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負三角度トカマク(NTT)は除熱第一優先を思想とした新規核融合炉概念である。本論文では、ELMが無いか弱く、大半径部にダイバータを置いたLモード端を保持した負三角度トカマクの炉概念について述べている。また、高い閉じ込め(HH=1.5)を得るか高い磁場(B=15.5T)でNTT核融合炉の主半径は9mから7mに縮めることが可能である。現在の物理課題と技術的実現可能性を議論している。

The effects of toroidal rotation and toroidal field (TF) ripple on the edge pedestal structure and edge-localized-modes (ELMs) were examined in JT-60U. The amplitude of the TF ripple was changed by the installation of ferritic steel tiles (FSTs). The profile of toroidal rotation VT became less counter with FSTs, particularly in the case where co-NBI was used in a large volume configuration. In this case, the plasma pressure was raised across the whole profile of the plasma. At the plasma edge, higher pedestal temperature was obtained with the growth of pedestal width. However, the effect of FSTs became less significant in a small volume configuration. As the VT became less counter at the pedestal, ELM frequency fELM was reduced and ELM energy loss ΔWELM was increased at fixed power crossing the separatrix Psep. The observed larger ELM energy drop at VT in a relatively co-direction involves the ELM perturbations of the electron temperature profile across an ELM that extends radially more inward. In addition, the inter-ELM transport loss is reduced and the pedestal pressure pped is weakly raised. The effect of FSTs appeared clearly in the large volume configuration where pped is raised even at a given VT at the pedestal while pped is not changed by FSTs at the small volume configuration. This suggests that the reduction of the TF ripple, the magnitude of which can affect the VT profile, plays a role in the increasing of pped. For increased pped due to the use of co-NBI and the reduction of TF ripple, the spatial width of the pedestal ion temperature became greater.

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

The properties of a spherical conductive dust particle on a plasma-facing wall and sheath formation with the dust particle are considered using the self-consistent two-dimensional Particle-In-Cell (2D PIC) simulations. We obtained the self-consistent simulated values of the dust charge on the wall position for different radii of the dust particle and values of the wall potential and made the comparison with the 1D model. These results allow us to evaluate the one-dimensional model and clarify the effect of dust shielding by plasma.

By conducting power scan experiments in three cases of toroidal field ripple configurations (δ r ≃ 0.4, 1.0 and 2.0%) with variation of the toroidal momentum source (co, balanced, counter) relative to the direction of the plasma current, the roles of the loss of fast ions and toroidal rotation on the H-mode pedestal structure were investigated. It was found that the pedestal pressure was increased with a decrease in fast ion loss power. On the other hand, the variation in toroidal rotation did not affect strongly the pedestal pressure, although the loss of fast ions due to existing large toroidal field ripple forced the toroidal rotation basically to the counter direction in JT-60U. Due to the enhanced counter toroidal rotation induced by the loss of fast ions, similar V tor profiles were obtained for discharges with different toroidal momentum source and loss of fast ions. When the loss of fast ions became smaller, pedestal density was raised remarkably while there was no significant difference in the temperature profiles.

Effects of plasma rotation and losses of fast ions on Type-I ELM characteristics have been systematically studied in the JT-60U tokamak, scanning combinations of NBI (tangential co-, balanced- and counter-NBI plus perpendicular NBI) in the three types of plasma configurations (corresponding toroidal field ripple at the plasma edge, δ r ∼0.4, 1.0 and 2.0%). New findings on the Type-I ELM characteristics are as follows: smaller ELM energy loss normalized by pedestal stored energy, ΔW ELM /W ped , and faster ELM frequency, f ELM , are confirmed in the counter-NBI in comparison with the co-NBI discharges. Nevertheless, the power loss due to ELM, P ELM (= ΔW ELM × f ELM ), normalized by heating power crossing the separatrix, P SEP , is constant regardless of the direction of the momentum injection at each plasma configuration. In contrast, the P ELM /P SEP decreases with increasing δ r . The relationship between ELMs, rotation and the losses of fast ions are discussed.

