FR - 4th International Conference
Materials Challenges for Sustainable Nuclear Fission and Fusion Technologies
ABSTRACTS
FR-1:IL02 Discrete and Continuum Models for Body Forces for Virtual Tokamak Reactor Simulations
S.L. DUDAREV, M. Boleininger, M.R. Gilbert, P.-W. Ma, D.R. Mason, L. Reali, UKAEA, Culham Science Centre, Oxfordshire, UK
Neutrons generated by nuclear reactions in the D-T plasma of a fusion tokamak device produce defects and dislocations in the structural materials, leading to significant changes in the physical and mechanical properties of materials. These changes stem from processes occurring at the atomic scale, however the effect of the changes manifests itself on the macroscopic scale. For example, fast neutrons initiate collision cascades, in which radiation defects are formed, and these defects produce highly localised fluctuating stress fields in the materials. Understanding the evolutions of stresses and strains, and their effect on reactor components requires developing multiscale models describing how the defects evolve and interact, and how the resulting microstructure on the one hand generates, and on the other hand responds to external mechanical stress or temperature gradients. The macroscopic parameters defining the evolution of irradiated materials to stress at the engineering scale are the operating temperature, irradiation dose and dose rate, as well as gravity and magnetic forces. In this presentation we explore how to relate the discrete and continuum models for the defect-induced body forces in the context of finite element analysis of a Virtual Tokamak Reactor.
FR-1:IL03 Rapid, Combinatorial Down-selection of Materials for Fusion Power
M. SHORT1, B. Dacus1, K. Woller1, G. Wallace1, E. Botica Artalejo1, K. Hattar2, C. Dennett 3, W. Zhou1, A. Peterkin1, W. Cairang1, Y. Yang4, A. Minor5, G. Zheng1, 1Massachusetts Institute of Technology, Cambridge, MA, USA; 2Sandia National Lab., USA; 3 Idaho National Lab., USA; 4Penn State University, USA; 5University of California at Berkeley, USA
It is generally accepted that the tried-and-true materials for some fusion power applications, from plasma-facing materials to structural materials, may not work for sustained use. The coupled phenomena of high temperature, radiation damage, plasma exposure, and corrosion all work to make fusion material down-selection a particularly difficult process. Here we will present a new modality of rapidly evaluating materials for fusion power, focused on the minimum viable product (MVP) of information for materials to proceed to the next selection step. Rather than focusing heavily on microstructural analysis and mechanical testing, we employ a far faster, though more information-sparse, suite of ion irradiation tests with in situ transient grating spectroscopy (TGS), micromechanical testing, and tracking *relative* material property changes to better refine material test matrices for longer, more destructive tests. Coupled effects testing will also be a focus of this talk, from tracking the fluence to the onset of fuzz growth in plasma-facing materials to the combined effects of irradiation and corrosion which sometimes shows a surprising deceleration in corrosion. The ultimate goal is to discover the best materials, as quickly as possible, to proceed to fast neutron exposure testing.
FR-2:IL01 EUROFER97 as Structural Material for the ITER Test Blanket Module and the DEMO Starter Blanket
G. AIELLO1, E. Gaganidze2, P. Lamagnère3, G. Pintsuk4, M. Rieth2, D. Terentyev5, M. Zmitko6, 1EUROfusion - Programme Management Unit, Fusion Technology Department, Garching, Germany; 2Karlsruhe Institute of Technology, Institute for Applied Materials, Eggenstein-Leopoldshafen, Germany; 3CEA, DES, IRESNE, Nuclear Technology Department, Cadarache, Saint Paul-Lez-Durance, France; 4Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Jülich, Germany; 5SCK CEN, Nuclear Materials Science Institute, Mol, Belgium; 6Fusion for Energy (F4E), Barcelona, Spain
EUROFER97 is a Reduced Activation Ferritic Martensitic steel developed by the European Union and selected as structural material for the Test Blanket Module (TBM) in ITER and for In-Vessel Components (IVCs) in the DEMOnstration fusion power plant. Material properties are already included in the “probationary rules” tome of the French nuclear code RCC-MRx used for the design and manufacturing of the TBM. Its full qualification to the code’s standards is still pending the characterization of additional batches to insure the stability of mechanical properties, the determination of the welded joints properties and the need for validation of selected design rules. Moreover, because of the much higher neutron fluence and different operational conditions, the qualification as structural material for DEMO IVCs will require supplementary investigations, in particular to quantify the additional embrittlement due to the increased number of transmutation products (H, He) as well as specific design rules for highly irradiated materials. This paper presents the EUROFER97 qualification program for RCC-MRx - as implemented by EUROfusion under agreement with F4E - and the experimental and theoretical activities carried out in parallel to enable its use in the DEMO starter blanket.
