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The JT-60SA superconducting tokamak is now under construction toward the first plasma in March 2019 as a joint project between the Broader Approach (BA) Satellite Tokamak Programme between Japan and Europe, and the Japanese national programme. The JT-60SA mission is to contribute to early realization of fusion energy by supporting ITER and by complementing ITER in resolving key physics and engineering issues for DEMO reactors. Before procurements of the major components, some R&Ds for key techniques were performed. By November 2014, 25 procurement arrangements (PAs) have been launched (JA: 14 PAs, EU: 11 PAs) covering 88% of the total cost of the BA Satellite Tokamak Programme, and the main components have entered the manufacturing stage. In addition, the JT-60SA tokamak assembly started since January 2013. This paper summarizes the recent progress of the JT-60SA project.

The JT-60SA divertor has to withstand a maximum heat flux of ∼15 MW/m2 for 100 s, and JAEA has developed the monoblock-type CFC divertor target for JT-60SA. To boost production yield of the targets, CFC cracking due to thermal expansion and contraction of the CuCrZr cooling tubes must be suppressed. CuCrZr properties depend strongly on the heat treatment, and the target is heat-treated during the manufacturing processes of brazing, solution treatment, and annealing. CuCrZr test specimens were heat-treated with the targets, and the strength and hardness of the specimens were almost the same for the quenching cooling rates of ∼0.65 °C/s and ∼1.5 °C/s around 800 °C. This indicates that the slower quenching cooling rate is acceptable, and consequently at least half of the mock-ups could remove the required high heat flux of 15 MW/m2.

Construction has just begun in the JT-60 Super Advanced (JT-60SA) tokamak project, which is part of the Japanese domestic program and the Japan – EU international program “ITER Broader Approach”. All plasma facing components (PFCs) will be actively cooled because they are exposed to high heat flux for long duration. A lower single-null closed divertor with a vertical target (VT) will be installed when experiments begin. A divertor cassette integrated with coolant pipe connections will be used for remote handling maintenance. Each cassette covers a 10° sector in the toroidal direction. Modularized PFCs such as VTs, baffles, and domes will be assembled on divertor cassettes. Static structural analysis shows that the displacement and stress in the divertor module are generally small for determining the dead weight, coolant pressure, and electromagnetic forces (EMFs). However, the support structure of the outer baffle requires improvement because eddy currents generate concentrated overturning EMFs.

Modification of JT-60 to a fully superconducting coil tokamak (JT-60 Super Advanced) has been started last year. Divertor plasma performance for high triangularity and low aspect ratio plasma in two types of divertor geometries was studied and compared by using 2D plasma fluid (SOLDOR) and neutral Monte-Carlo (NEUT2D) codes. ‘Vertical target with deep V corner’ is effective to reduce heat flux even in low density operation without additional gas puffing, but large pumping speed is required to change particle balance. On the other hand, the particle balance in ‘W-shaped with shallow V’ can be changed with small pumping speed and larger main plasma can be obtained due to its compactness. Performance of water-cooled mono-block carbon fiber composite (CFC) divertor target has been confirmed by heat load cycle test of small mockup. Obvious performance degradation was not found during 1450 cycles at 15 MW m−2 and 600 cycles at 20 MW m−2.

The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program “Broader Approach”. The maximum heat flux to JT-60SA divertor is estimated to ∼15 MW/m 2 for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R&D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R&D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace at the same time is also produced, and the half of the mock-ups could remove 15 MW/m 2 as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

Volume recombination of C4+ and e? into C3+ is observed for the first time in detached divertor plasmas with an X-point MARFE. The recombination zone is located around the X-point, and the electron temperature and density are evaluated to be 6.3?eV and 7.8 ? 1020?m?3, respectively. In this zone, the volume recombination flux is larger by two orders of magnitude than the ionization flux of C3+. However, the radiation power due to the recombination process is only 2% of the total radiation power, measured by a bolometer. In contrast, the radiation power due to the excitation process from the ground state of C3+ by electron collision dominates the total radiation power.

