Session UO4: Research In Support of ITER

ITER Research Plan

The ITER Research Plan (IRP) is being developed to provide a guide to the research activities which should be undertaken within the framework of the ITER Project and will encompass both physics and technology research during ITER construction and operation. The Plan provides a framework linking and integrating the current research priorities of the Project with the preparation for future exploitation. Recent analysis has focused on the adaptation of the IRP to the Project's updated construction and operation schedule with the aim of developing an operational programme which makes the most rapid transition, within the overall Project constraints, to DT fusion energy production. The programme evolves through a period of non-active (hydrogen and helium) operation to allow full commissioning of the facility, a phase of deuterium operation in which operation with all-metallic PFCs will be established, followed by the transition to full DT operation and the production of plasmas with significant fusion energy gain. Identification of the key elements in the experimental programme leading to burning plasma studies allows priorities for R{\&}D activities required to prepare the efficient exploitation of the ITER device to be defined.   

Disruption mitigation and avoidance at ASDEX Upgrade

Disruptions are a major concern for tokamaks and in particular for ITER. They cause high heat loads during the thermal quench and high mechanical forces during the subsequent current quench. The generation and loss of runaway electrons (highly accelerated electrons carrying large fractions of the plasma current) can produce damage to the vessel structures. Therefore, schemes are implemented in present tokamaks to mitigate or to even avoid them. Mitigation has been proven to be effective through the injection of noble gases causing a reduction of the thermal heat load by radiation and a reduction of the mechanical forces. In addition 25\% of the required density for the collisional suppression of runaways in ITER has been reached. For the trigger of the noble gas injection a locked mode detector is routinely used at ASDEX Upgrade. An extension to more complex precursors is planed. A different approach has been used for disruption avoidance by injecting ECRH triggered by the loop voltage increase before the disruption. The avoidance of an ongoing density limit disruption has been achieved when the ECRH is deposited at resonant surfaces where MHD modes, such as the m=2/n=1, occur. Present schemes for the mitigation and eventually avoidance of disruptions will be discussed.   

3-D radiation dynamics during gas jet mitigated disruptions on Alcator C-Mod

Demonstrating and understanding disruption mitigation (DM) techniques on present tokamaks is critical to the design of similar tools for ITER and beyond where a near zero-tolerance policy on unmitigated disruptions is envisioned. Efficient mitigation requires the bulk of the thermal and magnetic stored energy to be converted into radiation that is spread uniformly over the walls. Such uniformity has yet to be conclusively demonstrated or understood well enough to confidently design ITER's DM system. Using multiple AXUV diode arrays and spectroscopy, the toroidal and poloidal radiation distribution is investigated on Alcator C-Mod for DM experiments employing a high pressure gas jet. While large, time varying asymmetries in the radiation are observed, the total energy loss is shown to be nearly uniform. Prior to the onset of the thermal quench (TQ) it is shown that both radiation local to the gas jet and axisymmetric radiation are important in shedding the stored energy necessary to destabilize the MHD thought to trigger the TQ. The variation of this pre-TQ period with gas jet species and target plasma are discussed.   

Disruption Mitigation Experiments Carried Out on DIII-D

Disruptions are an important issue for ITER. Major and vertical displacement disruptions in ITER are predicted to generate multi-MeV runaway electron beams (RE) as well as high energy flux to the plasma facing components and high halo currents in the structures that could potentially damage the machine. To mitigate these phenomena, several techniques have been studied on DIII-D: massive gas injection (MGI), external magnetic perturbation, and injection of large shattered cryogenic pellets. MGI, which is proven to mitigate heat fluxes and halo currents, has been tested on DIII-D in terms of impurity radiation toroidal symmetry and optimization of the gas pulse length. Other techniques expected to mitigate the RE have also been tested on DIII-D: deconfinement of the RE using external magnetic fields and collisionally damping the avalanche multiplication process by rapidly increasing the electron density in the core by several orders of magnitude using a new shotgun pellet injector built by ORNL.   

