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Motivated by the recent experimental observation that plasma stability can be improved by external magnetic perturbations, the general problem of plasma response to external magnetic perturbations is investigated. Different (vacuum, ideal and resistive) plasma response models are considered and compared. Plasma response, in experiments where stabilization was achieved, is obtained through computation using the MARS-F code, with a plasma model that includes both plasma resistivity and rotation. The resultant magnetic field line stochasticity is much reduced from that obtained formerly using the vacuum plasma model. This reduced stochasticity is more consistent with the favourable experimental observation of enhanced stability. Examples are given for the response of an ITER plasma to perturbations generated by the correction coils; and the response of a plasma to external coils (antenna) up to the Alfvén frequency.

A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach {approximately} 1 Gw/m{sup 2} along the magnetic fieldlines and > 10 MW/m{sup 2} on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator inmore » a divertor program is not explored in this report.« less

Control simulations of ITER startup using 2D free-boundary equilibrium and 1D transport codes rely on accurate predictions of the electron and ion temperature profiles that determine the electrical conductivity and pressure profiles during the current rise. We present results of validation studies that apply the transport model used by the ITER team to DIII-D discharge evolution and compare predictions with data from similarity experiments. Results presented here detail difficulties and sensitivities associated with the modelling of time-dependent current profile evolution required to asses performance of the poloidal-field coil system and controllers on ITER.

Understanding the dynamics of plasma startup and termination is important for present tokamaks and for predictive modeling of future burning plasma devices such as ITER. We report on experiments in the DIII-D tokamak that explore the plasma startup and rampdown phases and on the benchmarking of transport models. Key issues have been examined such as plasma initiation and burnthrough with limited inductive voltage and achieving flattop and maximum burn within the technical limits of coil systems and their actuators while maintaining the desired q profile. Successful rampdown requires scenarios consistent with technical limits, including controlled H-L transitions, while avoiding vertical instabilities, additional Ohmic transformer flux consumption, and density limit disruptions. Discharges were typically initiated with an inductive electric field typical of ITER, 0.3 V/m, most with second harmonic electron cyclotron assist. A fast framing camera was used during breakdown and burnthrough of low Z impurity c...

Single pass absorption experiments using fundamental, O-mode ECH waves have been performed in the Microwave Tokamak Experiment (MTX) and found to be consistent with linear absorption theory, as predicted for the ECH source which is a 140 GHz, 400 kW CW gyrotron. These experiments provide the basis for comparison with future Free Electron Laser (FEL) experiments in which the absorption is predicted to be in the non-linear regime for P{sub FEL} {ge} 1 GW. The gyrotron power is coupled into a quasi-optical transmission line to the tokamak by a Vlasov antenna. The input power is measured by calorimeters located just outside the injection port and on the vacuum wall opposite the port. For plasma shots, the total transmitted power and its profile are measured. Temporal analysis of the thermistor signals compared with a heat diffusion model for the tiles gives the spatial profile of transmitted power. These measurements are compared with the stored energy increase by diamagnetism measurements, 3-D vacuum fields calculations and ray-tracing calculations of the transmitted power. 5 refs., 4 figs.

DIII-D experiments have investigated ITER startup scenarios, including an initial phase where the plasma was limited on low field side poloidal bumper limiters. In addition, l i feedback control has been tested with the goal of producing discharges in ITER within the capabilities of the poloidal field coil set and favourable to the intended mode of operations in the subsequent constant current (flattop) phase. These discharges have been modelled using the Corsica free boundary equilibrium code. High performance hybrid scenario discharges (βN = 2.8, H 98,y2 = 1.4) and ITER H-mode baseline discharges (βN > 1.6, H 98,y2 = 1–1.2) have been obtained experimentally in an ITER similar shape after the ITER-relevant startup. Studies have been initiated to develop a reliable scenario for exiting the burn phase and ramping down the plasma current in ITER without disruptions.

