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Estimation of neutron yields produced from SS316L, CuCrZr, and graphite targets of the mainly In-vessel components of a Tokamak fusion device for MeV-energy electrons has been performed by neutronics analysis using a 3D Monte Carlo code [PHITS], including an electron-gamma shower M.C. code [EGS 5]. Neutronics analysis for Tokamak fusion devices is performed mainly with neutrons and photons using evaluated nuclear data libraries with energy up to 20 MeV energy. Furthermore, we have experience of photoneutron emission due to runaway electrons at several hydrogen discharges in Tokamak fusion device. The energy of the runaway electrons reaches at several tens of MeV energy. A 3D Monte Carlo code [PHITS + EGS5] suitable for electron and photon transport at energies greater than 20 MeV is very attractive for the photoneutron analysis of Tokamak fusion devices.

In order to improve the spatial uniformity of the negative ion beam and to produce high current negative ion beams in a large negative ion source, a magnetic field configuration is modified from an original transverse filter to a tent-shaped filter, in combination with reducing the magnetic field strength in the JT-60 negative ion source. As a result, the beam uniformity is improved from 68% to 83% over an area of the whole extraction area of 450 × 1100 mm 2 . The improvement of the beam uniformity leads to the production of 32 A H − ion beams with the whole extraction area. The obtained beam current fulfills the requirement for JT-60SA.

A simulation study of the divertor for JT-60SA is discussed in both double-null (DN) and single-null (SN) configurations. In the DN case, the transition of peak heat load on the targets and divertor plasma states from CDN (connected double-null) to DDN (disconnected double-null) are shown in the simulation. The radiative cooling power in the divertor plasma is discussed in this study. The carbon impurity generated by the sputtering on the divertor target is included in the simulation. In the gas puffing cases of fueling gas (D2) and impurity gas (Ne), the reduction of heat load is confirmed consistently with increasing the radiative cooling loss power in the diverter plasma. The loss power and the distribution of radiative cooling in the divertor plasma are studied in this paper. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Plasma control on high-βN steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-βN exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected.

Using a 20 kA Nb/sub 3/Al cable-in-conduit (CIC) conductor, a two-turn D-shaped test coil with a height of 2.0 m and a width of 1.8 m was fabricated by a react-and-wind (R&W) method. After heat treatment, the conductor was formed to the D-shaped coil. Bending strain of the conductor is /spl plusmn/0.4% at critical current I/sub c/ test section. The I/sub c/ exceeded 31.4 kA at 7.3 T, 4.2 K. In spite of the stainless steel conduit, a critical current density J/sub c/ of 89% of the strand value was attained. The total strain of the D-shaped test coil was about -0.57%, the same as the triplex-CIC sample fabricated by a W&R method, indicating that bending did not influence the I/sub c/. The I/sub c/ at 7.4 T, 5.6 K including the margin of 1 K exceeds the operational current. Therefore, it was concluded that the R&W method is quite appropriate for the toroidal field coil of JT-60SC.

Modification of JT-60 as a full superconducting tokamak (JT-60SC) is planned. The objectives of the JT-60SC programme are to establish scientific and technological bases for steady-state operation of high performance plasmas and utilization of reduced-activation materials in an economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to the DEMO reactor have been developed for the superconducting magnet technology and plasma facing components of the JT-60SC design. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFCs) of JT-60SC. The R&D to demonstrate the applicability of the Nb3Al conductor to TFCs by a react-and-wind technique has been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low ac loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the carbon fibre composite target was successfully demonstrated on an electron beam irradiation stand.

The modification of JT-60 is programmed to be a superconducting tokamak (JT-60SC) that has superconducting toroidal field (TF) and poloidal field coils. The TF coil system, which consists of 18 D-shaped coils, makes the field of 3.8 T at plasma center (R = 2.8 m). The stored magnetic energy in total is 1.7 GJ. A squared Nb/sub 3/Al cable-in-conduit (CIC) conductor using SS316LN jacket allows us to fabricate the TFC by a react-and-wind (R&W) technique because of low strain sensitivity of Nb/sub 3/Al J/sub c/ characteristic. In order to support TF coils against the centering force, the inboard side of the TF coil case is wedged. Shear panels and keys between the TF coils are placed against the overturning force induced by poloidal fields. Stress analysis of the TF coil support structure was carried out. In the case of the maximum displacement caused by overturning force, the maximum stress was lower than the allowable stress of SS316LN at 4 K. The stress analysis of insulator and conduit at the inboard side of the TF coil with zooming model was also carried out. It is clear from zooming analysis that the stresses of conduit and insulator are within the allowable stress. Regarding a progress of R&D, demonstration of R&W technique using a D-shaped double layer coil with the full-scale CIC conductor is planned in this year in order to confirm its designed performance.