Transient transport experiments are performed in plasmas with and without internal transport barriers (ITB) on LHD and JT-60U. The dependence of χe on the electron temperature, Te, and on the electron temperature gradient, ∇Te, is analysed with an empirical non-linear heat transport model. In plasmas without an ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation: the χe depends on Te and ∇Te in JT-60U, while the ∇Te dependence is weak in LHD. Inside the ITB region, there is none or weak ∇Te dependence both in LHD and JT-60U. Growth of the cold pulse driven by the negative Te dependence of χe is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U).

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.

The dependence of the thermal diffusivity on the heat flux has been investigated in JT-60U plasmas by varying the neutral beam power. In positive magnetic shear (PS) plasmas, the ion thermal diffusivity (χi) in the core region generally increases with the heating power, as with the L mode at low heating powers. However, as a result of the intensive central heating, a weak internal transport barrier (ITB) is formed, and the χi value in the core region starts to decrease. Corresponding to a further increase in the heating power, a strong ITB is formed and the χi value is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of the χi value is observed in any of the heating regimes. The electron thermal diffusivity (χe) is correlated with the χi value in PS and RS plasmas. Furthermore, the dependence of the ion thermal diffusivity on the radial electric field (Er) shear has also been investigated. By considering the non-locality in the relation between the Er shear and the χi value, a threshold in the effective Er shear to change the state from a weak to a strong ITB is identified.

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H × βN / q 295, is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

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.

A substantial reduction of the L-H transition threshold power was observed under the W-shaped divertor in JT-60U, in comparison with the open divertor. Radiation power in the main region was also decreased in this case. These differences of radiation power from the main plasma between open and W-shaped divertor were not large enough to account for the apparent reduction of threshold power. In this paper, the emphasis is on the edge plasma parameters which relate to the L-H transition. It was found that edge plasma pressure just before the L-H transition in the W-shaped divertor case became smaller than that in the open divertor case. After the modification of divertor geometry, edge ion collisionality just before the L-H transition also became lower and it was established with lower auxiliary power input. It was suggested that reduction of edge ion collisionality arose from neutral particles near the X-point by the analysis of poloidal profile of neutral density. Therefore, neutral particles near the X-point prevented the threshold power for the L-H transition from more reduction at low plasma density, and caused the shift of plasma density which gave minimum threshold power.

In this paper, we compare various scalings for the H-mode edge pedestal width proposed so far empirically or based on theoretical or semi-empirical models, and examine possible inter-relations between them. Many of the scalings include poloidal or toroidal Larmor radius, beta-poloidal, normalized Greenwald density and machine size. Some of the scalings include explicitly the local magnetic shear, which is expected to play an essential role in determining the width of the transport barrier through MHD stability and turbulence suppression. Emphasis is placed on the comparison between those scalings without magnetic shear and those with shear. By considering the safety factor dependence of the magnetic shear and its spatial profile, some of the scalings without shear are shown to have a close relation with the scaling including shear.

Edge pedestal parameters (pedestal temperature and density, pedestal width) for JT-60U and DIII-D are studied and compared. For the discharges used in the study, the JT-60U H-mode operation space is in the high edge ion temperature and low edge electron density region, whereas DIII-D discharges have higher edge electron density and lower electron temperature. The pedestal width scaling has been studied separately in each machine. We tested these pedestal scalings against each other using JT-60U ion temperature pedestal width data and DIII-D electron temperature pedestal width data. Three previous scalings for pedestal width (Δe0.5ρpi, Δ/R(ρpi/R)0.66, Δ/R(βpPED)0.4) are tested using temperature pedestal widths from both machines. The relations between pedestal width and dimensionless parameters are examined. We propose a new pedestal width scaling for Te and Ti based on this study of the two machines. The new scaling includes normalized poloidal gyroradius and Greenwald density: Δaρ*0.4nG*0.3κ-1.5. The result of the comparison of these scalings is that the new scaling and Δ/R(βpPED)0.4 are well fitted and the fitting errors are almost the same as experimental error. However, further study is necessary for scaling parameters.