JET experiments on massive gas injection

We present JET experiments on massive gas injection for disruption mitigation. Stable plasmas were terminated by Ne, Ar or their D$_{2}$ mixtures. Reaction time of the method consists of vacuum gas flow and plasma edge cooling. This time depends on the injection and plasma parameters. In the following thermal quench the plasma is rapidly cooled, with up to $50$\% of the energy being radiated. Finally, the plasma current is ohmically dissipated. Since the current decay rate depends on the species, control of the halo currents is possible. Ar and Ne cause runaway generation in the current quench. The runaway conversion exceeds 50\% for Ar, while a lower one is found for Ne and no runaways are observed for mixtures. Runaways can locally heat the wall above $1500^{\circ}$C. For suppression of runaways the fueling efficiency is important. It decreases with the gas mass from 25\% for Ne/D$_2$ to 3\% for Ar. Our results suggest that for ITER: a further increase of the energy radiated in the thermal quench to $90$\% is necessary; the halo current control is feasible; and the runaway suppression remains an issue.   

NSTX ELM Pacing and L-H Threshold Experiments for ITER

We present a summary of recent edge-localized mode (ELM) pacing and L-H power threshold (P$_{LH})$ experiments performed in NSTX in support of ITER. ELM triggering using 3D magnetic perturbations was used to perform pacing during ELM-free H-modes induced by lithium conditioning, mitigating the impurity accumulation typically observed in these conditions. The waveform of the applied field has been tailored to provide high reliability triggering at frequencies of $>$60 Hz to reduce the average ELM size. ELM pacing was also performed using vertical position oscillations, with the ELM frequency increased to $\sim $30 Hz from a natural frequency of $\sim $15 Hz. P$_{LH}$ is reduced by $\sim $50{\%} at low triangularity, and also decreased by $\sim $50{\%} during discharge with thick lithium wall coatings. P$_{LH}$ was observed to increase strongly with plasma current during sustained H-modes. The influence of heating method, non-axisymmetric fields, and magnetic balance on P$_{LH}$ will be presented.   

Impurity Behaviour in the H-mode Edge Barrier: Neoclassical Transport and ELM Flushing

In H-mode discharges an edge transport barrier (ETB) exists and steep gradients are observed in density, temperature of electrons, ions and impurities forming an edge pedestal which is briefly but frequently disturbed and degraded by the occurrence of edge localized modes (ELMs). This edge pedestal sets the boundary conditions for the performance of a fusion plasma. In the actual work, the impurity transport in the ETB is measured by CXRS. It is found that the gradients and the dynamics in impurity densities for (He, C, Ne and Ar) are in agreement with neoclassical transport coefficients at the ETB (i.e. small diffusion and inward convection). Furthermore, a model for the ELM cycle was developed, which treats an impurity mix self-consistently, such that for each time step the neoclassical transport is reevaluated according to the actual impurity densities and gradients. For plasma radii inside the ETB turbulent transport is taken into account by additional diffusion. The impurities are expelled by the ELM and then sputter W at the plasma facing components which enters the plasma after losses by prompt re-deposition are taken into account. The model consistently combines the dominant effects into a picture which agrees with the experimental behavior of W in ASDEX Upgrade.   

Impact of large type I ELMs on plasma radiation in JET

To prevent an unacceptable erosion of divertor targets due to ELMs in ITER, the loss in plasma stored energy should be restricted to $\Delta $W$_{ELM}$ $\sim $ 1~MJ for a single ELM. Only the JET tokamak, thanks to its size, can produce ELMs in the order of 1~MJ with energy densities comparable to those found in ITER. This contribution examines the impact of large Type~I ELMs in high current H-mode JET discharges on plasma radiation and on power load. The production of large Type~I ELMs with $\Delta $W$_{ELM}$ in the range 0.25-1.3~MJ has been demonstrated. The ELMs provoke strong radiation losses, mostly confined to the inner divertor region. Large Type~I ELMs with $\Delta $W$_{ELM~}\ge $0.72~MJ show enhanced radiation losses which are associated with the ablation of carbon layers in the inner divertor. Such large ELMs are often followed by a phase of Type~III ELMs with an increased radiation in the plasma core. In this contribution, the impact of large Type~I ELMs on plasma radiation in JET is summarized and implications of these results for ITER are discussed.   