This paper describes the modification program of JT-60 towards the tokamak fusion DEMO-reactor in support of ITER. The JT-60 is planned to be modified for the superconducting tokamak, JT-60SC, to establish steady-state operation at high beta and ensure plasma applicability of ferritic steel as reduced activation materials in a reactor-relevant regime with a break-even class plasma. Main parameters of the JT-60SC device are the plasma current of 4 MA, the toroidal field of 3.8 T, the major and minor radii of 2.8 and 0.85 m, respectively, the elongation and triangularity of 1.8 and 0.35, respectively, a current flattop duration of 100 s and the total heating power of 15 MW for 100 s. Nb 3 Al and Nb 3 Sn superconducting strands with a high Cu/non-Cu ratio and NbTi strands with ∼10 μm diameter of filaments have been developed for the toroidal and poloidal superconducting coils requiring compactness and low AC losses. Full-scale cable-in-conduit cables were manufactured by automatic welding of SS316 conduits and jacketing the cables. An irradiation test for a mockup of the cost-effectively designed CFC divertor target plate has demonstrated the required high heat handling capability of ∼10 MW/m 2 on the JAERI Electron Beam Irradiation Stand.

The radiation shielding, nuclear heating and dose rate have been evaluated for JT-60 superconducting modification. The double-walled structure of a vacuum vessel was adopted so that the maximum nuclear heating at the inboard side of the superconducting TF coil could be suppressed to be lower than 2.5 mW/cm3. The 316 stainless steel (SS316L) boards outside the vacuum vessel are installed for effective gamma-ray shielding. The dose rate with a reduced activation ferritic steel pedestal for the first wall in the vacuum vessel was estimated. This pedestal reduced the in-vessel dose rate by 40% compared to SS316L pedestal. The dose rate outside the boron carbide doped concrete cryostat was estimated at shutdown after 10 years operation.

A flat carbon fiber composite (CFC) tile mock-up with screw tubes, which have helical fins like a nut, was fabricated aiming at further improvement of the heat removal performance of the cost-effectively manufactured divert or target for JT-60SC (modified JT-60 as a superconducting coil tokamak). The heat removal performance of the mock-up was successfully demonstrated on the JAERI Electron Beam Irradiation Stand. The estimated heat transfer coefficient of the screw tube at the non-boiling region was roughly three times higher than that of the smooth tube. This corresponds to 1.5 times that of the swirl tube. A heat cycle test of 10 MW/m2 showed that the mock-up with the screw tubes could withstand for 1400-cycles. These results indicate that the divertor target plate with the flat CFC tile and the screw tube can be a promising candidate for the JT-60SC divertor target.

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 x 10{sup 21} keVs/m{sup 3}, in the world on JT-60U.

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has b...

JT-60 is planned to be modified to a fully superconducting coil tokamak (JT-60SC) to establish scientific and technological bases for an attractive DEMO reactor. A divertor geometry for JT-60SC is designed to achieve enough exhaust capability for high triangularity configuration, which is essential to obtain high plasma pressure and type II ELMs. For the improvement of detachment control capability, vertical targets are applied with baffle plates into which 3?cm SOL flux surface on the outer midplane can be led. Cryopanels are installed under both a private dome and an outer divertor to allow separated pressure control for inner and outer divertor legs, which is important for the detachment control. The preliminary results of the numerical simulation with SOLDOR/NEUT2D code show that the pumping speed of ?60?m3?s-1 at an outer pumping slot would be enough to control the detachment at the outer divertor without an X-point MARFE under the condition of the edge density of 3.2?1019?m-3 and the SOL power flow of qi = qe = 6?MW.