The Sustained Spheromak Physics Experiment (SSPX) [Hooper et al., Nuclear Fusion 39, 863 (1999)] explores the physics of efficient magnetic field buildup and energy confinement, both essential parts of advancing the spheromak concept. Extending the spheromak formation phase increases the efficiency of magnetic field generation with the maximum edge magnetic field for a given injector current (B∕I) from 0.65T∕MA previously to 0.9T∕MA. We have achieved the highest electron temperatures (Te) recorded for a spheromak with Te>500eV, toroidal magnetic field ∼1T, and toroidal current (∼1MA) [Wood et al., “Improved magnetic field generation efficiency and higher temperature spheromak plasmas,” Phys. Rev. Lett. (submitted)]. Extending the sustainment phase to >8ms extends the period of low magnetic fluctuations (

Disruptions are observed in the Microwave Tokamak Experiment, MTX (nee Alcator C), over a wide range of plasma parameters. Indeed, disruptions often occur far from the boundaries of the operating space as defined by Hugill and l{sub i}-q plots. Despite this, the general behavior during the disruptive process is generally similar whatever the operating parameters. This report will describe one disruption in detail in order to provide a detailed anatomy of the event.

Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasing power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by th...

The present performance limits of tokamak discharges are strongly coupled to the stability and transport properties of the edge plasma. Both experimental and modeling efforts have shown a clear connection between the edge pressure pedestal height and core plasma confinement. The key to understanding the stability and performance limits of the pedestal revolves around an accurate knowledge of the plasma current in this region. Using the Zeeman effect in an injected 30keV lithium beam, we have measured the currents in the edge of the DIII-D [J. L. Luxon, Nucl. Fusion 42, 6114 (2002)] tokamak for various confinement modes. This method of determining j(r) is insensitive to the large electric fields which coexist in the pedestal region and which complicate motional Stark effect measurements. For the high confinement cases, where substantial pedestal pressures exist, we find large (~MA∕m2), localized (ΔR∼1–2cm) currents in the pedestal region, located near the maximum in the pressure gradient. These values are ...

Results are presented from comparisons of modeling and experiment in studies to assess the best choices of safety factor q profile, pressure profile, and discharge shape for high β, steady-state, noninductive advanced tokamak operation in the DIII-D device [J. L. Luxon, Nucl. Fusion 42, 614 (2002)]. These studies are motivated by the need for high qminβN to maximize the self-driven bootstrap current while maintaining high toroidal β to increase fusion gain. Modeling shows that increases in the normalized beta βN stable to ideal, low toroidal mode number (n=1,2), instabilities can be obtained through broadening of the pressure profile and use of a symmetric double-null divertor shape. Experimental results are in agreement with this prediction. The general trend is for qminβN to increase with the minimum q value (qmin) although βN decreases as qmin increases. By broadening the pressure profile, βN≈4 is obtained with qmin≈2. Modeling of equilibria with near 100% bootstrap current indicates that operation wit...

Thermal-barrier experiments in the Tandem Mirror Experiment Upgrade (TMX-U) are reported, along with progress made at the Lawrence Livermore National Laboratory in plasma confinement and central-cell heating. Thermal barriers in TMX-U improved axial confinement by two orders of magnitude over a limited range of densities, compared with confinement in single-cell mirrors at the same ion temperature. It is shown that central-cell radial nonambipolar confinement scales as neoclassical theory and can be eliminated by floating the end walls. Radial ambipolar losses can also be measured and reduced. The electron energy balance is improved in tandem mirrors to near classical, resulting in T/sub e/ up to 0.28 keV. Electron cyclotron heating (ECH) efficiencies up to 42%, with low levels of electron microinstability, were achieved when hot electrons in the thermal barrier were heated to average betas as large as 15%. The hot-electron distribution was measured from X-rays and is modeled by a Fokker-Planck code that includes heating from cavity radio-frequency (RF) fields. >

Since the start of TMX-U operations we have observed weak oscillations with a frequency approximately twice the ion-cyclotron frequency at the end-cell midplane. Recent results show that this instability is driven at the sloshing-ion outer turning point. We find that the sloshing-ion lifetime scales inversely with the instability amplitude. The sensitivity of this instability to the cold-ion density shows that this is a loss-cone instability. These results have important implications for mirror fusion experiments as well as solar and space plasmas.