FR-2:IL02 Radiological Response of Fusion Materials: Impact of Impurities Radwaste
M.R. GILBERT, G.W. Bailey, UKAEA, Culham Science Centre, Abingdon, UK
A key challenge for nuclear systems (fusion and fission) relates to the handling and disposal of radioactive materials generated when structural components are exposed to the neutrons in reactors. For fission, radiological mitigation has historically focussed on the hazard associated with spent fuel waste. In fusion (and next generation fission systems with closed fuel cycles), however, the radioactivity will instead be dominated by the response of materials used to build reactors. Combined neutron transport and inventory (burn-up) calculations demonstrate that minor impurities in materials can dominate the radiological response, especially at the decades and beyond timescales when waste disposal is planned. Inventory calculations can be used to define the maximum acceptable concentrations for these impurities, which can then be used to drive refinements in manufactured grades of materials. However, for some materials, most notably the reduced-activation ferritic martensitic steels envisaged for the bulk of in-vessel structural applications of fusion reactors, the realistically achievable impurity levels could be above the optimum to negate the radwaste problem. Decommissioning remains a fusion challenge requiring development of waste processing technologies and new materials.
FR-2:IL03 Development of Fabrication Technology for Low Activation Vanadium Alloys as Fusion Blanket Structural Materials
TAKUYA NAGASAKA1, 2, T. Tanaka1, J. Shen1, 2, S. Jiang3, K.-i. Fukumoto4, T. Toyama5, K. Yabuuchi6, 1National Institute for Fusion Science, Toki, Japan; 2SOKENDAI (The Graduate University for Advanced Studies), Toki, Japan; 3Qilu University of Technology (Shandong Academy of Sciences), Jinan, China; 4Research Institute of Nuclear Engineering, University of Fukui, Tsuruga, Japan; 5Institute for Materials Research, Tohoku University, Sendai, Japan; 6Institute of Advanced Energy, Kyoto University, Uji, Japan
A vanadium (V) alloy with the composition of V-4Cr-4Ti is an alternative to reduced-activation ferritic/martensitic steels (RAFMs) as the fusion blanket structural material, because of its low-activation and non-magnetization characteristics, better high-temperature strength and higher thermal conductivity. The self-cooled blanket with liquid Li breeder can provide high tritium breeding ratio without Li-6 isotope enrichment and neutron multiplier, such as Be and Pb, required for the other blanket concepts. Operating temperature range for the V alloy blanket is expected from 450 to 700°C, leading to higher thermal efficiency in power generation, compared with RAFM blanket with the range from 300 to 550°C. The present paper reviews and discusses recent progress in the development of fabrication technology for the blanket. The main topics are (1) neutron irradiation response of the V alloy and its welds including the effects of post-weld and post-irradiation heat treatment, (2) dissimilar-metals bonding to out-vessel component materials, such as 316L type stainless steel and Hastelloy-X Ni base alloy, (3) the first wall armor coating using W metals and (4) further purification of the base V metal by removal of high-activation impurities harmful for early materials recycling.
FR-3:IL01 Development of Tungsten Fibre-reinforced Copper Alloy Composites for Application as Heat Sink Materials in DEMO High Heat Flux Components
A. v. Müller, B. Böswirth, H. Greuner, K. Hunger, P. Junghanns, R. Neu, J. Riesch, J.H. You, Max-Planck-Institut für Plasmaphysik, Garching, Germany; V. Cerri, A. Moriani, S. Roccella, E. Visca, ENEA, Frascati RM, Italy; U. Siefken, Louis Renner GmbH, Bergkirchen, Germany
In future deuterium-tritium magnetic confinement fusion reactors divertor plasma-facing components (PFCs) will have to sustain intense particle, heat and neutron fluxes. In order that such components are capable of exhausting high heat fluxes during fusion operation reliably structural PFC heat sink materials with high thermal conductivity as well as sufficient neutron irradiation resistance have to be applied. Typically, PFC materials suffer from deterioration of mechanical and thermophysical properties under fusion neutron irradiation while this circumstance is eventually considered an issue regarding the realisation of magnetic confinement fusion as a viable energy source of the future. A potentially advanced material in this respect is a composite reinforced with high-strength tungsten (W) fibres that are embedded in a copper alloy matrix. In this context, the contribution will summarise recent results regarding the development of tungsten fibre-reinforced copper (Wf-Cu) composites for PFC heat sink application. In more detail, this will include results regarding material property characterisations as well as the progress with respect to the fabrication and testing of Wf-Cu pipes which are directly relevant in view of their application to the W monoblock divertor PFC design.
