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The ITER Plasma Control System (PCS) requires an extensive set of about 50 diagnostic systems to measure the plasma response and about 20 actuators to act on the plasma to carry out its control functions. The specifications and real limitations of the actuators and diagnostics are being assessed as part of the ongoing conceptual design of the PCS to understand the potential impact on plasma control. The actuators include magnetic coils (central solenoid (CS), poloidal field (PF), vertical stability (VS), edge localized mode (ELM), correction coils (CC)), heating and current drive (electron cyclotron (EC), ion cyclotron (IC), neutral beam injection (NBI), and possibly lower hybrid (LH)), glow discharge cleaning, fueling and impurity gas and pellet injection, vacuum pumping, and disruption mitigation systems. Diagnostic systems are prioritized according to their role in machine protection (MP), basic control (BC), advanced control (AC), and physics studies (PS). At the conceptual design phase, detailed control algorithms do not yet need to be specified, but conceptual solutions must be chosen that satisfy the PCS requirements for control within the real constraints of the diagnostics and actuators. The feasibility of the chosen solutions must be proven either through established control schemes on existing machines or through an R&D program to develop them before they will be required on ITER. The diagnostic and actuator requirements of the PCS will evolve from first plasma through the high performance DT phase. A comparison is made of the expected requirements to control vertical stability, sawteeth, neoclassical tearing modes (NTMs), edge localized modes (ELMs), error fields, resistive wall modes (RWMs), Alfven eigenmodes, and disruptions with the ITER baseline actuator and diagnostic specifications. (c) 2012 ITER Organization. Published by Elsevier B.V. All rights reserved.

We currently use the CORSICA integrated modeling code for scenario studies for both the DIII-D and ITER experiments. In these simulations, free- or fixed-boundary equilibria are simultaneously converged with thermal evolution determined from transport models providing temperature and current density profiles. Using a combination of fixed boundary evolution followed by free-boundary calculation, we determine the separatrix and coil currents. We use free-boundary evolution with a state-space controller representation and transport simulation to provide feedback modeling of shape, vertical stability and profile control. In addition to a tightly coupled calculation with simulator and controller imbedded inside CORSICA, we also use a remote procedure call interface to couple the CORSICA non-linear plasma simulations to the controller environments developed within the Mathworks Matlab/Simulink environment. We present transport simulations using full shape and vertical stability control with evolution of the temperature profiles to provide simulations of the ITER controller and plasma response.

Higher moments of the plasma shape than triangularity are found to significantly affect the pedestal pressure and the edge localized mode (ELM) characteristics in DIII-D. The shape dependence of the pedestal pressure was experimentally examined by varying the squareness in the proposed ITER configuration while holding the triangularity fixed. Over this scan the pedestal pressure increased by ~50% from highest squareness to lowest squareness. The variation of pedestal energy is found to be consistent with the stability analysis of the measured profiles. The ELM energy also varied with the shape to maintain a nearly constant fraction of the pedestal energy. Stability analysis using model shapes and pressure profiles indicates that much of the advantage of high triangularity for high pedestal pressure can be achieved in lower triangularity shapes by optimizing squareness and/or the distance of the secondary upper separatrix from the primary separatrix. In high beta discharges an increase in pedestal pressure is observed with higher global stored energy. The greatest pedestal pressure increase is at low squareness due to an increase in both the pressure gradient stability limit and the width of the pedestal. The variation in pedestal pressure with squareness was also used to optimize 'hybrid' discharges in DIII-D where a lower pedestal pressure was required for an improved overall performance. In the 'hybrid' regime low squareness resulted in a high pedestal pressure with large infrequent ELMs that eventually triggered an internal 2/1 tearing mode that locked, resulting in a disruption. At higher squareness the pedestal pressure was reduced with smaller and more rapid ELMs, resulting in the maintenance of a steady beneficial internal 3/2 tearing mode and good confinement. For all the cases studied, an increase in the pedestal width at low squareness appears to be a significant factor in the increase in the total pedestal pressure.

Quiescent double barrier (QDB) conditions often form when an internal transport barrier is created with high-power neutral-beam injection into a quiescent H-mode (QH) plasma. These QH-modes offer an attractive, high-performance operating scenario for burning plasma experiments due to their quasi-stationarity and lack of edge localized modes. Our initial experiments and modelling using ECH/ECCD in QDB shots were designed to control the current profile and, indeed, we have observed a strong dependence on the q-profile when EC-power is used inside the core transport barrier region. While strong electron heating is observed with EC power injection, we also observe a drop in the other core parameters, namely ion temperature and rotation, electron density and impurity concentration. At onset and termination of the EC pulse, dynamically changing conditions are induced that provide a rapid evolution of Te/Ti profiles accessible with 0.3 < (Te/Ti)axis < 0.8 observed in QDB discharges. We are exploring the correlation and effects of observed density profile changes with respect to these time-dependent variations in the temperature ratio. Increases in the measured ion thermal and particle diffusivities inside the barrier region during an ECH pulse correlate with electron heating and a rise in the core Te/Ti ratio as the ion temperature and density profiles flatten with this change in transport. The change in transport is consistent with a destabilization of ITG turbulence as inferred from the reduction of the stability threshold due to the change in Te/Ti.

