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JA DEMO selected the vertical maintenance scheme as the primary maintenance option. In order to improve reliability of Remote Handling (RH) and plant availability, update of in-vessel transferring mechanism of the segment, pipe unit structure on maintenance port were carried out. The RH equipment is composed of end-effectors (grippers) for the banana segment, a power manipulator, a telescopic guide and a carrier. The pipe structure in vertical maintenance port was updated considering the thermal expansion, easy handling (short maintenance time) and alignment of welding groove. Based on estimation of maintenance time, four works in parallel will be required for plant availability reachable for commercialization (>60%).

Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic steels have been investigated; however, oxidation of the steel surface is expected under an actual DEMO reactor condition, and then the tritium permeation behavior will be changed. In this study, deuterium permeation through the steels heat-treated under simulated environment conditions has been investigated for more precise predictions of tritium loss at DEMO reactor blankets. Reduced activation ferritic/martensitic steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen at 300, 400 and 500 °C for 100 and 200 h to simulate a solid breeder DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. An iron oxide layer was formed on the sample surface, and the thickness of the layer was 50 nm‒12 μm. The oxide layer on the sample surface heat-treated at 500 °C for 100 h decreased deuterium permeation by a factor of 5. After the permeation tests, dissipation of the oxide layers was confirmed.

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 5.7g-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 94kg/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 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, confinement concept of tritium release from PHTS is discussed under the condition of an 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.

A conceptual design of breeding blanket module with pressure tightness against in-box LOCA has been carried out, based on a pressurized water-cooled solid breeder blanket. The cooling water for DEMO is operated at the PWR water conditions of 15.5 MPa and 290 ºC-325 ºC. In this design, the breeding area of the module is divided into 0.1-m-squared cells with rib structure and has simple interior for mass production using a mixed pebbles bed of Li2TiO3 pebbles and Be12Ti ones. As a result, a rib with the thickness of 0.015m is needed to withstand the design pressure of 17.2 MPa by a stress analysis. The cooling system for the blanket module is designed by fluid dynamics analysis based on the PWR water conditions, and the outlet coolant temperature and the pressure drop are 321 ºC and 0.32 MPa, respectively. It was found that the self-sufficient production of tritium is likely to be satisfied with the blanket radial width thickness of 0.70 m or more and with an idea to improve TBR that the coolant is changed from existing light water to heavy water.

Considering the water cooling system in fusion reactors, there are several fusion-specific challenges in designs of in-vessel components and their cooling water systems (CWS). In this research, solutions of the challenges have been discussed and indicate the CWS concept of Japan's DEMO. The CWS transfers thermal energy of blanket and a part of divertor to a generator system, and thermal energy of the other part of divertor is used as heat utilization in plant. The required performance of system devices such as pumps is expected to be achieved by proven technologies and its extensions. It is also indicated that blanket can work as the low-pass filter, and can suppress the effect of fusion output fluctuation. The designed cooling water system requires 70 MW electric power and generates 620 MW power with a 1.5 GW fusion power plasma. This research indicates the basic concept of Japan's DEMO CWS that can be practically achieved.

In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such as tritium production rate. The influences of the cooling water and test cell wall were evaluated in order to decide its design. The DCDTM is one to achieve irradiation data of functional materials on diagnostic and control devices such as mirror and window. It was clarified from its nuclear analysis that the neutron fluence obtained in the DCDTM was enough to investigate the accumulated irradiation effects on the materials. In addition to the irradiation tests on the fusion reactor materials, various neutron applications are planned at A-FNS. One of the applications is medical isotope production. We clarified that enough amount of the medical isotope molybdenum-99 could be produced in comparison with the demand in Japan by using Radio-Isotope Production Module (RIPM). The RIPM affects little influence on the fusion reactor material irradiation tests.

The blanket modules in DEMO are made of reduced-activation ferritic martensitic (RAFM) steel F82H. This material is ferromagnetic and it drags the magnetic field lines into the first wall (FW). Because of this, the heat load by the plasma heat flux, which goes along the magnetic field line will become higher. In this research, the first analysis of such effect has been done. The extra magnetic field Bm made by RAFM wall becomes higher at inner midplane, and the heat load at the module front surface becomes 1.3 MW/m2 to 5 MW/m2. Additionally, near the toroidal gaps, BM becomes high. Thus, at the top of the FW, magnetic field lines are dragged into the toroidal gaps directly because, the magnetic flux surface is not closed. This makes high (about 10MW/m2) heat load concentration at the moduel edge. The effect of the NBI port is also analyzed. Also near the port, Bm becomes high and the orbit of the magnetic field lines are changed. The effect of this doesn't occur near the port, but far region such as inner midplane or top of the FW. The heat load becomes 6 MW/m2 at inner midplane. These results indicate that the effect of RAFM steel on the FW heat load is not negligible, and more detailed analysis is necessary.

Plasma heat flux in the peripheral plasma reaches the first wall (FW) along a magnetic field line, and it sometimes causes several MW/m2 orders of magnitude high heat flux concentration at narrow region, such as the edge of the blanket module. Thus, to assess and to reduce the heat load is key issue in the DEMO design activity. In this research, a new heat load analysis code is introduced based on the e-folding model. In this code, the decay length λ is changed depending on wall connection length of magnetic field line (defined as the length of magnetic field line from the wall to the wall), and parallel heat flux q// is calculated in each flux tube. This code can calculate the FW heat load simulating actual blanket module shapes. The 0.23 MW/m2 peak heat load at inner midplane in the case of the ideal FW (without gaps between blanket modules) is increased to 22 MW/m2 at toroidal module edge in the case of box shaped module. To shadow the edge and reduce such peak heat load, toroidal, and poloidal roof shaping is applied. Required roof height is analyzed from this code calculation. After shaping, peak heat load is reduced to 1.2 MW/m2. This value is under the allowable value 1.5 MW/m2, and in this case, surface temperature is also less than allowable temperature.