In JT-60U, Xe gas has been injected into reversed shear plasmas with internal transport barriers. During these discharges, spectra from highly ionized Xe atoms have been observed in a wavelength range of 4–16 nm. The observed spectra have been analyzed by calculating theoretical spectra for Xe XXXVII–XLV 3s–3p and 3p–3d transitions with relativistic atomic structure codes, a collisional radiative model, and an impurity transport code. Several lines for Xe XLI and XLII have been newly identified. Soft X-ray emission profiles have also been analyzed, and the results showed that the Xe ions accumulated inside the internal transport barrier.

Modification for long pulse operation up to 65 s with a neutral beam heating power of 12 MW (30 s) has provided an opportunity for studies on plasma-wall interaction on a time scale of tens of seconds. During a long pulse, ELMy H-mode discharge, the net wall-pumping rate, evaluated by a particle balance analysis, gradually decreases, and subsequently, becomes zero in ~20 s after the neutral beam injection starts. This situation is interpreted as 'wall saturation'. Wall saturation is observed in repetitive long pulse discharges. Although the outgas due to the increase in the divertor plate temperature and increase in the static retention with pulses also contribute to the particle balance, the dynamic retention process dominates the particle balance and determines the wall saturation. Under the condition of wall saturation, the plasma density continuously increases due to insufficient divertor-pumping rate. This density rise results in detachment of the outer divertor plasma, followed by an X-point MARFE. Results from the calculation of a 2-dimensional plasma transport code indicate that higher divertor-pumping efficiency by a factor of 1.5 is required to avoid the detachment. Note that no negative effects on impurity behaviour such as carbon bloom, or significant dilution of the core plasma has been observed even in the case of the total input energy reaching 350 MJ.

Characterization of Tokamak exhaust gas, gas species and their concentrations, is important for the design of the tritium fuel processing system. The exhaust gas from JT-60U during the experimental campaign has been investigated with micro gas chromatography (MGC). Gas species, such as hydrogen isotopes, CH 4 , C 2 H 2 + C 2 H 4 , C 2 H 6 and CO 2 , in the exhaust gas from a plasma discharge could be identified successfully. In this campaign, the total ratio of hydrogen isotopes and impurity species in exhaust gas was estimated to be 39 to 1. From the comparison between both amount of gas species and some plasma parameters, there was a tendency observed indicating that the amount of hydrogen isotopes and carbon compounds increased with maximum electron density. This relation suggests that the amount of hydrogen isotopes and carbon compounds in the exhaust gas is increases with the extension of plasma performance.

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high βp H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operating regime is extended to high density above the Greenwald density (nGW) with high confinement (HHy2 > 1) and high radiation loss fraction (frad > 0.9) by tailoring the internal transport barriers (ITBs). With a small plasma-wall gap, the radiation loss in the main plasma (inside the magnetic separatrix) reaches 80% of the heating power due to metal impurity accumulation. However, high confinement of HHy2 = 1.2 is sustained even with such a large radiation loss in the main plasma. By neon seeding, the divertor radiation loss is enhanced from 20% to 40% of the total radiation loss. In the high βp H-mode plasmas, high confinement (HHy2 = 0.96) is maintained at high density ( ) with high radiation loss fraction (frad ~ 1) by utilizing high-field-side pellets and argon (Ar) injection. The high is attributed to the formation of strong density ITB. Strong core-edge parameter linkage for confinement improvement is observed, where the pedestal pressure and the core plasma confinement increase together. The measured radiation profile including contributions from all impurities in the main plasma is peaked, and the central radiation is ascribed to the contribution from Ar accumulated inside the ITB. Impurity transport analyses indicate that the Ar density profile, twice as peaked as the electron density profile, which is the same level as that observed in the high βp H-mode plasma, can yield an acceptable radiation profile even with a peaked density profile in a fusion reactor.

Hydrogen particle behavior, especially H 2 molecule behavior, has been studied spectroscopically in attached and detached divertor plasmas of JT-60U. The decay lengths of the H 2 Fulcher line intensity in attached and detached divertor plasmas were roughly 1 cm and 4 cm, respectively. The fall in intensity of the H 2 Fulcher lines with distance from the divertor plates was reproduced by calculation using a neutral transport and a collisional radiative model code. Molecular assisted recombination was estimated to be as important as H + –e recombination in a detached divertor plasma.