The quiescent H (QH) mode, an edge localized mode (ELM)-free and high-confinement mode, combines well with an internal transport barrier to form quiescent double barrier stationary state, high performance plasmas. The QH-mode edge pedestal pressure is similar to that seen in ELMing phases of the same discharge, with similar global energy confinement. The pedestal density in early ELMing phases of strongly pumped counter injection discharges drops, and a transition to QH-mode occurs, leading to lower calculated edge bootstrap current. The plasmas current ramp experiment and ELITE code modelling of edge stability suggest that QH-modes lie near an edge current stability boundary. At high triangularity, QH-mode discharges operate at higher pedestal density and pressure and have achieved ITER level values of βPED and ν*.

Experiments are described that have increased our understanding of the transport and stability physics that set the H-mode edge pedestal width and height, determine the onset of Type-I edge localized modes (ELMs) and produce the nonlinear dynamics of the ELM perturbation in the pedestal and scrape-off layer (SOL). Models now exist for the ne pedestal profile and the pe height at the onset of Type-I ELMs, and progress has been made towards models of the Te pedestal width and nonlinear ELM evolution. Similarity experiments between DIII-D and JET suggested that neutral penetration physics plays an important role in the relationship between the width and height of the ne pedestal. Plasma physics appears to dominate in setting the Te pedestal width. Measured pedestal conditions including edge current at ELM onset agree with intermediate-n peeling?ballooning (P?B) stability predictions. Midplane ELM dynamics data show the predicted (P?B) structure at the ELM onset, large rapid variations of the SOL parameters and fast radial propagation in later phases, similar to features in nonlinear ELM simulations.

The quiescent high confinement mode (QH-mode) on DIII-D exhibits an H-mode like edge pedestal, with values of pedestal pressure and global energy confinement similar to those of the ELMing H-mode, but without ELMs. In many cases this mode is observed to reach a nearly stationary operating point, limited in duration only by hardware constraints. This mode is usually obtained in a plasma configuration that is strongly pumped with electron densities at the top of the pedestal below 0.3 of the Greenwald density. Electron temperatures at the top of the pedestal range from 1.2 to 2.2 keV. Ion temperatures at the pedestal are much higher, ranging up to 5 keV. In this paper, we will present edge profiles observed in the QH-mode. A selection of these profiles are used as input to the Corsica code to calculate the edge current profile, which is dominated by the bootstrap current. We discuss the implications of these edge profiles on the stability of the QH edge against ballooning/peeling modes. We also discuss results of experiments in which we attempt to expand the range of edge parameters, especially the edge density, achievable in the QH-mode.

Substantial progress has been made towards both understanding and control of internal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained high performance operating mode termed the quiescent double barrier (QDB) regime. The QDB regime combines core transport barriers with a quiescent ELM-free H mode edge (termed QH mode), giving rise to separate (double) core and edge transport barriers. The core and edge barriers are mutually compatible and do not merge, resulting in broad core profiles with an edge pedestal. The QH mode edge is characterized by ELM-free behaviour with continuous multiharmonic MHD activity in the pedestal region and has provided density and radiated power control for longer than 3.5 s (25τE) with divertor pumping. QDB plasmas are long pulse high performance candidates, having maintained a βN H89 product of 7 for five energy confinement times (Ti ≤ 16 keV, βN ≤ 2.9, H89 ≤ 2.4 τE ≤ 150 ms, DD neutron rate Sn ≤ 4 × 1015 s-1). The QDB regime has only been obtained in counter-NBI discharges (injection antiparallel to the plasma current) with divertor pumping. Other results include successful expansion of the ITB radius using (separately) both impurity injection and counter-NBI, and the formation of ITBs in the electron thermal channel using both ECH and strong negative central shear (NCS) at high power. These results are interpreted within a theoretical framework in which turbulence suppression is the key to ITB formation and control, and a decrease in core turbulence is observed in all cases of ITB formation.

Non-activation phase operations in ITER in hydrogen (H) and helium (He) will be important for commissioning of tokamak systems, such as diagnostics, heating and current drive (HCD) systems, coils and plasma control systems, and for validation of techniques necessary for establishing operations in DT. The assessment of feasible HCD schemes at various toroidal fields (2.65–5.3 T) has revealed that the previously applied assumptions need to be refined for the ITER non-activation phase H/He operations. A study of the ranges of plasma density and profile shape using the JINTRAC suite of codes has indicated that the hydrogen pellet fuelling into He plasmas should be utilized taking the optimization of IC power absorption, neutral beam shine-through density limit and H-mode access into account. The EPED1 estimation of the edge pedestal parameters has been extended to various H operation conditions, and the combined EPED1 and SOLPS estimation has provided guidance for modelling the edge pedestal in H/He operations. The availability of ITER HCD schemes, ranges of achievable plasma density and profile shape, and estimation of the edge pedestal parameters for H/He plasmas have been integrated into various time-dependent tokamak discharge simulations. In this work, various H/He scenarios at a wide range of plasma current (7.5–15 MA) and field (2.65–5.3 T) have been developed for the ITER non-activation phase operation, and the sensitivity of the developed scenarios to the used assumptions has been investigated to provide guidance for further development.