FR-3:IL04 Controlling Interfaces for Enhanced Thermal and Radiation Stability in Tungsten PFMs
J.R. TRELEWICZ, Stony Brook University, Stony Brook, NY, USA
Microstructural instabilities are a particular concern for tungsten as a plasma-facing material (PFM) due to the high heat loads coupled with aggressive particle and neutron fluxes found in the fusion environment. Exploiting nanoscale grain boundary engineering in the design of tungsten as a PFM provides a potential pathway for improving its thermal and radiation stability. Here, we present complementary modeling and experimental initiatives on the design of grain boundary engineered tungsten alloys with rational dopant selection synergistically enhancing stability and manufacturability. Dopants are identified through thermodynamic modeling and used to guide powder processing of ternary tungsten alloys. Optimized alloy chemistries containing nanoscale compositional heterogeneities consistent with model predictions are selected to produce fully dense nanoengineered tungsten alloys via sintering. Given the affinity for titanium to segregate at the grain boundaries, which enhances thermal stability, we employ in situ ion irradiation to study the coupling between defect accumulation and microstructural evolution. The addition of grain boundary dopants is shown to stabilize the nano-alloy against irradiation induced grain growth while limiting overall damage accumulation.
FR-3:IL05 Additive Manufacturing of Tailored Plasma-facing Fusion Wall Components
D. DOROW-Gerspach1, A. Kirchner2, M. Gipperich3, Th. Loewenhoff1, T. Weißgärber2, M. Wirtz1, G. Pintsuk1, Ch. Linsmeier1, 1Forschungszentrum Jülich, Institut für Energie- und Klimaforschung, Jülich, Germany; 2Fraunhofer-Institut für Fertigungstechnik und Angewandte Materialforschung IFAM, Dresden, Germany; 3Fraunhofer-Institut für Produktionstechnologie IPT, Aachen, Germany
The heat exhaust through the inner wall is a major challenge in fusion. Manufacturing plasma facing components (PFCs) that fulfill all the demanding requirements is a challenge on the road to a reliable, economic fusion power plant. We use and combine the new possibilities offered by different additive manufacturing (AM) techniques to face the issue of lifetime of PFCs. Stresses at the interface between plasma facing material (PFM) and heat sink/structural material are the main failure mechanism of PFCs. In order to reduce them three new concepts, which can be realized using AM, are presented: (1) A geometrically graded interface by building structured joints. (2) A flexible joint made out of an assembly of tungsten wires using AM to create a solid bond between the wires and the PFM/heat sink. (3) New and complex PFC geometries optimized with finite element simulations taking the limits of AM into account. The thermal performance and stability of the new designs and the AM tungsten are evaluated with comparative high heat flux tests in the electron beam facility JUDITH 2. The second major lifetime limiting issue is the surface erosion of PFCs. We investigate the possibility to regenerate eroded wall material by AM as a method to do in-machine maintenance in future reactors.
FR-3:IL06 Design and Challenges of Tritium Breeding Blanket Systems Tested in ITER
H. TANIGAWA, W. Guan, Y. Miyoshi, T. Katagiri, T. Hirose, Y. Kawamura, Rokkasho Fusion Institute, National Institutes for Quantum Science and Technology, Ibaraki, Japan
The blanket has three major functions of neutron shielding, tritium breeding and heat extraction for power generation in fusion reactor. In ITER, a majority of blanket has a limited function of neutron shielding and then is called as Shielding blanket. The Shielding blanket consists of conventional materials of stainless steel and copper. In two test ports in ITER, prototypical modules of the blanket with the three functions, named as Test Blanket Modules (TBMs), will be tested to demonstrate the functions in fusion environment. Participation parties in ITER have been developing individual TBMs with different concepts and relevant designs. The different designs of the TBMs need different materials. Structural parts of TBMs are made of reduced activation ferritic/martensitic steels. Inside the structural container, tritium breeding materials containing lithium and neutron multiplying materials containing beryllium or lead are installed. To cool the materials and to extract heat, coolants of water, helium or liquid breeding material (e.g. lithium lead) are introduced into TBMs. Simultaneously having the three functions with utilizing the different materials makes the designs of TBMs difficult. Mainly focusing on Japanese TBM and connected ancillary systems, the present status of the development and outstanding challenges will be reviewed.