Two types of cable-in-conduit conductors (CICCs) for the JT-60 toroidal field coil (TFC) have been developed. The operational conditions of this conductor are 7.4 T in a maximum magnetic field and 19.4 kA in operational current. Each conductor consists of a stainless steel conduit (17 mm in inner diameter) containing 216 high performance superconducting strand (Nb/sub 3/Al or Nb/sub 3/Sn) and 108 copper wires. The diameter of strand is 0.74 mm including 2-/spl mu/m Cr plating. In order to attain a compact and stable coil, strands with high Cu/non-Cu ratio and high critical current density (Jc) were developed. The Cu/non-Cu ratio of Nb/sub 3/Al and Nb/sub 3/Sn strands are 4.1 and 3.6, respectively. The noncopper J/sub c/ of developed Nb/sub 3/Al and Nb/sub 3/Sn strands were 1914 A/mm/sup 2/ and 1843 A/mm/sup 2/ at 7.4 T, 4.2 K, respectively. Measurements of critical current (I/sub c/) of full size conductors showed that the I/sub c/ of Nb/sub 3/Al and Nb/sub 3/Sn CICCs were 98% and 79% of I/sub c/ estimated from I/sub c/ in strands. These values exceed the target values indicating that the developed conductors are quite appropriate for TFC of JT-60.

The full-size samples of 20-kA Nb/sub 3/Sn and NbTi cable-in-conduit conductors for the superconducting poloidal field coils of the fusion experimental tokamak JT-60 were manufactured. Since these coils operate in pulse mode, reduction of ac losses is one of the key issues. For the Nb/sub 3/Sn conductor, in order to separate the strands sintered with Cr coat after heat treatment, bending strains were applied on it. As a result, ac loss was effectively reduced, and it was found that 0.2% bending strain is enough to realize the design value level of 60 ms. Its critical current was also measured, and it was estimated that the compression strain due to the difference between thermal contractions of the stranded wire and stainless steel jacket was -0.6 %. For the NbTi conductor, comparison between the ac losses of the SnAg- and Cr-coated conductors was carried out. The measured coupling time constant of the Cr-coated conductor satisfied the design requirement of 50 ms, however the SnAg-coated conductor was more than five times that of the Cr-coated one.

At JAERI a plan of remodeling JT-60 tokamak fusion device by using superconducting coil system is being considered to investigate performance of break-even plasma in long pulse operation. The coil system are composed of Toroidal Field (TF), Central Solenoid (CS) and Equilibrium Field (EF) coils. The TF coil is designed to have a maximum field of 7.4 T at winding with a nominal current of 19.4 kA and a magnetic stored energy of 1.7 GJ which is larger than that of every superconducting coil constructed so far. The coil uses a cable-in-conduit conductor that a Nb/sub 3/Al (or Nb/sub 3/Sn) strand cable is inserted into a square stainless steel conduit with a circular hole and is wound with a react-and-wind technique. The design operating temperature is 4.6 K. The CS which consists of 4 modules is operated with a maximum field of 7.4 T and a maximum ramp up rate of 1.0 T/s. For the conductor, a square Nb/sub 3/Sn cable-in-conduit conductor with stainless steel conduit is used. The EF coil consists of 6 coils that the maximum size coil has an outer diameter of 11 in. The EF4 coil which has a maximum field of 7.4 T and a ramp up rate of 2.2 T/s uses the same conductor as CS and other field coils use a NbTi cable-in-conduit conductor.

The 500 keV negative-ion based neutral beam injector for JT-60U has been operated since 1996 and has injected so far a neutral beam power of 5.2 MW at 350 keV for 0.7 s. In order to further increase the beam power and extend the beam pulse duration time, a few issues concerning the negative ion source have to be solved. The most serious issue of them is non-uniformity of source plasma in the arc chamber. Two out of some countermeasures for improving the non-uniformity have been tried. The first is changing the spatial distribution of the arc discharge through regulating arc current resistors which connect in series to each of eight filament groups. The second is masking both ends of the plasma grid with thin molybdenum plates to block beam extraction at the edges. Owing to these countermeasures, the acceleration efficiency of the deuterium negative ion beams in the accelerator has improved by more than 30% from 0.55 to 0.74.