The ITER Integrated Modelling & Analysis Suite (IMAS) will support both plasma operation and research activities on the ITER tokamak experiment. The IMAS will be accessible to all ITER Members as a key tool for the scientific exploitation of ITER. It will allow collective development of Integrated Modelling tools, by sharing data, code components and, ultimately, workflows based on coupling together various code components. Its design started in 2011 and a first prototype of the IMAS infrastructure has already been implemented at the ITER Organization (IO). The purpose of this paper is to describe the essential features of the IMAS design, the implemented prototype, as well as the first physics applications which have been developed under the IMAS infrastructure. The IMAS infrastructure is based on a standardized data model that covers experimental and simulated data with the same representation. The standard data model is device-generic and can be used to describe data from existing experiments. Since the data model will progressively cover a large number of areas (plasma, diagnostics, actuators, other tokamak subsystems, . . . ) and will be developed by many contributors, a set of data model design rules and guidelines have been established to ensure consistency and homogeneity of the data model. Physics components, once interfaced to the data model, can be coupled into an Integrated Modelling workflow orchestrated by a workflow engine. A first implementation of all these infrastructure elements has been carried out and is described in this paper. First applications have been integrated under the prototype IMAS infrastructure to allow their performance to be tested and to demonstrate the expected functionalities of the infrastructure. Transport solvers with free boundary equilibrium capabilities have been integrated to the IMAS infrastructure, namely CORSICA and DINA. ITER pulse simulations have been carried out by coupling the Plasma Control System (PCS), which in ITER has a dedicated and distinct simulation platform based on the Simulink r software, to the physics solvers. This has required the development of an original co-simulation technique between the plasma and plant simulator (under IMAS) and the dedicated PCS simulator. First results of full tokamak simulations under the IMAS infrastructure will be described in the paper.

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ∼2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5–10 MA depending on modelling assumptions.

Particle losses associated with edge localized mode (ELM) activity on the DIII-D tokamak [J. Luxon et al., Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1986, Vol. I, p. 159] are evaluated quantitatively using density profile data obtained from a Thomson scattering system. It is shown that up to 10% of the total core particle content is lost with each ELM. The particle loss varies inversely with ELM frequency. The temporally averaged ELM particle loss is shown to be about 25% of the total particle loss from the confined region under a wide variety of plasma conditions. Although this ELM loss is a small fraction of the total ion flux, it is large compared to the particle input from neutral beam heating. Hence ELM particle losses are sufficient to control the density rise associated with H-mode plasma operation with neutral beam heating. In addition to controlling the average density by enhancing the total ion fl...

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5>E with qmin > 1.5. (The normalized performance is measured by the product βNH89, indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes as previously observed. Confinement remains good despite qmin > 1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ≈75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with βNH89 ≈ 7 for up to 6.3 s or ≈34τE. These discharges appear to have stationary current profiles with qmin ≈ 1.05 in agreement with the current profile relaxation time ≈1.8 s.

Significant progress has been made in obtaining high-performance discharges for many energy confinement times in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. Normalized performance (measured by the product of βNH89 and indicative of the proximity to both conventional β limits and energy confinement quality, respectively) ∼10 has been sustained for >5 τE with qmin>1.5. These edge localized modes (ELMing) H-mode discharges have β∼5%, which is limited by the onset of resistive wall modes slightly above the ideal no-wall n=1 limit, with approximately 75% of the current driven noninductively. The remaining Ohmic current is localized near the half-radius. The DIII-D electron cyclotron heating system is being upgraded to replace this inductively driven current with localized electron cyclotron current drive (ECCD). Density control, which is required for effective ECCD, has been successfully demonstrated in long-pulse high-performance ELMing H-mode discharges with βNH89∼7 for up to 6.3 s. In plasma shapes compatible with good density control in the present divertor configuration, the achieved βN is somewhat less than that in the high βNH89=10 discharges.

Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection taking place when the heating power exceeds a threshold value of about 9 MW, contrasting to co-NBI injection, where Pthreshold

Discharges with negative central magnetic shear (NCS) hold the promise of enhanced fusion performance in advanced tokamaks. However, stability to long wavelength magnetohydrodynamic modes is needed to take advantage of the improved confinement found in NCS discharges. The stability limits seen in DIII-D [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] experiments depend on the pressure and current density profiles and are in good agreement with stability calculations. Discharges with a strongly peaked pressure profile reach a disruptive limit at low beta, βN=β(I/aB)−1⩽2.5 (% m T/MA), caused by an n=1 ideal internal kink mode or a global resistive instability close to the ideal stability limit. Discharges with a broad pressure profile reach a soft beta limit at significantly higher beta, βN=4 to 5, usually caused by instabilities with n>1 and usually driven near the edge of the plasma. With broad pressure profiles, the experimental stability limit is independent of the magnitude of negative she...