FR-4:IL02 Functional Materials - Polymer Type Materials for ITER Components
V. BARABASH, G. Kim, M. Loughlin, E. Polunovskiy, ITER Organization, St. Paul Lez Durance Cedex, France
The ITER Project is in the phase of manufacturing. For some ITER components the polymer type materials have been selected due to combination of the required electrical and mechanical properties. These materials are selected for components such as magnets, thermal shield, vacuum vessel gravity supports, anti-seismic bearings, cooling pump sealing, electrical connectors for neutral beam insulation elements, etc. Various types of materials have been considered such as PEEK, PET, silicon rubber, Viton, nitrile rubber, EPDM, polychloroprene, etc. In many cases the critical issue for application of these materials is mechanical and electrical properties degradation due to exposure to ionizing neutron and gamma radiation. The radiation conditions for various components are described in this paper. Additionally to the neutrons generated in fusion reaction in plasma the prompt photons produced in the neutron interactions with materials and the delayed photons and neutrons produced in the decay of 16N and 17N isotopes in the activated water shall be taken into account for specific components. Properties of several proposed polymeric materials are reviewed and radiation limits for these materials have been defined.
FR-4:IL03 Characterization of Advanced Dielectric and Optical Materials for Application in DEMO
R. VILA, D. Cruz, E. León, Laboratorio Nacional de Fusión, CIEMAT, Madrid, Spain
The test fusion power plant, DEMO, will have the crucial task of demonstrating reliability and very long pulse/steady-state operation, which calls for unique reliability of all diagnostic systems under high levels of neutron and gamma fluences. Consequently, diagnostics must be fully tested under DEMO conditions. The sub-project “Functional Materials” of EUROfusion has been dealing with the effects of radiation on candidate materials for optical diagnostics and control, such as windows, mirrors, lenses, etc. The key functional property is optical transmission/reflection from UV to Infrared range (optical diagnostics, visual inspection, temperature monitoring). Focus has been put on the effect of neutrons on the optical transmission of transparent materials (silica, sapphire, YAG, fluorites, diamond…) using fission reactor testing. These have been neutron irradiated up to 1 dpa. For first mirrors, results presented include important effects of H and He ions bombardment. Also, dielectric materials are used for high power transmission as well as general insulators. Results of neutron irradiation on dielectric loss are therefore presented. The implications for DEMO design are crucial, as the observed damage is much less than expected by extrapolation from low doses.
FR-5:IL01 Characterization of Innovative Advance Reactor Metallic Fuel Concepts
L. CAPRIOTTI1, D.L. Porter1, J.H. Harp2, D.M. Wachs1, 1Idaho National Laboratory, Idaho Falls, ID, USA; 2Oak Ridge National Laboratory, Oak Ridge, TN, USA
In sodium fast neutron spectrum nuclear reactors, metallic uranium-based alloys have often been chosen for their high fissile density, high thermal conductivity, and several reactor kinetic safety benefits.
In the US, R&D is on-going from the Advanced Fuels Campaign into enabling technologies that allow for improve metallic fuel performance and for transmutation of minor actinides. As such, candidate fuel compositions and designs are irradiated at the INL’s Advanced Test Reactor (ATR), and subsequently examined at the Material Fuel and Complex facilities. In addition to and complementary to ATR experiments, characterization of experiments irradiated in true fast reactors (e.g. Phenix, EBR-II) are performed.
In recent years, experiments have explored new alloys and forms beyond what was historically irradiated (U-[20Pu]-10Zr, 75% smeared density, sodium bonded fuel) to overcome limiting performance factors, such as: high swelling rate (design tested: annular fuel, low smear density, alternative alloying metals) and fuel cladding chemical interaction from fission products (design tested: additives, liners / coating) and to assess the performance of adding minor actinides (Am, Np) to the fuel systems.
