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Progress of the JT-60SA Project

Y. Kamada1, P. Barabaschi2, S. Ishida3, the JT-60SA Team and JT-60SA Research Plan Contributors 1JT-60SA JA Home Team, Japan Atomic Energy Agency and 3 JT-60SA Project Team, Naka, Ibaraki, Japan, 2 JT-60SA EU Home Team, Fusion for Energy, Garching, Germany E-mail contact of main author: kamada.yutaka@jaea.go.jp Abstract. The JT-60SA project by Japan and the EU is progressing on schedule towards the first plasma in Mar. 2019. After careful R&D, procurements of the major components have entered their manufacturing stages. In parallel, disassembly of JT-60U is also progressing on time, and the JT-60SA tokamak assembly is expected to start in Dec.2012. The JT-60SA device, a highly shaped large super conducting tokamak with variety of plasma control actuators, has been designed in order to contribute to ITER and to complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. Detailed assessment and prediction studies of the JT-60SA plasma regimes have confirmed these capabilities: In ITER- and DEMO-relevant plasma regimes and heating conditions, JT-60SA enables an integrated studies on MHD stability at high beta, heat/particle/momentum transport, high energy ion physics, pedestal physics including ELM control, and divertor physics. By integrating these studies, the project provides ‘simultaneous & steady-state sustainment of the key performance characteristics required for DEMO’ with integrated control scenario development.

1. Introduction

Construction of JT-60SA (Fig.1) is in progress toward its first plasma in Mar. 2019 (Fig.2) as a joint project between the Broader Approach (BA) Satellite Tokamak program by Europe and Japan, and the Japanese national program. The project mission is to contribute to early realization of fusion energy by supporting ITER and by complementing ITER in resolving key physics and engineering issues for DEMO reactors [1]. The JT-60SA device is a highly shaped large super conducting tokamak (the major radius Rp~3 m, the aspect ratio A~2.6, the maximum plasma current Ip=5.5 MA) with variety of plasma control actuators including high power (41 MWx100 s) heating by NB and ECRF. This device is capable of confining break-even-equivalent class deuterium plasmas lasting for a duration (typically 100 s) longer than the time scales characterizing the key plasma processes, such as current diffusion and particle recycling. JT-60SA also pursues fully non-inductive steady-state operations with high plasma pressure exceeding the no-wall ideal MHD stability limits. The JT-60SA experiments explore ITER and DEMO-relevant plasma regimes in terms of non-dimensional plasma parameters. JT-60SA has been designed to complement ITER in all areas of fusion plasma development necessary to decide DEMO construction. In particular, the most important goal of JT-60SA is to decide the practically acceptable DEMO plasma design including practical and reliable plasma control schemes [2].

FIG.1. JT-60SA Tokamak

FIG.2. Schedule of the JT-60SA project

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2. Progress of Procurement and R&D

By Sep. 2012, 18 Procurement Arrangements (PAs) have been launched (JA: 10PAs, EU: 8PAs) covering 75% of the total cost of the BA Satellite Tokamak Program, and the main components have entered the manufacturing stage. Disassembly of the JT-60U tokamak is also progressing on schedule, and the JT-60SA tokamak assembly starts in Dec. 2012.

2.1. Poloidal Field Coils (JAEA)

The poloidal field (PF) coil system consists of the central solenoid (CS) (4 modules, Nb3Sn, 20kA, maximum field 8.9T, operation temperature 5.1K, flux 40 Wb), and 6 equilibrium field (EF) coils (NbTi , 20kA, EF3&4: 6.2T, 5.0K, EF1,2,5&6: 4.8K, 4.8T) [3] (see Fig.12 for location of each EF coil). Manufacture of conductor is in its mass production phase, and 55% of the EF conductor and 25% of the CS conductor have been manufactured at Naka site by Sep. 2012. All types of conductors satisfied their design requirements. The critical current (Ic) of the CS conductor and the resistance of CS and EF coil joints were investigated under collaboration of JAEA and NIFS [3, 4]. Degradation of Ic by cyclic excitation was evaluated up to 4,000 cycles (with warm-up and cool-down after the 2,000 cycle), and no degradation was observed (Fig. 3). It was found that 180 kN/m in electro-magnetic force, which is 1/3 of ITER, does not degraded the Nb3Sn conductor. The all coil joints showed the resistance below the requirement of 5 nΩ. Fabrication of the first EF coil winding pack (EF4) was completed in Mar. 2012 (Fig.4). The manufacturing error (in-plane ellipticity) of the current center is 0.6 mm which is sufficiently small, 10 %, compared with the requirement (6.0 mm) [5]. The EF4 winding pack is made by ten double pancake (DP) coils. For each DP coil, careful measurement of winding error was conducted during the winding process. Fig. 5 shows the measured current centre of each DP coil and the averaged current centre as a winding pack. The maximum error for one DP coil is 2.5 mm. By optimizing the positions of these ten DP coils, the error as a winding pack became 0.6 mm. Fabrication of larger-sized EF5 and EF6 coils are now underway at Naka. As for CS, winding and heat treatment of a model coil was finished.