Some preparations for pulse operation tests of the ITER CS model coil in the JT-60 power supply system have been performed. Since the ITER CS model coil takes a longer current ramp-up time, the pulse duration is extended up to 70 s from 15 s by decreasing the output voltage of the motor-generator. In the protection system a new grounding resistor with higher resistance has been installed and a new protection make-switch has also been developed. An analysis for the local resonances in the coil has been performed, and measurements of the coil impedance is planned to confirm the validity of the analysis results.

The JT-60 project concentrates research efforts to contribute to the ITER EDA and to establish the basis of a steady-state tokamak reactor. Negative shear, as well as high-βp-mode, are the major operation schemes in the recent JT-60 experiments. Equivalent-QDT=1.05 with a negative shear discharge and the fusion triple product nD·τE·Ti=1.5×1021 m−3·s·keV with a high-βp H-mode have been achieved. A high recycling and strong radiative boundary plasma has been obtained in negative shear high confinement discharges. Long sustainment of high performance discharges has satisfied the conditions required for a steady-state tokamak reactor concept such as SSTR. JT-60 started operation in 1985 and had a significant modification to JT-60U in 1989–1990. The most frequent machine troubles occurred due to control software bugs, air-leaks of the vacuum vessel, and electric insulation breaks in the tokamak machine. Machine behaviors beyond our expectations have been the strong membrane vibrations of the vacuum vessel in the original machine JT-60, the cooling water leaks from the TF coils, divertor tile breaks due to halo current, and the cooling pipe damages due to high energy electrons penetrating under the divertor tiles. These should be considered the important engineering issues even with regard to future fusion research.

As an R&D task of the ITER engineering design activities (EDA), we carried out the irradiation tests on the basic materials of transmission components such as ceramics, windows, fiber optics and mirrors, and in-vessel diagnostics sensors such as magnetic probes and bolometers. Ceramics will be used as general insulating materials. Here we investigated the radiation induced conductivity (RIC) of ceramics for 14 MeV neutrons. Optical elements are the most sensitive to radiation damages among the transmission components. Radiation induce transmission loss of window materials and fiber optics have been measured in the Japan material testing reactor (JMTR). Off line irradiation tests were carried out for molybdenum corner cube reflector (CCR) and gold coated mirrors in JMTR also. Fourteen MeV-neutron-induce luminescence of window material was evaluated in the fusion neutronics source (FNS). Bolometers should be installed inside the vacuum vessel to measure the radiation losses in the infra red to soft X-ray range. We demonstrated the performance of the JT 60 type bolometer under 60Co gamma-rays.

Ripple enhanced banana drift loss of fast ions is studied on the basis of heat load measurements at the first wall near the outer equatorial plane in JT-60 U. Hot spots in the ion cyclotron range of frequency (ICRF) and neutral beam injection (NBI) heating are observed at the outboard wall by using an infrared camera. In ICRF heating, the total heat load to the wall is around 5% of the injected ICRF power when the plasma-wall clearance gap is less than 0.1 m. The heat flux on the wall increases with the ICRF power and decreases strongly with the plasma density, which is consistent with Fokker-Planck calculations. In NBI heating, hot spots are observed at almost the same positions, and the heat flux and position agree with the banana drift loss as predicted by an orbit following Monte Carlo code

Damage to a TiC coated Mo divertor plate used in JT-60 for about two years was re-examined from a material viewpoint to obtain data concerning utilization of high- Z plasma facing components in future fusion reactors. Serious damage was observed on the edge of several divertor plates located next to a toroidal gap. Due to very high heat fluxes a apart of the edge was melted and a large crack of about 12 cm wide and 2.5 cm deep was caused near the melted edge along a lamination of the Mo plate. Recrystallization by the heat loading embrittled the subsurface region and resulted in cracking and melting at a lower heat loading. The present work suggests that the maximum temperature of the powder metallurgical Mo should be maintained below the recrystallization temperature to prevent material degradation and plasma radiation losses.