FR-5:L02 Mixed Actinides Oxides Synthesis by Solution Combustion Synthesis
A. Hautecouverture, C. Rey, P. Estevenon, X. Deschanels, ICSM UMR 5257, CEA Marcoule, Bagnols sur Cèze, France
A part of fourth generation of nuclear power plant’s aim is to recycle plutonium, preserve uranium supplies and produce homogenous plutonium-rich fuels. Therefore, various routes of nitrates to oxides conversion are studied ; within them Solution Combustion Synthesis can be used to synthesize various actinides oxides with a low energy supply and a powder characteristics control. The first step is to obtain a network gel dissolving nitrates precursors (surrogates or actinides) and organic fuel in distilled water and deshydrate. The sol-gel reaction can be obtain in a furnace or on hot plate, according to the powder characteristics wanted. Various fuel and nitrates, fuel-to-nitrate ratio, heating modes and crucible geometry were studied in order to obtain a powder which characteristics are compatible with pellet shaping and sintering. These parameters seem to allow to control actinide oxidation state to some extent (uranium/glycine system for example). It is also possible to obtain powders with low carbon content, high specific surface and a controlled oxydation step without any calcination step. Before igntion, precursors gel were studied by IR and Raman spectroscopy. Ignition was studied in TG-DTA and final oxides were characterized by XDR, BET, SEM, carbon content.
FR-5:L03 Dislocation Nucleation in UO2 Spent Fuels evaluated by Synchrotron-based Laue Diffraction Method
S. Bhattacharya, G. Kuri, J. Bertsch, Nuclear Energy and Safety Department, Paul Scherrer Institut, PSI-Villigen, Switzerland
UO2 is the most used nuclear fuel for energy production in light water reactors. The UO2 lattice is disturbed during irradiation by the loss of fissile U-235 atoms and the fuel microstructure is restructured as the burnup proceeds. A key detrimental concern is the formation of so-called high burnup structure characterized by ultrafine UO2 grains and the accumulated radiation damage in the irradiated fuel matrix. The present work appraises the nucleated extended defects and evaluation of dislocation densities in irradiated UO2 microstructures, which is of significant interest for both understanding and predicting the in-service fuel behavior. Micron-sized UO2 spent fuel particles have been analyzed by synchrotron-based micro-beam X-ray diffraction. By analyzing the streak-length of Laue spots imaged in a diffraction pattern, we have quantified the geometrically necessary dislocation content in four selected samples. The estimated dislocation densities range from 1-5 × 10^14 m-2 covering a local burnup range from 30-85 MWd/kgU in spent fuels. The experimental results obtained using the present evaluation procedure have been compared with both theoretical simulations and limited experimental observations reported in the literature. All these results will be presented and discussed.
FR-5:L04 Microstructural Effect of p(O2) Variation in Mixed Oxides (MOX) Sintering
G.C.C. Miranda, CEA/DES/ISEC/DMRC, Marcoule, France and Université Grenoble Alpes, Saint-Martin d’Hères, France; G. Bernard-Granger, L. Ramond, F. Lebreton, CEA/DES/ISEC/DMRC, Université de Montpellier, Marcoule, France; A. Ndiaye, T. Gervais, Orano Melox, Chusclan, France
The atmosphere imposed during sintering of mixed oxide (MOX) pellets can influence the diffusion phenomena involved for densification, grain growth and to ensure chemical homogeneity. In this work the microstructural effects of the sintering atmosphere p(O2) variation was studied for (U,Pu)O2±x containing 11 mol% of Pu/(U+Pu). Polished cross sections were characterized using electron probe microanalyses. They showed a chemical homogeneity of the material and the presence of a single phase for all the p(O2) values used. With electron backscatter diffraction (EBSD) mappings, the average grain size measured by intercepts method showed an increase with the increase of p(O2). Following the same behavior, the distribution of grain size also presented a greater homogenization. The pellets were at least 95% of the theoretical density (TD) with a good reproducibility in all atmospheres. However, the gradual decrease of the relative density with the increase of p(O2) indicates the existence of an optimum value of the (Density/Average grain size) couple.
