Tungsten is plasma-facing material (PFM) for the divertor and test blanket module to be deployed in ITER. During operation, sub-components made of tungsten will be subjected to cyclic heat exposure, plasma particles loads and bombardment by fast neutrons. The optimization of tungsten grades, in terms of mechanical properties, hopefully offering enhanced resistance against high heat flux and irradiation damage is ongoing, and recently A. L.M.T. has proposed an extra cross-rolling step for the industrial fabrication of thick tungsten plates. Here, we present the results of recent experiments aimed at characterizing the impact of high flux plasma load, thermal shock and neutron irradiation on the damage induced in newly released grades and conventional ITER specification tungsten. The assessment included the tensile mechanical properties before and after the irradiation as well as microstructural investigation of surface modification due to the high heat flux testing. The neutron irradiation and high heat flux tests were performed in a wide range of temperatures (400 - 1200 degree celsius) and covered typical exposure conditions that will be met by PFM in the ITER divertor.

Tritium retention in the castellated structure of beryllium limiters used in JET with the ITER-like wall (ILW) during the first (ILW1), third (ILW3) and all three (ILW1-3) campaigns were examined and evaluated. Tritium was deposited on the surfaces inside the castellation grooves together with deuterium, beryllium, oxygen, carbon and small amounts of metallic impurities such as nickel, copper and tungsten. The tritium content after the ILW1 campaign was greater than after the ILW3 campaign. This is attributed to the steadily decreasing amount of carbon impurities in JET from campaign to campaign. The majority of tritium was retained in shallow regions in the grooves, up to 2 mm from the entrance to the gap. It was comparable on all sides of the castellation, i.e. no difference has been detected between the toroidal and poloidal gaps. Secondly, the tritium retention in the gaps was similar on all specimens independent of their position in the tokamak, while the retention on the plasma-facing surfaces clearly depended on the tile position. The tritium deposition patterns in the castellation were also compared with the deuterium distribution determined in earlier studies.

The existing nano-indentation method and developmental micro-tensile testing method are more attractive in accurately defining the irradiation effect on the mechanical behavior and adopted in this study to identify the detailed irradiation behavior of pure tungsten. The pure tungsten (A.L.M.T. Corp., Japan) produced by the powder metallurgy and hot-rolling process has a purity of 99.99 wt%. The irradiation was conducted to a maximum of 18.0 dpa by the TIARA (Takasaki Ion Accelerators for Advanced Radiation Application) at 500 and 800 ℃. The micro-tensile specimens with the gauge section of about 0.8 µm (width) × 0.8 µm (thickness) × 10 µm (length) were machined by a focused ion beam device. The nano-indentation results showed the saturation of irradiation hardening and its irradiation temperature dependence. In the case of 500℃, the indentation hardness of pure tungsten irradiated in TIARA increased with increasing dose, and it reached 6.5 GPa after irradiation to 1.0 dpa. Beyond 1.0 dpa, indentation hardness approached a constant. In the case of 800℃, indentation hardness reached 8.1 GPa by irradiation to 3.0 dpa with an irradiation hardening saturation. Moreover, with the stress vs. strain curve by the micro-tensile tests, we found a similar tendency to the nano-indentation results. Additionally, tungsten micro specimens showed a ductile fracture at all conditions in these tests. The total elongation was independent of irradiation dose and irradiation temperature.

The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokkasho Fusion Institute is reported. In order to advance the Japanese DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach activity and ITER-related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D; design of a fusion neutron source and development of an accelerator, ITER Test Blanket System; tritium handling technology; and information technology infrastructures, including a supercomputer system and a remote experimentation system, are carried out for a Fusion DEMO reactor.

Based on the results of the IFMIF/EVEDA project in the broader approach (BA) activities, the conceptual design of the Advanced Fusion Neutron Source (A-FNS) to obtain irradiation data of DEMO blanket materials is under consideration in Japan. This paper describes the progress of the A-FNS conceptual design, which consists of a 40 MeV, 125 mA deuteron beam, a liquid lithium target system, and a test and post-irradiation test facilities. the main objectives of the A-FNS are the acquisition of RAFM irradiation data up to 2035 and the tritium recovery test on the blanket. In addition, it is expected to be used for various neutron researches. The technical design activities of the A-FNS are focused on improvement of the lithium target design, review of impurity criteria, remote handling, and irradiation application plan.

This work addresses an overview of the recent progress, associated risks and development plans of structural materials for in-vessel components (IVCs) in DEMO plants. Reduced Activation Ferritic Martensitic Steels form the primary structural materials for most DEMOs IVCs and will be the focus of the paper. An overview of EU- and J-DEMO programs RAFM steels developments is presented. We highlight the present status in their development for use in DEMO plants, with a focus on the structural materials’ operational, and project-oriented, requirements. The development of materials property handbooks from high-quality data is illustrated. A process to validate these steels for operation in DEMO IVCs is summarised, revealing the pragmatic procedures ongoing, and limitations of this approach due to the synergistic operational effects on IVC materials; the use of in-situ or surveillance monitoring is outlined as a method to accommodate this limitation in synergistic effects and allow confidence in future DEMO operations. The development of advanced modifications to F82H and EUROFER are highlighted, with minor modifications leading to improved low and high-temperature operational design space being open for DEMO reactors.