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

Results of an aerodynamic design study for the multistage axial helium compressor of a 300 MWe class nuclear gas turbine are presented. Helium compressor aerodynamics is challenged by the characteristically narrow and numerous-stage flow path, which enhances loss effects of blade surface and end wall boundary layer growth, secondary and clearance flows, and any occurrence of flow separation and stage mismatch. To meet the high efficiency and reliability requirements of the nuclear application, base line and advanced aerodynamic design techniques are incorporated with the intent to mitigate the flow path adverse working condition and losses. Design validation is carried out by test and test-calibrated 3D viscous CFD analyses of a subscale model compressor. In addition to verifying the success of the design intent, the data and computational insights of overall performance and internal flow behavior are used to establish a performance model based on Reynolds number and used for the full compressor performance prediction. The model applicable to all geometrically similar designs shows sensitive responses of helium compressor aerodynamic efficiency to Reynolds number and surface roughness. Presented in the paper is the first modern design with experimental validation for multistage axial helium compressor that concerned itself with a difficult past but which has strong current interest in countries now developing thermal and fast nuclear gas reactors.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

The Fusion Experimental Reactor (FER) is a D-T-burning tokamak machine currently being designed. It is expected to provide physical information and technical experiences that will be sufficient to proceed towards the DEMO Fusion Reactor which will demonstrate electric power generation by fusion energy. An efficient ash exhaust, a hybrid current drive operation, the use of a 3% ripple field, the technological achievements in R&D of the magnets, and the negative-ion beam system are expected to allow the FER to achieve its cost-effectiveness. >

Our understanding of the physics of internal transport barriers (ITBs) is being advanced by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and two-dimensional profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E x B flow shear rate is close to the linear growth rate of the ion temperature gradient (ITG) modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the two-dimensional profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E x B shear rate is comparable to the linear growth rates at the location of the ITB.

A conceptual design study of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The capability of the present JT-60 facility will be fully utilized for this upgrade. The mission of JT-60SU is to establish integrated basis of physics and technology for steady-state tokamak reactors. In JT-60SU, steady-state physics will be evaluated in the intermediate parameter region between the present tokamaks and steady-state tokamak reactors. Technology development for long pulse operation and research for engineering safety will also be pursued.

On the basis of a high bootstrap current fraction observation with JT-60, the concept of a Steady-State Tokamak Reactor, the SSTR, was conceived and was developed with the design activity of the SSTR at JAERI (Japan Atomic Energy Research Institue). Results of ITER/FER (International Thermonuclear Experimental Reactor/Fusion Experimental Reactor) design activities have enhanced the SSTR design. Moreover, the progress of R&D for fusion reactor engineering, especially in the development of superconducting coils and negative-ion-based NBI at JAERI, has promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor. Notably, a reduction of the circulating power of the power station facility, owing to spontaneous bootstrap current, may ensure the high Q around 50 necessary for a commercial power reactor. >

The physics of ITER (International Thermonuclear Experimental Reactor) is based on demonstrated tokamak physics and credible extrapolations of that physics. Assessment of the energy confinement and MHD stability requirements led to the choice of the major plasma parameters of 22 MA for the plasma current, a toroidal field of approximately 5 T, an aspect ratio of approximately 3, and an elongation of approximately 2. Among the major accomplishments of the physics group has been the development of a database and an empirical scaling for L-mode energy confinement and the facilitation of an H-mode database and scaling. The divertor heat loads have been estimated by using experimentally validated models. The thermal and mechanical loads due to off-normal events such as disruptions have been based on analysis of the data from other tokamaks. To achieve the required availability of 10%, the pulse length has been extended by the use of current drive using 75 MW 1.3 MeV neutral beam and 45 MW lower hybrid systems. A relatively complete set of plasma diagnostics is planned for ITER. Finally, a physics R and D program has been developed. >

The 'stiffness' of thermal transport in ELMy H-modes [edge localized high-confinement modes] is examined in a series of carefully chosen JT-60U plasmas, and measured temperatures are compared with the predictions of several transport models. A heating power scan with constant T(subscript ''ped''), a scan of pedestal temperature, T(subscript ''ped''), with constant heating power, and an on-axis/off-axis heating comparison are presented. In the power scan a 45% increase in heating (and a 12% density rise) produces an approximately fixed core temperature profile in a group of five plasmas with the same pedestal temperature. With fixed heating power, we find that a 30-40% increase in T(subscript ''ped'') is associated with similar increases in core temperature. Heating in the deep core is varied by employing different groups of neutral beams that deposit their power near the magnetic axis and farther from the axis. In these plasmas, on-axis heating produces slightly more peaked temperature profiles, although they have 60% more heating power inside r = a/2. Transport models are tested by solving the power balance equations to predict temperatures, which are then compared to the measurements. Predictions of the RLWB and IFS/PPPL models generally agree with the measured temperatures outside r approximately 0.3a, but the multimode model uniformly predicts temperatures that are too high except in the central region. Tests based on these discharges are not able to discriminate between the transport models of varying stiffness, so we conclude that larger changes are needed in the P(subscript ''heat'') and T(subscript ''ped'') scans.

