JT-60 Plasma Research Articles

Status and prospect of JT-60 plasma control and diagnostic data processing systems for advanced operation scenarios

The large tokamak fusion device JT-60 is expected to explore more advanced tokamak discharge scenarios toward the ITER and a future power reactor. Since various experimental issues are to be adequately discussed, and possibly to be solved in JT-60, the plasma real-time control system has been drastically improved with remodelling in hardware as well as in software. To satisfy the requirements, a “multiple networks” structure is employed as a basic principle of the remodelling. Distributed processors for diagnostics, actuators, and supervisory controllers are linked through a reflective memory (RM) network for fast, real-time communication. Similarly, advanced and complex calculations to reproduce plasma shape and profiles are performed by several processors connected to the same RM network. In this report, we discuss the developments to improve the JT-60 plasma control and data processing systems. In addition, a future plasma control system leading to a standard design for a power reactor is outlined, which is based on the 20-year plasma operation experience.

Neutron Diagnostics for the Energetic Ion Transport Analysis

Neutron diagnostic is one of the powerful tools for the fast ion study in the DD and DT burning plasmas. Neutron production processes in burning plasmas and the neutron diagnostic techniques are briefly reviewed. The transport analysis of the neutron yield in the NB heated JT-60 plasma provides the fast ion contribution in the neutron production process. The fast ion slowing down process has been studied by the NB blip experiments at JT-60. Triton burn-up measurement has been carried out with the 14-MeV neutron measurement in the DD plasma, which shows that the behavior of 3.5-MeV tritons is almost classical. Neutron emission during the TAE mode driven by 400-keV negative-ion-based NB has been investigated at JT-60. The knock-on α spectroscopy is tried in the DT campaign of JET by the magnetic proton recoil spectrometer.

Operation and Development on the Positive-Ion Based Neutral Beam Injection System for JT-60 and JT-60U

The positive-ion based neutral beam injection (NBI) system for JT-60, which consists of 14 beamline units and has a beam energy of 70 to 100 keV, started operation in 1986 with hydrogen beams and injected a neutral beam power of 27 MW at 75 keV into the JT-60 plasma. In 1991, the NBI system was modified to be able to handle deuterium beams as part of the JT-60 upgrade modification. After executing some research and developments, deuterium beams of 40 MW at 95 keV were injected in 1996. As a result, NBI has contributed to the achievement of the highest performance plasmas, a DT-equivalent fusion power gain of 1.25 and a fusion triple product of 1.55 × 1021 keVs/m3, in the world on JT-60U.

Beam performance of negative-ion based NBI system for JT-60

The negative-ion based NBI system (N-NBI) for JT-60 is being operated for a high energy neutral beam injection into the JT-60 plasma, in parallel with increasing the beam power. The estimation of the beam characteristics is very important for the enhancement and optimization of the beam power. The following four items have been used as the characterization of the beam: neutron yield from the beamline, beam divergence, heat load on the beamline components and beam profile on the armor plate in the tokamak vacuum vessel. Since the beam power accelerated in the ion source is proportional to the neutron yield in the beamline components, the operation status of deuterium negative ion current can be monitored easily. The beam divergence estimated at the drift duct and ion dumps, and the heat load on the beamline components are used for the optimization of the ion source operation parameters and the evaluation of neutral beam injection power. The beam divergence measured at the drift duct is 4 mrad in the horizontal direction, 6 mrad in the vertical, which are in agreement with the design value of 5 mrad. The deposition profile on the armor plate is used for monitoring of the neutral beam profile.

Power flow in the negative-ion based neutral beam injection for JT-60

The negative ion based neutral beam injection system for JT-60 has operated since 1996 injecting neutral beam into JT-60 plasmas. A power flow measurement in the beam line and ion source with a water calorimeter had shown that 40%–50% of accelerated beam particles were intercepted on the two accelerator grids and the grounded grid at an ion source gas pressure of 0.2–0.3 Pa. Much of the beam loss was not caused by stripping loss of the negative ions, but rather by direct impingement of the negative ions onto the grids. After reducing the acceleration area by masking the edge area (about 13% of the extraction area) of the accelerator grid so as to minimize the edge effect of magnetic field in the arc chamber, the loss in the accelerator decreased by roughly 25%. In comparing a deuterium beam with a hydrogen beam, the neutral beam power with deuterium is lower by 30% than that of hydrogen at the same arc power, although the heat load onto the grounded grid does not change so much. The power deposition ratios along the beamline were as follows: the beam scraper in the ion source tank and the neutralizer cell received about 3% and about 7%, respectively, of the accelerated beam power, while the ion dumps for both D− and D+ received 20%–30% in total, and 30%–34% reached the neutral beam calorimeter.