In a collaboration of eight experiments the dependence of the chemical erosion yield of carbon on the ion flux, Φ , was established to Φ −0.54 at high ion fluxes. With this flux dependence a comprehensive description for chemical erosion is available as function of energy, temperature and flux. With this description the erosion and re-deposition of carbon in the ITER divertor can be calculated using the ITER steady-state plasma scenario and the ERO code. The resulting gross and net erosion rates during steady-state phase of discharges are compared to previous estimates using a constant erosion yield of 1.5%. The use of the complete parameter dependence results in an order of magnitude lower erosion, and, accordingly, of the T codeposition inventory.

The chemical erosion of carbon in interaction with a hydrogen plasma has been studied in detail in ion beam experiments, and erosion yield values are available as a function of ion energy and surface temperature. However, the conditions in the ITER divertor cannot be simulated by ion beam experiments, especially as far as ion flux is concerned.Therefore, a joint attempt was made through the EU Task Force on plasma-wall interaction and the international tokamak physics activity involving seven different fusion devices and plasma simulators to clarify the flux dependence. For each data point the local plasma conditions were normalized to an impact energy of 30 eV, care was taken to select data for a surface temperature close to the maximum yield or room temperature and the calibration of the diagnostic was performed in situ. Through this procedure the previous large scatter was significantly reduced, revealing a clear trend for a decreasing yield with increasing ion flux, Phi. After the attribution of an error to each data point a fit using Bayesian probability analysis was performed, yielding a decrease in the erosion yield with Phi(-0.54) at high ion fluxes.

The measurements of the SOL flow and plasma profiles both at the high-field-side (HFS) and low-field-side (LFS), for the first time, identified the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the HFS and LFS separatrix of the main plasma for the ion ∇B drift direction towards the divertor. 'Flow reversal' at the main SOL was reproduced numerically using the UEDGE code with the plasma drifts included although Mach numbers in measurements were greater than those obtained numerically. Particle fluxes towards the HFS and LFS divertors produced by the parallel SOL flow and Er × B drift flow were evaluated from the measured profiles of Mach numbers, the density and the radial electric field. The drift flux in the private flux region was also evaluated, and it was found that its contribution to the HFS-enhanced asymmetry of the divertor particle flux was larger than the ion flux from the HFS SOL. The ion flux for the intense gas puff and divertor pump ('puff and pump') was investigated, and it was found that both the Mach number and density were enhanced, in particular, at the HFS. Ion flux at the HFS SOL can be enhanced so as to become larger than the drift flux in the private flux region.

As part of study to establish in-vessel tritium reduction methods in carbon plasma facing fusion devices such as ITER, measurement of tritium removal characteristics of glow discharge methods, usually used for wall conditioning, have been applied to and examined in the vacuum vessel of JT-60U. Release rates of hydrogen isotopes (tritium and deuterium) as well as hydrocarbons from the JT-60U vacuum vessel induced by glow discharge cleaning (GDC) with He and H2 were measured. Release characteristics of hydrogen isotopes could be classified into three release processes, and each process was well described by a simple exponential decrease with time. It was found that H2 GDC showed the superior hydrogen isotope release characteristics to the He GDC, probably because of chemical processes, such as isotope exchanges assisted by the chemical sputtering process between discharged hydrogen and hydrogen isotopes in the plasma facing carbon tiles were enhanced by the H2 glow discharge. Based on the release kinetics observed in the present work, it is estimated that it takes several days to reduce tritium (deuterium) inventory within the removal depth by the H2 GDC in JT-60U to a half by continuous H2 GDC at 573 K.