A conception of Neutron Science Research Project (NSRP) has been proposed in Japan Atomic Energy Research Institute (JAERI) since 1994 for its future big science project. The project aims at exploring new basic science and nuclear energy science by using a high-intensity proton accelerator. NSRP is a complex composed of a powerful superconducting proton linac, the target systems which convert the proton beam to neutrons or other particles, and the facilities for scientific research programs. The proton linac is required to supply a high-intensity proton beam with an energy up to 1.5 GeV and an average current around 10 mA. The scientific research programs are as follows: In the area of basic science, structural biology and material science with slow neutron scattering method, neutron nuclear physics and spallation radioisotope physics, and in the area of nuclear energy science, the experimental feasibility studies of incineration for the nuclear waste transmutation and material developments with a neutron irradiation facility. Other scientific research programs are also proposed such as meson science for meson and muon physics, radioisotope production for medical use. Research and development (RD an ion source, an RFQ linac and a part of DTL linac. The conceptual design work and R&D activities for NSRP have started in the fiscal year, 1996. The first beam of 1.5 GeV and 1 mA is expected to be extracted from the proton linac by 2004 and finally a 10 mA is to be obtained in 2007 by reflecting the results of technological developments.

A conceptual design of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The present JT-60 facility will be fully utilized for this upgrade. The major objective of JT-60SU is to establish an integrated basis of physics for steady-state tokamak reactors such as SSTR. JT-60SU aims at investigation of steady-state operation relevant to the reactor regime. The technology development for long pulse operation and research for fusion safety improvement are also aimed at. The toroidal field coils are designed with using an advanced disk structure to improve the force transmission to the structure material. A modified three-in-hand grading using a circular Nb 3 Al and NbTi cable-in-conduit conductors is also adopted. A 40 cm thick water tank-type vacuum vessel made of Ti-6Al-4V, and boron (1% of 10 B) doped water is chosen for the vacuum vessel to reduce activation and to increase neutron shielding. Through these studies, we plan to conduct advanced tokamak experiments in JT-60SU for steady-state tokamak reactors. These studies will also contribute to the steady-state bum in ITER as an ultimate goal.

With the use of r.f. electric fields, the heating of the plasma and the improvement in the energy confinement are studied for an open-ended toroidal quadrupole. From an analysis of the loss mechanisms of the plasma and the heating rate, both theoretical and numerical, some optimizations have been made. The resulting heating and energy confinement are exhibited in computer experiments. The plasma can be heated easily and an improvement in the energy containment time of approximately 10% is obtained.

Emphases in JT-60 experiments are placed on (1) lower-hybrid (LH) current drive characteristics with a multi-junction type launcher, and (2) the confinement study with combination of neutral beam injection LH current drive and pellet injection. The new multi-junction LH launcher provides a sharp N/sub /// spectrum with high directivity for N/sub ///=1-3.4. The current drive efficiency and the radial distribution of high energy electron production show clear correlation with injected N/sub ///: the current drive efficiency has the maximum at low N/sub ///( approximately 1.3) while flattening of plasma current is more effective in large N/sub ///. A broad radial distribution of high energy electron current and approximately 30% reduction in sawtooth inversion radius were obtained by high N/sub /// ( approximately 2.5) LH injection. To fully suppress the sawtooth activity, low N/sub /// ( approximately 1.3) injection was found to be more effective. Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(0)/(ne) approximately 5 and ne(0) approximately 2.0*1020 m-3. The improved discharges are characterized by a strongly peaked pressure profile within the q=1 magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3 mm, 4 mm) and fast (2.2 km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14 MW. Further investigations of IDC characteristics have been made. The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of approximately 0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region.

