High Heat Flux Materials Research Articles

Materials for in-vessel components

The EUROfusion materials research program for DEMO in-vessel components aligns with the European Fusion Roadmap and comprises the characterization and qualification of the in-vessel baseline materials EUROFER97, CuCrZr and tungsten, advanced structural and high heat flux materials developed for risk mitigation, as well as optical and dielectric functional materials. In support of the future engineering design activities, the focus is primarily to assemble qualified data to supply the design process and generate material property handbooks, material assessment reports, DEMO design criteria and material design limits for DEMO thermal, mechanical and environmental conditions.Highlights are provided on advanced material development including (a) steels optimized towards lower or higher operational windows, (b) heat sink materials (copper alloys or composites) and (c) tungsten based plasma facing materials. The rationale for the down-selection of material choices is also presented. The latter is strongly linked with the results of neutron irradiation campaigns for baseline material characterization (structural, high heat flux and functional materials) and screening of advanced materials.Finally, an outlook on future material development activities to be undertaken during the upcoming Concept Design Phase for DEMO will be provided, which highly depends on an effective interface between materials’ development and components’ design driven by a common technology readiness assessment of the different systems.

European materials development: Results and perspective

This paper addresses the EUROfusion DEMO materials research program and highlights the characterization of the in-vessel baseline materials EUROFER97, CuCrZr and tungsten as well as advanced structural and high heat flux materials developed for risk mitigation. In support of engineering design activities the focus primarily is to assemble a data base, to supply material property handbooks, material assessment reports, DEMO design criteria and material design limits for DEMO thermal, mechanical and environmental conditions.For advanced candidate material including (a) steels optimized towards lower or higher operational windows, (b) heat sink materials (copper alloys or composites) and (c) tungsten based plasma facing materials data representative for DEMO are presented.In view of preparing for future licensing and clarification of performance limits, as major step nine neutron irradiation campaigns were started recently. The campaigns include (i) baseline materials irradiated up to medium neutron dose levels, typical for DEMO 1stphase, to fill gaps in the materials property handbook and improve design limits (ii) advanced material irradiated at lower fluence for screening and down-selection and (iii) a huge temperature-dpa matrix at low fluence for material modelling verification and increased fundamental knowledge.

Development of advanced high heat flux and plasma-facing materials

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.

Baseline high heat flux and plasma facing materials for fusion

In fusion reactors, surfaces of plasma facing components (PFCs) are exposed to high heat and particle flux. Tungsten and Copper alloys are primary candidates for plasma facing materials (PFMs) and coolant tube materials, respectively, mainly due to high thermal conductivity and, in the case of tungsten, its high melting point. In this paper, recent understandings and future issues on responses of tungsten and Cu alloys to fusion environments (high particle flux (including T and He), high heat flux, and high neutron doses) are reviewed. This review paper includes; Tritium retention in tungsten (K. Schmid and M. Balden), Impact of stationary and transient heat loads on tungsten (J.W. Coenen and Th. Loewenhoff), Helium effects on surface morphology of tungsten (Y. Ueda and A. Ito), Neutron radiation effects in tungsten (A. Hasegawa), and Copper and copper alloys development for high heat flux components (C. Hardie, M. Porton, and M. Gilbert).

High particle and heat load material testing device

A brief overview of the high heat flux material testing facilities is presented which forms background for a new small-scale laboratory device capable of high heat flux material testing. The scheme of the device is presented and its working principles are described. High heat and particle loads are achieved by focusing of either electron or ion beam on the sample being tested. The sample is fixed to an actively cooled copper table. Experiments with hydrogen ion beam irradiation of an MPG-8 graphite sample are presented, in which heat loads of 270 MW/m2 were achieved. SEM images of the sample surface after testing are presented.