Discharges exhibiting the highest plasma energy and fusion reactivity yet realized in the DIII-D tokamak have been produced by combining the benefits of a hollow or weakly sheared central current profile with a high confinement (H-mode) edge. In these discharges, low power neutral beam injection heats the electrons during the initial current ramp, and {open_quotes}freezes in{close_quotes} a hollow or flat central current profile. When the neutral beam power is increased, formation of a region of reduced transport and highly peaked profiles in the core often results. Shortly before these plasmas would otherwise disrupt, a transition is triggered from the low (L-mode) to high (H-mode) confinement regimes, thereby broadening the pressure profile and avoiding the disruption. These plasmas continue to evolve until the high performance phase is terminated nondisruptively at much higher {beta}{sub T} (ratio of plasma pressure to toroidal magnetic field pressure) than would be attainable with peaked profiles and an L-mode edge. Transport analysis indicates that in this phase, the ion diffusivity is equivalent to that predicted by Chang-Hinton neoclassical theory over the entire plasma volume. This result is consistent with suppression of turbulence by locally enhanced E x B flow shear, and is supported by observations of reduced fluctuations in the plasma. Calculations of performance in these discharges extrapolated to a deuterium-tritium fuel mixture indicates that such plasmas could produce a DT fusion gain Q{sub DT} = 0.32.

In this paper the authors present measurements of the global power and particle balance in the high-performance phase of negative central magnetic shear (NCS) discharges and compare with reference VH-mode discharges. The principal differences observed are that NCS has a much lower fraction of the total input power flowing into the boundary, less core radiation, and larger rate of stored energy increase as a fraction of total power. Scrape-off layer (SOL) temperature and divertor heat flux profiles, and radiation profiles at the midplane, are similar to VH-mode. Due to the good core particle confinement and efficient fueling by neutral beam injection (NBI), with little gas puffing, the gas load on the walls and the recycling are very low during the NCS discharges. This results in a rate of density rise relative to beam fueling at the L to H transition time which is 1/3 of the value for VH transitions, which is in turn 1/2 that for L-to-ELMing-H-mode transitions.

Strong shaping, favourable for stability and improved energy confinement, together with a significant expansion of the central region of improved confinement in negative central magnetic shear target plasmas, increased the maximum fusion power produced in DIII-D by a factor of 3. Using deuterium plasmas, the highest fusion power gain, the ratio of fusion power to input power, Q, was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma, which is similar to values achieved in tokamaks of larger size and magnetic field. A simple transformation relating Q to the stability parameters is presented

Plasma discharges with negative central magnetic shear (NCS) and access to the ballooning second stability regime have been produced on DIII-D with edge plasma conditions similar to those in the standard L mode confinement regime. Confinement enhancement factors up to H identical to tau E/ tau ITER-89P~2.5 are obtained while maintaining the L mode edge. Compared with discharges with monotonic q profiles, highly peaked toroidal rotation (fphi (0) approximately=30-70 kHz), ion temperature (Ti(0) approximately=15-22 keV) and density (ne(0)/ne approximately=2.2) profiles are observed. This regime has yielded the highest DD neutron rates observed to date on DIII-D. Steep pressure gradients drive a large bootstrap current, with up to ~75% total non-inductive current drive. These features make this an attractive advanced tokamak operating regime for further study

Reliable operation of discharges with negative central magnetic shear has led to significant increases in plasma performance and reactivity in both low confinement, L‐mode, and high confinement, H‐mode, regimes in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159]. Using neutral beam injection early in the initial current ramp, a large range of negative shear discharges have been produced with durations lasting up to 3.2 s. The total noninductive current (beam plus bootstrap) ranges from 50% to 80% in these discharges. In the region of shear reversal, significant peaking of the toroidal rotation [fφ(0)∼30–60 kHz] and ion temperature [Ti(0)∼15–22 keV] profiles are observed. In high‐power discharges with an L‐mode edge, peaked density profiles are also observed. Confinement enhancement factors up to H≡τE/τITER‐89P∼2.5 with an L‐mode edge, and H∼3.3 in an edge localized mode (ELM)‐free H mode, are obtained. Transp...

The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor () profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, , of at least 5. This paper presents candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers.

Experiments to explore the long‐time evolution of noninductive, high βp plasmas in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159], have identified a new, quiescent, high performance regime. The experiments were carried out at low current (400–800 kA) with medium power neutral beam injection (3–10 MW). This regime is characterized by high q0 (≳2) and moderate li(∼1.3). It is reached by slow relaxation of the current profile, on the resistive time scale. As the profiles relax, q0 rises and li falls. When q0 goes above 2 (approximately), magnetohydrodynamic (MHD) activity disappears, and the stored energy rises. Most dramatic is the strong peaking of the central density, which increases by as much as a factor of 2. The improved central confinement appears similar to the PEP/reversed central shear/second stable core modes seen in tokamak experiments, but in this case without external intervention or transient ...

This paper outlines the principles of magnetic confinement of fusion plasma as implemented in the ITER device. The configuration of the magnet power supplies is described together with coil current scenarios for a typical plasma discharge.

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96-L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down. © 2011 IAEA, Vienna.

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300–500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode and a late H-mode onset. Static equilibrium analyses for this scenario, which determine PF coil currents to produce a given plasma configuration, indicate that the original PF coil limitations do not allow low li(

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation. © 2009 IAEA, Vienna.