Simulation of the ITER Rampdown Scenario on \mbox{DIII-D}

Safe termination of a tokamak discharge becomes increasingly important as the energy stored in the plasma and in the poloidal magnetic field increases ($>$750 MJ in ITER). In DIII-D, we simulate the initiation, rampup, and rampdown phases of the proposed ITER discharge scenario, to identify potential issues and problems, to mitigate difficulties and to improve performance. The issues for rampdown are maintaining the separatrix strike point locations on the armored divertor, avoiding additional consumption of transformer flux, and maintaining the density and internal inductance within controllable limits while reducing the current to a negligible level. In these experiments dimensions are reduced by a factor of 3.7 and the current by 10 relative to ITER. We have simulated the ITER rampdown scenario to $I_p <\,$1~MA (equivalent) while controlling the strike points to better tolerance than required. The density falls fast enough to avoid density limits, and $\ell_i$ remains within control limits. The current ramp rate must be increased beyond the ITER reference value to avoid additional flux consumption.   

Recovery of retained fuel through disruptions: implications for ITER

Single discharge retention of injected gas in current tokamaks ranges from 3-50{\%} of the injected gas raising concerns of excessive tritium retention for ITER. As part of a recent study of fuel retention in Alcator C-Mod it was found that, averaged over a run period, the fuel retained, normalized by ion fluence to divertor surfaces, was 100-1000x lower than for the same normalized retention in a single, non-disruptive discharge. Analyzing all disruptions for a run campaign it was found that the average disruption during plasma current flattop (15{\%} of all discharges), led to fuel recovery 5-6x that retained in a single, non-disruptive discharge; Disruptions appear to remove all the fuel retained in non-disruptive discharges. Analysis of the fuel recovery dependence on disruption characteristics gave a scaling linear in plasma thermal energy and as the square of the magnetic energy. In this presentation we review the above information, discuss the role of the high-Z plasma facing components, and examine possible scalings to the use of disruptions for fuel recovery in ITER.   

Physics Requirements for the ITER Plasma Control System

The Plasma Control System (PCS) on ITER will control the evolution of the plasma parameters necessary to operate ITER throughout all phases of the discharge including plasma termination following off-normal events and plant system faults. The PCS is composed of six closely coupled subsystems that control specific physical quantities comprising: 1) wall conditioning and tritium removal, 2) plasma axisymmetric magnetic control, which includes plasma initiation, inductive plasma current, position, and shape control, 3) power and particle flux control to the first wall and divertor, 4) plasma kinetic control, including fuelling, non-inductive plasma current, plasma pressure and fusion burn control, 5) non-axisymmetric stability control, which includes sawteeth, edge localized modes (ELMs), neoclassical tearing modes (NTMs), error fields and resistive wall modes (RWMs), Alfven eigenmodes, etc., and 6) disruption mitigation and controlled plasma termination. The PCS is the first of a three layered system for machine protection that will decide the course of action to take following off-normal events or plant system faults during plasma operation together with the Central Interlock System (CIS) and the Central Safety System (CSS) for safety related events. Only if the PCS cannot control the plasma within specified operating limits and conditions will the disruption mitigation system by triggered by the CIS.   