2.2. Toroidal Field Coils (F4E, ENEA, CEA, SCK CEN)

The toroidal field (TF) coil system [6] consists of D-shaped 18 (+ 1 spare) coils (6 double pancake, NbTi, 25.7KA, 5.8T, total magnetic energy 1.5GJ). The coils are supported by wedge & key at the inboard, and by the outer inter coil structure at the outboard.

FIG.3. Results of repetition excitation test of

CS conductor [4].

FIG.4. EF4 winding pack (JAEA)

FIG.5. Current center radius for each pancake

and average as a winding pack for EF4 coil

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Fabrication of both superconducting NbTi and copper strand [7] (F4E) is progressing on schedule, reaching about 60% of the total production. The first two full size conductor lengths were pressure- and flow- tested, and delivered in time for calibration/commissioning of the winding machines. The conductor for the current feeders was manufactured and tested. The first conductor for TF coils was also manufactured. The two qualification samples for TF conductor and joint were successfully tested in SULTAN facility. The joint resistance was found to be 0.84 nΩ, which is below the target value of 2 nΩ. As for the TF coil winding (ENEA, CEA), contracts were placed for the manufacture of all 18 coils and the associated structural components. Many qualification activities were carried out, including the manufacture and testing of the internal joints, the helium inlets and impregnation samples of the complete winding pack. The manufacture of the first winding line, which will be used for the manufacture of 9 coils, was completed, and entered the commissioning phase (Fig.6). The second winding line for the remaining 9 coils will be completed by the end of 2012. The first complete winding packs will be available in early 2013 for installation into the casing structures. All superconducting TF coils will be tested at operational conditions at a dedicated test facility at CEA in Saclay. Each coil will undergo high-voltage tests, flow tests, leak tests and current tests at the operation current of 25.7 kA. The tests at helium temperatures ask for a dedicated, large cryostat to thermally insulate the coils. This cryostat has dimensions of 11mx7mx5m and is equipped with a LN2 cooled nitrogen shield. The cryostat (SCK CEN) (Fig.7) has been delivered to the test site.

2.3. High Temperature Superconducting (HTS) Current Leads (KIT)

The superconducting CS, EF and TF coils require special current connections between the current supplies and the cold coil feeders. In order to minimize the heat input to cryogenic temperatures the current leads incorporate high-temperature superconducting inserts. The design of HTS current leads for JT-60SA is based on the design which is currently integrated in W7-X. Material purchasing is progressing as scheduled, preparation of test facility at KIT is progressing and the detailed design of superconducting cable jumpers is ongoing.

2.4. Vacuum Vessel (JAEA)

The JT-60SA vacuum vessel (VV) has a double wall structure in high rigidity withstanding a large electromagnetic force at disruption and to keep a high toroidal one-turn resistance. In addition, the double wall structure fulfills i) reduction of the nuclear heating in the superconducting magnets by boric-acid water circulated in the double wall, and ii) effective baking by 200°C nitrogen gas flowing between the double wall. The outer diameter and

FIG.6. TF coil winding line (left) and TF coil winding pack vacuum testing chamber (right) (ENEA)

FIG.7. TF coil test-facility cryostat during leak tests in Belgium (SCK CEN)

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height of VV are 9.95 m and 6.63 m, respectively. The Inner/outer shell and stiffening ribs of the double wall have 18 mm and 22 mm thicknesses, respectively The manufacturing procedure was established through the welding R&D and a trial manufacturing of the 20-degree upper-half mock-up. By March 2012, manufacture of the first three 40-degree-sectors was completed (Fig.8). The key dimension was controlled within the tolerance of ±5 mm [8].