Development of Negative-Ion Based NBI System for JT-60

The negative-ion based neutral beam injector (N-NBI) for JT-60 has been developed for plasma core heating and neutral beam current drive in higher density plasmas. Construction of the N-NBI system was completed in 1996, and just after this completion, the efforts to increase beam power and beam energy started. The N-NBI system has already operated with negative ion beams with 14.3 A at 380 keV of deuterium and with 18.5A at 360 keV of hydrogen. During N-NBI experiments on JT-60, a deuterium neutral beam power of 5.2MW at 350keV has been injected for 0.7s stably, and the response of the JT-60 plasma to high energy beam injection with the N-NBI has been confirmed to be in agreement with a theoretical prediction.

Beamline performance of 500 keV negative ion-based NBI system for JT-60U

Construction of the 500 keV negative-ion-based NBI system (N-NBI) with its maximum capability of 10 MW and 10 s for NB current drive and plasma core heating in high-density plasmas in JT-60 was completed in March 1996. The neutral beam injection into JT-60 with N-NBI has been performed since then. One of the key issues in the injection was how to control the bending magnetic field for residual ion beams. In N-NBI for JT-60, a stray magnetic field from the tokamak was actively used for bending of the residual ion beams. Combination of the stray magnetic field from JT-60 and the magnetic field of the deflecting coils was controlled sufficiently to successfully deposit residual ion beams at a constant point on the ion dump surface. The deposited positive and negative beam profiles on the ion dump were measured with an array of thermocouples brazed on the dump surface. This report describes the performance test of bending the residual ion beams and the measurement of the heat load on the ion dump for JT-60 N-NBI experiments.

Development of Negative-Ion Based NBI System for JT-60.

The negative-ion based neutral beam injector (N-NBI) for JT-60 has been developed for plasma core heating and neutral beam current drive in higher density plasmas. Construction of the N-NBI system was completed in 1996, and just after this completion, the efforts to increase beam power and beam energy started. The N-NBI system has already operated with negative ion beams with 14.3 A at 380 keV of deuterium and with 18.5A at 360 keV of hydrogen. During N-NBI experiments on JT-60, a deuterium neutral beam power of 5.2MW at 350keV has been injected for 0.7s stably, and the response of the JT-60 plasma to high energy beam injection with the N-NBI has been confirmed to be in agreement with a theoretical prediction.

ＪＴ‐６０用負イオンＮＢＩ装置の開発

A 500keV/10MW negative-ion based NBI system for JT-60 has been completed in March 1996 and an injection experiment has started. This system is the first NBI system based on negative-ions, which is considered to be the first candidate as a plasma heating and current drive device for the ITER (International thermo-nuclear fusion experimental reactor), and the purpose of the system is to prove physically and technically the feasibility of realizing the negative-ion based NBI system. Prior to the completion of the whole NBI system, an ion source-power supply coupling test has been conducted to confirm the performance of negative-ion generation and beam acceleration with a part of the NBI system completed in March 1995. In the test between June and October 1995, a deuterium negative-ion beam has reached an acceleration current of 13.5A, ion beam power of 5.4MW at the beam energy of 400keV, which is the world highest negative-ion beam power. The whole NBI system has completed in December 1995. After confirming the performance of two ion sources, beamline and power supply, a deuterium beam of 180keV, -0.1MW has been injected into the JT-60 plasma successfully in March 1996.

Ion temperature profile simulation of JT-60 and TFTR plasmas with ion temperature gradient mode transport models

Ion temperature profiles of neutral beam heated plasmas in JT-60 and TFTR have been simulated using models of ion thermal diffusivity, χi, based on ion temperature gradient mode (ηi mode) turbulence and drift wave turbulence (trapped electron mode and circulating electron mode). The ion temperature profiles measured by charge exchange recombination spectroscopy are compared to predicted Ti profiles calculated from three theoretical models for ion heat transport by Dominguez and Waltz (1987), Lee and Diamond (1986), and Romanelli (1989). All three χi models can reproduce the measured Ti profile over a wide range of parameters in JT-60 L mode plasmas, except for two 1 MA limiter cases. With the use of transport models, which were adjusted to fit JT-60 plasmas, it was found that the Dominguez and Waltz and Romanelli models agree with measured Ti, profiles in TFTR L mode discharges in the region a/4 < r < a/2. The observed central peaking of Ti near the magnetic axis of TFTR L mode plasmas cannot be reproduced. It was found that the Lee and Diamond model does not fit me data as well. The χi models studied cannot consistently reproduce the measured Ti in the peripheral region of either JT-60 or TFTR plasmas. In the high ion temperature (high Ti) plasmas of JT-60 and supershot plasmas of TFTR, the predicted Ti profiles are much broader than the measured profile. In supershot plasmas, the measured central ion temperature greatly exceeds the predicted temperature, although there is reasonable agreement near the edge