The recent experimental results on H-mode in JT-60U and JFT-2M are reported. The enhanced confinement mode with the edge transport barrier ( H-mode) in JT-60U is characterized by the high edge ion temperature. The confinement improvement is well correlated with the edge ion temperature. The H-mode at low magnetic field ( approximately 2 T) in JT-60U has similar features to the H-mode in JFT-2M: rapid reduction of Dalpha emission. However, in the H-mode at high magnetic field ( approximately 4 T) and at low density, no clear transition is observed in JT-60U, although the confinement is still improved with the edge transport barrier. Collisionality and the width of the transport barrier are also investigated in both tokamaks, and are compared with Shaing's theory (1989). Improvement of the high-pp mode discharge with the H-transition was demonstrated in JT-60U. Owing to this improvement, the record value of the fusion triple product (1.1*1021 m-3 s keV) has been achieved. Quasi-steady H-mode operation with H-factor of approximately 1.8 has been also demonstrated in JT-60U by controlling the repetition rate of edge localized modes (ELMs). In JFT-2M, ELM activity has been successfully induced by a new and active method using helical magnetic field perturbation, and as a result, a steady H-mode has been achieved.

A review of the results from JT-60 ICRF heating experiments is presented. The JT-60 ICRF experiments (120 and 131 MHz) are characterized by higher-harmonic heating with phase control of the compact 2*2 loop antenna array. Second-harmonic heating is most intensively investigated with antenna phase control for various cases where the resonant species (hydrogen) are either a majority component or a minority component, and with or without hydrogen HBI heating. It is found that antenna phase control plays a key role in optimizing antenna power injection characteristics as well as heating efficiency. The out-of-phase mode (( pi ,0) mode) can avoid RF sheath and parametric decay instabilities which cause harmful effects on antenna power injection capability with the in-phase mode ((0,0 ) mode) is some operating conditions. The incremental energy confinement time ( tau Einc) of the ( pi ,0) mode is about 50% better than that of the (0,0) mode. Minority hydrogen second-harmonic heating with ( pi ,0) phasing in helium discharges shows the best results over a wide range of electron densities (ne=2.5-6.5*1019 m-3) and plasma currents (Ip=1.5-2.4 MA (qa=4.2-2.6)). The best confinement enhancement factor ( tau E/ tau E(L-mode)) is 1.3, which is obtained with ohmic target plasmas. Higher-harmonic beam ion acceleration is observed up to fourth harmonics in a combined ICRF (( pi ,0) mode) and NBI heating scenario. Strong central-electron heating associated with beam ion acceleration is observed in combined third-harmonic ICRF and NBI heating.

In JT-60, hot ion modes (Ti(0) = 9-13 keV, H0 beam into H+ plasma) exhibiting peaked profiles of the ion temperature, Ti(r), and the toroidal rotation velocity, Vt(r) have been obtained with an effective safety factor qeff ~ 10. Experiments have been performed to study the characteristics of these peaked profiles in which qeff was scanned over the range 4.5 < qeff < 15. Peaking of the profiles of Ti(r) and Vt(r) occurs within a limited region. Moreover, the boundary of the region of improved performance is well characterized by the q = 1, 2 and 3 surfaces. The improvement in global confinement in JT-60 hot ion modes is attributed to the substantial increase in the volume of the enhanced confinement region within the q = 3 surface, in which region the peaked Ti profile is generated

Plasma coupling tests of the RF heating system in JT-60 were successfully performed at JAERI prior to the substantial injection for the plasma heating experiment. About 24 MW LHRF at 2 GHz band and 6 MW ICRF at 120 MHz band system were installed in JT-60. The total system is remotely controlled via a mini-computer, which is linked with the central control system named “ZENKEI” and “CAMAC” systems. The RF power and phase difference can be controlled according to a set preprogram and the plant data can be collected and recorded during RF injection. These are carried out at both LHRF and ICRF injections and simultaneous NBI injection. It is shown that this RF heating system can contribute to both LHRF and ICRF heating experiments for JT-60.

The JT-60 tokamak was modified wherein the original outer-single null vacuum vessel and poloidal field coils, were exchanged with those of large dee-shaped cross sections. The basic dimensions of the new vessel allows plasmas of 6 MA, 70 to 100 m 3 volume and vertical elongations of 1.4 to 1.7. The vessel was designed along a new concept of a continuous chamber in a double skin construction. Change of the gas to deuterium from hydrogen is being prepared and will enable neutral beams of 40 MW. The first wall was designed to utilize carbon fiber composite for the divertor and isotropic graphite for the other regions. The divertor tiles are cooled by water, while the other regions are cooled by conduction to the vessel wall.