FR-6:L02 Effects of Solutes on Radiation-induced Segregation at Grain Boundaries in Fe-Cr and Fe-Ni Alloys
A. RAHMOUNI1, 2, O. Tissot1, E. Meslin1, A. Fraczkiewicz3, 1CEA, DEN, Service de Recherches Métallurgiques Appliquées, Laboratoire d’Analyse Microstructurale des Matériaux, Gif-sur-Yvette, France; 2Université Paris 6 Pierre et Marie Curie, formation doctorale : ED 488 – Sciences, Ingénierie, Santé (EDSIS), France; 3Mines Saint-Etienne, ENSM-SE, Département RMT, Campus de Saint-Étienne, Saint-Étienne, France
Under the extreme functioning conditions of a nuclear power plant reactor, the study of the microstructural evolution altered by damage profile is a key point to ensure its safety. Given their low swelling properties, based Ferritic Martensitic (F/M) steels are primary candidates for structural materials anticipated for the GEN IV reactors. However, predicting its behavior in terms of stress corrosion cracking (SCC) is not yet fully investigated. In fact, one of the phenomena that is responsible for the SCC is the radiation-induced segregation (RIS), which changes the chemical composition at defect sinks such as grain boundaries. The main goal of this study is to quantify the effect of the solutes on the radiation-induced segregation on Σ3 grain boundaries, a disorientation of 60° along <111> axe in cubic centered (CC) crystallographic structure.
During this work, the quantification of intergranular segregation in under-saturated ferritic binary alloys, Fe-3%at Cr and Fe-3%at Ni irradiated with Fe5+ ions (5MeV) at 450°C, 2pa were reported. The characterization was done using Atom probe Tomographic (APT) in order to obtain the chemical and special distribution along the Σ3 grain boundaries. The identification of the grain boundaries and their disorientation was done using Electron BackScatter Diffraction (EBSD) mapping on the massive sample as well as Transmission Kikuchi Diffraction (TKD) on the APT tips. The discussion will focus on the quantification of the segregation for each solute before and after irradiation.
FR-6:L03 Interstitial Loop Evolution in Alpha-Fe under Irradiation: Effects of C15 Cluster Stability and Loop One-dimensional Movement
JIE GAO, E. Gaganidze, J. Aktaa, Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Eggenstein-Leopoldshafen, Germany
We intensively used cluster dynamics simulations to investigate the evolution of 1/2<111> and <100> interstitial loops in alpha-Fe under irradiation. The cluster dynamics model implemented the kinetics of C15 clusters formation and subsequent collapse into loops of different types. The loop energetics at high temperatures were assessed using anisotropic elasticity theory. Following results were obtained in this work: 1) C15 clusters essentially act as SIA buffers by accommodating SIAs and immobilizing them. The larger the collapse size, the larger the capacity of the buffers; 2) Loop energetics calculation revealed that collapsing C15 clusters more likely reconstruct into <100> configurations with increasing temperatures; 3) The stability of C15 clusters defines the early environments for defect long-term evolutions. The influence of C15 clusters on loop evolution weakens with the increasing irradiation dose; 4) The reaction “1/2<111>+<100>” essentially provides a path for 1/2<111> loops transferring into components of <1oo> loops; 5) Temperature strongly affects the loop relative population via influencing defect diffusivities. Most of 1/2<111> loops are merged into components of <100> loops during their long-range movements when the temperature is over (equal) 500°C.
FR-6:L06 Numerical and Experimental Study of Stress Impact on Point Defect Absorption by Dislocations
D. DA FONSECA, T. Jourdan, Université Paris-Saclay, CEA, Service de Recherches de Métallurgie Physique, Gif-sur-Yvette, France; F. Onimus, Université Paris-Saclay, CEA, Service de Recherches Métallurgiques Appliquées, Gif-sur-Yvette, France; F. Mompiou, CEMES, CNRS et Université de Toulouse, Toulouse, France
Irradiation is known to accelerate creep, but underlying mechanisms are still not well established. In this study, we focus on creep due to dislocation climb, based on the different absorption rates of point defects by dislocations of various orientations with respect to applied stress. We couple simulations and experiments to quantify the impact of dislocation climb on creep. The chosen material is pure aluminium. An Object kinetic Monte Carlo model taking into account the effect of local strain on point defect diffusion is parametrized on Density Functional Theory calculations and then used to compute the absorption rates of point defects by dislocations. This allows us to determine physical parameters playing the most important role on creep by dislocation climb. Experimentally, we perform in-situ electron irradiation in a 200 keV transmission electron microscope. It enables us to observe dislocation loop growth under external stress. Microstructure characterization reveals that we mainly create interstitial Frank loops. An anisotropic distribution of loops is observed, we explain this phenomenon in light of models proposed in the literature.