ICRF Current Drive and Helicity Injection

The relation between RF-driven current and wave helicity is analytically and numerically studied for a tokamak plasma. The helicity conversion ratio from the wave to the plasma is discussed and numerically examined for the waves in the range of ion cyclotron frequency in JT-60 plasma. The wave propagation equation which includes kinetic effects is solved as a boundary-value problem with one-dimensional inhomogeneities. Both the RF helicity-driven current and the RF momentum driven current are calculated. It is shown that the efficiency of the current drive by the wave helicity is small compared with the conventional scheme for the fast wave excitation of the present-day plasma parameters.

Eddy current effect study on the JT-60 plasma equilibrium control

The A-ф method (A is the magnetic vector potential and ф is the scalar potential) was applied to eddy current analysis in the JT-60 tokamak fusion device. This analysis agrees with the coil excitation experiments much better than the filament-current-coil approximation. The accuracy of the A-ф method is estimated by comparing it with the analytical solution of magnetic field penetration into an infinitely long cylindrical conductor. A JT-60 plasma equilibrium-control-dynamics (ECD) model including eddy currents induced in the vacuum vessel is proposed for control system design and analysis. Considering the plasma-vessel interaction, the one-point plasma model and its equilibrium force balance are assumed. Comparison of the simulation and experimental data identifies the essential points for reproduction of the plasma ECD: (a) The magnetic field induced by the plasma motion is proportional to its velocity. (b) Magnetic field penetration is regarded as a response of a first-order differential equation. Application limits of the employed model are also identified: (a) Plasma volume effects are not accurately involved. (b) The influence of eddy current on shape evolution is not taken into consideration.

Improvement of confinement and remote radiative cooling of beam heated plasmas in IT-60 with ion Del B drift towards the X-point

Improvement of both energy and particle confinement was observed in beam heated hydrogen plasmas in JT-60 when the heating power and the line averaged density exceeded thresholds. The radiation power from the divertor plasma increased to as much as 50% of the heating power. This regime was termed improved divertor confinement (IDC). The discharges with improvement of the energy confinement by up to 20% and enhanced divertor radiation lasted for several seconds in a quasi-steady state without significant impurity accumulation in the core. The threshold line averaged electron density for the IDC regime increased linearly with power and dropped with increasing safety factor. The improvement was observed only when the ion Del B drift was towards the X-point. The behaviour of light impurities and in-out asymmetries in particle recycling changed with reversal of the toroidal magnetic field. These findings are consistent with the theoretical model in which collisional transport near the separatrix is connected with the onset of the IDC regime

Anomalous ion transport of L-mode plasmas in JT-60

Local heat transport in auxiliary heated JT-60 plasmas has been analysed, with emphasis on the understanding of the deteriorated confinement observed in the L-mode. The dependence of the local thermal transport on plasma current, density and heating power have been studied in both divertor (single null, lower X-point) discharges and limiter discharges with hydrogen beam injection into hydrogen plasmas. The deterioration of the energy confinement time with increasing auxiliary heating power-the so-called power scaling-is mainly due to the stronger impact of the ion heat transport. It is found that the ion thermal diffusivity has a favourable density dependence. Experimentally obtained profiles of the ion temperature have been compared with the predictions of the theoretical model, especially with regard to the transport coefficients derived from the theory of the ion temperature gradient instability. For L-mode plasmas in JT-60, the theory has been found to be in good quantitative agreement with measurements over a wide range of plasma parameters, with the exception of limiter discharges with a plasma current of 1 MA

Central magnetohydrodynamic activity in pellet-fueled JT-60 plasmas

The central magnetohydrodynamic (MHD) activities are strongly affected by hydrogen pellet injection. The change in the sawtooth characteristics (viz., crash time, crash mechanisms, and sawtooth period) seem to be dependent on the density (and pressure) peakedness. With deepening pellet penetration, the sawtooth frequency becomes longer. At the sawtooth emerging after the deep pellet penetration into high-Ip limiter discharges, only a small amount of the central kinetic energy is released and the crash does not follow the fully reconnecting style. The sawtooth crash after the pellet injection tends to have more ideal-like characteristics for higher density and pressure peaking factors. At each sawtooth, the m=1 rotation frequency changes suddenly to the ion-diamagnetic direction or the codirection (parallel to the plasma current).