Low-frequency (f

A conceptual design of a D—T fusion facility for the continuous production of 14 MeV neutron wall loading from 5 to 10 MW/m 2 at the plasma surface is presented. In this design, D—T neutrons are produced in a linear, two-component plasma formed by neutral beam irradiation of a fully ionized warm plasma target. The beam energy, which is deposited in the center, is transferred to the warm plasma mainly by electron drag and is conducted along the target plasma column to end regions where it is absorbed in neutral gas at high pressure. The target plasma is operated in a regime where electron thermal conduction along the column is the controlling energy-loss process. The loss rate is minimized by adjusting the diameter and length of the plasma column. A substantial gradient in T e along the column results in recombination of the plasma to gas in the end regions before impact on the end walls. The resultant hot gas is cooled by contact with large-area heat exchangers. In this way, the large steady-state heat load from the injected neutral beams is diffused and removed at tolerable heat flux levels. The reacting plasma is essentially an extrapolation of the 2XIIB high-β plasma to higher magnetic field, ion energy, and density.

The design and performance of a relatively low-cost, plasma-based, 14-MeV D-T neutron source for accelerated end-of-life testing of fusion reactor materials are described in this article. An intense flux (up to 5×1018 n/m2·s) of 14-MeV neutrons is produced in a fully-ionized high-density tritium target (n e ≈ 3×1021 m−3) by injecting a current of 150-keV deuterium atoms. The tritium plasma target and the energetic D+ density produced by D0 injection are confined in a column of diameter ⩽ 0.16 m by a linear magnet set, which provides magnetic fields up to 12 T. Energy deposited by transverse injection of neutral beams at the midpoint of the column is conducted along the plasma column to the end regions. Longitudinal plasma pressure in the column is balanced by neutral gas pressure in the end tanks. The target plasma temperature is about 200 eV at the beam-injection position and falls to 5 eV or less in the end region. Ions reach the walls with energies below the sputtering threshold, and the wall temperature is maintained below 740 K by conventional cooling technology.

A tandem-mirror configuration with an axisymmetric central cell, similar to the geometry of MARS (Mirror Advanced Reactor Study) or the Kelley-TDF mode of MFTB-B, can be generated by inserting a 6-tesla, throttle coil in each end of the TMX-U central cell. The throttle coil geometry of TMX-U will test the physics issues associated with axisymmetric tandem-mirror reactors, such issues as: (1) increased radial confinement time for central-cell ions confined by axisymmetric mirror cells and electrostatic potentials; (2) theoretical limits set by the trapped particle instability for the required passing density between the central cell and the end-cell anchor; and (3) pumping of trapped particles within the thermal barrier and transition regions with methods other than neutral beams. The central-cell plasma parameters for the throttle coil geometry are evaluated for two operating points. The first requires heating hardware (neutral beams and ECRH) and vacuum performance at the TMX-U proposal level, yielding plasma parameters, central-cell betas, and plasma confinement exceeding those of the original TMX-U proposal. The second operating point, requiring approximately half the ECRH end-cell performance of the first, is predicted to equal the beta and to exceed the plasma pressure and confinement time of the central cell in the standardmore » TMX-U geometry.« less

This report examines each of the instruments now on the Tandem Mirror Experiment (TMX-U) and identifies significant tasks required to use them on either GAMMA 10 or on the Mirror Fusion Test Facility (MFTF-B).

TMX experimental data on ambipolar potential control and on the accompanying electrostatic confinement are reported. New results on the radial dependence of the central-cell confining potential are given. Radial and axial particle losses as well as scaling of the central-cell axial confinement are discussed.

This report summarizes initial results from TMX-U carried out from June through August 1982. In these successful experiments we operated TMX-U in the sloshing-ion mode. We generated sloshing ions, measured improved energy confinement, and observed improved microstability compared to TMX. The experiments operated about as we expected and we are pleased with the results. During this period many additional achievements were also recorded. The magnetically confined sloshing ions constitute one of the two ingredients needed to build a thermal barrier. The second ingredient consists of magnetically confined electrons, which will be studied in the next series of TMX-U experiments using microwave heating of the electrons. Later, the hot ions and electrons will be combined to form thermal barriers.

The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma performance must apply multiple control functions simultaneously with a limited number of actuators. A sophisticated shared actuator management system is being designed to prioritize the goals that need to be controlled or weigh the algorithms and actuators in real-time according to dynamic control needs. The underlying architecture will be event-based so that many possible plasma or plant system events or faults could trigger automatic changes in the control algorithms or operational scenario, depending on real-time operating limits and conditions. (C) 2014 ITER Organization. Published by Elsevier B.V. All rights reserved.