FR-6:IL07 Helium Induced Degradation Scenarios for Fusion Structural Steels
A. BHATTACHARYA, Materials Science and Technology Division, Oak Ridge National Lab, Oak Ridge, TN, USA
Reduced activation ferritic-martensitic (RAFM) steels and ODS steels are the most promising structural material candidates for fusion first-wall/blanket structures. In addition to high neutron damage (50-80 dpa for DEMO and FNSF) at elevated temperatures, these steels will suffer high generation rate of transmutation induced helium (He) production – currently estimated to be ~10-12 appm He/dpa. The high He generation rate is expected to degrade the performance of steels over the entire temperature range of the blanket operations. For T≤0.4Tm, He is expected to further degrade the low temperature hardening-embrittlement scenarios. In the intermediate temperature range, between 400-500 °C, uncertainties remain on the effect of He on cavity swelling. For T≥0.4Tm, the presence of He high temperature and applied stresses means structural materials can suffer from high temperature He embrittlement (HTHE) – a non-hardening embrittlement occurring due to preferential formation of grain boundary bubbles. The susceptibility of RAFM and ODS steels to HTHE is not fully evaluated for fusion in-vessel applications. In this talk, the different degradation scenarios associated with He generation in fusion structural steels are presented to guide future irradiation studies.
FR-6:IL08 Development of Small Specimen Test Technique for Master Curve Fracture Toughness Measurements of Eurofer97 and F82H
XIANG CHEN, M.A. Sokolov, Y. Katoh, Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA; S. Gonzalez De Vicente, International Atomic Energy Agency, Vienna, Austria
Eurofer97 and F82H are two leading reduced-activation ferritic-martensitic (RAFM) steels for fusion blanket applications. The harsh environment of fusion reactors can result in severe degradation of materials fracture toughness (FT). Thus, the post-irradiation evaluation of FT is critical to understand the material behavior. Due to the space constraint of irradiation facilities, the development of small specimen test technique (SSTT) is necessary to evaluate the performance of irradiated materials. In this study, we evaluated the specimen size and geometry effects on FT of Eurofer97 and F82H. The specimen thickness ranged from 1.65 to 12.7 mm and the geometries included 1.65 mm bend bar, 4 mm compact tension (CT), and 0.5T CT specimens. Testing and analysis were performed using the Master Curve method in the ASTM E1921 standard. No specimen size effect was observed in the 4 mm and 0.5T CT specimens on the Master Curve reference temperature T0 while the bend bars yielded a higher T0. A strong effect of fatigue precrack front straightness on T0 for 0.5T CT specimens was observed. The minimum number of specimens needed for each specimen geometry has been determined. Recommendations will be given for developing and standardizing SSTT for FT characterization of RAFM steels.
FR-6:L10 Phase Separation under Irradiation in Fe-Ni and Low-alloyed Steels
Q. TENCE1, E. Meslin1, M. Nastar1, I. Mouton2, B. Décamps3, I. Joliot-Curie4, 1Université Paris-Saclay, CEA, Service de Recherches de Métallurgie Physique, Gif-sur-Yvette, France; 2Université Paris-Saclay, CEA, Service de Recherches de Métallurgie Appliquée, Gif-sur-Yvette, France; 3Laboratoire de Physique des 2 infinis; 4(IJCLab), Université Paris-Saclay, Orsay, France
Lattice point defects (PD) induced by irradiation are recognized to have a significant effect on the solute redistribution and stability of phases in alloys. We investigate their kinetic and thermodynamic effects in Fe-Ni model alloys of ferrite and austenite. We conduct ion irradiations at JANNUS facilities on FCC Fe-Ni alloys with Ni concentrations ranging from 30 to 50 at.% under several radiation fluxes, fluences, and temperatures. TEM, STEM-EDS and APT analysis revealed several kinds of solute redistribution, which include Ni-rich loops and precipitates but also the formation of SFT. Among the irradiation parameters investigated, it appears that the phase decomposition mechanism essentially depends on the alloy composition and the PD sinks, composed mostly by radiation induced defects. To model the effect of a permanent supersaturation of point defects on the dynamic equilibrium between phases, we rely on a CALPHAD database and introduce a non equilibrium point defect contribution. The resulting metastable compositions are consistent with the experimental observations. We extend this approach to Fe-based dilute model alloys of low-alloyed steels, in order to gain a better understanding of the Late Blooming Phase (LBP) formation mechanism.
