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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]

The superconducting tokamak JT-60SA is currently being assembled at the QST laboratories in Naka (Japan). Within the European contribution in the framework of the Broader Approach, Spain has been responsible for providing JT-60SA cryostat. The cryostat is a large vacuum vessel made up of 304 stainless steel which encloses the tokamak providing the vacuum environment to reduce thermal loads on the components at cryogenic temperature. It must withstand the external atmospheric pressure during normal operation and the internal overpressure in case of an accident. Due to functional purposes, the cryostat has been divided in three assemblies: the Cryostat Base (CB), the Cryostat Vessel Body Cylindrical Section (CVBCS) and the Top Lid. For transport and assembly reasons the cryostat is made up of 20 main parts: 7 making up the CB and 13 making up the CVBCS (including the top lid). The joints between them rely on bolted flanges together with light seal welds, non-structural fillet welds performed from inside and/or outside of the cryostat. The single wall is externally reinforced with ribs to support the weight of all the ports and port plugs and also to withstand the vacuum pressure. The material is SS 304 (Co < 0.05 wt%) with a permeability (μ rel) below 1.1. The CVBCS made of a single wall is a stainless steel shell with a thickness of 34 mm. The CB was manufactured and assembled in-situ in 2013, while the CVBCS has been manufactured, assembled, measured by a Spanish company (ASTURFEITO S.A) and delivered to Japan in November 2017. The paper summarizes the assembly and final measurement of the CVBCS at the factory. [ABSTRACT FROM AUTHOR]

• The overall plant system related to the water-cooling system has been developed for JA DEMO. • Management of tritium concentration in the primary cooling system is clarified for JA DEMO. • Tritium permeation from the primary to the secondary cooling system was found to be less than the Japanese PWR restriction value. • A pressure suppression pool is effective to suppress the tritium release to the environment at the ex-vessel LOCA. • The system diagram of water coolant and tritium was also delineated for JA DEMO. The conceptual design of Japan's fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA. [ABSTRACT FROM AUTHOR]

This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier. [ABSTRACT FROM AUTHOR]

In 2013, the IFMIF, the International Fusion Material Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase, framed by the Broader Approach Agreement between Japan and EURATOM, accomplished in 2013 its mandate to provide the engineering design of the plant on schedule [1] . The IFMIF aims to qualify and characterize materials that are capable of withstanding the intense neutron flux originated in D-T reactions of future fusion reactors due to a neutron flux with a broad peak at 14 MeV, which is able to provide >20 dpa/fpy on small specimens in this EVEDA phase. The successful operation of such a challenging plant demands a careful assessment of the Conventional Facilities (CF), which have adequate redundancies to allow for the target plant availability [2] . The present paper addresses the design proposed in the IFMIF Intermediate Engineering Design Report regarding the CF, particularly the IFMIF's Nuclear and Industrial HVAC design. A preliminary feasibility study, including the initial configuration, calculations and reliability/availability analysis, were performed. The nuclear HVAC design was developed progressively; first, by establishing a conceptual design, starting from the system functional description, followed by the identification of the corresponding interfacing systems and their technical requirements. Once the technical requirements were identified, safety zones were identified based on the radiation classification, frequency dose and parameter of Derivate Air Contamination (DAC). The zone color was determined to match the room's radiation classification. The system design was further developed by defining and creating a Block Diagram with basic and additional information, eventually resulting in a Process Flow Diagram concurrent with the equipment layout definition. Subsequently, we studied and developed the various Piping & instrumentation diagrams (P&ID's), air duct layout and equipment list for different air handling units, air ducting as well as a layout plan of the equipment piping, which was eventually integrated into the 3D model of the building and coordinated with others subsystems of the IFMIF. [ABSTRACT FROM AUTHOR]

• Evaluation of error field based on the accuracy of manufacturing of superconducting coils on JA DEMO were investigated. • The error field with all errors is 0.52 mT (ΔB TMEI /B T < 10-4). • The EFCC current was 200 kAT using two types of EFCC which are different toroidal angles outside the vacuum vessel. • The tolerance in coil fabrication would be mitigated by a factor of 2.5-5 compared with that of ITER on JA DEMO. Generally, DEMO requires larger toroidal field (TF) coils than ITER, resulting in one of the major difficulties, the tolerance in TF coil fabrication. This paper presents the possible solutions based on the design study on Japan's DEMO (JA DEMO). It was confirmed that, in the case of adopting a mitigated tolerance by a factor of 2.5–5 compared with that of ITER, the resulting error field of TF coils is correctable to an acceptable level in terms of locked mode avoidance. In addition, the design of the error field correction coil (EFCC) on JA DEMO was investigated. [ABSTRACT FROM AUTHOR]

Both the China Fusion Engineering Test Reactor (CFETR) and Japan's DEMO solid water-cooled blanket have proposed a mixed pebble bed for neutron multiplier and tritium breeder, which has higher initial packing factor with the special size ratio of the mixed pebble and therefore improves the effective thermal conductivity and tritium breeding comparing with the single particle pebble. However, the study of the behavior of the mixed pebble bed under various disturbances is still lacking. It is found that the mixed pebble bed under the disturbance will produce the phenomenon of "sorting" in the experiment. The effect of sorting on the mixed pebble bed has not been considered for the evaluation of TBR (tritium breeder rate). The pebble bed will inevitably suffer from various forms of disturbance during service of the blanket in the fusion reactor. Therefore, it is significant to consider the influence of the sorting phenomenon caused by the disturbance on the design of the mixed pebble bed structure of the fusion reactor. This paper presents the sorting phenomenon with mixed pebble bed under disturbance, the results indicate that: 1). The sorting degree is sensitive to the structure of the pebble bed. 2). The sorting behavior of the pebble bed is mostly influenced by the disturbance intensity, and also influenced somehow by the density ratio and size ratio in different way and magnitude. As a result, it is necessary to pay attention to this phenomenon with the design of the blanket using the mixed pebble bed. [ABSTRACT FROM AUTHOR]