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In series of experiments in 2000 and 2008, helium wall conditioning discharges by ECRF heating (ECH) were extensively studied in JT-60U with BT = 3.6 T and PECRF ⩽ 2.5 MW. Homogeneous ECH conditioning discharges were reproducibly obtained by applying the horizontal field as large as 0.2% of BT. It was found that the horizontal field is effective to extend He plasma toward high field side from the fundamental resonant surfaces of ECH. The ECH wall conditioning discharge with PECRF = 2.5 MW and tpulse = 1.5 s was applied after the plasma disruption with W = 3.3 MJ and the successful recovery of the current ramp-up with 0.5 V/m of ohmic electric field was obtained.

The neutron penetration and the activation characteristics of the boron-doped low activation concrete were investigated for irradiation of 2.45 and 14 MeV neutrons. The shielding property of the 2 wt% boron-doped low activation concrete is superior to that of the 1 wt% boron for the thermal neutron, on the contrary to no clear difference for the fast neutron. The total activity detected in the boron-doped low activation concrete was about one hundredth of that in the geostandard sample at more than 30 days cooling time. The total activity of the boron-doped concrete by major nuclei does not depend on the boron density for the 14 MeV neutron irradiation.

Japan Atomic Energy Research Institute (JAERI) plans to construct a tokamak fusion device called JT-60SC in which superconducting magnet systems will be used. The purpose in this paper is to qualify type 316LN stainless steel (SS316LN) for use as the conduit material of the Nb3Sn cable-in-conduit conductor for the central solenoid (CS) of JT-60SC. Tensile properties, fracture toughness and fatigue crack growth rates of the as-welded metal and the aged (923 K × 240 h) one of SS316LN were evaluated at 4 K. The tensile properties and fatigue crack growth rates were adequate to ensure the design requirements for JT-60SC. However, fracture toughness of the aged weld metal could not be validated due to unstable crack growth. It was concluded that improvement of fracture toughness after aging was required to ensure the structural integrity of the CS conduit.

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.

In ELMy H-mode plasmas of Japan Atomic Energy Research Institute Tokamak-60 Upgrade, extension of a high confinement regime toward high density and high radiation-loss-power fraction has been investigated by Ar injection. With injecting Ar and putting the outer strike point on the top of divertor dome, confinement improvement, electron density, and radiation-loss-power fraction reached the ITER relevant regime; the thermal energy confinement time was as high as that the International Tokamak Experimental Reactor ELMy H-mode scaling IPB98(y,2), the electron density is ∼80% of the Greenwald density, and the radiation-loss-power fraction is ∼0.8. Although electron density profile is slightly peaked with Ar injection, the confinement improvement is mainly attributed to the improvement of ion transport. Moreover, the maximum divertor heat load due to ELMs has been reduced by a factor of 3–5 with keeping high confinement. ELM activity can be controlled by injecting Ar and changing the strike point position.

Two-dimensional (2D) reconstruction software was developed for bolometer signals of the JT-60U tokamak. Radiative divertor plasma with emissivity of 10 MW m −3 was found previously to extend 10 cm from the target plates. When the density is increased further by larger gas puffing, the emissivity imbalance between both divertor legs becomes less significant. The radiative zones move upward along the divertor legs and eventually merge into a single bright peak near the X-point at density around 6×10 19 m −3 called X-point multifaceted asymmetric radiation from edge. Upward movement of the radiation peak during detachment is sometimes observed directly by the in-vessel bolometer. Total radiated power at the divertor saturates at a certain level while the core radiation continues to increase with the density. Although the 2D profile of the divertor radiation undergoes dramatic changes near the saturation, the energy confinement appears to degrade with density increase irrespective of these changes.

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.

The divertor geometry of JT-60U was modified in 1998 to add a pumping slot near the outer strike point for improved pumping. In the both-side pumping divertor, the x-point MARFE onset density is ∼10% higher with pumping than without pumping. Radiated power near the x-point is decreased by pumping, and the divertor pumping is effective in suppression of the x-point MARFE. The effect of outer pumping is investigated in L-mode plasmas. The onset densities of the x-point MARFE are the same in both divertors. The fraction of total radiated power over absorbed power at the x-point MARFE onset is ∼20% higher in the both-side pumping divertor than in the inside pumping divertor, but it does not depend on the onset density and on a distance between the private dome and the outer leg. Carbon penetration from the divertor region to the main plasma has a weak dependence on the gap-out in the both-side pumping divertor.

Effects of the divertor pumping on the SOL plasma flow and neutral recycling were investigated under high density and detached divertor conditions, using reciprocating Mach probes at the outer midplane and just below the divertor x-point. For the divertor pumping from inner and outer sides of the private dome (both-sides pumping), plasma flow in the outer divertor was increased compared to that for the inner-side pumping case (i.e., closing the outer pumping slot) when the plasma detachment appeared on the divertor target. Partially detached plasma was maintained at the inner and outer divertor without causing x-point MARFE, and in-out symmetrical profiles of D α brightness and ion flux were observed.

Transport coefficients of impurities in reversed shear plasmas are evaluated and compared with those in ELMy H-mode plasmas. The diffusivity is significantly reduced (∼1/10) at the internal transport barrier. An inward pinch velocity of −3 m/s is also observed at the ITB. The inward pinch velocity increases as charge number of impurity increases. Effects of Ar seeding to the main plasma are investigated in ELMy H-mode. Argon seeding maintain improved confinement HITER89=1.4, the clear edge pedestal, the radiation fraction of ∼80% and the divertor detachment up to 70% of Greenwald limit. Moreover, argon seeding reduces type I ELM frequency less than half, but the heat load of each ELM on the outer target plate does not change. Therefore, the reduction of energy losses due to ELMs probably causes the improvement of edge confinement in Ar seeded cases, though the radiated power inside the separatrix increases.