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high βp mode plasmas of JT-60U. The electron effective diffusivity is well correlated with the ion thermal diffusivity in the ITB region. The ratio of particle flux to electron heat flux, calculated on the basis of the linear stability analysis, shows a similar tendency to an experiment in the RS plasma with a strong ITB. However, the calculated ratio of ion anomalous heat flux to electron heat flux is smaller than the experiment in the ITB region. Helium and carbon are not accumulated inside the ITB even with ion heat transport close to a neoclassical level, but argon is accumulated. The helium diffusivity (DHe) and the ion thermal diffusivity (χi) are 5–15 times higher than the neoclassical level in the high βp mode plasma. In the RS plasma, DHe is reduced from 6–7 times to a 1.4–2 times higher level than the neoclassical level when χi is reduced from 7–18 times to a 1.2–2.6 times higher level than the neoclassical level. The carbon and argon diffusivities estimated assuming the neoclassical inward convection velocity are 4–5 times larger than the neoclassical value, even when χi is close to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying electron cyclotron heating (ECH) in the high βp mode plasma, where both electron and argon density profiles become flatter. The flattening of the argon density profile is consistent with the reduction of the neoclassical inward convection velocity due to the reduction of the bulk plasma density gradient. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density gradient control in suppressing impurity accumulation.

Three reciprocating Mach probes were installed at the high-field-side (HFS) baffle, low-field-side (LFS) midplane and just below the X-point in JT-60U tokamak in order to investigate the SOL flow pattern. Profiles of particle fluxes produced by parallel SOL flow and Er×B-drift flow were evaluated from Mach number, electron temperature, density and floating potential. The drift particle flux was dominant near the separatrix, while it became smaller than the parallel particle flux at the outer flux surfaces. Total particle fluxes towards HFS and LFS divertors were investigated at various plasma densities. The drift flux was comparable to the parallel flux at relatively low densities, while the parallel flux became dominant at high densities. Drift flux in the private flux region was also evaluated, and it contributed to produce in–out asymmetries in the divertor ion flux and particle recycling under attached divertor condition.

In order to extend the high-confinement reversed shear regime with an internal transport barrier (ITB) toward high radiation loss power, Ne and Ar injection have been investigated in JT-60U. With Ne injection, high confinement ( H 89PL >2.4, HH 98(y,2) >1.6) and high radiation loss power fraction (>80%) were simultaneously obtained at high density (>0.7 n GW ). The radiation loss power was enhanced both in the main plasma and in the divertor plasma. With divertor plasma detachment and an X-point MARFE, the ITB was pronounced and the H 89PL factor reached 1.8. However, Ar injection resulted in excessive radiation loss enhancement inside the ITB. The Ne and the C density profile were similar to the electron density profile, and their accumulation was not significant. On the other hand, Ar showed accumulation inside the ITB.

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.

Boronization effects have been systematically investigated using discharges with identical conditions in JT-60U. In ∼50 shots after boronization using 70 g of B 10 D 14 , the carbon content decreased due to suppression of the chemical sputtering of carbon tiles while the boron content increased. In contrast to the boron and carbon contents, the oxygen content gradually increased from ∼0.7% to ∼1.3% in ∼500 shots. Continual boronization using 10–20 g of B 10 D 14 every ∼200 shots successfully kept the oxygen content less than ∼1.0%. However, the durability of boronization using 20 g of B 10 D 14 is much shorter than that of boronization using 70 g after vacuum vessel ventilation. Due to the reduction of the carbon and oxygen contents, Z eff in 50 shots after boronization lies in the lowest range of the database.