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

The findings of the EU ‘Materials Assessment Group’ (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision.A DEMO phase I with a ‘Starter Blanket’ and ‘Starter Divertor’ is foreseen: the blanket being capable of withstanding ⩾2MWyrm−2 fusion neutron fluence (∼20dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5MWyrm−2 first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650°C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290–320°C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200–350°C) that could be realised, as a baseline-concept, using tungsten on a copper-alloy substructure. The difficulty of establishing design codes for brittle tungsten puts great urgency on the development of a range of advanced ductile or strengthened tungsten and copper compounds.Lessons learned from Fission reactor material development have been included, especially in safety and licensing, fabrication/joining techniques and designing for in-vessel inspection. The technical basis of using the ITER licensing experience to refine the issues in nuclear testing of materials is discussed.Testing with 14MeV neutrons is essential to Fusion Materials development, and the Roadmap requires acquisition of ⩾30dpa (steels) 14MeV test data by 2026. The value and limits of pre-screening testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams are evaluated to help determine the minimum14MeV testing programme requirements.

Joining of HHF materials applying electroplating technology

Tungsten will be used as armor material for blanket shielding and is designated as high heat flux material for divertors, beyond application of improved W composite alloys as structural material. Independent from design (water- or helium-cooled), a successful development is inherently correlated with joining of tungsten with functional components.Depending on the design variants, the fabricated joints have to guarantee specific functional or structural properties, e.g., good thermal conductivity or mechanical load transmission. Tungsten shows lacks in adapted joining due to its metallurgical behavior ranging from immiscibility over bad wetting to brittle intermetallic phase formation. Electroplating has shown to overcome such drawbacks and that homogeneous functional (e.g., Ni or Pd) and filler (e.g., Cu) layers can be deposited.In this paper the progress achieved in development of electroplating processes for joining W to W or steel to steel will be shown. The main focus will be the characterization of the processed joints applying metallurgical investigations including SEM/EDX analyses and non-destructive testing. The mechanical stability of the produced joints is demonstrated by presenting recent shear test data. The W–W joints failed due to cracking in W, whereas the steel–steel joints cracked in the brazing zone at about 200N/mm2 load.

Materials R&D for a timely DEMO: Key findings and recommendations of the EU Roadmap Materials Assessment Group

The findings of the EU Fusion Programme's ‘Materials Assessment Group’ (MAG), assessing readiness of Structural, Plasma Facing (PF) and High Heat Flux (HHF) materials for DEMO, are discussed. These are incorporated into the EU Fusion Power Roadmap [1], with a decision to construct DEMO in the early 2030s.The methodology uses project-based and systems-engineering approaches, the concept of Technology Readiness Levels, and considers lessons learned from Fission reactor material development. ‘Baseline’ materials are identified for each DEMO role, and the DEMO mission risks analysed from the known limitations, or unknown properties, associated with each baseline material. R&D programmes to address these risks are developed. The DEMO assessed has a phase I with a ‘starter blanket’: the blanket must withstand ≥2MWyrm−2 fusion neutron flux (equivalent to ∼20dpa front-wall steel damage). The baseline materials all have significant associated risks, so development of ‘Risk Mitigation Materials’ (RMM) is recommended. The R&D programme has parallel development of the baseline and RMM, up to ‘down-selection’ points to align with decisions on the DEMO blanket and divertor engineering definition. ITER licensing experience is used to refine the issues for materials nuclear testing, and arguments are developed to optimise scope of materials tests with fusion neutron (‘14MeV’) spectra before DEMO design finalisation. Some 14MeV testing is still essential, and the Roadmap requires deployment of a ≥30dpa (steels) testing capability by 2026. Programme optimisation by the pre-testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams is discussed along with the minimum 14MeV testing programme, and the key role which fundamental and mission-oriented modelling can play in orienting the research.

Simulation of transient heat loads on high heat flux materials and components

In order to simulate transient events on beryllium, thermal shock experiments have been carried out in the electron beam facility JUDITH.Different grades of Chinese and Russian beryllium have been loaded in comparison to the ITER reference grade S65C. The pulse length was 5ms, and energy densities covered the range from 1.2 to 5MJ/m2. All tests have been performed at room temperature.In a second series of tests, the influence ELM-like conditions has been investigated. Multiple shot experiments with up to 10,000cycles on beryllium S65C were carried out at energy densities below the ones for the onset of crack formation in single shot experiments. Most loading parameters were similar to the single shot tests, but in addition the experiments were carried out on hot beryllium surfaces of 250°C.Post mortem examinations of the samples were carried out by optical microscopy, SEM, metallography and other diagnostic methods.