The design and performance of a relatively low-cost, plasma-based, 14-MeV deuterium-tritium neutron source for accelerated end-of-life testing of fusion reactor materials are described. An intense flux (up to 5 [times] 10[sup 18] n/m[sup 2][center dot]s) of 14-MeV neutrons is produced in a fully ionized high-density tritium target (n[sub e] [approx] 3 [times] 10[sup 21] m[sup [minus]3]) by injecting a current of 150-keV deuterium atoms. The tritium plasma target and the energetic D[sup +] density produced by D[sup 0] injection are confined in a [

High density (non-thermal-barrier) operation in the Tandem Mirror Experiment Upgrade (TMX-U) is found to be restricted by a stability limit. This limit is observed in the ratio of the neutral beam sustained central cell beta βc to the hot ion beta βih in the minimum-B anchor cells at both ends of the central cell, qualitatively consistent with a flute interchange stability limit. The ratio is βc/βih = 5, over the range of 0.03 ≤ βc ≤ 0.22, with no apparent reduction due to ballooning at high βc. This is a factor of six below the standard magnetohydrodynamic (MHD) m = 1 stability theory prediction of βc/βih = 33 at βc ≤ 0.1, where ballooning corrections to flute modes are small. The discrepancy could be due to approximations in the theory; however, experimental data indicate that the stability limit is due to drift wave turbulence or to large-m MHD flute or ballooning modes. The experimental beta limit is nearly independent of the hot electron beta in the anchor cells, which is compatible with theoretical predictions that the hot electron beta will decouple from MHD activity.

The next-generation of tokamaks, ITER (International Thermonuclear Text Reactor) or BPX (Burning Plasma Experiment), will be characterized by an even greater emphasis on joint operation and experimentation. Preparations are being made for such shared facilities by developing a systematic approach to remote, joint physics operation involving experimental teams at several locations. The local area network of computers used for control and data acquisition on present and future experiments can be extended over a wide area network to provide a mechanism for remote operation of subsystems required for physics experiments. The technology required for high-bandwidth (>or=45-Mb/s) connections between multiple sites either exists or will be available over the next few years. With the rapid development of high-performance workstations, network interfaces, distributed computing, and video conferencing, one can proceed with the development of a system of control and analysis sites to provide for consistent, efficient, and continuing collaborations. Early establishment of such sites could also enhance existing joint design and development efforts. >

Collaboration is increasing in importance as physics research continues to become concentrated on a few large and expensive facilities. This is particularly true in magnetic fusion energy research, where the next generation experiments will be operated as facilities with national and international participation. In addition, these facilities are designed for steady state operation, for which interactive, real-time experimentation becomes an important issue. We are developing tools to provide for the establishment of geographically distant centers for interactive operations; such centers would allow scientists to participate in experiments from their home institutions. We are developing a testbed for a Remote Experimental Environment (REE), a "Collaboratory" The testbed will be used to evaluate the ability of a remotely located group of scientists to conduct research on the DIII-D Tokamak at General Atomics (GA). The REE will serve as a testing environment for advanced control and collaboration concepts applicable to future experiments. Process-to-process communications over high speed wide area networks provide real-time synchronization and exchange of data among multiple computer networks, while the ability to conduct research is enhanced by adding audio/video communication capabilities. The Open Software Foundation's (OSF) Distributed Computing Environment (DCE) is being used to test concepts in distributed control, security, naming, remote procedure calls and distributed file access using the Distributed File Services (DFS). We are exploring the technology and sociology of remotely participating in the operation of a large scale experimental facility.

The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed ‘error fields’. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER’s diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible.

Sustainment of Q ~ 10 operation with a fusion power of ~500 MW for several hundred seconds is a key mission goal of the ITER Project. Past calculations and simulations predict that these conditions can be produced in high-confinement mode operation (H-mode) at 15 MA relying on only inductive current drive. Earlier development of 15 MA baseline inductive plasma scenarios provided a focal point for the ITER Design Review conducted in 2007–2008. In the intervening period, detailed predictive simulations, supported by experimental demonstrations in existing tokamaks, allow us to assemble an end-to-end specification of this scenario consistent with the final design of the ITER device. Simulations have encompassed plasma initiation, current ramp-up, plasma burn and current ramp-down, and have included density profiles and thermal transport models producing temperature profiles consistent with edge pedestal conditions present in current fusion experiments. These quasi-stationary conditions are maintained due to the presence of edge-localized modes that limit the edge pressure. High temperatures and densities in the pedestal region produce significant edge bootstrap current that must be considered in modelling of feedback control of shape and vertical stability. In this paper we present new results of transport simulations fully consistent with the final ITER design that remain within allowed limits for the coil system and power supplies. These self-consistent simulations increase our confidence in meeting the challenges of the ITER program.