The first experiment simulating helium ash exhaust in a DT fusion reactor has been performed in the JT-60 tokamak with {ital L}-mode divertor discharges. He was injected with a 30-keV neutral beam to realize core fueling. Divertor neutral pressures of hydrogen and He increased in proportion to the cube of the electron density of the main plasma. The He enrichment factor was 0.5 at {ital {bar n}}{sub {ital e}}=6{times}10{sup 19} m{sup {minus}3} and increased in proportion to {ital n}{sub {ital e}}. In an {ital L}-mode plasma of a fusion reactor, thermalized {alpha} particles (He ash) generated in the core plasma can be easily exhausted with a pumping speed of several tens m{sup 3}/s.

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.

The density limit for a series of gas fuelled and pellet fuelled limiter discharges in JT-60 has been studied. With pellet injection into high current/low q (qcyl = 2.1–2.4) discharges, the Murakami factor reaches (10–13) × 1019 m−2·T−1. The values are factors of 1.5–2.0 higher than those for gas fuelled discharges. For pellet fuelled discharges the central density is high, whereas in the outer region (a/2 < r) the electron density is limited to the same level as that for gas fuelled discharges. The density limit is confirmed to be an edge density limit; this can be explained by the power balance in the outer region of the plasma, for which the plasma purity (Zeff), the heating power (Pabs) and the electron temperature are the key parameters. The onset of a disruptive event can occur when Prad(total) ~ 12–80% of Pabs(total), and the disruptive limit of the density can be explained by Pabs and ne2 (r = 50 cm) Zeff (r = 50 cm). The thermal stability in the edge region is weaker for lower (

The JT-60 LHRF launcher was designed with broad frequency capability (1.74 to 2.23 GHz) to inject multi-megawatt, multi-second RF waves into plasmas. The launcher is characterized by 32 long vacuum waveguides of ∼ 8 m without front windows to avoid plasma bombardment of the ceramic windows in the large tokamak. The front of the launcher is of a detachable grill so it can be changed easily to optimize the wave spectrum. Three types of grills have been developed, and one of them employs a multijunction technique to drive the plasma current efficiently. Three launchers have been installed on JT-60, and the total injection power has reached up to ∼ 10 MW. A good coupling of reflection coefficients, ∼20%, has been obtained for yarious plasma configurations. The development and operation of the LHRF launcher on JT-60 are described.

The tokamak, JT-60, was upgraded to double the plasma current to 6 MA by replacing the vacuum vessel and poloidal field coils with those of large D-shaped cross-sections. The vessel is unique since it employs a three-dimensionally curved double-skin construction. The vessel, as well as the poloidal field coil, are nestled very closely to the original toroidal field coil, and minimization of the wali thickness to below 40 mm is inherent in order to ensure that the plasma current exceeds 6 MA. In order to design a thin but strong vessel wali, detailed structural analyses were conducted, including dynamie effects, and a multi-arc-shaped vessel of double-skin construction with a midplane support was found to be the best and was adopted. The poloidal field coils support structures were used for the vessel support. Three-dimensional shell analysis results, including effects from the support condition, are described. Local stresses near the base of ports and support rods were highest, and were confirmed...

The J1-60 neutral beam system has been successfully operated for 4 years under a wide range of operation conditions: beam energy of 30-75 keV, beam pulse up to 6 s, injection power up to 26 MW with hydrogen beams. The maximum injection power of 26 MW was obtained at 73 keV with a two-stage accelerator. In a lower energy beam injection with a single-stage accelerator, the beam power at 38 keV reached 18 MW. The system could routinely inject a nominal power of 20 MW with high reliability. The beam energy could be changed during a beam pulse, e.g. from 40 keV to 70 keV for 1.5 s. Helium beams were injected with one of the beamlines for a simulation experiment of helium ash, and the injected power was 0.4 MW at 31 keV. The helium beams could deposit in the vicinity of the plasma center column. Helium gas in the beamline was evacuated by SF6 gas condensed cryo-sorption pumps whose pumping speed was about 800 m3/s.