2.5. Cryostat (CIEMAT)

The Cryostat Base, the base of the tokamak with 13.4m in diameter and 250 tonnes, is expected to be delivered to Naka site around the end of 2012. Machining of the three large “lower structure” components and welding of the upper “double ring” components were completed in Augusts 2012. As for the Cryostat Vessel Body Cylindrical Section, drawings and specification were completed, and all the material supplied by JAEA, which is around 300 tonnes of stainless steel plates, were delivered in June 2012.

2.6. Cryogenic Systems (CEA)

The cryogenic system for JT-60SA has a refrigeration capacity equivalent of about 7000 W if measured at 4.5 K. The refrigeration loads at 80 K and the helium supply for the current leads of the superconducting coils are rather constant. The heat loads on the coils (about 2250 W during idle operation of the coils) increases to about 7350 W during a plasma pulse. This large variation asks for excess refrigeration capacity stored in a buffer. Between plasma pulses the buffer reservoir is refilled. Design and specification have been completed in Sep. 2012, and the contract of manufacture is expected within 2012.

2.7. Power Supply (CNR-RFX, ENEA, CEA)

Specification of the whole JT-60SA power supply system has been completed including modification of the existing power supply system by JAEA. The Quench Protection Circuit (QPC) (CNR-RFX) was designed to provide a fast discharge of the energy stored in the superconducting coils in case of quench of the superconducting coils. This function is obtained by opening a dc circuit breaker and by diverting the coil current into a discharge resistor. For a backup protection, a pyrobreaker is

FIG.8.Vacuum Vessel design and the first three 40-degree sectors (JAEA)

FIG.9. Drawing of the Cryostat Base, and the first Lower Structure sector in final machining. (CIEMAT)

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added in series to the dc circuit breaker. The three QPCs for the TF coil are interrupt the nominal current of 25.7kA, and assuring a maximum I2t in the coil of 4.6GA2s from the QPC activation time. The ten QPCs for the PF coil interrupt a maximum current of ±22.5 kA, assuring a maximum I2t of 2GA2s. The design of the dc circuit breaker, based on a mechanical By-Pass Switch (BPS) connected in parallel to a static CB based on Integrated Gate Commutated Thyristor technology, was successfully tested with a full scale prototype (Fig. 10), proving the validity of the solution that allows combining the low power dissipation of the mechanical BPS with the fast arc-less current interruption of the static CB. The ongoing type tests of the complete QPC prototypes will be completed in Sep. 2012. The Switching Network Units (SNU) (ENEA) is required in order to provide a fast current variation in the CS for creating the electric field needed for plasma breakdown. The four SNU for CS are designed to insert a discharge resistor in the coil circuit and then to by-pass it after plasma initiation. They are designed to provide up to 5kV at the nominal current of 20kA. During plasma initiation, it is possible to modify the discharge resistor value. The call for tender of the contract was issued in July 2012. As for the Superconducting Magnet Power Supply (SCMPS) (CEA & ENEA), completion of the contracts is expected by March 2013. The TF coil converter is rated for a nominal current of 25.7 kA and a nominal voltage of 80V. The 10 base converters for the PF coil are rated for a nominal current of ±20kA and a nominal voltage of ±1kV. In order to be able to reverse the coil current without any discontinuity across zero, circulating current operation is foreseen for the base converters for the PF coils when their current is close to zero. The RWM control coil Power Supply (RWM PS) (CNR-RFX) should have a very fast dynamic performance (maximum delay time between reference and output 50us) in order to control the Resistive Wall Mode (RWM) by the 18 control coils. In order to assure the maximum flexibility in RWM control, current in each coil will be supplied by an independent inverter. The technical specification is in the final stage.

2.8. Divertor (JAEA)

JT-60SA allows studies of power and particle handling at the full injection power of 41 MWx100 s using the water-cooled divertors compatible with the maximum heat flux of 15 MW/m2. The W-shaped configuration and the vertical target with V-corner at the outer target (Fig.11) enhance radiation from the divertor area. Material of the plasma facing components (divertor and first wall) is initially Carbon. Metallic divertor targets and first wall together with an advanced shape divertor will be installed in a later phase of the JT-60SA project in order to demonstrate the high integrated performance with metallic wall. Cryo-panels are installed behind the divertor at nine toroidal sections, and the divertor pumping speed can be changed by 10 steps between 0 - 100 m3/s. The divertor cassettes with fully water-cooled plasma facing components were designed to have compatibility with remote maintenance, since they are equipped inside the highly radio-activated vacuum vessel. Fabrication of the divertor cassettes was started after the trial manufacture which ensured the manufacturing procedure with the required accuracy of ±1

