Advanced Tokamak Research Articles

Status of the KSTAR tokamak construction

The KSTAR is a superconducting tokamak under construction at the National Fusion Research Center (NFRC) in Daejeon, Korea. The project mission aims at a steady-state operation and advanced tokamak physics. At present, the project is in the peak of fabrication and assembly phase. The fabrication of the major tokamak structures such as vacuum vessel, cryostat, port system, thermal shields, and gravity support, has been completed. The manufacture and testing of the 30 superconducting magnets are rigorously progressing. As of September 2005, 16 toroidal field (TF) coils and four large poloidal field coils have been completed. To verify the operational feasibility of the KSTAR coils, cool-down and current charging tests of a real sized prototype TF coil and a pair of central solenoid (CS) model coil have been carried out. The assembly of the device has begun in the beginning of 2004. Now, the vacuum vessel body, thermal shields and 11 toroidal field magnets were assembled on the tokamak position. Assembly finish is scheduled for August 2007. This paper describes the manufacture and assembly progress of the KSTAR tokamak.

A simulation study on inductive ITB control in reversed shear tokamak discharges

A self-consistent simulation, including a model for improved core energy confinement, demonstrates that externally applied, inductive current perturbations can be used to control both the location and strength of internal transport barriers (ITBs) in a fully non-inductive tokamak discharge. We find that ITB structures formed with broad non-inductive current sources such as LHCD are more readily controlled than those formed by localized sources such as ECCD. Through this external control of the magnetic shear profile, we can maintain the ITB strength which is otherwise prone to deteriorate when the bootstrap current increases. The inductive current perturbation, which can be implemented by a weak Ohmic power, offers steady-state, advanced tokamak reactors an external means of efficient ITB control for regulating the fusion-burn net output and spatial profile.

Advanced Fueling System for Use as a Burn Control Tool in a Burning Plasma Device

Steady-state advanced tokamak (AT) scenarios rely on optimized density and pressure profiles to maximize the bootstrap current fraction. Under this mode of operation, the fueling system must deposit small amounts of fuel where it is needed and as often as needed, so as to compensate for fuel losses, but not to adversely alter the established density and pressure profiles. Conventional fueling methods have not demonstrated successful fueling of AT-type discharges and may be incapable of deep fueling long-pulse edge-localized-mode-free discharges in ITER. The capability to deposit fuel at any desired radial location within the tokamak would provide burn control capability through alteration of the density profile. The ability to peak the density profile would ease ignition requirements, while operating ITER with density profiles that are peaked would increase the fusion power output. An advanced fueling system should also be capable of fueling well past internal transport barriers. Compact toroid (CT) fueling has the potential to meet these needs, while simultaneously providing a source of toroidal momentum input. Experimental data needed for the design of a CT fueler for ITER could be obtained on NSTX using an existing CT injector.

Status and prospect of JT-60 plasma control and diagnostic data processing systems for advanced operation scenarios

The large tokamak fusion device JT-60 is expected to explore more advanced tokamak discharge scenarios toward the ITER and a future power reactor. Since various experimental issues are to be adequately discussed, and possibly to be solved in JT-60, the plasma real-time control system has been drastically improved with remodelling in hardware as well as in software. To satisfy the requirements, a “multiple networks” structure is employed as a basic principle of the remodelling. Distributed processors for diagnostics, actuators, and supervisory controllers are linked through a reflective memory (RM) network for fast, real-time communication. Similarly, advanced and complex calculations to reproduce plasma shape and profiles are performed by several processors connected to the same RM network. In this report, we discuss the developments to improve the JT-60 plasma control and data processing systems. In addition, a future plasma control system leading to a standard design for a power reactor is outlined, which is based on the 20-year plasma operation experience.

Electron drift waves in an advanced tokamak plasma

The influence of details of an international thermonuclear experimental reactor (ITER)-like geometry on drift wave stability is studied. The eigenvalue problem for electrostatic electron drift waves is solved numerically by following the ballooning mode formalism and using a standard shooting technique. The real frequencies and growth rates of the most unstable modes and their eigenfunctions are calculated for two specific magnetic flux surfaces. For the equilibrium under investigation, the modes are found to be unstable for peak density profiles and their stability is found to be strongly affected by the local magnetic shear (LMS). The presence of positive LMS is found to be destabilizing on the magnetic surface where global magnetic shear is reverse. The stability behaviour is however different for a positive magnetic shear surface where the effect of large positive LMS is found to be stabilizing. The eigenfunctions are more localized in the regions where normal curvature is bad and magnetic field is weak.