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, Tediv, 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.

Modification of the divertor to a W-shaped pumped divertor and installation of pumping has enabled demonstration of the `puff and pump' effect in injected neon, albeit the pumping speed was modest (∼14 m3/s). It was demonstrated in ELMy H-mode plasmas with PNB⩽25 MW that neon seeding and pumping was useful to control the radiation profiles in the main and divertor plasmas, while the large radiation loss from the divertor MARFE was sustained. It was suggested that the neon enrichment in the divertor plasma, which was enhanced by the main D2 puff, effectively improved the pumping efficiency of neon.

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, Zeff 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.

Performance of inner leg pumping of JT-60U W-shaped divertor is presented. Inner leg pumping works well as the dome separates inner and outer divertor channels for neutral particles in the private flux region. In ELMy H-mode with NBI heating power of 18 MW, the operational regime where inner leg pumping is effective extends to the divertor radiation range of about 60% of SOL power. The fraction of inner particle flux pumped is estimated to be about 3%. It is qualitatively shown that a combination of main gas puff and inner leg pumping reduces carbon impurities in the main plasma, suggesting the existence of SOL flow effect on impurity reduction.

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.

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.

The density limit is increased by the application of an external helical field in the JFT-2M tokamak. The effect of the magnetic stochasticity due to the external field is investigated to study the mechanism of the improved density limit related to the edge plasma behaviour. The improvement is correlated with the retardation of the increase in the plasma inductance. At the improved density limit, local radiation loss is modified by the helical field, in which that from the vicinity of separatrix X-point is remarkably reduced, while that from outboard edge is slightly increased. The formation of a positive radial electric field at the plasma edge is also observed in the presence of the helical field.

The effects of divertor biasing and edge ergodization on the divertor plasma have been investigated in the JFT-2M tokamak. Experimental results show; (1) The differential divertor biasing can change the in/out asymmetry of the divertor plasma. It especially changes the density on the ion side divertor plasma. The in/out electron pressure difference has a good correlation with the biasing current. (2) The unipolar divertor biasing can change the density profile of divertor plasma. The radial electric field and shear flow are the cause for this change. (3) The electron temperature of the divertor plasma in the H-mode with frequent ELMs induced by edge ergodization is lower than that of usual H-mode. That is due to the enhancement of the radial particle flux by frequent ELMs.

Particle and heat balance in the scrape-off layer and the divertor region were studied in the JFT-2M tokamak. Using particle and energy conservation laws, particle and heat diffusivities perpendicular to the flux surface were evaluated just outside the magnetic separatrix. It was found that the particle diffusivity decreases with increasing electron density in the scrape-off layer and decreases by a factor of 2–3 in the H-mode phase as compared with that in L-mode. The heat diffusivity has almost the same dependence on the electron density. The ratio of the heat diffusivity to the particle diffusivity is about 2.

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.

The effects of plasma edge modification on confinement have been investigated on the JFT-2M tokamak since 1985. In the course of investigations, a new enhanced confinement regime, limiter H-mode, was discovered in 1986. This paper is a review of investigations into edge recycling control methods which extend the H-mode operational regime. In addition to Ti-gettering, the following methods have been shown to reduce edge neutral buildup, to extend the operational regime and to improve n τ of H-mode discharges: (1) optimization of the gas-puff programming (decreased or terminated at the start of auxiliary heating), (2) a toroidally distributed gas-puff system, (3) edge heating by ECH, and (4) particle removal with a pump limiter.

The impurity behaviour during the ICRF heating phase was studied in the JFT-2M tokamak. In this experiment, impurity behaviour during the ICRF heating phase was studied with special attention to the boundary electron temperature. From these experimental results, it was concluded that carbon (limiter material) impurity was released by an ion sputtering accelerated by the sheath potential. Metal impurities such as iron and titanium released from the antenna and/or near the antenna region are caused by some additional mechanism by applying the RF during the ICRF heating phase.

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.

A pump limiter is applied to the JFT-2M tokamak in order to reduce edge gas buildup. About 12% of the effective loss flux is removed by the pump limiter and the fueling efficiency is improved by 30%. In the high density regime, the gas buildup is successfully reduced and edge cooling by cold gas-puffing is prevented by the pump limiter. The pump limiter also improves the density limit and the energy confinement time. Improvement in the nτ value is about 34%. This pump limiter is shown to be compatible with the limiter H-mode.

Thermal desorption spectroscopy (TDS) has been applied during discharge cleaning in the JFT-2M tokamak to measure hydrogen fluxes. The TDS carbon sample, thickness 0.13 mm, was heated to 1000°C by direct current and the temperature distribution of the sample surface measured by infrared thermography. The probe was exposed to three types of plasma: Taylor-type discharge cleaning (TDC), ECR discharge cleaning (ECR-DC), and glow discharge cleaning (GDC). The TDS spectra show peak desorption at around 800°C. The hydrogen flux, obtained by integration of the TDS spectrum, decreases exponentially in the radial direction with decay length 7.4 cm and 5.8 cm in TDC and ECR-DC, respectively. The relation between hydrogen fluxes and water vapour production was investigated. In TDC, the amount of water vapour depends more strongly on the electron temperature of the plasma than on the hydrogen flux. In ECR-DC, the production of water vapour increases approximately linearly with the hydrogen-flux. In GDC, hydrogen fluxes were measured by TDS but no water vapour could be detected in the residual gases during the discharge.