In 1992, our group began exploring the requirements for a comprehensive simulation code for toroidal magnetic fusion experiments. There were several motivations for taking this step. First, the new machines being designed were much larger and more expensive than current experiments. Second, these new designs called for much more sophisticated control of the plasma shape and position, as well as the distributions of energy, mass, and current within the plasma. These factors alone made it clear that a comprehensive simulation capability would be an extremely valuable tool for machine design. The final motivating factor was that the national Numerical Tokamak Project (NTP) had recently received High Performance Computing and Communications (HPCC) Grand Challenge funding to model turbulent transport in tokamaks, raising the possibility that first-principles simulations of this process might be practical in the near future. We felt that the best way to capitalize on this development was to integrate the resulting turbulence simulation codes into a comprehensive simulation. Such simulations must include the effects of many microscopic length- and time-scales. In order to do a comprehensive simulation efficiently, the length- and time- scale disparities must be exploited. We proposed to do this by coupling the average or quasistatic effects from the fast time-scales to a slow-time-scale transport code for the macroscopic plasma evolution. In FY93-FY96 we received funding to investigate algorithms for computationally coupling such disparate-scale simulations and to implement these algorithms in a prototype simulation code, dubbed CORSICA. Work on algorithms and test cases proceeded in parallel, with the algorithms being incorporated into CORSICA as they became mature. In this report we discuss the methods and algorithms, the CORSICA code, its applications, and our plans for the future.

In-vessel, non-axisymmetric, control coils have proven to be an important option for mitigating and suppressing edge-localized modes (ELMs) in high performance operating regimes on a growing number of tokamaks. Additionally, an in-vessel non-axisymmetric ELM control coil is being considered in the ITER baseline design. In preparing for the initial operation of this coil set, a comprehensive study was carried out to characterize the linear superposition of the 3D vacuum magnetic field, produced by the ELM coil, on a series of equilibria representing nine standard ITER operating scenarios. Here, the spatial phase angle of toroidally distributed currents, specified with a cosine waveform, in the upper and lower rows of the ITER ELM coil (IEC) set is varied in 2? steps while holding the current in the equatorial row of coils constant. The peak current in each of the three toroidal rows of window-frame coils making up the IEC is scanned between 5?kAt and 90?kAt in 5?kAt steps and the width of the edge region covered by overlapping vacuum field magnetic islands is calculated. This width is compared to a vacuum field ELM suppression correlation criterion found in DIII-D. A minimum coil current satisfying the DIII-D criterion, along with an associated set of phase angles, is identified for each ITER operating scenario. These currents range from 20?kAt to 75?kAt depending on the operating scenario being used and the toroidal mode number (n) of the cosine waveform. Comparisons between the scaling of the divertor footprint area in cases with n?=?3 perturbation fields versus those with n?=?4 show significant advantages when using n?=?3. In addition, it is found that the DIII-D correlation criterion can be satisfied in the event that various combinations of individual IEC window-frame coils need to be turned off due to malfunctioning components located inside the vacuum vessel. Details of these results for both the full set of 27 window-frame coils and various reduced sets, using either n?=?3 and n?=?4 perturbation fields, are discussed.

The linear ideal magnetohydrodynamic (MHD) limits of the pedestal in ITER scenarios associated with the preparation and realization of the nominal fusion gain Q?=?10 (inductive scenario at 15?MA/5.3?T, half-field/half-current and intermediate H-mode scenario at 10?MA/3.5?T), as well as the hybrid scenario at 12?MA/5.3?T, are investigated in this work. The accessible part of the MHD stability diagram is determined by computing the bootstrap current and self-consistently evaluating the corresponding pedestal current. This procedure shows that only a small part of peeling?ballooning diagrams is physically accessible. Uncertainties about the foreseen plasma profiles motivate studies evaluating the impact of various parameters on the pedestal limits. We have addressed issues such as the pedestal width, the global performance, pressure peaking, edge current density, internal inductance and plasma shaping. A scaling law for the maximum pedestal pressure in the ITER scenarios is proposed, highlighting that the main dependences are on the plasma current, the edge safety factor, the pedestal width and the internal inductance.

Intrinsic toroidal plasma rotation due to the neoclassical toroidal plasma viscosity (NTV) effect induced by a three-dimensional helical magnetic field ripple in tokamaks is investigated in this paper. The intrinsic rotation is determined self-consistently by searching for the roots of the ambipolarity constraint, after evaluation of the particle fluxes from the numerical modelling. In the low-collisionality case, there are three roots, in which two are stable roots. One corresponds to the ?ion root? in the counter-current direction, and the other stable one corresponds to the ?electron root? in the co-current direction, near which the electron flux is dominant. Both of the two stable roots scale like the diamagnetic frequency. In the high-collisionality case, there is only one ?ion? root. The application of this modelling for International Thermonuclear Experimental Reactor (ITER) cases is discussed. In a large range of plasma radii, there are three roots. The NTV torque drives plasma rotation in ITER towards one of the stable roots, depending on the initial condition. The amplitudes of the electron roots near the pedestal in both baseline and steady-state scenarios are much larger than that of the ion roots. The amplitudes of the NTV torque density and the electron roots near the pedestal increase with increasing height of the temperature pedestal in the ITER baseline scenario.