Advanced tokamak research on long time scales in JT-60 Upgrade

The duration of advanced tokamak plasma operation in JT-60 Upgrade (JT-60U) [A. Kitsunezaki et al., Fusion Sci. Technol. 42, 179 (2002)] has been extended on long time scales exceeding the current diffusion time (τR) and close to the wall saturation time. A very high bootstrap current fraction (fBS) of 75% was maintained for 7.4s (2.7τR), while a normalized beta of 2.5 was maintained for 15.5s (∼9.5τR). The current profile reaches stationary conditions in ∼τR for the small fBS regime, while a longer time is required for the large fBS regime. A plasma with a weak shear q profile, similar to requirements for the ITER steady-state operational scenario, was successfully maintained. The particle inventory in the wall was observed to saturate in repeated long-pulse (∼30s) H-mode discharges with edge localized modes. The analysis of neutral particles in the scrape-of-layer plasma indicates different time scales are involved for the wall saturation. Ferritic steel tiles have been installed in the vacuum vessel to reduce the toroidal field ripple towards extending the advanced tokamak regime further in the coming operational campaign.

Progress toward fully noninductive, high beta conditions in DIII-D

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario.

Impact of the edge pedestal characteristics on the integrated performance in advanced tokamak operation modes in JT-60U

Characteristics of the edge pedestal has been studied for the high βp ELMy H mode and the reversed shear ELMy H mode in JT-60U. The contribution of the pedestal is around 20% for the confinement improvement factor, the normalized beta, and the bootstrap fraction. The upper boundary of the fusion performance measure increases with the pedestal poloidal beta, βp-ped. At high triangularity, δ, βp-ped increases with the total βp (βp − tot) almost linearly for positive shear type I ELMy discharges. This dependence is not due to the profile stiffness, since the dependence is the same for the discharges both with and without an internal transport barrier (ITB). In the reversed shear ELMy H mode, βp-ped increases with βp-tot on the same line as the positive shear cases except at high q95 = 8–9.3. In the high plasma current regime, the final structure with both an ITB and edge transport barrier seems to be determined by the balance between the expanding ITB-foot radius and deepening ELM penetration: the ELM penetration radius deepens with increasing pedestal stored energy and then reaches the ITB-foot radius. The ITB-foot seems to behave as a barrier against the ELM crash penetration, and shrinks after a few ELM attacks, and the ELM penetration follows the shrinking ITB-foot.

Comparison of H-mode pedestals in different confinement regimes in DIII-D

A survey of global performance parameters and their correlation with pedestal parameters is performed for standard H-mode, QH-mode and the enhanced confinement regimes of VH-mode, hybrid and advanced tokamak in the DIII-D tokamak. This study shows that there is a trend for global confinement quality or global beta to increase as the pedestal electron pressure or beta increases. However, there are also improvements in core confinement and beta, observed at fixed pedestal pressure or beta, which indicate that factors other than pedestal parameters also contribute to the best core performance. Several other pedestal structure parameters are found to be similar among these regimes. The scale lengths for electron pressure in the pedestal are in the range 0.8–1.6 cm at the outer midplane, most ηe values are in the range 1–3 in the middle of the Te pedestal and the Te and ne pedestals tend to penetrate the same distance into the plasma.

Overview of the National Centralized Tokamak programme

An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.

Identifying Alfvén eigenmodes in the early phase of advanced tokamak plasmas

We examine the fast-ion driven Alfvén eigenmodes (AEs) in the frequency range 0–500 kHz during the current ramp-up phase of advanced JET plasma before the main heating starts (the preheating phase). Modelling of AEs using equilibrium reconstruction and the ideal MHD spectral code MISHKA shows that three types of mode commonly found during the preheating phase correspond to: global toroidal Alfvén eigenmodes (TAEs) found in regions of moderate to high magnetic shear; core-localized TAEs found in regions of low magnetic shear; and Alfvén cascades found in regions of zero shear. MHD spectroscopy diagnosis of the time evolution of the plasma current within the central region of the plasma and of the minimum safety factor value based upon the frequency evolution of these unstable AEs is discussed.