Based on results from the IFMIF/EVEDA project in the Broader Approach (BA) activities, a conceptual design of the Advanced Fusion Neutron Source (A-FNS) in Rokkasho, Aomori aiming at obtaining material irradiation data for a fusion DEMO reactor is presented in this paper. The A-FNS is composed of an accelerator facility with a 40 MeV and 125 mA deuteron beam, a test facility including a liquid lithium target system and a post irradiation examination facility. The prime objective of A-FNS is the data acquisition of RAFM irradiation data by 2035 and the tritium recovery test on the blanket and durability test on the fusion diagnostic and control devices. A particular attention in the design is paid on an integration of the test facilities by adopting a newly designed for A-FNS. Furthermore, as a unique usage of A-FNS neutron, the multipurpose usage has been investigated. For the mission achievement, the A-FNS/CDA will continue until 2020 and then the EDA will be continuously implement between 2020 and 2024. After the EDA, the A-FNS facility is to start constructing from 2025 toward the neutron operation around 2030. As current activity for A-FNS conceptual design, the analyses and investigations for test module, Li target and loop, remote maintenance, neutron monitor and multipurpose usage has been progressed to solve the design issues.

Silicon carbide (SiC) fiber reinforced SiC matrix composites continue to undergo development for fusion applications worldwide because of inherent advantages of the material including low activation, high temperature capability, relatively low neutron absorption, and radiation resistance. This paper presents an international overview of recent achievements in SiC-based composites for fusion applications. Key subjects include applications in fusion reactors, high-dose radiation effects, transmutation effects, material lifetime assessment, and development of joining technology (processing, test method development, irradiation resistance, and modeling capability). This paper also discusses synergy among research for fusion materials and non-fusion materials (for fission and aerospace applications). Finally, future research directions and opportunities are proposed.

The diameter of pressurized tubes of F82H and B-doped F82H irradiated up to similar to 6 dpa have been measured by a non-contacting laser profilometer. The irradiation creep strains of F82H irradiated at 573 and 673K were almost linearly dependent on the effective stress level for stresses below 260 MPa and 170 MPa, respectively. The creep strain of (BN)-B-10-F82H was similar to that of F82H IEA at each effective stress level except 294 MPa at 573K irradiation. For 673K irradiation, the creep strain of some (BN)-B-10-F82H tubes was larger than that of F82H tubes. However, the generation of similar to 300 appm He did not cause a large difference in the irradiation creep behavior at 6 dpa.

In situ measurements of the volume electrical conductivities of chemical-vapor-deposited silicon carbide (CVD-SiC) samples were carried out under irradiation by 2.5- and 14-MeV fast-neutron beams in air at room temperature. A slight radiation-induced conductivity (RIC) was detected under fast-neutron irradiation. A radiation-induced electrical degradation (RIED)-like behavior was observed in the form of degradation of the base conductivity in the absence of irradiation with increase in the fast-neutron fluence. The SEM micrograph images and XPS analysis of the surface of the fast-neutron-irradiated CVD-SiC samples revealed that the dissociation of carbon from existing SiOxC3−x compositions did not only occur via displacement damage, but also via ionizing effects (radiolysis), thereby leading to the observed RIED-like behavior as well as RIC in the electrical properties of the CVD-SiC samples.

The present study reports the compatibility of a reduced-activation ferritic steel F82H and Ni with liquid Pb–Li under rotating flow conditions at 600°C. Cross-sectional observation of Ni using field emission electron probe micro-analyzer (FE-EPMA) after exposure for 100h revealed that severe grain boundary penetration of Pb into Ni occurred up to approximately a depth of 700μm, causing liquid metal embrittlement (LME). In contrast, the results for F82H after exposure for 500h showed the formation of pitting holes and a Cr-depleted layer at the surface with an approximate maximum depth of 10μm. Oxide particles were also found in the Pb–Li region in the F82H specimen after exposure. The radio-frequency glow discharge spectrometers successfully detected Li and indicated Li oxide formation at the surface.

The silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite system for fusion applications has seen a continual evolution from development a fundamental understanding of the material system and its behavior in a hostile irradiation environment to the current effort which is directed at a broad-based program of technology maturation program. In essence, over the past few decades this material system has steadily moved from a laboratory curiosity to an engineering material, both for fusion structural applications and other high performance application such as aerospace. This paper outlines the recent international scientific and technological achievements towards the development of SiC/SiC composite material technologies for fusion application and discusses future research directions. It also reviews the materials system in the larger context of progress to maturity as an engineering material for both the larger nuclear community and broader engineering applications.