FR-6:IL13 Critical Evaluation of High Temperature Helium Embrittlement Phenomena in Structural Materials
S.J. ZINKLE1, 2, Zehui Qi1, 1University of Tennessee, Knoxville, TN, USA; 2Oak Ridge National Laboratory, Oak Ridge, TN, USA
Pronounced reductions in ductility associated with the growth of grain boundary cavities during exposure to mechanical stress at high temperatures has been recognized for ~60 years as a potentially serious issue in structural materials. Moderate to high levels of helium introduced during neutron irradiation can accelerate the rate of grain boundary cavity embrittlement in structural materials for fission and fusion energy systems. In general, this high temperature helium embrittlement (HTHE) has been considered to be the key degradation process that (along with chemical compatibility and corrosion issues) controls the maximum possible operating temperature for structural materials in fusion reactors. The most extensive experimental HTHE data has been generated on austenitic stainless steels (solid solution and precipitation strengthened alloys); these experimental studies and related modeling work has provided a good mechanistic understanding of the importance of parameters such as strain rate, test temperature, overall helium content, and He trapping at nanoscale matrix precipitates. Limited experimental studies on ferritic/martensitic steels and other body centered cubic (BCC) alloys have generally reported less significant HTHE effects compared to austenitic (face centered cubic, FCC) stainless steels and Ni alloys. Potential causes of this reported difference in HTHE susceptibility between FCC and BCC alloys will be critically assessed, and recommendations for near-term research to objectively assess HTHE sensitivity will be outlined. The crucial roles of helium generation per unit time (rather than He/dpa) and performing application-relevant slow strain rate testing for assessing HTHE susceptibility will be discussed.
FR-8:IL01 Challenges for Nuclear Fusion Science and Technology
M. ABDOU, Mechanical and Aerospace Engineering Dept, University of California-Los Angeles, Los Angeles, CA, USA
Fusion Nuclear Science and Technology (FNST) is the science, engineering, technology& materials for the fusion nuclear components that generate, control and utilize neutrons, energetic particles & tritium. FNST is the remaining principal challenge and “time -controlling step” for demonstrating the practicality of fusion. The core nuclear components: blanket/FW, divertor/ PFC, and vacuum vessel. Other FNST subsystems include: tritium fuel cycle, remote maintenance, I&C, heat transport. The fusion nuclear environment is complex and unique. The combined loads of neutrons, bulk and surface heating, magnetic field, and mechanical & electromagnetic forces, all intense with steep gradients, lead to thermal-chemical-mechanical-electrical-magnetic-gravitational-nuclear interactions and multiple/synergistic effects. Recent results show these multiple/synergistic effects cannot be adequately simulated in laboratory facilities but require a DT-plasma-based facility, called FNSF, which is yet to be built and new initiative for complex 3-D modelling. The most challenging FNST issues include tritium self-sufficiency, high power density, high temperature, MHD for liquid breeders/coolants, and tritium control. Reliability/Availability/Maintain./Inspect. (RAMI) is a serious challenge.
FR-8:IL02 Challenges for Fusion Materials and Potential Solutions
J.W. COENEN1, 2, Y. Mao1, A. Litnovsky1, 3, A. Houben1, V. Ganesh1, A. Terra1, D. Dorow-Gerspach1, J. Riesch4, R. Neu4, 5, Ch. Linsmeier1, 1Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Partner of the Trilateral Euregio Cluster (TEC), Jülich, Germany; 2Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI, USA; 3National Research Nuclear University MEPhI, Moscow, Russian Federation; 4Max-Planck-Institut für Plasmaphysik, Garching, Germany; 5Technische Universität München, Garching, Germany
For the first wall of a fusion reactor materials face unique challenges. These challenges require outstanding properties in lifetime, erosion, fuel management and safety. Tungsten (W) is the main candidate material as it is resilient against erosion, has the highest melting point of any metal and shows benign behavior under neutron irradiation. However, W is intrinsically brittle and faces operational embrittlement by neutrons. Use of pure tungsten may cause an issue during a Loss-of-Coolant-Accident accompanied with air ingress due to (formation and) release of radioactive oxides. New advanced materials are being developed in order to tackle these issues. Metal Matrix Composites that incorporate extrinsic toughening mechanisms are aimed at crack resilience. Alloying of Tungsten is considered to mitigate release of radioactive oxides during accidental air ingress. Both had recent success towards upscaling. Advanced armor materials like µ-structured W are under development, here cracks are inherently suppressed. Qualification of these materials for of mechanical, oxidation and plasma-wall interaction properties is ongoing. For the development of components, the issues of joining, power exhaust, fuel retention and permeation need also be considered requiring an integrated approach.