N. Suzuki

Global energy confinement H-mode database for ITER

Describes the content of an H-mode confinement database that has been assembled for the ITER project. Data were collected from six machines of different sizes and shapes: ASDEX, DIII-D, JET, JFT-2M, PBX-M and PDX. A detailed description of the criteria used in the selection of the data and the definition of each of the variables is given. The authors also present an analysis of the conditions of the database, the scalings (power law and offset linear) of the data with both dimensional and dimensionless variables, and predictions of the expected confinement time for ITER

ITER H mode confinement database update

This paper describes an update of the H mode confinement database that has been assembled for the ITER project. Data were collected from six machines of different sizes and shapes: ASDEX, DIII-D, JET, JFT-2M, PBX-M and PDX. The updated database contains better estimates of fast ion energy content and thermal energy confinement times, discharges with RF heating, data using boronization, beryllium and pellets, more systematic parameter scans, and other features. The list of variables in the database has been expanded, and the selection criteria for the standard dataset have been modified. We also present simple scalings of the total and thermal energy confinement time to the new dataset

H mode power threshold database for ITER

The ITER Threshold Database, which at present comprises data from nine divertor tokamaks, is described. The main results are presented and discussed. The properties and dependences of the power threshold in individual devices are reviewed. In particular, the analysis shows a rather general linear dependence on magnetic field, but a non-monotonic density dependence that varies from device to device. Investigation of the combined database suggests that the threshold dependence Pthres approximately=0.3neBT2.5 shows reasonable agreement with the data. This expression yields Pthres approximately=150 MW at a density of 0.5*1020 m-3 for ITER. Other expressions with weaker size dependence, and therefore lower threshold power for ITER, are also discussed. Their agreement with the present data is poorer than that of the above expression. In addition, the database is investigated by statistical discriminant analysis. The edge data included at present are described and discussed. Finally, there is a discussion of the implications of the results for ITER

Improvement of energy confinement time by continuous pellet fuelling in beam-heated Doublet III limiter discharges

A centrifuge injector that repetitively fires 1.3 mm deuterium pellets (1 torr⋅L per pellet) at a rate of 32 pellets per second was used to build up and maintain a Doublet III 2.4 MW neutral-beam-heated limiter discharge at a line-averaged density of 1 × 1014 cm−3. When compared to a conventional gas-fuelled plasma at similar density, the pellet-fuelled plasma was characterized by a factor-of-three reduction in edge neutral density and limiter recycling, a centrally peaked profile, a 70% increase in global energy confinement, and a tenfold increase in the fusion reaction rate.

Confinement and fueling studies during additional heating phase in the JFT-2M tokamak

Increments of peripheral hydrogen/deuterium neutral gas pressures (PH2/PD2) during the additional heating phase (neutral beam and RF heatings) have been observed in a D2 gas-puff fueled JFT-2M tokamak with H2-absorbed graphite limiters/divertor plates. In the beam heating phase, a large increment of PH2 raises the plasma density 2 times or more without the degrading energy confinement time. The D2 gas-puff valve is closed in this phase. This improvement is interpreted as being due to “wall fueling”, i.e., fueling by desorbed hydrogen from the graphite wall by energetic particles. The first observation of the H-mode in the INTOR-type stubby open divertor with a short divertor channel (1–8 cm) at the high density regime (4–7 × 1013 cm−3) enhanced by the wall fueling is presented. The improved energy confinement time is comparable to or higher than that of ohmically heated discharges.

Effects of ergodization on plasma confinement in JFT-2M

A steady-state H-mode plasma with frequent ELMs (edge localized modes) was obtained by applying magnetic fields with high poloidal mode numbers resonant in the plasma edge. The induced ELMs limit the density and impurity accumulation normally observed during ELM-free H-mode. An EML coil with numerous resonant modes ( m 10) in the plasma edge appears the most effective. This method of control is extended to higher auxiliary heating power by increasing the magnitude of the magnetic perturbations.

Transition from the L-mode to the H-mode by electron cyclotron heating of a tokamak edge plasma

Transitions of L-mode plasmas to the H-mode have been induced by an electron cyclotron heating (ECH) pulse. The transitions occur when ECH is applied to plasmas preheated either by a neutral beam or by waves in the ion cyclotron range of frequency with power levels well below their own threshold power for the H-mode transition. The position of the electron cyclotron resonance layer has been scanned and it has been shown that edge heating rather than central heating is effective in inducing the transition to the H-mode.

Efficient lower hybrid current drive using a multijunction launcher on JT-60

A multijunction launcher was designed and constructed to improve the directivity of launched waves with a narrow wave number spectrum by dividing each main waveguide into three secondary waveguides. The coupling characteristics agree fairly well with theory, and a power of up to 2 MW was obtained after a few days of conditioning. The dependence of the current drive efficiency ηCD (= neIRFR/PLH) on the wave spectrum was studied by changing the phase difference between the adjacent main waveguides. It was found that high current drive efficiency is obtained by waves interacting with fast electrons as far into the plasma as the waves travel. A maximum current drive efficiency of 2.8 × 1019 AW−1m−2 was achieved with this launcher at a plasma current of 1 MA. The efficiency of the multijunction launcher was higher by 40% than that of the previous conventional launcher on JT-60.

Simple multijunction launcher with oversized waveguides for lower hybrid current drive on JT-60U

Abstract A multijunction technique with oversized waveguides has been developed for the lower hybrid current drive launcher on JT-60U. The launcher consists of 4 (toroidal) × 4 (poloidal) multijunction modules. RF power in the module is divided toroidally into 12 sub-waveguides at a junction point through an oversized waveguide. This method simplifies the structure of the multijunction launcher with a large number of subwaveguides. A maximum power density up to 25 MW m −2 has been achieved with a low reflection coefficient of less than 2%. The coupling and current drive efficiency are well explained by the designed wave spectra without taking account of higher modes in the oversize waveguides. Thus, the simple multijunction launcher has been demonstrated to excite expected wave spectra with high power handling capability.

Current drive and confinement studies during LHRF experiments on JT-60

Results are presented of the first Lower Hybrid Current Drive (LHCD) experiments in JT-60. 2 MA of RF driven current is successfully produced for the first time in a reactor grade tokamak. The magnetic divertor works quite well in eliminating the impurities released by the current carrying fast electrons which have allowed the generation of the reactor relevant RF current in a very low density plasma. The efficiency which is defined as ηCD = eRIRF/PLH(1019 m−3 AW−1), reaches values of 0.8 to 1.7. NBI heating enhances the current drive efficiency by a factor of 1.5, and LHCD improves the confinement time of high power NBI heated plasma. The key to confinement improvement is found to be the active control of the current profile by LHCD.