Dynamic Simulation of Multiplier Effects of Helium Plasma and Neutron Irradiation on Microstructural Evolution in Tungsten

Plasma-facing and high heat flux materials in a fusion reactor suffer two types of damage: displacement damage caused mainly by high- energy neutrons and surface damage, such as erosion, sputtering, and blistering, caused by hydrogen and helium plasma. Usually, these two kinds of damage are investigated separately. In the present study, multiplier effects of helium plasma and neutron irradiation on microstructural evolution in tungsten were investigated using computer simulations based on a rate theory. Neutron irradiation with 10 � 6 dpas � 1 at 873 K and a helium flux of 10 18 m � 2 � s � 1 with 10-keV energy were used as typical irradiation conditions. The effects of irradiation temperature and defect production rate on the interaction between helium and defects were also investigated. The simulation results revealed the rapid diffusion of helium in tungsten. The helium concentration was saturated in tungsten 0.067 mm thick within 0.01 s at 873 K, and the formation of helium- vacancy clusters was nearly uniform in the matrix except at the surfaces facing towards and away from the irradiation after 0.01 s. This meant that the 10-keV helium plasma could enhance formation of helium-vacancy clusters in the region without damage produced by helium. The concentration of 6He-V clusters reached a very high level (10 � 6 ) even after a 1-s irradiation with a defect production rate of 10 � 6 dpas � 1 at 873 K. The concentration of helium-vacancy clusters decreased with increasing irradiation temperature and decreasing defect production rate. The accumulation of helium and the formation of helium-vacancy clusters depended on not only the vacancy concentration, which was determined by the irradiation temperature and defect production rate, and helium concentrations, but also irradiation time.

Plasma facing and high heat flux materials – needs for ITER and beyond

Plasma-facing materials (PFMs) have to withstand particle and heat loads from the plasma and neutron loads during reactor operation. For ITER knowledge has been accumulated by operation experience and dedicated tests in present tokamaks as well as by laboratory experiments and modelling. The rationale for the selection of PFMs in ITER (Be, W, carbon fibre reinforced carbon) is described with regard to the critical issues concerning PFMs, esp. erosion during transient heat loads and the T-inventory in connection with the redeposition of carbon. In the fusion reactor generation after ITER the very stringent conditions of increased surface power to be removed from the plasma, a lifetime requirement of several operational years, high neutron fluences and increased operation temperature are exerting even more severe constraints on the selection of possible materials. Comparing these boundary conditions with materials under development and their further potential, only a narrow path is left regarding heat sink and PFMs. In this context the investigations on W as first wall material carried out e.g. in ASDEX Upgrade are being discussed as well as laboratory results on W-based material systems. The implications of these results are the starting point of what should form a consistent programme towards plasma-facing and heat sink materials for a fusion reactor.

Overview on fabrication and joining of plasma facing and high heat flux materials for ITER

This paper presents the results of the R&D program on the development of the joining technologies, including non-destructive inspection, for the high heat flux components carried out by the three ITER Participating Teams. Reliability requirements and design criteria are also discussed. The large amount of R&D performed within the ITER project has resulted in the development of suitable technologies, which meet or even exceed the design requirements and form a solid basis for the possible future construction of the ITER machine. However, further developments are still needed for the consolidation of the positive results with improving the reliability of the joints and reducing the manufacturing cost.

High heat load properties of TiC dispersed Mo alloys

A high heat load experiment of TiC dispersed Mo alloys using electron beams was carried out to investigate fundamental properties of thermal process due to high heat load as high heat flux materials such as the divertor tiles. As a result, although characteristic damage and gas emission of the TiC dispersed Mo alloys around the melting point are inferior to the PM-Mo, but those of the Mo-0.5wt%TiC and Mo-1.0wt%TiC below a temperature of about 2000°C were superior to polycrystalline Mo. Thus, their materials are expected to be suitable as high heat flux materials used below the recrystallization temperature. Results also indicate that it is necessary to release internal strain energy induced by the manufacturing process in order to decrease damage by grain growth and crack formation at high temperature, which was caused by migration of grain boundary and stress, respectively. Characterization of microstructure produced by the rolling process during manufacture is required to design improvement of the thermal conductivity.

