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Conceptual Design Report on JT-60SA ___1. JT-60SA

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Conceptual Design Report on JT-60SA ___1. JT-60SA Conceptual Design Report on JT-60SA ＿＿＿＿＿＿＿＿ 1. JT-60SA Mission and Program 1.1 Introduction Realization of fusion energy requires long-term research and development. A schematic of fusion energy development is shown in Fig. 1.1-1. Fusion energy development is divided into 3 phases before commercialization. The large Tokamak phase achieved equivalent break-even plasmas in JET and JT-60 and significant DT Power productions in TFTR and JET. A programmatic objective of the ITER phase is demonstration of scientific and technical feasibility of fusion energy. A primary objective of the DEMO phase is to demonstrate power (electricity) production in a manner leading to commercialization of fusion energy. The fast track approach to fusion energy is to shorten its development period for fusion energy utilization by adding appropriate programs (BA program) in parallel with ITER. Program elements are advanced tokamak/simulation studies and fusion technology/material development. Fig. 1.1-1 Schematic of fusion energy development To specify program elements needed in parallel with ITER, we have to identify the concept of DEMO. Typical DEMO concepts of Japan and EU are shown in Fig.1.1-2. Although size spans widely, operation mode is unanimously “steady-state”. Ranges of the normalized beta are pretty close each other, βN=3.5 to 4.3 for JA DEMO and 3.4 to 4.5 for EU DEMO. Fig. 1.1-2 Cross section and parameters of JA-EU DEMO studies ave 2 The neutron wall load of DEMO exceeds that of ITER (Pn =0.57MW/m ) by a factor of 3-6. The neutron fluence of DEMO (order of 10MWa/m2) is larger than that of ITER (0.3MWa/m2). 1. JT-60SA Mission and Program Sec.1 Page 1 Conceptual Design Report on JT-60SA ＿＿＿＿＿＿＿＿ Neutron damage of first wall is important for DEMO. There are good candidate materials for such high flux and fluence such as F82H in Japan and EUROFER in EU. These materials showed good characteristics for low energy fission neutron irradiation. It is necessary to confirm its properties in 14MeV neutron environment while the use of the ferritic steel in front of high performance plasmas should be checked in a reactor relevant plasma environment. The blanket development is also important for DEMO. It is quite important to decide whether the blanket structure can have a stabilizing effect on RWM since minimization of the plasma-blanket coupling is required for structural soundness against disruption force even in DEMO. If the stabilizing shell should be located far from the plasma, stabilization of higher n RWM may not be practical. The divertor PFC development is also an important element for DEMO to handle high heat and particle fluxes incident to its surface. It is believed that carbon-based materials may not be usable in the DEMO environment. If so, a metal-based divertor must be developed and tested in the reactor relevant tokamak configuration. Otherwise, a usable condition of carbon-based PFC must be identified and tested in the reactor relevant tokamak configuration. Steady state and high beta are important features of DEMO concepts. However, the achievement of steady state operation with Q > 5 in ITER requires further physics R&D in other tokamaks. Sustainment of high beta above the no-wall limit has not been demonstrated in large tokamaks and not foreseen in the early phase of the ITER operation. Confirmation and expansion of the steady state operation, and exploration of the high beta operation for DEMO are important missions of the satellite tokamak. As for configuration optimization of DEMO, the advanced tokamak configuration not foreseen in ITER could be explored for concept development of DEMO in the satellite tokamak. ITER is an integrated test bed for DEMO relevant fusion plasma science and technology. But the application of various ideas to ITER is difficult. Especially, any risky idea not proven in a break-even class tokamak would be difficult. In the Large Tokamak phase, there were three devices and various ideas were tested and compared each other. The role of Satellite Tokamak in providing confidence of any idea for application to ITER is quite important. Also ITER would require physics investigation in well-diagnosed devices if ITER encounters any difficulty to fulfill its missions. Such research is called ITER support program. The Satellite Tokamak Program is a joint Program of Japan and EU under the BA Program using JT-60 facility modification shared with the JAEA’s program for national use. The scientific missions of satellite tokamak are described in Section 1.2. 