H. Tamai

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

Fast plasma shutdown scenarios in the JT-60U tokamak using intense mixed gas puffing

Fast plasma shutdown without runaway electron generation by gas puffing is investigated in the JT-60U tokamak. Argon-only injection enables a fast shutdown; however, it induces runaway electron generation. Hydrogen-only injection generates much less runaway electrons; however, the shutdown time is considerably longer. Mixed injection of hydrogen and argon achieves a fast plasma shutdown without runaway generation. Argon atoms contribute to radiate the energy of plasma leading to a fast plasma shutdown, whilst hydrogen atoms contribute to increase the electron density for avoiding runaway electron generation.

Role of divertor geometry on detachment and core plasma performance in JT60U

Experimental results related to the divertor geometry such as divertor plasma detachment, neutral transport and plasma energy confinement, were compared in the open and W-shaped divertors. The ion flux near the outer strike point was larger than in the open divertor, and the electron temperature at the target, T e div , was reduced. Divertor detachment and x-point MARFEs occurred at n e 10-20% lower than that for the open divertor. Although the leakage of neutrals from the divertor to the main chamber decreased, a neutral source in the main chamber due to an interaction of the outer scrape-off layer (SOL) plasma to the baffle plates became dominant above the baffle. Degradation in the enhancement factor of the energy confinement was observed similarly in the open and W-shaped divertors. The neutral density inside the separatrix was estimated to be a factor of 2-3 smaller, which did not affect the energy confinement.

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

Conceptual Design of Superconducting Magnet System for JT-60SA

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) has been decided by both parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of 18 toroidal field (TF) coils, a Central Solenoid (CS) with four modules, seven Equilibrium Field (EF) coils. The TF case encloses the winding pack and is the main structural component of the magnet system. The CS consists of independent winding pack modules, which is hung from the top of the TF coils through its pre-load structure. The seven EF coils are attached to the TF coil cases through supports which include flexible plates allowing radial displacements. The CS modules operate at high field and use Nb3 Sn type superconductor. The TF coils and EF coils use NbTi superconductor. The magnet system has a large heat load from nuclear heating from DD fusion and large AC loss. This paper describes the technical requirements, the operational interface and the outline of conceptual design of the superconducting magnet system for JT-60SA.

Study of plasma termination using high-Z noble gas puffing in the JT-60U tokamak

Argon, krypton and xenon were puffed with and without simultaneous hydrogen gas puffing into Ohmically heated plasmas of the JT-60U tokamak with low plasma currents in order to study the capability of disruption mitigation. It was found that krypton gas puffing can provide a plasma termination with smaller amounts of runaway electrons in comparison to argon and xenon gas puffing.

High radiation and high density experiments in JT-60U

In order to obtain improved confinement plasmas with high radiation at high density, Ar gas was injected into ELMy H mode plasmas in JT-60U. A confinement improvement of HH98(y,2) ≈ 1 was obtained with a high radiation loss power fraction (~80%) at an electron density of ~0.65nGW. The HH factor was about 50% higher than that in plasmas without Ar injection.

Conductor Design of CS and EF Coils for JT-60SA

The conductor for central solenoid (CS) and equilibrium field (EF) coils of JT-60 Super Advanced (JT-60SA) were designed. The conductor for CS is Nb3Sn Cable-In-Conduit (CIC) conductor with JK2LB jacket. EF coil conductors are NbTi CIC conductor with SS316LN jacket. The field change rate (3.9 T/s), faster than ITER generates the large AC loss in conductor. The analyses of current sharing temperature (Tcs)margins for these coils were performed by the one-dimensional fluid analysis code with transient heat loads. The margins of these coils are 1 K for the plasma standard and disruption scenarios. The minimum Tcs margin of CS conductor is 1.2 K at plasma break down (BD). The margin is increased by decreasing the rate of initial magnetization. It is found that the disruption mainly impacts the outer low field EF coil. The disruption decreases the Tcs margin of the coil by >1 K. A coupling time constant of <100 ms, Ni plating, and a central spiral are required for NbTi conductor.

Simulation of divertor detachment characteristics in JT-60 with superconducting coils

The capability of detachment control in JT-60SC is demonstrated with a divertor code (SOLDOR/NEUT2D). Under the standard operation of the core edge density at 95% of the minor radius, n edge = 3.2 x 10 19 m 3 , the power flow from the core plasma Q c = Q i = 6 MW, and the conductance of inner and outer cryopanels C in = C out = 50 m 3 /s, the inner divertor plasma is partially detached and the outer divertor plasma attached. The inner divertor can be changed from partially detached to attached plasma by inner pump of 150 m 3 /s. The fully detached inner divertor is obtained in high density operation with n sep > 4 x 10 19 m 3 , where n sep is the density at the separatrix outer midplane, The elastic collision in a dense and cold divertor plasma (n c ≥ 5 x 10 20 m 3 . T c < 5 eV) play an important role on detachment.