FIG.10. Toroidal Field Coil QPC full scale prototype during testing (CNR-RFX)

FIG.11. The JT-60SA divertor structure (left) and preassembly of three divertor cassettes (JAEA)

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mm [8]. Figure 11 shows the preassembly of the first three divertor cassettes including heat sinks for inner/outer targets, private dome and inner/outer baffles. Manufacture of the whole 36 cassettes will be completed in March 2013. In addition, manufacture of brazed CFC monoblock targets for the vertical divertor target has been started with precise control of tolerances inside CFC blocks. A method of the infrared thermography test has been developed as the simplified inspection for mass production of the monoblock targets [8].

2.9. In-vessel Coils (JAEA)

JT-60SA allows exploitations of MHD stability control studies in particular at high beta with the stabilizing plates, the first plasma position control coils (FPCC), the error field correction coils (EFCC), and the resistive wall mode control coils (RWMC) (Fig.12). The stabilizing plates are installed in order to improve the plasma positional stability and the ideal beta limit. A pair of FPCC with cupper conductor is installed for fast plasma position control. Each coil has 23 turns and the maximum current is 120kAT. Independent power supplies will be connected to each coil for simultaneous control of vertical and horizontal fields. The EFCCs (3-poloidal x 6-toroidal, 30kAT each) are primarily used for minimization of error fields due to manufacturing and assembly errors of the TF, PF, and NB correction coils. The error field should be reduced in order to keep a sufficiently large field-null area at plasma initiation and to avoid locked modes during Ip ramp-up. The EFCCs are also utilized for resonant magnetic perturbation (RMP: Bresonant~9 G ~4x10-4BT) for ELM control. The RWMCs (3-poloidal x 6-toroidal, 2.2kAT each) are installed on the plasma side of the stabilizing plate in order to minimize the shielding effect of the stabilizing plate on high frequency magnetic field modulation needed for RWM control. Each coil will be connected to an independent amplifier and the “mode control” feed back scheme will be applied in order to stabilize n=1, 2 and 3 RWMs simultaneously. The detailed study of the feedback scheme is underway by JAEA-RFX collaboration.

2.10. NBI System (JAEA)

The JT-60SA neutral beam (NB) system having the full injection power of 34 MW x 100 s allows a variety of heating/ current-drive/ momentum-input combinations (Fig.13). The negative ion source based neutral beam (N-NB) provides 10 MW/500 keV co-tangential injection. The positive ion source based neutral beams (P-NBs) at 85 keV consist of 2 units of co-tangential beams (4 MW), 2 units of counter-tangential beams (4 MW), and 8 units of near perpendicular beams (16 MW). The N-NB system consists of two beams (5MW each) with different injection trajectory; one is relatively on-axis and the other is pretty off-axis in order to sustain/control the weak / negative magnetic shear plasmas. In order to realize a stable acceleration of the high current D- ion beams of 500 keV for 100 s without breakdowns, the vacuum voltage holding capability (VHC) of the large-area multi-aperture grid was measured in the JT-60 negative ion source. In this study, the gap

FIG.12. The stabilizing shells and the in vessel coils

FIG.13. The heating system for JT-60SA and NB injection trajectories

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length (G), the grid area (S) and number of the apertures (N) were widely varied [9]. As shown in Fig.14, the sustainable voltage normalized to G0.5 scales with S-0.13 and N-

0.15. These results indicate that i) the VHC in the grid of the JT-60SA, and also ITER, negative ion source is 50-60% of that in small electrodes, and ii) the VHC seems to depend on the electric field integrated over the grid. Based on these results, the gap length between the acceleration grids of 70mm was tested for a single gap. As expected, 200 kV was successfully sustained for 100 s without breakdown. This result suggests that stable long-pulse acceleration at 500 keV is reasonably expected with a 20% of margin (=600keV) for the three acceleration stages in the JT-60SA negative ion source. In parallel, we have started upgrade of the acceleration power supplies for N-NBs.

