JT-60U Divertor Research Articles

Imaging plate technique for determination of tritium distribution on graphite tiles of JT-60U

The tritium imaging plate technique was applied to determine surface tritium distributions on graphite tiles used as the first wall and W-shaped divertor in JT-60U, in which tritium produced by the D–D nuclear reaction in the plasma was implanted and/or deposited depending on the incident energy. Measured samples were isotropic graphite (IG-430U) and CFC graphite (CX-2002U), used as divertor tiles and/or baffle plates just outside the divertor. Tritium areal distributions on graphite divertor tiles, dome units and baffle plates of JT-60U were successfully measured for the first time. Tritium distributions observed in JT-60U tiles can be explained by homogeneous implantation of high energy tritium which is influenced by redeposited layers and redistributed by the temperature increase due to the plasma heat load. The tritium retention in graphite heated above 800 K was significantly small.

Edge plasma parameters at the L-H transition under the conditions of open and W-shaped divertor in JT-60U

A substantial reduction of the L-H transition thresholdpower was observed under the W-shaped divertor in JT-60U, in comparisonwith the open divertor. Radiation power in the main region was alsodecreased in this case. These differences of radiation power from themain plasma between open and W-shaped divertor were not large enough toaccount for the apparent reduction of threshold power. In this paper,the emphasis is on the edge plasma parameters which relate to the L-H transition. It was found that edge plasma pressure just before the L-Htransition in the W-shaped divertor case became smaller than that in theopen divertor case. After the modification of divertor geometry, edgeion collisionality just before the L-H transition also became lower andit was established with lower auxiliary power input. It was suggestedthat reduction of edge ion collisionality arose from neutral particlesnear the X-point by the analysis of poloidal profile of neutral density.Therefore, neutral particles near the X-point prevented the thresholdpower for the L-H transition from more reduction at low plasma density,and caused the shift of plasma density which gave minimum thresholdpower.

Helium exhaust in divertor–closure configuration with W-shaped divertor of JT-60U

The W-shaped divertor of JT-60U was modified from inner-leg pumping to both-leg pumping. After the modification, the pumping rate was remarkably improved with both-leg pumping in a divertor-closure configuration, which means both separatrixes close to the divertor slots. In steady state, efficient helium exhaust was realized in a divertor-closure configuration with both-leg pumping in ELMy H-mode plasmas. A global particle confinement time of τ He = 0.4 s and τ He /τ I = 3 was achieved in attached plasmas, well within the range generally considered necessary for successful operation of a future fusion reactor, such as ITER-FEAT. The helium exhaust efficiency in the divertor-closure configuration was extended by 45% as compared to the one with inner-leg pumping. In the high X-point configuration with both-leg pumping, the helium exhaust efficiency and the pumping rate of deuterium deteriorated because of the backflow through the outer slot.

Measurement of natural plasma flow along the field lines in the scrape-Off layer on the JT-60U divertor tokamak

The plasma profile and parallel plasma flow in the scrape-off layer (SOL) were systematically measured using reciprocating Mach probes installed at the outer midplane and near the divertor magnetic null (x point) in the JT-60U tokamak with a single null divertor. For the ion vertical drift due to the toroidal magnetic field gradient (ion nablaB drift) directed towards the divertor, SOL plasma flow along the magnetic field lines away from the divertor ("flow reversal") was discovered at the midplane far from the divertor. A quantitative evaluation of the ion "Pfirsch-Schluter flow," wherein the parallel flow is naturally produced in a toroidal plasma, was consistent with the measurement.

Plasma characteristics near the X-point in W-shaped divertor of JT-60U

A reciprocating Langmuir probe was installed below the X-point of W-shaped divertor in JT-60U. Plasma profiles were measured from the outer divertor nose to the X-point. A large peak was observed in the density profile near the X-point as the main plasma density exceeded about 60% of MARFE onset density. The electron density at the peak exceeded the average density of the main plasma. The difference in electron temperature between the X-point and the outer target was smaller at the separatrix than outside the separatrix. The electron temperature near the separatrix decreased rapidly from 60 to 20 eV as the density peak increased. The electron temperature profile became flat and decreased down to ∼10 eV during a MARFE.

Carbon impurity behavior in w-shaped pumped divertor of JT-60U

The divertor geometry of JT-60U was modified from an open type to a w-shaped pumped configuration. This paper discusses behavior of carbon impurity in the w-shaped divertor. Carbon generation by chemical sputtering is important in deuterium plasmas. The reduction of carbon generation in helium discharges is consistent with this conclusion. CD-band intensities in the private flux region were smaller in the w-shaped divertor than in the open divertor, suggesting that the divertor dome is effective in decreasing chemically sputtered hydrocarbon molecules. The observed CD-band intensity was fairly reproduced by simulation using the IMPMC-code. When deuterium gas was puffed into the divertor, Z eff value in the main plasma became about 15% larger than with puffing from the main chamber. For reducing impurity in the main plasma, the gas puffing from the top of main plasma was effective.