Simulation study of pellet fuelled plasmas in JT-60

The transport of pellet fuelled plasmas in JT-60 has been investigated with a predictive tokamak transport code. Peaked density profiles have been observed in pellet fuelled plasmas during the sawtooth free phase. Simulation analysis shows that an inward pinch of ∼0.2 m/s at the plasma half-radius and a reduced particle diffusion coefficient of 0.1 m2/s in the central region are necessary to explain the peaked density profile. These particle transport properties yield improved energy confinement in a plasma with a strong particle source in the central region under sawtooth free conditions, i.e. in a plasma fuelled by pellet injection. The dependence of the improved stored energy on the plasma current can be explained by reduced particle diffusion within the q = 1 surface.

Ideal MHD stability of pellet fuelled plasmas in JT-60

Stability against ideal infinite-n ballooning and low-n internal kink modes is analysed for pellet fuelled plasmas in JT-60, which have strongly peaked pressure profiles within the q = 1 resonance surface. The observed pressure gradient in pellet fuelled plasmas locally reaches the marginal value for the infinite-n ballooning mode limit just inside the q = 1 surface. After reaching the marginal value on the magnetic flux surface, the marginally stable region grows inside the q = 1 surface. However, the poloidal beta within the q = 1 surface, βp1, does not reach the limit imposed by the ideal ballooning mode. The results of the analysis of the low-n kink mode using experimental observations show that the maximum βp1 attained is consistent with the beta limit imposed by the n = 1 internal kink mode.

MARFE phenomena on neutral beam heated JT-60 plasmas

In JT-60, MARFEs have been observed frequently in high- I p and high density limited discharges with neutral beam heating after replacement of the first wall from TiC coated molybdenum tiles to graphite ones. The MARFEs occur near the density limit of the JT-60 plasmas. The NB heating is effective to suppress the MARFE. The empirical scaling of the threshold density for MARFE onset taking account of the NB power is obtained, which is useful to predict the MARFE onset. The radiated power from the plasma with MARFE is about 90% of the absorbed power. Both stored energy and central electron temperatures do not change with the MARFE onset in spite of the such intense radiation loss. So the MARFE does not affect the core plasma. A simple model based on the thermal equilibrium in a peripheral plasma can explain the onset condition of the MARFE.

Measurements of neutral beam stopping for hydrogen and helium in the JT-60 plasma

Shine-through for hydrogen and helium beams in a plasma was measured. Shine-through for an atomic hydrogen beam with f1:f2:f3=7%:32%:61% (where f1, f2 and f3 represent fractions of the full, half and third energy, respectively) and a full energy of 140 keV, was smaller than the value predicted by usual cross-sections without inclusion of excitation processes at an electron density (ne) above 3*1013 cm-3. A fractional beam stopping increment ( delta ), which was defined by delta =( sigma s- sigma s(0)) sigma s(0) (here sigma s was an experimental stopping cross-section and sigma s(0) was the cross-section without including multistep processes via excited states), was about 0.2 at ne=(3.1-5.0)*1013 cm-3 and Zeff=1.6-2.5. The value of delta was consistent with the calculation by Janev et al. (1989, Nucl. Fusion 29, 2125). For a 160-190 keV helium beam, the enhancement of the stopping cross-section was not observed, as expected.

Radiation losses and global power balance of JT-60 plasmas

The radiation losses and the global power balance for Ohmic and neutral beam heated plasmas have been investigated in different JT-60 configurations. Discharges with a TiC coated molybdenum wall and with a graphite wall, with limiter, outer and lower X-point configurations have been studied by bolometric measurements, thermocouples and an infrared TV camera. In neutral beam heated outer X-point discharges with a TiC coated molybdenum first wall, the radiation loss of the main plasma was very low (10% of the absorbed power). The radiation loss due to oxygen was dominant in this case. On the contrary, in discharges with TiC coated molybdenum limiters the radiation loss was very high (>60% of the absorbed power). In the discharges with a graphite wall the radiated power from the main plasma was 20-25% for both limiter and lower X-point configurations. In lower X-point discharges the main contributor to the radiation loss was oxygen, whereas in limiter discharges the loss due to carbon was equal to the loss due to oxygen. The radiation loss from the lower X-point divertor increased with increasing electron density of the main plasma.