The electron gun of the 1 MW-klystron for JT-60 was redesigned to suppress the degradation of the withstand voltage due to deterioration with age. The key aims were to reduce the concentration of the electric field near the electrodes by improving the configuration of the electrodes, and to suppress the barium evaporation from the cathode by replacing the non-metal-coated dispenser cathode (S-type) with an iridium (Ir)-coated one allowing a iower operating temperature. The maximum beam operating voltage was inereased from 84 kV to 94 kV and an RF output power of 1.4 MW was generated for 10 sec at 2.17 GHz. Preliminary results of the klystron operation in JT-60 experiments showed that DC breakdown frequency of the improved klystron was reduced to less than 10% of the previous one.

Insufficient helium exhaust can seriously degrade the performance of any fusion reactor and has recently been identified as a potentially significant issue for the International Thermonuclear Experimental Reactor (ITER). The effects of variations of the ratios DHe/χi, DHe/DD and vHe/DHe and of the sawtooth period on helium exhaust in ITER have been studied with BALDUR code transport simulations. Recent measurements of DHe/χi in TFTR, of DHe/DD in TEXTOR and of vHe/DHe in TFTR, TEXTOR and JT-60 are found to be compatible with the requirements of sustained ignition and helium exhaust for ITER. Ignition is found to be critically sensitive to the ratio vHe/DHe, particularly at the plasma edge, and to a reduced sawtooth period. These critical dependences are moderated for reduced helium recycling at the separatrix and for hollow particle diffusivity profiles.

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter × 3.0 mm long for two pellets and 4.0 mm in diameter × 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector.

The two-electron transfer reaction was applied to the detection of helium ions in the thermal and epithermal energy range in JT-60. Using this technique, the helium ion density and the energy distribution were determined. The ion density and energy distribution obtained were consistent with the results from other diagnostics and from theory. The potential of this technique for measurements of the helium ash density and the energy distribution in fusion plasmas has been demonstrated.

Operational experiences on the performance and material behavior of the JT-60 divertor plates are presented. The outside X-point operation using TiC-coated Mo (TiC/Mo) plates was performed with absorbed power up to 20 MW and duration of 1 s without degradation of the plasma performance by sweeping the separatrix to control the surface temperature below the melting point. The use of graphite divertor plates extended the absorbed power up to 25 MW with duration of 4 s in the lower X-point operation. However, the graphite was eroded and redeposited on the divertor plates and the carbon content in the plasma was increased compared to the outer X-point operation with carbon tiles. A few lower graphite tiles were seriously damaged and radioactivated. Improvements in thermal shock resistance and erosion characteristics of the graphite are necessary. High heat concentration at tile edges and disruption-induced heat loads should be avoided in the future devices.

Results are presented of the second harmonic heating experiments in the ion cyclotron range of frequencies (ICRF) on JT-60. Heating and confinement characteristics are investigated by using the hydrogen second harmonic (H-majority 2ωcH scheme and the hydrogen minority second harmonic (H-minority 2ωcH scheme (nH/nHe ≈0.1) in two toroidal phasing modes of antenna currents, in phase ((0,0)) and out of phase ((π, 0)), at PIC ≤ 3 MW, ne =(1.3−6.6) × 1019 m−3 and Ip = 1–2.4 MA. Efficient plasma heating has been demonstrated in both heating schemes. Intense non-Maxwellian ion tails are observed at low density. Sawtooth oscillations with an enhanced period in the electron temperature are found over a wide range of electron densities. The incremental energy confinement time, τEinc, for (π, 0) phasing is larger than for (0,0) phasing, irrespective of the heating scheme. In particular, H-minority 2ωcH heating with (π, 0) phasing shows excellent plasma heating with τEinc ≈ 100–120 ms, which is larger than that of neutral beam injection (NBI) heating by a factor of about two. Different dependences of τEinc on Ip are obtained in the two heating schemes. The (0,0) phasing data of τEinc in H-majority 2ωcH heating increase with Ip up to 1.9 MA, while the (π, 0) phasing data in H-minority 2ωcH heating are almost constant with Ip up to 2.4 MA. The global energy confinement shows an L-mode behaviour. The energy confinement time for H-minority 2ωcH heating at high density agrees well with the Shimomura-Odajima scaling.

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.