Development on JET of advanced tokamak operations for ITER

Recent research on advanced tokamak in JET has focused on scenarios with both monotonic and reversed shear q-profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide internal transport barriers (ITBs), r/a ∼ 0.7, are formed at ITER relevant triangularity δ ∼ 0.45 and moderate plasma current, IP = 1.5–2.5 MA, with ne/nG ∼ 60% when ELMs are moderated by Ne injection. At higher current (IP ≤ 3.5 MA, δ ∼ 0.25) wide ITBs sitting at r/a ≥ 0.5, in the positive shear region, have been developed. Generally MHD events terminate these barriers otherwise limited in strength by power availability. ITBs with core density close to Greenwald value, Te ∼ Ti and low toroidal rotation (4 times lower than standard ITBs) are obtained in plasma target preformed by opportune timing of lower hybrid current drive (LHCD), pellet injection and a small amount of NBI power. Wide ITBs, r/a ∼ 0.6, of moderate strength, can be sustained without impurities accumulation for a time close to neoclassical resistive time in 3 T/1.8 MA discharges that exhibit reversed magnetic shear profiles and type-III ELMy edge. These discharges have been extended to the maximum duration allowed by JET subsystems (20 s) bringing to the record of injected energy in a JET discharge: E ∼ 330 MJ. Portability of ITB physics has been addressed through dedicated similarity experiments. The ITB is identified as a layer of reduced diffusivity studying the propagation of the heat wave generated by modulating the ICRF mode conversion (MC) electron heating. Impressive results, QDT ∼ 0.25, are obtained in these deuterium discharges with 3He minority when the MC layer is located in the core. The ion behaviour has been investigated in pure LHCD electron ITBs optimizing the 3He minority concentration for direct ion heating. Preliminary results of particle transport, studied via injection of a trace of tritium and an Ar–Ne mixture, will be presented.

Status of the KSTAR Project

Synopsis: The KSTAR project was started in 1996 and conducted in three phases, the conceptual design phase (1996–1998), the R&D and engineering design phase (1998–2002), and the construction phase (2002–2007). The mission of the Korea Superconducting Tokamak Advanced Research (KSTAR) Project is to develop a steady-state-capable advanced superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Since the KSTAR mission includes the achievement of a steady-state-capable operation, the use of superconducting coils is an obvious choice for the magnet system and the long pulse current drive and heating systems are also the important aspect of the KSTAR design features. The advanced tokamak aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizer, internal control coils. All the major components are in the stage of the fabrication and assembly for the completion of the KSTAR construction in the year 2007.

Long Pulse Operation of High Performance Plasmas in JT-60U

Recent experimental progress in JT-60U advanced tokamak research is presented:sustainment of the normalized beta (βN)∼ 3 in a normal magnetic shear plasma, the bootstrap current fraction (fBS) ∼ 45% in a weak shear plasma and ∼ 75% in a reversed magnetic shear plasma in a nearly fully non-inductive current drive condition for longer than the current relaxation time. Achievement of high-density, high-radiation fraction together with high-confinement in advanced plasmas is demonstrated. Achievements and findings in long pulse operations after system modification are presented as well. A 65 s discharge of Ip = 0.7 MA was successfully obtained. As a result, high-βN of 2.3 was successfully sustained for a very long period of 22.3 s. In addition, a 30 s standard ELMy H-mode plasma of Ip up to 1.4 MA was also obtained. Effectiveness of divertor pumping to control particle recycling and the electron density under the saturated wall retention was demonstrated. These achievements and issues in development are discussed.

A Shear-Flow Instability Relevant to Advanced Tokamak Operation

An ideal magnetohydrodynamic instability which is driven by sheared flow and which may be relevant to reversed-shear advanced tokamak operation is described. If there is flow shear at the qmin surface, relatively weak velocity shear can drive this instability, with a time scale of the flow. The flows may be significantly slower than the ambient poloidal Alfven velocity, and no inflection point is needed. Thus the time scale of the instability may be significantly longer than that of the poloidal Alfven transit time, and it might account for disruption of reversed-shear discharges recently observed in JT-60U.