Particle control and behaviour of neutrals in the pumped W-shaped divertor of JT-60U

As a trial for an active particle control, simultaneous feedback control of the divertor neutral pressure and main plasma density by the dual locations of gas puff is performed in the W-shaped divertor of JT-60U. At a relatively low gas-puffing rate, the divertor neutral pressure and main density are well controlled as pre-programmed. Lowering the X-point height is effective for isolation of the main plasma from divertor neutral particles. Simultaneous control has some difficulty at a high gas-puffing rate due to an increase of the cross-term coupling with each gas puff. Particle flux from the neutral beam also makes it difficult to control divertor neutral pressure and density. The simultaneous controllability depends on the operational region of the divertor neutral pressure and the main density. The coupling of the divertor neutral pressure with gas puff becomes larger in the high flux region, so that core fuelling in order to minimise the coupling is necessary for an effective particle isolation.

Helium exhaust in ELMy H-mode plasmas with W-shaped pumped divertor of JT-60U

By injecting a neutral beam of 60 keV helium (He) atoms as central fueling of helium into the ELMy H-mode plasmas, helium exhaust has been studied in the W-shaped pumped divertor on JT-60U. Efficient He exhaust was realized by He pumping using argon frosted cryopumps in the JT-60U new divertor. The He fueling rate by He beam injection was balanced with He pumping. In steady state, good He exhaust capability was successfully demonstrated in attached ELMy H-mode plasmas. A global particle confinement time of τ* He=0.7 s and τ* He/ τ E=4 ( τ E: the energy confinement time) was achieved in attached plasmas, well within the range generally considered necessary for successful operation of a future fusion reactor, such as ITER. The enrichment factor of He was obtained to be about 1.0, which is five times larger than the ITER requirement of 0.2. Good He exhaust capability was also obtained in detached ELMy H-mode plasmas, which was comparable to one in attached plasmas. This result of the helium exhaust is sufficient to support a detached divertor operation on ITER.

Study of Intensity Ratios of He I Lines (668nm, 706nm and 728nm) for Measurement of Electron Temperature and Density in the JT-60U Divertor Plasma.

Emission of He I lines (668nm, 706 nm and 728nm) has been simultaneously observed in the JT-60U divertor plasma. From calculations using a collisional-radiative model, intensity ratios of these He I lines are considered to be useful to measure electron temperature and density. In order to evaluate the feasibility of the intensity ratio method for temperature and density measurements, the measured intensity ratios are compared with those calculated from the temperature and density measured with Langmuir probes. As a result, the measured intensity ratios are well reproduced by the calculations. The discrepancies in the ratios I (728nm)/I (706nm) and I (668 nm)/I (728nm) between the measurements and calculations are about 12%.

Development of a compact W-shaped pumped divertor in JT-60U

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current.

Comparison of edge neutral effect on the condition of H-mode transition before and after the modified divertor in JT-60U

The JT-60U divertor geometry was recently modified to a W-shaped structure in order to reduce the back flow of neutrals from the divertor. The H-mode threshold power decreased subtly compared with the open divertor case. As regards density dependence, the exponent of density was 0.5 to 0.75. In addition, we found that the outflux from the outer baffle plate should be considered for the evaluation of neutrals.

Fluctuations in Plasma Equilibrium Control on JT-60U

Experimental analyses are presented of fluctuations observed in a plasma-current-center, I p -center, location and the inward clearance between the outermost flux surface and the first wall at plasma perturbations (e.g., β p collapse and disruption) for JT-60U divertor plasmas. The neutral point for vertical and horizontal movement of the I p center at plasma disturbances is confirmed, and attachment of the plasma to the inward first wall is avoided because of the increase in l i at fast plasma-current quench. These experimental observations will support the design of fusion reactors like the International Thermonuclear Experimental Reactor (ITER).

Spectroscopic study of radiative losses in the JT-60U divertor plasma

Spatial distributions of absolute line intensities of D I and C II-IV in the divertor region have been measured simultaneously with a 38-channel optical fibre array in JT-60U. The radiative losses in the divertor region have been derived from the intensities. In L-mode discharges without the MARFE, the fractions of the radiative losses due to deuterium and carbon are 36% and 64%, respectively. The carbon radiation is related to the carbon influx by a simple impurity model based on the confinement time of the carbon ion on the assumption of the physical sputtering. The power radiated due to carbon accounts for a large increase in the radiative losses by the MARFE.

Radiative divertor in JT-60U

Characteristics of the divertor radiation loss are investigated with the spatially resolved bolometer arrays. In discharges with the ion grad- B drift toward the divertor, a radiative region is formed between the outer separatrix strike zone and the X-point. The radiative state is accompanied by a gradual decrease in heat flux to the divertor plates, and lasts for a few seconds with large particle fluxes to the separatrix strike zones, finally leading to a MARFE at the X-point. This radiation loss fraction of the input power depends on the electron density, NB heating power and safety factor, and can reach the level of radiation loss found in the MARFE. In the discharges with the ion grad- B drift away from the divertor, the MARFE has never been generated at the X-point. However, there is no significant difference in the maximum radiation loss level. The radiation loss is nearly in-out symmetric when the radiation loss is large for both types of discharges.