The study of operation with raised strike points on the first ITER divertor target plates is motivated by the need to gain experience with operation with strike points on tungsten (W) surfaces during the non-active phases (in the case of an initial carbon fibre composite (CFC)/W divertor); or (if ITER begins with a full-W divertor), to gain experience with plasma control and transients while operating with raised strike points to avoid damaging the baseline strike regions in preparation for the nuclear phase, and to provide a means for operation should damage occur in the baseline strike zone. For operation with raised strike points, we use the Corsica code to investigate the range of possible H- and L-mode equilibria, with emphasis on the maximum plasma current, achievable shapes, etc. With raised strike points the maximum achievable plasma current is close to 14 MA. The operating space (βp − li) for raised strike points has been studied. The size of the βp − li operating space shrinks (compared to using standard strike-point positions) at 14 MA. For 12 MA, however, the operating space is not affected when using raised strike points. For equilibria with elevated strike points (at roughly the CFC/W transitions, following the 2007 ITER Design Review) the vertical-instability growth-rates at high plasma current (14 MA) are somewhat high but are within the 20 s−1 which studies indicate are controllable in ITER. At lower currents (12 MA) in H-mode, the vertical-instability growth rates stay below 10.0 s−1 for most of βp − li space. At 12 MA in H-mode, multiple equilibria which meet our constraints have been found in overlapping regions of the βp − li operating space.

This work demonstrates that simulations of advanced burning plasma operation scenarios can be successfully parallelized in time using the parareal algorithm. CORSICA -an advanced operation scenario code for tokamak plasmas is used as a test case. This is a unique application since the parareal algorithm has so far been applied to relatively much simpler systems except for the case of turbulence. In the present application, a computational gain of an order of magnitude has been achieved which is extremely promising. A successful implementation of the Parareal algorithm to codes like CORSICA ushers in the possibility of time efficient simulations of ITER plasmas.

Author(s): Lazarus, EA; Navratil, GA; Greenfield, CM; Strait, EJ; Austin, ME; Burrell, KH; Casper, TA; Baker, DR; DeBoo, JC; Doyle, EJ; Durst, R; Ferron, JR; Forest, CB; Gohil, P; Groebner, RJ; Heidbrink, WW; Hong, R; Houlberg, WA; Howald, AW; Hsieh, C; Hyatt, AW; Jackson, GL; Kim, J; Lao, LL; Lasnier, CJ; Leonard, AW; Lohr, J; La Haye RJ; Maingi, R; Miller, RL; Murakami, M; Osborne, TH; Perkins, LJ; Petty, CC; Rettig, CL; Rhodes, TL; Rice, BW; Sabbagh, SA; Schissel, DP; Scoville, JT; Snider, RT; Staebler, GM; Stallard, BW; Stambaugh, RD; St John HE; Stockdale, RE; Taylor, PL; Thomas, DM; Turnbull, AD; Wade, MR; Wood, R; Whyte, D | Abstract: Fusion power has been increased by a factor of 3 in DIII-D by tailoring the pressure profile to avoid the kink instability in H-mode plasmas. The resulting plasmas are found to have neoclassical ion confinement. This reduction in transport losses in beam-heated plasmas with negative central shear is correlated with a dramatic reduction in density fluctuations. Improved magnetohydrodynamic stability is achieved by controlling the plasma pressure profile width. In deuterium plasmas the highest gain Q (the ratio of fusion power to input power), was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma. © 1996 The American Physical Society.

During the past year, LLNL efforts in the DIII-D experimental program have expanded to include time-dependent modeling of advanced tokamak (AT) operating modes. Consistent with our involvement in experimental operations, we have directed our initial efforts at modeling the negative central shear (NCS) configuration, an important and attractive mode of operation for reducing the size and cost of future tokamak experiments without sacrificing performance. In this endeavor, we have brought into use the Corsica modeling code as a tool for investigating the time-dependent evolution and control of various operating modes. In our current efforts, we are contributing to the analysis of the NCS experimental data using analysis tools such as the EFIT equilibrium code and the ONETWO and TRANSP transport codes. Results of these analyses are being used for comparisons with the Corsica modeling. Future directions include the modeling of startup and sustaining of NCS (and other AT) configurations, the understanding of current drive effects, the development of current drive scenarios and control algorithms, and the design of experiments and prediction of experimental results. We are currently in the early stages of applying this powerful modeling tool to the DIII-D experimental program.

This report summarizes LLNL`s involvement in recent high {beta}{sub p} experiments on the DIII-D tokamak at General Atomics. These experiments were done in collaboration with several members of the DIII-D physics staff from GA and from other collaborating institutions and could not have succeeded without this joint effort. In this report, the authors summary a specific, limited set of experiments to extend high {beta}{sub p} operation with enhanced core confinement to higher plasma currents. The interest in these experiments stems from the non-inductive current drive requirement for steady-state advanced tokamak regimes which can most reasonably be met by operation with a high bootstrap current fraction.