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of qeff=2.5 and high density discharges of 4.8*1019 m-3 in line-averaged density were obtained in the divertor configuration. In divertor discharges radiated loss from the main plasma can be kept at 20-30% of the ohmic input.

Essential divertor functions — density and energy control — have been investigated in the JT-60 divertor with metal walls. Neutral pressure in the divertor chamber increases in proportion to n2e and high recycling state for particle exhaust is realized. At ne ≤ 6 × 1019m−3, wall pumping by the divertor plates is dominant in particle exhaust. However, at ne ≥ 6 × 1019m−3, particle exhaust by active divertor pumping systems becomes essential. Because of the effectiveness of the divertor, radiation loss in the main plasma is reduced to 5–10% of the absorbed power and impurity concentrations are significantly suppressed at very low level (Zeff = 1.5-2.0). In the experiments with graphite walls, short periods of H-mode phases were found in the outside X-point divertor discharges although improvement in energy confinement is limited to within 10%. In limiter discharges, combination of high vessel temperature (210–300 °C) and low current limiter discharges (1.5 MA) without gas puffing is effective as a wall conditioning in obtaining reproducible high density discharges. The maximum parameters, Ip = 3.2 MA, ne = 1.2 × 1020m−3, qeff = 2.2 at Bt = 4.8T, a Murakami parameter of 7.5, and a stored energy of 3.1 MJ have been attained. The maximum density is limited by the occurence of the MARFE. During high power neutral beam heating, enhanced carbon influx was observed possibly due to chemical sputtering. The wall pumping by the inner wall is still effective in decreasing the electron density from the high density region of ne ~ 1 × 1020m−3. An empirical energy confinement scaling of JT-60 has been drawn in terms of an offset linear function as τE = 0.19 I1.9PPabs + 0.062 a1.8p

From June to October 1987, JT-60 achieved fusion product (ne(0). tau E*.Ti(0)) of 6*1019 m-3.keV.s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of approximately 20 MW. The central electron density of 1.3*1020 m-3 was obtained at plasma current of 3 MA with 13 approximately 20 MW neutral beam power and the confinement time reached 0.14-0.18 s. An offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection approximately 40 keV, confinement degradation was found. Many short periods (0.05 approximately 0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analyses were also made for the outside X-point divertor equilibrium in connection with H0-phase capability. Heating powers of 9.5 MW and 1.9 MW were obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. High energy beam ions were accelerated by ICRF in the central region of the plasma.

The possibility of achieving a high effective Q plasma (Q is the fusion power multiplication factor) by increasing the power transfer from alpha particles to the background plasma through heating in the ion cyclotron range of frequencies (ICRF) is examined. The effective Q-value, Qeff, which is defined as the ratio of the indirect plasma heating power from ICRF heated alpha particles to the direct heating power, is evaluated by using the rate of wave absorption by alpha particles on the basis of the energy moments of the quasi-linear RF diffusion operator. The fourth harmonic alpha cyclotron wave, which is absorbed mainly by alpha particles in the absence of hydrogen, should be favourable for achieving a high Qeff plasma. In the high magnetic field, Qeff can also be quite well increased by the third harmonic wave. Effects of tail formation by ICRF waves and of wave absorption by hydrogens on Qeff are also discussed.

The input power requirements for accessing H-mode at low density and maintaining it during the density ramp in ITER is addressed by statistical means applied to the international H-mode threshold power database. Following the recent addition of new data, the improvement of existing data and the improvement of selection criteria, a revised scaling law that describes the threshold power required to obtain an L-mode to H-mode transition is presented. Predictions for ITER give a threshold power of ~52MW in a deuterium plasma at a line average density ne = 0.5×10 20 m -3 . At the nominal ITER H-mode density, ne = 1.0×10 20 m -3 , the threshold power required is ~86MW. Detailed analysis of data from individual devices suggests that the density dependence of the threshold power might increase with the plasma size and the magnetic field. On the other hand, the density at which the threshold power is minimal is found to decrease with the plasma size and increase with magnetic field. The influence of these effects on the accessibility of the H-mode regime in ITER plasmas is discussed. Analyses of the confinement database show that, in present day devices, H-modes are generally maintained with powers exceeding the threshold power by a factor larger than 1.5, and that, on the other hand, good confinement can be obtained close to the threshold power although rarely demonstrated.