Heat and particle control has been studied under the reactor-relevant high-power heating in the large tokamak of JT-60U with an open divertor and progressively a W-shaped pumped divertor. Heat and particle control is crucial for reduction in heat load onto the divertor plates, control of density in the main plasma, effective exhaust of helium ash, and reduction in impurity contamination. For the reduction of heat load, radiative divertor concept was developed based on understanding of heat and particle transport in scrape-off layer and divertor plasmas, which contributed to establishment of divertor concept in ITER. With argon injection, the total radiation loss power reached up to 80% of the net heating power with high confinement of HH{sub y2} {approx} 1, where HH{sub y2} is a confinement enhancement factor over the IPB98(y,2) ELMy H-mode scaling, at high density of 80% of the Greenwald density in the ELMy H-mode plasma. For the density control, the dependence of particle confinement on plasma parameters was systematically studied with two confinement times for center- and edge-fueled particles, which enabled discussion of density controllability. Core fueling using a high-field-side pellet injection extended the operation range of high confinement (HH{sub y2} {approx} 1) from 60 to 70% ofmore » the Greenwald density in the high {beta}{sub p} ELMy H-mode plasma. Efficient helium ash exhaust of {tau}{sub He}*/{tau}{sub E} = 2.8 was demonstrated in the ELMy H-mode plasma with the pumping from the private flux region, which is the same pumping geometry as that in ITER design. Reduction in Z{sub eff} by puff-and-pump scheme was demonstrated, and chemical sputtering yields were estimated with the consideration of not only methane but also heavier hydrocarbons. Their sputtering yields showed strong dependence on the wall temperature and weak dependence on the particle flux. The measured profiles of C II and C IV line intensities were well reproduced by the Monte Carlo impurity transport simulation code (IMPMC code). The estimation of sputtering yields and development of the simulation code enabled reliable predictions for impurity behavior in a fusion reactor.« less

The diagnostic system of JT-60U (JT-60upgrade) is composed of about 50 individual diagnostic devices. Recently, the detailed radial profile measurements of plasma parameters have been improved, so ...

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...

The chemical sputtering yields of CH4/CD4 and C2Hx/C2Dx have been measured at the divertor plates of JT-60U. Spectroscopic measurements for CH/CD and C2 spectral bands are applied to estimate the CH4/CD4 and the C2Hx/C2Dx flux. At the surface temperatures of 380, 440 and 560 K, the CH4 yield is, respectively, ~0.8%, 1-2% and 2-3%, the C2Hx yield 1-2%, 3-4% and 4-5%, and the total sputtering yield by hydrogen ions 3-4%, ~8% and ~10%. With increasing ion flux to the divertor plates (Γion), the sputtering yields (Y) decrease, i.e. Y∝Γion(-0.05 to -0.40). With increasing electron temperature (Te), the sputtering yields increase, i.e. Y∝Te0.5. It is concluded from the result of regression analysis of Y∝Te0.5 that the negative dependence of the yields on the ion flux is attributed to the incident ion energies to the carbon plates. The ratio of the sputtering yields by deuterium ions and hydrogen ions is estimated to be ≥1.5 based on the ion flux measurement by Hα/Dα intensity. The C2Hx/C2Dx sputtering yield accounts for ~80% of the total number of sputtered carbon atoms.

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.

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.

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.

Particle confinement and transport have been systematically analysed for improved confinement modes in JT-60U. A scaling law for the total number of ions in the main plasma of ELMy H mode plasmas was proposed for the first time, with confinement times of the particles fuelled by NBI (centre fuelling) and by recycling and gas puffing (edge fuelling) separately defined. The confinement time increases and decreases with density for centre and edge fuelling, respectively. A comparison between the experimental data and the scaling was made for high βp ELMy H mode and reversed shear plasmas. In the reversed shear plasma, particle confinement was enhanced by a factor of about 2 compared with the scaling, when the enhancement factor of the energy confinement time over the ELMy H mode scaling was larger than 1.2. Density controllability is discussed on the basis of the scaling in ELMy H mode, and it was found that additional fuelling with a confinement time of 0.1-0.4 s is necessary for density control in JT-60U. The particle diffusivity and convection velocity were evaluated by means of a perturbation technique using modulated helium gas puffing in ELMy H mode and reversed shear plasmas. The particle diffusivity and convection velocity were estimated to be 0.2-2 m2/s and from -4 (inward) to 0 m/s for the reversed shear plasma, and 0.4-2 m2/s and from -5.5 (inward) to +2.5 (outward) m/s for the ELMy H mode plasma, respectively. The particle diffusivity around the internal transport barrier was reduced by a factor of 5-6 compared with that in the inside and outside regions in the reversed shear plasma.