2.11. ECRF System (JAEA)

The ECRF system for JT-60SA is composed of 9 gyrotrons (dual frequency at 110GHz and 138GHz) enabling 7MW injection for 100 s, transmission lines with the total length of 80 m and linear-motion launchers enabling ≥1 MW x 100 s injection with a wide coverage in both toroidal (typically −15◦ to +15◦ with respect to the radial direction) and poloidal (40◦ downward to 20◦ upward) directions. Toward this goal, development of the ECRF system has significantly progressed [10]. Gyrotron operations using a newly installed 60.3 mm diameter transmission line enabled 1 MW output for 70 s and 1.4 MW output for 9 s (Fig.15). These sustainment times are longer than the previous records by a factor of 2. Design and fabrication of a new dual-frequency gyrotron was competed, and 1 MW output for 0.1 s was achieved at 110 GHz. Oscillations at 138 GHz was also confirmed in short-pulses. Mock-up test of a novel linear-motion antenna showed preferable optical and mechanical characteristics.

2.12. Disassembly of JT-60U Torus

Disassembly of the JT-60U torus started in 2010 is progressing on schedule toward its completion in Nov. 2012 (Fig.16). Since this work is the first experience of disassembling a radio activated large fusion device in Japan, very careful treatment of the activated materials and safety work are being recorded as an important knowledge base for fusion research. Various effective tools have been utilized such as the ‘diamond wire-saw’ for cuttng large and hard activated components (such as the vacuum vessel

FIG.15. Progress of the output power and duration of the 110GHz Gyrotron for JT-60SA.

FIG.16. Progress of JT-60U torus disassembly

FIG.14. (a) Normalized voltage as a function of the grid area, (b) Normalized voltage as a function of the number of the apertures

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and the poloidal field coils). So far more than 10,000 components have been removed from the torus hall and stored safely.

3. JT-60SA Plasma Regimes

The JT-60SA device is a highly shaped large super conducting tokamak designed for contributing to ITER and complementing ITER in all areas of fusion plasma development necessary to decide construction of DEMO reactors [2]. Towards this goal, the JT-60SA Research Plan has been discussed widely both in JA and EU fusion research communities, and the JT-60SA Research Plan ver. 3.0 [11] was completed in Dec. 2011 with 332 co-authors (JA 145 + EU 182 + Project Team 5, from 38 institutes). This chapter introduces the JT-60SA plasma regimes based upon the Research Plan.

3.1. Plasma Parameters of Representative Scenarios

The JT-60SA device has been designed to realize a wide range of plasma equilibria covering a DEMO- equivalent high plasma shape parameter S (= q95Ip/(aBt)) ~7 and a low aspect ratio of A~2.5. The typical plasma parameters of the representative scenarios of JT-60SA are shown in Table 1. The maximum plasma current is 5.5 MA with the highly shaped configuration (Scenario 2: lower single null, A=2.5, κx=1.87, δx=0.50). The achievable plasma current at q95=3 for an ITER-shaped configuration (κx=1.81, δx=0.43) is 4.6MA (Scenario 4-1). In Fig.17, the plasma equilibria for these two scenarios are shown. In DEMO reactors, we have to sustain high values of the energy confinement improvement factor (the HH-factor), the normalized beta βN, the bootstrap current fraction fBS, the non-inductively driven current fraction, the plasma density normalized to the Greenwald density nGW, the fuel purity, and the radiation power normalized to the heating power simultaneously in the steady-state. However, such a high ‘integrated performance’ has never been achieved so far. The most important goal of the JT-60SA research for DEMO is to demonstrate and sustain this integrated performance. Assuming HH=1.3-1.4, the expected Ip for high βN (=4.3) and high fBS (=70-80%) with full non-inductive current drive is 2.1-2.3MA at the Greenwald density ratio of ~1 (Scenarios 5-1, 5-2). In Scenario 5-2 with 31MW heating, controllability of the high βN and high fBS plasmas can be studied by utilizing the remaining power of the N-NB and P-NB. This plasma regime satisfies the highly integrated performance as shown in Fig.18,

TABLE. 1. Plasma parameters of representative scenarios

FIG.17. The full bore single null shape (Scenario 2) and the ITER-like shape (Scenario 4-1)

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which compares JT-60SA, DEMO Slim CS [12], ITER steady-state, and a simultaneous achievement in JT-60U (q95=5.5 – 6 in all cases). The central reference of the DEMO reactor for JT-60SA is the compact steady-state reactor [12]. However, it should be emphasized that the most important goal of JT-60SA is to decide the practically acceptable DEMO plasma design including practical and reliable plasma control schemes. Therefore the JT-60SA research project has to treat the ‘DEMO regime’ as a spectrum spreading around the reference design including the advanced inductive operations (scenario 4-2), and has to assess reliable DEMO design targets as shown in Fig.19.