Scientific research is now often conducted on large and expensive experiments that utilize collaborative efforts on a national or international scale to explore physics and engineering issues. This is particularly true for the current US magnetic fusion energy program where collaboration on existing facilities has increased in importance and will form the basis for future efforts. As fusion energy research approaches reactor conditions, the trend is towards fewer large and expensive experimental facilities, leaving many major institutions without local experiments. Since the expertise of various groups is a valuable resource, it is important to integrate these teams into an overall scientific program. To sustain continued involvement in experiments, scientists are now often required to travel frequently, or to move their families, to the new large facilities. This problem is common to many other different fields of scientific research. The next-generation tokamaks, such as the Tokamak Physics Experiment (TPX) or the International Thermonuclear Experimental Reactor (ITER), will operate in steady-state or long pulse mode and produce fluxes of fusion reaction products sufficient to activate the surrounding structures. As a direct consequence, remote operation requiring robotics and video monitoring will become necessary, with only brief and limited access to the vessel areamore » allowed. Even the on-site control room, data acquisition facilities, and work areas will be remotely located from the experiment, isolated by large biological barriers, and connected with fiber-optics. Current planning for the ITER experiment includes a network of control room facilities to be located in the countries of the four major international partners; USA, Russian Federation, Japan, and the European Community.« less

A stability analysis using equilibria from CORSICA transport simulations finds that the maximum stable pedestal pressure in ITER 15 MA baseline plasma is 110 kPa corresponding to a pedestal temperature of 5.9 keV. The height of the stable pedestal is robust for the assumption of the pedestal height varying only by about 10% if the width of the pedestal is varied by 30%. A conducting first wall has a stabilizing effect on the peeling–ballooning modes that limit the edge pressure. However, the stabilization is unlikely to significantly change the stability limits, but could affect the ELM dynamics by lowering the growth rate of the ELM triggering peeling–ballooning modes. The entire pedestal region is stable against n = ∞ ballooning modes for all studied pedestal temperatures. This is due to the high bootstrap current keeping the magnetic shear in the region of large pressure gradient.

The non-linear reduced four-field RMHD model in cylindrical geometry was extended to include plasma rotation, neoclassical poloidal viscosity and two fluid diamagnetic effects. Interaction of the static resonant magnetic perturbations (RMPs) with the rotating plasmas in tokamaks was studied. The self-consistent evolution of equilibrium electric field due to RMP penetration is taken into account in the model. It is demonstrated that in the pedestal region with steep pressure gradients, mean flows perpendicular to the magnetic field, which includes and electron diamagnetic components plays an essential role in RMP screening by plasma. Generally, the screening effect increases for lower resistivity, stronger rotation and smaller RMP amplitude. Strong screening of central islands was observed limiting RMP penetration to the narrow region near the separatrix. However, at certain plasma parameters and due to the non-linear evolution of the radial electric field produced by RMPs, the rotation can be compensated by electron diamagnetic rotation locally. In this case, RMPs can penetrate and form magnetic islands. Typical plasma parameters and RMPs spectra on DIII-D, JET and ITER were used in modelling examples presented in the paper.

Intense pulse electron cyclotron heating (ECH) experiments have been carried out on the MTX tokamak. Rf pulses at 140 GHz with peak power of 1–2 GW and 25 ns pulse length were generated by the ETA‐II/IM FEL and transported quasi‐optically to MTX for O‐mode launch. Because of the intense rf electric fields (∼250 kV/cm), reduction of plasma absorption by nonlinear effects was predicted and several rf beam geometries (kII gradient) were investigated to study their effect on the absorption. Measurements of beam transmission showed increases, compared to low power (2 kW), which agreed with theory to within the data scatter. For these experiments x‐ray, ECE, and Thomson diagnostics showed evidence for localized absorption at the cyclotron resonance and hot electron production. A comparison of these results with calculations from the orbit following code ORPAT will be presented.

A minimum set of equations based on the peeling–ballooning (P–B) model with nonideal physics effects (diamagnetic drift, E × B drift, resistivity and anomalous electron viscosity) is found to simulate pedestal collapse when using the BOUT++ simulation code, developed in part from the original fluid edge code BOUT. Linear simulations of P–B modes find good agreement in growth rate and mode structure with ELITE calculations. The influence of the E × B drift, diamagnetic drift, resistivity, anomalous electron viscosity, ion viscosity and parallel thermal diffusivity on P–B modes is being studied; we find that (1) the diamagnetic drift and E × B drift stabilize the P–B mode in a manner consistent with theoretical expectations; (2) resistivity destabilizes the P–B mode, leading to resistive P–B mode; (3) anomalous electron and parallel ion viscosities destabilize the P–B mode, leading to a viscous P–B mode; (4) perpendicular ion viscosity and parallel thermal diffusivity stabilize the P–B mode. With addition of the anomalous electron viscosity under the assumption that the anomalous kinematic electron viscosity is comparable to the anomalous electron perpendicular thermal diffusivity, or the Prandtl number is close to unity, it is found from nonlinear simulations using a realistic high Lundquist number that the pedestal collapse is limited to the edge region and the ELM size is about 5–10% of the pedestal stored energy. This is consistent with many observations of large ELMs. The estimated island size is consistent with the size of fast pedestal pressure collapse. In the stable α-zones of ideal P–B modes, nonlinear simulations of viscous ballooning modes or current-diffusive ballooning mode (CDBM) for ITER H-mode scenarios are presented.