Hydrogen isotope retention and impurity deposition of carbon based components used in JT-60U

Various types of carbon based components, such as isotropic graphite, carbon fiber composite (CFC-2D felt type) and B 4C-overlaid CFC/graphite were used for divertor and first wall tiles in JT-60U. Surface analysis of tiles exposed to discharge shots in JT-60U, was performed using MeV region ion beam. The retention of deuterium for each tile was distributed between ∼ 6 × 10 20 and ∼ 6 × 10 21 m −2. Highest deuterium retention was observed on the outboard, lower part of first wall tiles. Metallic impurity measurement was performed by X-PIXE. The dominant species were Fe, Cr, Ni, Cu, Mo and Zn of which total concentration as high as 10 21 m −2. This value is one order of magnitude smaller than that of JT-60. Highest oxygen concentration examined by RBS was observed on the B 4C converted graphite tile. Carbon tile containing B 4C is shown to be useful for oxygen gettering compared with an other.

Latest results from JT-60U divertor research

The JT-60U divertor research program is designed to address key issues for the boundary plasma of future tokamaks. The work is organized into five key areas of activity. These research areas include: (1) heat and particle exhaust in the divertor and particle confinement in the main plasma, (2) formation of a radiative divertor and impurity production processes, (3) disruption and its effects, (4) development of theoretical and computer models to predict the operational parameters of future machines, and (5) the conceptual design study of a new radiative pump divertor. The divertor and scrape-off layer (SOL) play a crucial role in establishing the overall performance capabilities of the tokamak. In particular, the formation of stable radiative divertor is one of the most critical issues to develop the concept as a method to reduce and control heat flux to the divertor target. This paper highlights the recent major results from the divertor and boundary physics research in JT-60U.

The softening of current quenches in JT-60U

Current quenches caused by density limit disruptions have been investigated in JT-60U divertor plasmas in order to develop general methods of reducing the current decay rate and of suppressing the generation of runaway electrons. The reduction of the impurity influxes at an energy quench and the direct neutral beam (NB) heating of the plasma core during a current quench were beneficial for reducing the speed of the current quench. The low stored energy just before the energy quench and the high safety factor at the plasma edge had the advantage of decreasing the impurity influxes at the energy quench. The detached plasma state was useful for degrading the energy confinement for both joule and NB heated plasmas within a short time period. Runaway electrons were not generated at plasma densities of more than 5*1019 m-3 measured just before the energy quench. Fast controlled plasma shutdown with energy quenches but without a current quench was demonstrated successfully from 2 MA to zero, with a value of dIp/dt of -6 MA/s. This shutdown was obtained by lowering the stored energy within a short time period of 20 ms, actively using the detached plasma state produced by intense helium gas puffing

Transport analysis of divertor plasma in JT-60U

A simple divertor model has been developed in order to analyze the divertor transport based on experimental data. The fluid equations are solved numerically in one dimension along the magnetic field lines. In contrast to the conventional divertor code, the boundary conditions are given by experimentally determined plasma parameters at the divertor plate, which are measured by Langmuir probes. The neutral particle transport is simulated in two dimensions by a Monte Carlo code. The interaction between the plasma and the neutral particles is solved self-consistently using an iterative procedure; the plasma parameters converge in a few iterations. The little computational time needed facilitates a systematic analysis of the divertor transport. By applying this model to the JT-60U divertor plasmas, the particle confinement time in the main plasma and the heat diffusivity in the scrape-off layer are evaluated.

Impurity generation mechanism and remote radiative cooling in JT-60U divertor discharges

The impurity behavior and the remote radiative cooling have been studied in the JT-60U divertor discharges with high power NB heating. NB at a power level of 20 MW was injected for 2 s, and no carbon bloom has been observed. For the first time, the measurement and the calculation for the carbon influx from the divertor plates are compared quantitatively. The carbon generation mechanism can be explained by the sputtering with deuterium (physical sputtering), oxygen and carbon. It is found that the contribution of the chemical sputtering with deuterium is small and that of the sputtering with oxygen is important. Remote radiative cooling can reduce the heat flux onto divertor plates. The dependence of the radiative cooling power on the electron density and the safety factor is presented. The favorable operation regime for discharges with high power auxiliary heating is discussed based on the experimental results.

Heat load on the divertor plate of JT-60 during normal and abnormal operations

Temperature measurements and inverse thermal analyses of the heat flux on the divertor plates of JT-60 have been performed in both outside X-point and lower X-point operations. These investigations showed that heat load behavior during normal operation is quite different depending on parameters such as plasma density, the shape of the magnetic surface near the divertor plate, polarity of the plate (ion-side vs. electron-side) and the plate materials (TiC-coated Mo vs. graphite). Due to contracting magnetic fluxes near the divertor plate in the outside X-point operation, the peak heat flux on the outside divertor plate was higher than that on the lower divertor plate. Large heat concentration was produced at the edge of the plate due to a small gradient of the magnetic field line incident on the plate, which resulted in serious damage to the divertor plates. Heat deposition due to high energy runaway electrons produced during disruption was observed on the lower divertor plate. Based on these experimental results, heat load and design issues of the JT-60U divertor plate are discussed.