Recent progress of small specimen test technique (SSTT) and the engineering design and engineering validation tests of high flux test module (HFTM) for the IFMIF (International Fusion Materials Irradiation Facility) test cell is mainly summarized and evaluated in the IFMIF/EVEDA (Engineering validation and engineering design activities) projects under Broader Approach (BA) Agreement between EURATOM and Japan. The evaluation of the optimization of shape and size of specimen and the arrangement is very important for SSTT of IFMIF. Effects of specimen size on mechanical properties such as impact properties and ductile-to-brittle transition temperature (DBTT) are known to occur in ferritic/martensitic steels, and some parts of them have been prepared in the guideline and standard of mechanical tests by ASTM-international and ISO. However, our research of ferritic/martensitic steel F82H shows that it does not match with our data, i.e., master curve method for fracture in ductile-to-brittle transition behaviour of F82H steel. Accordingly, we need to modify and develop these standards for the tests including small size specimens of fusion materials in IFMIF. In the design of HFTM, two types are preparing for RAF/M steels by EU KIT team and for the advanced materials by JA team, respectively.

Silicon carbide has been studied for, and utilized by, nuclear systems for decades. Historically, it has been used as the micro pressure vessel in high-temperature gas-cooled reactor fuels, although it has been studied, and is currently being developed, in its composite form for structural application in gas-cooled and light water fission reactors. This article discusses in some detail the effects of neutron irradiation on the thermophysical properties of the currently known radiation-stable forms of silicon carbide.

A SiC/SiC composite is an attractive candidate material but it is a challenge to apply it to the practical components because of the inherent brittle-like failure and structural anisotropy. This study aims to evaluate the failure behavior of SiC/SiC composites by various test modes. Comparison between tensile and compressive strengths revealed the clear axial anisotropy of failure strength. The in-plane shear strength by the off-axial tensile method is invalid unless considering the mixed failure modes. Alternatively, it was demonstrated that the in-plane shear strength can successfully be obtained by the Iosipescu method. The true inter-laminar shear strength can be identified by the diametral compression method.

Development of small specimen test techniques is one of the key issues for irradiation studies. This study primarily aims to develop a new tensile test method using small notched tensile specimens. For that purpose, the crack extension behavior of the notched tensile specimens was first evaluated. Observing the crack propagation behavior, the apparent notch insensitivity of the axial tensile properties was identified. Due to the notch insensitivity, it was found that the key tensile properties such as the proportional limit stress and ultimate tensile strength can easily be predicted from the test results on the notched specimens. In contrast, for the off-axial tensile tests, the apparent notch sensitivity was identified. (C) 2011 Elsevier B.V. All rights reserved.

As a part of the Broader Approach activities, R&D on blanket related materials, reduced-activation ferritic martensitic (RAFM) steels as a structural material, SiCf/SiC composites for flow channel insert in the liquid blanket and/or use as advanced structural material, advanced tritium breeders and neutron multiplier, has been initiated directed at DEMO. As part of the RAFM steel mass production development, a 5ton heat of RAFM steel (F82H) was procured by Electro Slag Re-melting as the secondary melting method, which was effective in controlling unwanted impurities. An 11ton heat of EUROFER was also produced. For the SiCf/SiC composite development, NITE- and CVI-SiCf/SiC composites were prepared as reference materials and preliminary mechanical and physical properties were measured. Also compatibility tests between SiC and Pb–17Li have been prepared, related to the He-cooled Li–Pb blanket concept. For the beryllide neutron multiplayer Be–Ti alloy development, large size rods of about 30mm diameter were fabricated successfully in EU.

Silicon carbide was first proposed as a low activation fusion reactor material in the mid 1970s. However, serious development of this material did not begin until the early 1990s, driven by the emergence of composite materials that provided enhanced toughness and an implied ability to use these typically brittle materials in engineering application. In the decades that followed, SiC composite system was successfully transformed from a poorly performing curiosity into a radiation stable material of sufficient maturity to be considered for near term nuclear and non-nuclear systems. In this paper the recent progress in the understanding and of basic phenomenon related to the use of SiC and SiC composite in fusion applications will be presented. This work includes both fundamental radiation effects in SiC and engineering issues such as joining and general materials properties. Additionally, this paper will briefly discuss the technological gaps remaining for the practical application of this material system in fusion power devices such as DEMO and beyond.

Breeding blankets are the most important components in DEMO. The DEMO blanket has to withstand high neutron flux typically 15–30 dpa/year under continuous operation. Therefore integrated and effective development of blanket structural materials and breeding/multiplying materials is essential in the blanket development for DEMO. In parallel to the ITER program, broader approach (BA) activities are initiated by EU and Japan. Based on the common interest of each party towards DEMO, R&D on reduced activation ferritic martensitic (RAFM) steels as a DEMO blanket structural material, SiCf/SiC composites which have potential for use in DEMO blankets, advanced tritium breeders and neutron multiplier for DEMO blankets, and tritium technologies including tritium behavior studies in advanced materials for DEMO blanket applications will be carried out as a part of the BA activities.

The design progress in a compact low aspect ratio (low A) DEMO reactor, ‘SlimCS’, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.