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Minimum sets of in-vessel components were installed for initial operation of JT-60SA in order to product and control the plasma and to know the plasma basic information, including inboard first wall, upper divertor, protection limiter, glow electrodes and magnetic sensors. All of magnetic sensors were installed within the required accuracy with in-situ measurement with laser tracker. The inboard first wall and the upper divertor were installed for the limiter configuration at the plasma initiation and the divertor configuration, respectively. They have carbon tiles as the first wall, which must be installed with accuracy of ± 1 mm in order to avoid the heat concentration. However, the vacuum-vessel wall had some deformation. Therefore, we measured in advance with laser tracker the position of the vacuum vessel where the bases of the inboard first wall and upper divertor would be welded. By using these position data and 3D CAD, we customized all bases. Owing to these procedures, we could install the first wall with required accuracy. Also, all other components were installed with required accuracy. [ABSTRACT FROM AUTHOR]

Resistive wall mode control coils (RWMCs) will be installed in the vacuum vessel of JT-60SA. Since 200 °C baking of vacuum vessel on JT-60SA is performed to remove the impurity in the plasma facing components, in-vessel components must be made with heat-resistant materials such as metal and ceramic or must be cooled with water. We looked for the heat-resistant organic insulator for JT-60SA RWMC and performed the heat resistant test for several resins. However, no resin passed the test (Sukegawa et al., 2015). We again performed the heat resistant test of several organic insulator in the circumstances without air and finally found that the some Bismaleimide–Triazine (BT) resin has enough heat resistance without air. We designed the eight-turn RWMC with the BT resin. Based on this design, the specification of RWMC and power supply requirement are decided. For the design of RWMC, the finite element (FEM) analysis of electromagnetic force during disruption and thermal stress during baking and operation were performed. Manufacturing of the RWMC has already launched. [ABSTRACT FROM AUTHOR]

Abstract: For a tokamak fusion DEMO reactor with the fusion output of 2.95GW, neutronic and thermal design of blanket is under way to find a feasible blanket concept. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li4SiO4, Li2TiO3 and Li2ZrO3) and Be multipliers (Be and Be12Ti). Based on a neutronics-heat coupled analysis, the tritium breeding ratio was evaluated so as to satisfy constraints of the operating temperature of ≤900°C for Li ceramics and Be12Ti, and ≤600°C for Be. Cooling water condition was assumed to be 23MPa and 290–360°C. The result indicates that surplus tritium production in lower neutron wall load (P n ) blanket compensates a shortfall in higher P n blanket and thus the overall tritium production can marginally satisfy fuel self-sufficiency. [ABSTRACT FROM AUTHOR]

Abstract: Fluid flow and heating tests on the full-scale Test Blanket Module (TBM) First Wall (FW) mock-up under the TBM-relevant operating condition were carried out in an ion beam irradiation system with a pressurized high temperature water loop in JAEA to verify the fabrication process and design of the TBM First Wall and demonstrate its heat removal capability. The TBM FW mock-up is made of reduced activation ferritic/martensitic (RAFM) steel, F82H, and its parts such cooling channels and plates were assembled with Hot Isostatic Pressing (HIP) method. In the fluid flow test, the cooling water at room temperature is supplied to 15 parallel flow paths in the mock-up. The flow distribution in each path inside the First Wall mock-up is compared with a numerical simulation. This result shows that no severe cross-sectional deformation of the entire flow path in the mock-up takes place during the fabrication process. In the heating test, the mock-up was exposed to repetitive hydrogen ion beam irradiation with the maximum heat flux of 0.5MW/m2. No indication of joint defect of the HIP joint like a hot spot was observed during 80 irradiation cycles. [ABSTRACT FROM AUTHOR]

Abstract: Our idea on the DEMO plant is that it must demonstrate (1) an electric power generation of one GW level, (2) self-sufficiency of tritium fuel (TBR is more than 1.05), (3) year-long continuous operation, etc. At the same time, DEMO is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. The design guidelines for the blanket are defined in order to meet the mission of the DEMO plant as mentioned above. Major design conditions are surface heat flux of 0.5MW/m2 with peaking factor of 2, a neutron wall load of 3.5MW/m2 with peaking factor of 1.5 and a neutron fluence of about 10MW/m2. To moderate the technological extrapolation, reduced activation ferritic steel (F82H) structural material, Li2TiO3 and Be neutron multiplier are considered. To improve the economical aspect, supercritical water with inlet/outlet temperatures of 280/510°C is chosen as coolant material, with coolant pressure of 25MPa. As a result, a thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirements (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components. [Copyright &y& Elsevier]