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• The blanket testing facility for WCCB TBM was newly built in Rokkasho Fusion Institute, Japan. • An apparatus for flow accelerated corrosion testing under coolant condition of TBM was installed. • An apparatus for high heat flux testing under cooling with pressurized water using electron beam gun was installed. • An apparatus for ingress testing of pressurized water inside TBM was installed. • An apparatus for evaluating characteristics of beryllium-water reaction was installed. Intensive research and development activities have been conducted for designing a water-cooled ceramic breeder (WCCB) blanket system. In this system 325 °C/15.5 MPa of pressurized water removes heat injected and accumulated in blanket containers made of a reduced activation ferritic/martensitic steel, F82H. The container has lithium ceramics and beryllium pebbles as tritium breeder and neutron multiplier, respectively. In order to demonstrate performance of the system, especially for high-temperature pressurized water coolant, experimental apparatus have been installed in Rokkasho Fusion Institute, National Institutes for Quantum Science and Technology, Japan. These apparatuses enable to conduct the following experiment on full-scale mockups: 1) high heat flux (HHF) test to demonstrate heat removal of surface heat flux by electron beam gun simulating plasma, 2) flow assisted corrosion, 3) coolant ingress into dummy pebble beds, and 4) beryllium-water reaction between the coolant and the pebbles. In this paper, overviews, specifications, early results of these apparatus are reported. [ABSTRACT FROM AUTHOR]

The Broader Approach (BA) activities aim to complement the ITER project and to promote the early realization of fusion energy through research, development, and tests of technologies supporting the future demonstration fusion reactor (DEMO). These activities are implemented under the BA agreement, which was signed and ratified in 2007 between Euratom and the Government of Japan. In essence, the BA activities consist of three projects: a) The Satellite Tokamak Programme Project JT-60SA, the world's largest superconducting tokamak until ITER starts, aims to support the assembly, commissioning and preliminary operation of ITER, and carry out demonstration and optimisation of steady-state operation of advanced plasma configurations for DEMO; b) The International Fusion Energy Research Centre (IFERC), which comprises three different sub-projects: the DEMO Design Research and Development Coordination Centre to coordinate design and R&D on materials and components for DEMO, the Fusion Computer Simulation Centre for the simulations of fusion plasmas, the analysis of experimental data, modelling of ITER operation, and contribution to the design of DEMO, and the Remote Experimentation Centre to allow scientists to participate remotely in fusion experiments from its control room in Japan; c) The Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) focused on the design and validation of key components needed for the future Fusion Neutron Source facilities to characterize materials envisioned for DEMO. After 13 years of fruitful collaboration, Euratom and Japan launched the second phase of the BA activities in 2020. It is focused on exploiting and enhancing the facilities that have already been built and on working more closely than ever with ITER as the latter advances towards the first plasma. This paper will present a synthesis of the achievements already obtained thanks to the three BA projects and will address the future developments planned in 2023 and beyond. [ABSTRACT FROM AUTHOR]

In a fusion reactor, the plasma based on a deuterium-tritium reaction generates high energy neutron flux interacting with the materials of the plasma facing components. In order to study the irradiated behavior of these materials, the International Fusion Materials Irradiation Facility (IFMIF) was conceived to generate these fusion relevant neutrons through d -Li stripping source. IFMIF is presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach agreement signed between EURATOM and Japanese Government in 2007. This agreement mandates the IFMIF/EVEDA project to validate the design of the different systems of IFMIF and to produce an integrated engineering design of IFMIF, together with the data necessary for future decisions on its construction and operation. While the Engineering Validation Activity (EVA) of the Lithium Target Facility and the Test Facility was completed by constructing prototypes, the EVA of the Accelerator Facility with the Linear IFMIF Prototype Accelerator (LIPAc) is still on-going at the Rokkasho Fusion Institute, Japan. This paper overviews the achievements in the previous phase ended on 31 March 2020 and the progress in the current phase to complete the commissioning of the LIPAc. [ABSTRACT FROM AUTHOR]

JT-60SA is a joint international fusion experiment being built and operated in Japan under the framework of the Broader Approach Agreement and Japanese National Fusion Programme, aiming at an early realization of fusion energy by conducting supportive and complementary work for the ITER project, all towards supporting the basis for DEMO. In this context, a collaboration has been established via EUROfusion between Fusion for Energy and CEA to develop actively cooled targets of the JT-60SA divertor. The first Actively Cooled Divertor (ACD) target, planned to be operated in 2029, has graphite as armour material and TZM as heat sink material. To be DEMO relevant, ACD target with tungsten armour is planned to be operated in 2033, with either CuCrZr or TZM as heat sink material. This paper focusses on the choice of the design (material, dimensions) of these targets. Results of thermal and structural analysis of these targets are presented. The calculations, carried out in the environment of Ansys 2021R2 code, have taken into account the heat loading under plasma operation (with a peak heat flux of 10 MW/m2) and the electromagnetic loads from a representative plasma disruption (VDE 30 ms). In order to check the compliance with nuclear requirements, RCC-MRx 2015 has been used for the mechanical assessment of metallic components, while for the graphite the comparison between the maximum stress and the yield stress has been considered. The performed analyses show that the thermal load on target during the plasma heat loading drives target designs due to differential expansion at the interface between the heat sink and the flat tiles. Additionally, on-going assessments on these divertor targets are presented, focussing on experimental High Heat Flux testing, as well as computational thermal-hydraulic studies on enhanced target configurations. [ABSTRACT FROM AUTHOR]