-treated samples suggest a good bonding strength. Each interface was investigated in more detail to verify the absence of a brittle intermetallic layers using nanoindentation and transmission electron microscopy. The gains from this microstructural investigation of an SPS sample can readily be applied to many other techniques, including additive manufacture of complex structural parts for fusion reactors.
Cancelled
Tensile Properties and Microstructure of Neutron Irradiated Tungsten Fibers for Fusion Materials Application: Lauren Garrison1; John Echols1; Johann Riesch2; Hans Gietl1; Maxim Gussev1; 1Oak Ridge National Laboratory; 2Max-Planck-Institut für Plasmaphysik, Garching
While tungsten is desired as a plasma-facing material for fusion, powder metallurgy produced tungsten is brittle under fusion operating conditions. Tungsten fiber reinforced tungsten matrix composites are a promising material because of their improved toughness. The composite behavior strongly depends on the properties of the fibers, with a lesser contribution from the matrix. Thus, unalloyed and potassium-doped (60 ppm) tungsten fibers with 150 µm diameter were neutron irradiated in the High Flux Isotope Reactor at temperatures between 400–1100°C to doses of ~0.2–0.7 dpa. The fibers had two different pre-irradiation straightening treatments (mechanical and heat treatment) for easier handling. Initial tensile tests showed the fibers retained some ductility after neutron irradiation, and the pre-heated fibers had a lower ultimate strength. Subsequent tensile testing utilized digital image correlation to uncover the details of the deformation. Additionally, the fiber microstructure before and after irradiation was examined with scanning electron microscopy.
Effect of He Plasma Exposure on Recrystallization and Properties of W: Dhriti Bhattacharyya1; Calvin Hoang2; Matthew Thompson3; Cormac Corr3; 1Australian Nuclear Science and Technology Organization; 2University of New South Wales; 3Australian National University
Tungsten is the primary material used in the manufacture of plasma facing diverter components of Tokamak fusion reactors. It is known that exposure to He ash from the plasma can cause surface changes in the tungsten, causing degradation of its morphology and properties. The diffusion of He also causes changes in the recrystallization characteristics of W, thus affecting its properties in the long run as well. In this study, W samples have been exposed to He through the experimental plasma device called Magpie, and its effect on recrystallization has been studied through subsequent annealing. The exposed and annealed samples were characterized by scanning and transmission electron microscopy, electron backscatter diffraction and nanoindentation. It is apparent from these studies that exposure to He at higher temperatures (above 300°C) causes a delay in recrystallization, as evident from the higher annealing temperatures required for a drop in hardness.
Anomalous Precipitation of Cr in Fe-rich Ferritic Steels under Irradiation in Presence of C and N Impurities: First Principles Modeling and Experimental Observations: Mark Fedorov1; Jan Wróbel1; Andrew London2; Krzysztof Kurzydłowski3; Sergei Dudarev2; Duc Nguyen-Manh2; 1Warsaw University of Technology; 2Culham Centre for Fusion Energy, United Kingdom Atomic Energy Authority; 3Białystok University of Technology
Cluster Expansion (CE) model based on Density Functional Theory (DFT) calculations was developed for Fe-Cr alloys with interstitial C and N, representing steels impurities, and vacancies, representing radiation defects. The model is constructed from two sublattices: (i) Fe-Cr matrix with vacancies on bcc sublattice and (ii) the interstitial defects on octahedral sublattice. Binding energies obtained in DFT calculations reproduce the results from previous studies for Fe-C(N) systems. The cross-validation error between DFT and CE model is 5 meV. MC simulations for Fe-rich alloys show that interstitial defects are strongly attracted to each other and to vacancies, forming gas bubbles/graphite precipitations and/or alpha-cementite-like precipitations with Cr concentration increased as compared to the nominal Cr concentration.Our modeling results at 650K are in agreement with the recent APT experimental observations of the anomalous precipitation of Cr in Fe-5.8% Cr ferritic steels under irradiation, which is spatially correlated with the C/N defect clusters.
First-principles Calculations of Tungsten-based Alloys under Fusion Power Plant Conditions: Yichen Qian1; Mark Gilbert2; Lucile Dezerald3; David Cereceda1; 1Villanova University; 2Culham Centre for fusion Energy; 3Universite de Lorraine
Tungsten and tungsten alloys are being considered as leading candidates for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for the design of magnetic and inertial fusion energy reactors. Significant neutron-induced transmutation happens in these tungsten components during nuclear fusion reactions, creating transmutant elements including Re, Os and Ta. Density functional theory (DFT) calculations that allow the calculation of defect and solute energetics are critical to better understand the behavior and evolution of tungsten-based materials in a fusion energy environment. In this study, we present a novel computational approach to perform DFT calculations on transmuting materials. In particular, we predict elastic and plastic mechanical properties (such as bulk modulus, shear modulus, ductility parameter, etc.) on a variety of W-X compositions that result when pure tungsten and tungsten alloys are exposed to the EU-DEMO fusion conditions for ten years.
Multi-scale Model for Segregation of Transmutation-generated Solutes in Neutron Irradiated Tungsten: Duc Nguyen-Manh1; Matthew LLoyd2; Jan Wrobel3; Michael Klimenkov4; Luca Messina5; Sergei Dudarev1; Enrique Martinez6; Charlotte Becquart7; Christophe Domain8; 1UK Atomic Energy Authority; 2University of Oxford; 3Warsaw University of Technology; 4Karsruhe Institute of Technology; 5CEA Cadarache; 6Clemson University ; 7Univ. Lille; 8EDF-R&D
Understanding how the properties of materials change due to nuclear transmutations initiated by exposure to neutrons is a major challenge for the structural components of a fusion reactor. In this study, the multiscale modelling approach is used to investigate segregation of transmutation-generated solutes in neutron irradiated tungsten. Monte-Carlo simulations using a cluster expansion energy model for W-Re-Os-Vacancy system shows the formation of voids decorated by both Re and Os at concentrations of only 1.5% and 0.1%, respectively. This anomalous segregation is analysed in terms of the correlations between Re and Os and found in agreement with APT and STEM experimental observations of W exposed at 1200K in HFR. A systematic DFT investigation of the interaction between solutes and a <111 > dislocation loop in W demonstrates that Os has the strongest binding energy. The new energy model was employed to predict the formation of solute denuded zones near grain boundaries.
Liquid Metal Compatibility Evaluations for Fusion Applications: Bruce Pint1; Marie Romedenne1; Jiheon Jun1; 1Oak Ridge National Laboratory
Liquid metals continue to be of interest for fusion applications in heat transfer, tritium breeding and plasma-facing components (PFCs). For Li, liquid metal embrittlement is a concern and an initial study at 200°C showed decreased ductility with hollow F82H (Fe-8Cr-2W) specimens when Li was inside. For PFCs, Sn is attractive because of its low vapor pressure but capsule testing showed poor compatibility with F82H at 400°C. More promising results were observed with pre-oxidized FeCrAlMo at 400°C-500°C in static Sn. A flowing Sn thermal convection loop (TCL) was conducted for 1000 h with a peak temperature of 400°C. Large mass losses for pre-oxidized ODS Fe-(10-12)Cr-6Al and FeCrAlMo suggest that Sn will be challenging to use. Finally, a series of TCL experiments have been conducted with commercial Pb-17Li and FeCrAlMo. The most recent experiments included CVD SiC specimens. Research sponsored by the U.S. Department of Energy, Office of Fusion Energy Sciences.