A radiative divertor experiment with neon gas puff was carried out with the aim of investigating impurity behaviour, after the open divertor was modified to the pumped W-shaped divertor. To evaluate neon radiation, we analysed lines from the neon ion measured with a VUV spectrometer in the divertor plasma. As a result, lines from Ne IV-Ne VIII were identified. By combining the measurement of an absolutely calibrated multichannel interference filter spectrometer with that of the VUV spectrometer, the radiation loss through neon line emission was estimated in the divertor plasma. In the case of pumping-off, the detached plasma evolved into multifaceted asymmetric radiation from the edge (MARFE). It was observed that the line intensities of highly ionized neon (Ne VII, Ne VIII) increased simultaneously with the formation of the MARFE and furthermore increased after the MARFE formation. It is considered that after the divertor plasma was detached, the plasma flow velocity and the friction force were weaker in the case of pumping-off than in the case of pumping-on and more impurities moved toward the X-point region. These observations imply impurity back flow from the divertor upstream by the reduction of friction force.

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.

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%.

The particle transport for several impurity ion species was investigated in reversed shear and ELMy H-mode plasmas on JT-60U. Particle diffusivity for neon was evaluated from gas-puffing modulation experiments; the diffusivity was smaller by one order of magnitude at the internal transport barrier in reversed shear plasmas than in ELMy H-mode plasmas. An increase of the inward pinch velocity for neon was observed at the internal transport barrier in the reversed shear plasma. The particle transport coefficients at the internal transport barrier were compared for neon, carbon and helium. The particle diffusivity was almost the same for all ions, while the inward pinch velocity was larger for ions having higher charge state.

In order to investigate the behaviour of helium atoms in the divertor region of JT-60U, the spectral profile of a He I singlet line (667.82 nm) has been observed with a high-resolution visible spectrometer. In attached divertor plasmas, the spectral profile is narrow in front of the divertor plates, because the line is mainly emitted from low-energy helium atoms desorbed from the plates. In the upper stream of the divertor plasma, as the electron density increases, the profile becomes broader and the temperature of helium atoms corresponding to the broadening increases from 0.2 to 1.5 eV. In detached divertor plasmas, the temperature is 0.9-1.5 eV. The profiles in attached divertor plasmas have been reproduced by simulation with a neutral transport code considering elastic scattering by proton collision. The broadening of the profile is attributed to the elastic scattering. The elastic scattering affects helium concentration in the main plasma and helium pumping efficiency. In the simulation for an attached divertor plasma, the penetration probability of helium atoms from the divertor plates to the main plasma and the helium atom flux to the gap for pumping increase by 30% due to the elastic scattering.

In order to understand the recycling and emission processes of deuterium atoms, spectral profiles of the line emitted from the divertor region of JT-60U have been observed with a high-resolution spectrometer and analysed by simulation with a three-dimensional neutral particle transport code. The profile has been explained as composed of narrow and broad components; the narrow component is attributed to dissociative excitation and electron collisional excitation of the atoms produced by dissociation, and the broad component is attributed to electron collisional excitation of the atoms produced by reflection and charge exchange. In low-density plasmas, the simulated line profile agrees reasonably well with that observed, although the component attributed to the atoms reflected at the divertor tiles is overestimated by a factor of about two. Dissociative excitation from deuterium molecules and molecular ions plays an important role for the line intensity. The ratio of the line intensity to the deuterium atom flux for high-energy deuterium atoms, which are produced by the reflection and charge exchange, is reduced, because the fast atoms readily escape from the divertor plasma. The width of the narrow component in a low-density case corresponds to a temperature of deuterium atoms of 1.3 eV, and that in a high-density case corresponds to a temperature of 2.2 eV.

