Osamu Mitarai

Blanket and divertor design for force free helical reactor (FFHR)

Abstract Conceptual design of blanket and divertor for a force free helical reactor (FFHR) is presented. The demonstration-relevant FFHR is a heliotron-type helical reactor having superconducting helical and poloidal coils based on the large helical device (LHD) which is now under construction in the National Institute for Fusion Science. The main feature of FFHR is force free configuration of helical coils, which allows us to simplify the coil supporting structure and to use high magnetic field instead of high plasma β. For the goal of a self-ignited D—T reactor of 3 GW thermal output, the design parameters for FFHR are investigated under the LHD scaling for energy confinement and density limit. In particular, to satisfy the reactor lifetime of 30 years, the engineering issues in FFHR are discussed by focusing on selection of structrual materials for 500 dpa, optimization of tritium breeding system with neutron multiplier, cooling with molten-salt Flibe and operation temperature in the blanket, radiation shielding to achieve a reduction of more than 5 orders of magnitude at superconducting coils, and steady state helium ash removal with an efficiency of around 30%.

Design studies of helical-type fusion reactor FFHR

Abstract The main feature of FFHR is force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt Flibe as a self-cooling tritium breeder from the main reason of safety. Collaboration works based on the LHD project have made great progress in the reactor studies by focusing on engineering aspects of the high magnetic field and Flibe system design. Encouraging positive results are shown on ignition access, mechanical stress in coils supporting structures, improvement in the blanket system including materials selection and tritium recovery. Critical issues on fundamental safety analysis and maintainability of reactor components are also discussed, and many subjects are pointed out as future works.

Overview of steady state tokamak plasma experiments in TRIAM-1M

An overview of steady state tokamak studies in TRIAM-1M (R0 = 0.8 m, a × b = 0.12 m × 0.18 m and B = 8 T) is presented. The current ramp-up scenario without using centre solenoid coils is reinvestigated with respect to controllability of the current ramp-up rate at the medium density region of (1–2) × 1019 m−3. The plasma is initiated by ECH (fundamental o-mode at 170 GHz with 200 kW) at B = 6.7 T, and the ramp-up rate below the technical limit of 150 kA s−1 for ITER can be achieved by keeping the LH power less than 100 kW during the current ramp-up phase. The physics understanding of the enhanced current drive (ECD) mode around the threshold power level has progressed from a viewpoint of transition probability. A transition frequency, ftrans, for the ECD transition is determined as a function of PCD. At ~70 kW no transition occurs for an ftrans value of ~0.017 Hz, meaning almost zero transition probability. With increasing PCD > Pth, ftrans increases up to 10 Hz, and the transition tends to occur with high probability. The record value of the discharge duration is updated to 3 h 10 min in a low and low power (<10 kW) discharge. The global particle balance in long duration discharges is investigated, and the temporal change in wall pumping rate is determined. Although the density was low, the gas supply had to be stopped at 30 min after the plasma initiation to maintain the density constant. After that the density was sustained by the recycling flux alone until the end of the discharges. In addition to the recycling problem, in the high power and high density experiments, the localized PWI affects the SSO of the tokamak plasma. The effects of enhanced influx of metal impurities (Fe, Cr, Ni, Mo) on sustainment of the high performance ECD plasma are investigated. In order to evaluate the helium bombarding effects on the plasma facing component and hydrogen recycling in the future burning plasma, microscopic damage of metals exposed to long duration helium discharges was studied. The total exposure time was 128 s. From thermal desorption experiments for the specimens the amount of retained helium was evaluated as 3.9 × 1020 He m−2 and the scale length to be ~1 mm in the SOL.

Alternating current tokamak reactor with long pulses

Alternating current (ac) tokamak operation in the reactor parameter range is studied by considering the volt-second consumption. A simple condition for obtaining ac operation with nearly constant pulse length is given by 1/sub rho//R/sub rho/

RF start-up and sustainment experiments on the TST-2@K spherical tokamak

Plasma start-up and sustainment without an inductive field have been studied in the TST-2@K spherical tokamak using high power RF sources (8.2?GHz/up to 170?kW). Steady state discharges with a plasma current of 4?kA were achieved. The line integrated density was about 3 ? 1017?m?2 and the electron temperature was 160?eV. A truncated equilibrium was introduced to reproduce magnetic measurements. It was found that a positive Pfirsch?Schl?ter current in the open field line region at the outboard boundary makes a significant contribution to the current. Insensitivity of the current to variations in the vertical field and RF power variation was also found.