The ARIES-AT advanced tokamak, Advanced technology fusion power plant

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% ( β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT ( κ x = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.

Compatibility of advanced tokamak plasma with high density and high radiation loss operation in 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.

Solenoid-free toroidal plasma start-up concepts utilizing only the outer poloidal field coils and a conducting centre-post

Eventual elimination of the in-board Ohmic heating solenoid is required for the spherical torus (ST) to function as a compact component test facility and as an attractive fusion power plant. An in-board Ohmic solenoid, along with the shielding needed for its insulation, can dramatically increase the size and, hence, the cost of the plant. Advanced tokamak reactor designs also assume no or a small in-board solenoid to reduce the size and cost of the plant. In addition, elimination of the in-board solenoid greatly reduces the coil stresses and simplifies the coil design. Here, we investigate using static as well as dynamic codes in ST geometries with two complementary solenoid-free plasma start-up approaches: one utilizes only the outer poloidal field coils to create a relatively high quality field null region while retaining significant poloidal flux, and the other takes advantage of the poloidal flux stored in the conducting centre-post to create a start-up condition similar to that of the conventional Ohmic solenoid method. We find that it is therefore possible to come up with a promising configuration, which produces a quality multi-pole field-null and sufficient loop-voltage needed for plasma initiation and significant poloidal flux for subsequent current ramp-up. The present solenoid-free start-up concept, if proved feasible, can be readily extended to higher field devices due to relatively simple physics principles and favourable scaling with the device size and toroidal field.

Development of internal transport barrier scenarios at ITER-relevant high triangularity in JET

The development of scenarios characterized by H-mode confinement and internal transport barriers (ITBs) in high triangularity, δ ∼ 0.4–0.5, discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours H-modes which are ELM-free or develop type I edge localized mode (ELM) activity, which inhibits long lasting ITBs. The recent experiments reported here concentrate on integrated optimization of edge and core conditions. The stability of the edge pedestal was controlled using gas injection, deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, resulting in L-mode. In parallel, the conditions for triggering and sustaining ITBs encompassing a large proportion of the plasma volume (outer ITBs) were optimized, as opposed to less performing ITBs located closer to the plasma centre (inner ITB). These plasmas have deeply reversed target current profiles with qmin ∼ 3 and a narrow inner ITB, located typically at a small normalized radius ρ < 0.5 and close to the reversed shear region, is routinely observed. Large radius outer ITBs are only triggered at an input power in excess of 20 MW, but they do not usually survive the transition into H-mode. The best results, in terms of sustained high confinement, have been obtained with neon injection; an outer ITB is triggered during the phase with L-mode edge and survives into H-mode for about 2 s, corresponding to ∼10 times the global energy confinement time τE, at q95 = 7.5, H89βN ∼ 3.5–4 and ∼60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios.

100% noninductive operation at high beta using off-axis ECCD in DIII-D

The advanced tokamak programme on DIII-D is to develop the scientific basis for steady state, high performance operation in future devices. We report on experiments attempting to demonstrate sustainment of 100% noninductive current for several seconds at high beta, using up to 2.5 MW of off-axis electron cyclotron current drive and up to 15 MW of neutral beam injection with q95 ≈ 5. A 100% noninductive condition was achieved with βT = 3.6%, βN = 3.5, H89 = 2.4 and improved current drive alignment. However, the duration of this phase was limited by the pressure profile evolution leading to magnetohydrodynamic instabilities after about 0.7 s. In a separate discharge, a nearly (∼90%) noninductive, stationary condition was maintained for one current relaxation time (1.8 s) only limited by the duration of the hardware system. These experiments have achieved normalized fusion performance with bootstrap current fraction fBS ≈ 60%, consistent with requirements for the ITER Q = 5 steady-state scenarios. The modelling tools that were successfully employed to devise experiments in DIII-D are applied to ITER, indicating that full noninductive operation is plausible for an ITER steady-state scenario with Q ≈ 5.