Development of a Spatially Resolving X-Ray Crystal Spectrometer (XCS) for Measurement of Ion-Temperature (Ti) and Rotation-Velocity (v) Profiles in ITER

An imaging XCS is being developed as a US-ITER activity for Doppler measurement of Ti and v profiles of impurities (W, Kr, Fe) with $\sim $7 cm (a/30) and 10 ms resolution in ITER; the instrument is modeled after the very successful imaging XCS on the C-Mod tokamak, which was developed through a PPPL-MIT collaboration and is a prototype for the ITER instrument. The imaging XCS uses a spherically bent crystal and 2d imaging x-ray detectors to achieve high spectral resolving power ($>$6000) in the horizontal dimension and spatial imaging vertically. Two XCS arrays will measure Ti and both poloidal and toroidal rotation velocity profiles. Simultaneous measurement of many spatial views permits tomographic inversion for determination of local parameters. The design of the ITER instrument, predictions of performance, and interesting measurements from C-Mod will be presented.   

On the JET ITER-Like ICRF antenna and implications for the ICRF system for ITER

A new ``ITER-Like'' Ion Cyclotron Resonance Frequency (ICRF) antenna was installed on the JET tokamak in 2007 and extensively operated on plasma since May 2008 for a wide range of conditions (frequencies: 33, 42 and 47 MHz, L- and ELMy H-mode plasmas, antenna strap - plasma separatrix distances from 9 to 17 cm). Aspects relating to the potential performance and design of the ITER system, will be discussed: (i) the wave coupling performance and validation of the TOPICA modelling code used to predict the coupled power in ITER; (ii) the operation at high coupled power density (up to 6.2 MW/m$^{2}$ in L-mode, 4.1 MW/m$^{2}$ in H-mode) and high RF voltage on the antenna structure (up to 42 kV); (iii) the coupling of ICRF power during fast variations (ms) in coupling occurring during ELMs and (iv) antenna control in the presence of high mutual coupling between antenna straps.   

Simulations of ITER-like Discharges on Alcator C-Mod

Discharges in Alcator C-Mod are being used to examine the plasma evolution under ITER-like conditions in order to benchmark calculations for ITER. These include equivalent current diffusion timescales, q95, betaN, lower single null and ITER shape, at ITER's toroidal field of 5.4 T, and with ICRF heating. Rampup experiments with both ohmic and with ICRF heating were performed. Simulations, with the time-dependent free-boundary plasma evolution code TSC, of these discharges showed that predictions of the Coppi-Tang L-mode energy transport gave rise to too low a temperature near the plasma edge, which consequently caused the current profile to be too peaked. The model, based on profile consistency, was adjusted to broaden the electron temperature profile producing much better agreement with the temperature profile, sawtooth onset, and li. Simulations with GLF23, which has only a boundary condition, showed some difficulties in the L-mode phase. ICRF and LH in the rampup phase of C-Mod was examined with simulations showing that volt-second savings were resistive, similar to the result from ITER simulations. In rampdown experiments, the transition from H-mode to L-mode and the ramp rate were controlled but demonstrated some complex H-L-H behavior with ICRF power reductions. Supported by DE-AC02-09CH11466, DE-FC02-99ER54512.   

Progress in Developing ITER Operational Scenarios on DIII-D

The DIII-D program has initiated an effort to provide experimental evaluations of four ITER operational scenarios: baseline ELMy H-mode, advanced inductive, hybrid, and steady-state. Discharges in 2008 matched the anticipated ITER design for plasma shape, aspect ratio and value of I/aB, with size reduced by a factor of 3.7, while matching key performance targets for $\beta_N$ and $H_{98y2}$. In 2009, attention has focused on improving the match to anticipated ITER parameters: Baseline scenario plasmas have been operated with reduced densities to match the anticipated ITER edge pedestal collisionality, while maintaining target values for $\beta_N$ and $H_{98y2}$. These plasmas have enabled a demonstration of 2/1 NTM mode suppression at low $q_{95}$ using ECCD, as planned for ITER. Additional experiments are planned to investigate operation of baseline scenario plasmas with low torque input (low rotation), as well as rf-dominated operation. Comparison of experimental profiles to those from transport modeling will be shown.