Improved energy confinement for pellet fuelled plasmas on JT-60 is mainly due to the peaked density and pressure profiles inside the q = 1 rational surface, where the confinement characteristics seem to be better than those in the outer (q > 1) region. In centrally fuelled pellet injection discharges, the sawtooth activity can be completely suppressed during 0.4−1 s, or the frequency of the sawteeth can be reduced by up to one order of magnitude during 0.5−1.5 s after pellet injection. For high current, low-q (Ip = 2.5−3.1 MA, q(a) < 3) discharges, reduction of the sawtooth frequency is strongly related with the enhanced confinement and the peakedness of the electron density profile. The contribution of the sawtooth activity to the global energy confinement increases systematically with decreasing safety factor at the plasma surface.

Beam acceleration by heating in the second-harmonic ion cyclotron range of frequency (ICRF) in combination with heating by hydrogen neutral beam injection (NBI) was investigated in the JT-60 tokamak. The energy spectra of the accelerated fast ions were measured by a charge exchange neutral energy analyser whose line of sight was intersected by specific beam lines in the plasma core in order to obtain the required information. The dependences of the tail ion temperature on various parameters (electron density, NBI power and the toroidal phasing of the antenna) were examined. Optimized conditions for beam acceleration were evaluated. For combined ICRF heating and NBI heating, an incremental energy confinement time of 210 ms was achieved, which was three times higher than that obtained with NBI heating alone or with ICRF heating alone. This improvement of the energy confinement during combined NBI and ICRF heating can be explained by a build-up of the fast ions accelerated by the ICRF wave. The scaling of the incremental energy confinement time during combined ICRF+NBI heating was obtained.

This paper reports that in the JT-60 neutral beam injectors (NBIs), an active method using a set of coils is adopted to shield unneutralized beams from stray magnetic fields, while the usual passive method using high-mu materials is used to shield the ion sources and neutralizers. This active shielding method is a unique solution for the unneutralized beams in the JT-60 NBI under the constraints of the limited space available and minimizing the error file induced by the shielding. A passive shielding method is permissible for the ion sources and the neutralizers because the space to be shielded is limited. The active shielding system is designed by making a one-fourth model of the magnetic system and calculating ion orbits using magnetic fields measured in the model. The shielding characteristics are checked by arrays of thermocouples buried in the beam dump where the unneutralized beams are thermalized. The thermocouple outputs are consistent with those predicted from the ion orbit calculations.

Helium ash exhaust characteristics have been studied in the JT-60 lower divertor with ohmic heating (OH) and neutral beam heating (NBH) of PNB ≦ 18 MW and Ip = 1 MA. Helium fueling was done by gas puff. The divertor neutral pressures of hydrogen (H2) and helium were measured by a residual gas mass analyzer. The helium enrichment in the divertor was also measured. In OH discharges with helium gas puff, no buildup of helium neutral pressure was observed and helium was depleted. On the other hand, in NBH discharges with helium gas puff, although enrichment of helium is not observed, the neutral pressures of H2 and helium increase in proportion to . The density dependence of the helium neutral pressures in NBH discharges is promising for efficient helium ash exhaust in fusion reactors.

This paper describes the experimental investigation of the heat load on a lower hybrid (LH) launcher in divertor plasma. It was found that a good coupling of the reflection coefficient, ρ = 15–20%, was obtained by pushing the launcher head a few mm from the first wall even though the distance δ between the separatrix and the launcher head was more than 10 cm. In this condition, the heat load on the launcher head rapidly decreased at δ > 5 cm and was estimated to be about 0.4–0.6 kW/cm 2 at δ = 10 cm, n ¯ e 13 cm −3 during only LH injection. A higher heat load was observed on the ion side of the launcher guard limiter during combined NB + LH heating at n ¯ e > 3 × 10 13 cm −3 when fast ions were observed at the plasma edge due to parametric decay instability. We discuss the implication of these results with respect to LH application on a high power heated plasma.

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.

RF heating system with total power of 24 MW fed by 1 MW klystrons at around 2 GHz have been developed and operated successfully in the JT-60 tokamak. System performance has been demonstrated in a real plasma load. A super high power system has worked stably against changeable plasma load without output circulators. A maximum power input into plasmas reached 11 MW from three array launchers of four row by eight line waveguides, which corresponded to 17 MW generator power.