Non-inductive current start-up assisted by energetic electrons in Q-shu University experiment with steady-state spherical tokamak

After intensive discharge cleaning of the chamber wall, non-inductive current start-up experiments have been successfully performed in QUEST in moderate vertical fields of about 1.0–1.5 mT with positive n-index. Simultaneously, with increasing plasma current, an asymmetric toroidal flow of energetic electrons was observed and direct measurements of current driven by this asymmetric flow were taken with a newly developed Langmuir probe technique. A numerical study of the energetic electron orbits indicates that the total current is enough to play a dominant role in the formation of a closed flux surface in QUEST.

Plasma Current Rampup by the Outer Vertical Field Coils in a Spherical Tokamak Reactor

To find a solution of the plasma current rampup problem in a low aspect ratio spherical tokamak (ST) reactor, the effect of the outer vertical field coils on plasma current rampup is studied with noninductively driven current and bootstrap current but without the OH transformer during the fusion power rampup phase. As a lower elongation of [kappa] = 2 does not allow a large poloidal beta, a low density discharge with high heating/current drive power is necessary to increase the noninductive plasma current up to 30 MA, and then the vertical field can ramp the plasma current up to 48 MA just before reaching the steady state fusion burn phase. A higher elongation of [kappa] = 3 can reach a higher βp value, in which case the plasma current of 48 MA can be achieved by the vertical field without powerful heating/current drive. As the level of the vertical field current drive depends on the plasma energy, which is determined by the confinement time, a lower confinement factor can only produce a lower plasma current. The current rampup time can be chosen arbitrarily, shorter or longer, facilitating a flexible ST reactor operation.

An alternating current tokamak reactor with ohmic ignition and bootstrap current

In this paper an alternating current (ac) tokamak reactor with ohmic ignition and long pulses due to bootstrap current is proposed as a simple and quasi-continuous fusion power plant. An ohmic plasma current of 23 MA with a high toroidal field of {approximately} 10 T in the alternating Current Tokamak Reactor-Upgrade (ACTR-U) (10-m major radius and 2-m minor radius) provides the ohmic ignition. After entering the ignition regime, the plasma current is reduced by one-half to enhance the bootstrap current with a high-beta poloidal field ({beta}{sub p} {approximately} 2) to prolong the pulse length. When the ohmic transformer reaches the maximum flux, the plasma current is ramped down and reversed; ac operation follows. The authors thus demonstrate that an ohmic transformer alone is in principle sufficient for a quasi-continuous deuterium-tritium fusion reactor.

Development of completely solenoidless tokamak operation in JT-60U

Plasma current start-up to 100 kA was achieved successfully in the JT-60U tokamak without the use of the centre solenoid (completely solenoidless tokamak operation). Only poloidal field coils located on the outboard side of the torus were used, in combination with strong ionization by electron cyclotron (EC) power. The presence of a field null was not necessary for plasma current start-up, but the flux conversion efficiency was low in such a case. In a nearly solenoidless start-up, low neutral pressures were favoured, and the optimum location of the EC resonance was slightly to the high field side of the vacuum vessel centre. The required EC power for efficient utilization of flux swing in JT-60U was about 1 MW. A plasma current of 260 kA was maintained for 1 s by NB only, and plasma current ramp-up from 215 to 310 kA was achieved by EC and neutral beam (NB) only (without lower hybrid current drive (LHCD)). However, the ramp-up efficiency was much lower compared with LHCD. Recharging of the centre solenoid was observed with only counter and perpendicular NB injection, indicating bootstrap overdrive. Integration of these elements can lead to the achievement of a completely solenoidless tokamak operation.

Plasma density at the current reversal in the STOR-1M tokamak with AC operation

The plasma density behaviour in the STOR-1M tokamak with alternating current (AC) operation is described using the Murakami-Hugill diagram (1/qa, nR/Bt). At the current reversal, Ip = 0 (1/qa = 0), the plasma density remains finite and the Murakami parameter is nR/Bt = (0.66 ± 0.22) × 1018m-2.T-1. Gas puffing before the current reversal does not noticeably increase the plasma density at the current reversal, but allows AC operation with larger currents and improves its reproducibility. A qualitative explanation for the finite plasma density at the current reversal is given on the basis of a short circuit effect by the limiter