Research on plasma-facing component materials

An irradiation target chamber with rotatable table and cooling systems and a set of quadrupole magnetic lenses associated with the 14 MeV, short-pulse electron linear accelerator was used to irradiate high heat flux materials. The electron gamma shower simulation code (EGS4) was modified to calculate the displacement atom concentration, deposited energy density and damage dose with 14 MeV electron irradiation. The SiC/SiC composite specimens prepared by CVI using ethyl-trichloro-silane (ETS) as SiC source and SiC fiber in Nicalon NL-200, were irradiated with a cyclic 14 MeV electron beam. The thermal diffusivity and resistance of specimens were measured before and after irradiation as a function of temperature.

High-temperature mechanical and material design for SiC composites

Silicon carbide (SiC) fiber reinforced composites (FRCs) are strong potential candidate structural and high heat flux materials for fusion reactors. A concise discussion of the main material and design issues related to the use of SiC FRCs as structural materials in future fusion systems is given in this paper. The status of material processing of SiC/SiC composites is first reviewed. The advantages and shortcomings of the leading processing technology, known as chemical vapor infiltration are particularly highlighted. A brief outline of the design-relevant physical, mechanical, and radiation data base is then presented. SiC/SiC FRCs possess the advantage of increased apparent toughness under mechanical loading conditions. This increased toughness, however, is associated with the nucleation and propagation of small crack patterns in the structure. Design approaches and failure criteria under these conditions are discussed.

Recent Results on Fusion Reactor Materials from FFTF/MOTA Program. Neutron Irradiation Effects in Molybdenum as High Heat Flux Materials.

Neutron irradiation effects in molybdenum as high heat flux materials are surveyed. Microstructures including small dislocation loops, dislocations, voids and void lattices were developed depending on irradiation conditions.Neutron irradiation induced severe loss of ductility, but the degree of embrittlement depended on materials conditions.Radiation resistance of the metal may be improved by choosing proper heat treatment and fabrication method.

照射位置で一様なビーム強度を得るためのビームレット偏向の最適化

Flat beam intensity at a target surface is required to study high heat flux materials. An optimum beam steering for this purpose is formulated and then numerically solved for the case that a multi-aperture ion source is used. It is found that the optimum distribution of the beam axes at the target surface is discrete when the beam divergence is comparable with the required flat width. The distribution gradually becomes continuous as the required width is enlarged.

Breeder and test blankets in ITER

An overview of ITER efforts is presented in this paper in the area of blanket development and ITER test program preparation. The design of ceramic and eutectic options of the driver blanket were developed aimed on a reliable operation and adequate tritium production. With an estimated net tritium breeding ratio of 0.8–0.95 ITER is able to achieve testing fluence goal of 1 MWa/m 2 during 8 years of technology phase operation. The test program was developed to ensure realistic extrapolation to DEMO on the basis of ITER experience. Submodules for the blanket, high heat flux components and material testing are to be inserted in especially designed test ports. Tests are provided for DEMO-relevant blanket design concepts.

Engineering design conditions and constraints affecting experimental fusion reactors and power-reactors

In the design of advanced technology products such as fusion reactors it is important to make clear the criteria which depend on engineering conditions and constraints. For fusion reactor design studies such as ITER, the conditions and constraints for superconducting magnets, high heat flux materials, and so on, have been examined. Using the present database, these issues are reviewed and major considerations for fusion power reactors are discussed. The effect of improving the engineering design constraints is studied by the tokamak system analysis code. It is shown that the development of high heat flux materials, high field magnets and special materials for magnets is the key to the future of fusion reactors.

Impact of engineering constraints on fusion reactors.

Engineering design conditions and constraints regarding with neutron wall load, high heat flux, superconductor magnet, and so on, play important role to design fusion reactors. Present database of first wall materials, high heat flux materials, superconductor etc., are assessed. The impact of these constraints on the next step devices and a first generation of fusion reactor is investigated by the tokamak systems analysis code.