The operation of reversed shear plasmas in JT-60U has been extended to the low-q, high-Ip region keeping a large radius transport barrier, and a high fusion performance has been achieved. Record values of deuterium-tritium (DT)-equivalent power gain in JT-60U have been obtained: QDTeq = 1.05, τE = 0.97 s, nD(0) = 4.9 × 1019 m-3 and Ti(0) = 16.5 keV. A large improvement in confinement resulted from the formation of an internal transport barrier (ITB) with a large radius, which was characterized by steep gradients in electron density, electron temperature and ion temperature just inside the position of qmin. Large negative shear regions, up to 80% of the plasma minor radius in the low-qmin regime (qmin~2), were obtained by plasma current ramp-up after the formation of the ITB with the pressure and current profiles being controlled by adjustment of plasma volume and beam power. The ITB was established by on-axis beam heating into a low density target plasma with reversed shear that was formed by current ramp-up without beam heating. The confinement time increased with the radius of the ITB and the decrease of qmin at a fixed toroidal field. High H factors, up to 3.3, were achieved with an L mode edge. The effective one fluid thermal diffusivity χeff had its minimum in the ITB. The values of H/q95 and βt increased with the decrease of q95, and the highest performance was achieved at q95 ~3.1 (2.8 MA). The performance was limited by disruptive beta collapses with βN~2 at qmin~2.

Fusion performance of reversed shear discharges with an {ital L}-mode edge has been significantly improved in a thermonuclear dominant regime with up to 2.8 MA of plasma current in the JT-60U tokamak. The core plasma energy is efficiently confined due to the existence of persistent internal transport barriers formed for both ions and electrons at a large minor radius of r/a{approximately}0.7 near the boundary of the reversed shear region. In an assumed deuterium-tritium fuel, the peak fusion amplification factor defined for transient conditions involving the dW/dt term would be in excess of unity. {copyright} {ital 1997} {ital The American Physical Society}

As a result of an experiment at JT-60U on H mode transition power threshold scaling, which is one of the urgent ITER Physics R&D issues, it has been found that: (a) the derived scaling law of Pth(MW)=0.18ne0.5 (1019 m-3) BT1.0 (T) R1.5 (m) with presumed non-dimensional constraints predicts a threshold heating power of ≤ 100 MW at 5 × 1019 m-3 for the ITER EDA design, and (b) an increase of the neutral particle density at the plasma edge results in a reduction of the edge ion collisionality just before the transition below unity and thereby an increase of the threshold heating power. It has also been found that both the low density boundary, below which the transition cannot occur, and the apparent density dependence of the threshold power scaling are related to the edge neutrals. It is suggested that a charge exchange process might well be the potential mechanism of the neutral effect

In JT-60U, a visible spectroscopic system with high wavelength resolution has been developed to observe Doppler broadening of spectral lines emitted from divertor plasmas. An echelle grating is used to obtain the high wavelength resolution and the nominal wavelength resolution is 5.3 pm. Using optical fibre arrays, the Littrow mounting and a charge-coupled device camera, it is possible to observe 10 sight lines simultaneously with a spatial resolution of 1.2 cm.

Isolation of the core with improved confinement (inside the internal transport barrier) from high recycling and strongly radiative boundary plasmas was demonstrated for the first time in the reversed shear plasmas of the JT-60U tokamak. The internal transport barrier remained intact, while edge radiation and recycling was enhanced with neon and hydrogen gas puffing. The internal transport barrier reduced the particle transport by a factor 10{endash}15, while shrinking in the time scale of resistive diffusion towards the center. {copyright} {ital 1997} {ital The American Physical Society}

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.

A two-dimensional impurity code based on Monte Carlo techniques (IMPMC) has been developed, in order to study the impurity behavior in the divertor plasma. The model includes (1) impurity generation at the divertor plate, (2) ionization of sputtered neutrals, (3) parallel motion of impurity ions along field lines, (4) Coulomb scattering, (5) cross-field diffusion, and (6) atomic processes. This model is applied to the carbon impurity behavior in a JT-60U NB heated plasma. The impurity generation mechanism, the effect of the thermal force on impurity sheilding, and the contribution of charge exchange recombination to C 3+ ion density near the plates are clarified. The measured spatial profile of the CIV line indicates that the diffusion coefficient, D ⊥ is around 1 m 2 /s. The ion temperature measured by Doppler broadening of the CIV line is found to be shifted to lower temperature side due to the overspread distribution of C 3+ ions in the SOL by the thermal force.