1.2 Objectives and Mission of Satellite Tokamak Program The objectives of the Satellite Tokamak Program are, a) the participation in the upgrade of the JT-60 owned by JAEA to a superconducting tokamak JT-60SA; and 1. JT-60SA Mission and Program Sec.1 Page 2 Conceptual Design Report on JT-60SA ＿＿＿＿＿＿＿＿ b) the participation in its exploitation to support the exploitation of ITER and the research towards DEMO by addressing key physics issues for ITER and DEMO. The main missions in support of ITER are to optimize operation scenarios for ITER, to optimize ITER auxiliary systems, which come later in the construction of ITER, to train, in an international environment, scientists, engineers and technicians in view of integrated operation and scientific exploitation of ITER. The main mission during ITER operation are, to support further development of operating scenarios and the understanding of physics issues, and to test possible modifications before their implementation on ITER. The main missions in support of DEMO are to explore operational regimes and issues complementary to those being addressed in ITER. In particular these will include: • steady state operation • advanced plasma regimes (higher normalized β) • control of power fluxes to walls. 1. JT-60SA Mission and Program Sec.1 Page 3 Conceptual Design Report on JT-60SA ＿＿＿＿＿＿＿＿ 1.3. JT-60SA Facility 1.3.1. Outline of JT-60SA Facility JT-60SA is designed consistently with maximum utilization of existing JT-60 facilities such as plasma heating and current drive systems, power supplies, diagnostics and cooling system. Figure 1.3.1-1 shows a bird's eye view of the JT-60SA tokamak with heating and current drive systems. A cross sectional view of the tokamak is shown in Fig. 1.3.1-2. Plasma with aspect ratio down to 2.6 can be formed to determine the optimum plasma shape for a cost effective DEMO reactor [1.3-1]. Major components of tokamak will be installed inside the spherical cryostat with a diameter of about 14 m for thermal shielding of superconducting magnets. The maximum plasma current is 5.5 MA for a full-bore (~127 m3) plasma with double-null or single-null divertor and 3.5 MA for an ITER-like shaped plasma with single null divertor as summarized in Table 1.3.1-1. The divertor geometry was optimized to produce both high triangularity double-null/single-null plasmas and ITER-like shaped single-null plasmas in one machine. A semi-closed vertical divertor with a flat dome was adopted to maintain high flexibility for plasma shaping. The divertor geometry will reflect further investigation. Heating and current drive (H&CD) systems are upgraded from those reported in Refs. [1.3-2, 1.3-3] to enable power densities in excess of ITER and to allow independent control of heating, current and rotation profiles, which are essential for expanding advanced tokamak regime. The high power plasma heating results in a remarkable increase of DD neutron yield to about 2×1019/shot. Significant efforts were made in the design of radiation shield for TF coils, vacuum vessel and cryostat structure. Borated acid water is introduced into the double walled space of vacuum vessel to reduce nuclear heating in TF coils. The space between the double walls of the cryostat was filled with boron doped concrete for bio-shielding of the torus hall. However, remote handling is considered to be absolutely necessary for maintenance and repair of in-vessel Fig. 1.3.1-1. Bird's eye view of JT-60SA with heating and current drive systems. 1. JT-60SA Mission and Program Sec.1 Page 4 Conceptual Design Report on JT-60SA ＿＿＿＿＿＿＿＿ devices such as divertor modules and first wall, because the expected dose rate at the vacuum vessel may exceed 1 mSv/hr after 10 years operation and three months of cooling. The increased heating power has also affected the design of divertor targets and cooling system significantly. Figure 1.3.1-3 shows the layout of the JT-60SA torus hall showing major components. Four large ports are devoted to remote handling devices. In addition, four beam towers of perpendicular positive ion based NBI (P-NBI), two beam tanks of tangential P-NBI, and a negative ion based NBI (N-NBI) located in the assembling hall will be reused. The ECH waveguides will be introduced from a neighboring room as illustrated. Current lead box, valve boxes Fig. 1.3.1-2. Cross sectional view of the and other major equipments are located JT-60SA tokamak. TF and PF magnets, near the cryostat. Plasma operation is vacuum vessel, upper and lower divertors, planned mainly during the day time, stabilizing plates, in-vessel position control because the regeneration of divertor coils and sector coils for RWM control are cryo-panel is required every night in order shown. to cope with high plasma density operation. Each component of the JT-60SA facility is described briefly in the following subsections. ECH(140GHz) ECH(110GHz) Notice: P-NBI P-NBI RH: Remote Handling VB: Valve Box CL: Current Leads P-NBI Ctr-NBI N-NBI Power Supply RH VB N-NBI Co-NBI P-NBI CL Assembling Hall Torus Hall 40m Fig.
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