3.2. Non-dimensional parameters, plasma heating conditions, and current diffusion

JT-60SA enables explorations in ITER- and DEMO-relevant plasma regimes in terms of the non-dimensional parameters. Figure 20 shows (a) the normalized poloidal gyro radius ρ* and the normalized collisionality ν*, (b) the normalized collisionality at the top of the pedestal, (c) the normalized beta βN and the shape parameter, and (d) the first ion β and the fast ion velocity normalized to the Alfven velocity. In JT-60SA, confinement and transport (heat, particle, momentum) characteristics can be clarified at the ITER- and DEMO- relevant vales of ρ* and ν* (Fig.20(a)). The pedestal structure, inter-ELM transport, and the ELM physics including ELM mitigation by RMP and pellet injection can be studied at the ITER-relevant low pedestal collisionality (Fig.20(b)). The MHD stability physics and control at high βN are studied at the DEMO-relevant shape parameters (Fig.20(c)). The high energy N-NB is a powerful tool for study on the fast ion physics, for example the Alfven Eigenmodes (AEs) at the ITER- and DEMO-equivalent values of the fast ion β (Fig.20(d)).

FIG.18. Integrated plasma performance of JT-60SA [2].

0

1

2

3

4

5

2 3 4 5 6 7q95

#1

DEMO(Slim CS)

DEMO(CREST)

ITER Q=10

ITER Steady-state

JT-60SA

#4-1

#2

#3

#4-2

#5-2#5-1(a)

0

0.2

0.4

0.6

0.8

1

2 3 4 5 6 7

#2#3

#4-1

#4-2

#5-1

#5-2

ITER Q=10

ITER Steady-state

#1

DEMO(CREST)

DEMO(Slim CS)

q95

(b) JT-60SA regime for DEMO - R&D

FIG.19. The JT-60SA research regime for assessment of DEMO designs.

FIG.20. (a) the normalized poloidal gyro radius ρ* and the normalized collisionality ν*, (b) the normalized collisionalityat the top of the pedestal, (c) the normalized beta βN, and the shape parameter, and (d) the first ion β and the fast ion velocity normalized to the Alfven velocity.

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These studies have to be conducted with ITER- and DEMO- relevant heating conditions; such as dominant electron heating and low central fueling enabled by N-NB and ECH, and low external torque input enabled by N-NB, ECH, perpendicular P-NBs and balanced injection of co- and counter- tangential P-NBs. Effects of the electron heating fraction and the plasma rotation can be also clarified by changing combinations of these heating systems. As for the study on electron heating, for example, the electron heating power fraction can be scanned from 20% to 60% at a fixed total heating power of ~20 MW by choosing combination of the N-NB power and the ECH power (Fig.21). In order to complete the studies mentioned above, the plasma flat-top length, τflat-top, has to be sufficiently longer than the resistive diffusion time of the plasma current τR. For this purpose, τflat-top/τR>2-3 is required. Figure 22(a) shows Ip and τflat-top/τR for the representative operation scenarios (τR=11-34 s, τflat-top=100 s). All the scenarios satisfy τflat-top/τR>~3. We have carried out a systematic modelling study for time evolutions of the safety factor profile, q(r), in these representative scenarios by METIS code [13]. One example is shown in Fig.22(b) for the advanced inductive scenario (Scenario 4-2), where q(r) reaches its steady-state profile before t=50s with q(0)>1 which is needed for the ‘advanced inductive’ operation. It should be emphasized that plasmas in DEMO are highly self-regulated. In order to improve the plasma research toward these plasmas, mutual interactions among core transport, pedestal transport & stability, core MHD stability, and fast ion behaviours have to be clarified and controlled under the DEMO relevant heating conditions with the steady-state safety profile. Figures 20 – 22 indicate that such a study is realized in JT-60SA.

3.3 Plasma Controllability

In order to accomplish the studies treated above, plasma profile controllability is essentially important. The JT-60SA NB system allows various combinations of heating/ current-drive/ momentum-input (Sec.2.10). Figure 23 shows (a) profiles of the NB driven current for three cases of N-NB injection (upper unit 5MW, lower unit 5MW, upper 2.5MW + lower 2.5MW), and (b)

FIG.21. Fraction of the electron heating power to the total heating power

FIG.22. (a) Ip and τflat-top/τR for the representative JT-60SA operation scenarios. (b) Evolution of the safety factor profile for Scenario 4-2.

FIG.23. (a) Profiles of the NB driven current for three cases of N-NB injection. (b) Profiles of the torque density normalized to the absorbed power for co-P-NB, counter-P-NB, perpendicular-P-NB, and N-NB.

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profiles of the torque density normalized to the absorbed power for co-P-NB, counter P-NB, perpendicular P-NB, and N-NB. Using this profile controllability, we have conducted predictive integrated modelling studies using the 1.5D transport code TOPICS (with the CDBM transport model) coupled with F3D-OFMC [14]. Figure 24 shows examples for the

high-βN discharge scenario at Ip=2.3MA with an internal transport barrier (based on Scenario 5-1): By changing the lower N-NB to the upper N-NB, the q profile is changed, and the minimum q value is increased above 1.5. By changing combinations of tagnential P-NBs (co, counter, and balanced) with N-NB, the tproidal rotation profile is changed significantly. Capability of the ECRF system was evaluated both for 110 GHz and 138 GHz [10]. Figure 25 shows the EC driven current density profile per 1MW injection with the toroidal injection angle of 15◦ for 110GHz (Scenario 5-1, Bt=1.7T) and for 138GHz (Scenario 2, Bt=2.25T). In the 110GHz case, the range of deposition covers the locations of the internal transport barrier and the minimum q-value. The total amount of EC-driven current is typically 15 kA at ρ = 0.5 for 1 MW injection. For 7 MW (full power) injection, the peak EC driven current density becomes comparable to the bootstrap current density at the deposition location. Thus, effective current profile modification is expected. In case of the 138 GHz injection, the deposition range covers the region where NTMs with m/n = 3/2 and 2/1 are expected to appear. Calculations by GRAY/GRE [15] indicated that the suppression of an m/n = 3/2 NTM seems possible by injection of 3 MW (the initially prepared EC power at the first plasma), while the suppression of an m/n = 2/1 NTM is marginal. Simultaneous achievement and control of the radiative divertor with the high core performance is the goal of JT-60SA toward DEMO (Fig.18). In addition, the allowable heat flux onto the divertor target is 15MW/m2 in JT-60SA. A SOL/divertor simulation code suite SONIC has been used to evaluate how to control and

FIG.26. (a) The separatx density against Ar fraction with Carbon concentration of 2, 3 and 4%, with the outer divertor heat flux ~10 MW/m-2. (b) The outer divertor heat load for cases with (nAr/ni =1%, nC/ni =4%) and (nAr/ni =2%, nC/ni =2%).

FIG.24. Controllability of (a) the safety factor profile and (b) the toroidal rotation profile with different combinations of neutral beams.

FIG.25. EC driven current density profile per 1MW injection with the toroidal injection angle of 15◦ for 110GHz (Scenario 5-1, Bt=1.7T) and 138GHz (Scenario 2, Bt=2.25T).

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reduce the divertor heat flux down to the acceptable level. We investigated effect of Argon gas seeding for Scenario 5-1 (high βN plasma of 85% nGW at Ip=2.3MA). In Fig. 26(a), the separatrix density necessary to maintain the peak heat flux onto the outer divertor target <~10MW/m2 is plotted against Argon concentration at the outer divertor for three cases of Carbon concentration of 2, 3, and 4%. In Fig. 26(b), the outer divertor heat load for two cases with (nAr/ni =1%, nC/ni =4%) and (nAr/ni =2%, nC/ni =2%) are plotted. In Scenario 5-1, the separatrix density is expected to be ~1.6x1019m-3. Therefore, the white region in the Fig.26(a) would be the acceptable. As shown in this figure, even if the carbon concentration varies, the heat flux can be managed by controlling Ar gas puffing. In addition to such heavy impurity gas puffing, fuel gas puffing and pellet injection are to be utilized combining with the divertor pumping [2] for study on the integrated plasma control with radiative divertor in JT-60SA.

4 . Summary

The JT-60SA project by Japan and the EU is progressing on schedule towards the first plasma in Mar.2019. After careful R&D, procurements of the major components have entered their manufacture stages. In parallel, disassembly of JT-60U is also progressing on time, and the JT-60SA tokamak assembly is expected to start in Dec.2012. The JT-60SA device has been designed in order to contribute to ITER and to complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. Detailed assessment and prediction studies of the JT-60SA plasma regimes have confirmed these capabilities.

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