Wall conditioning of fusion devices involves removal of desorbable hydrogen isotopes and impurities from interior device surfaces to permit reliable plasma operation. Techniques used in present devices include baking, metal film gettering, deposition of thin films of low-Z material, pulse discharge cleaning, glow discharge cleaning, radio frequency discharge cleaning, and in situ limiter and divertor pumping. Although wall conditioning techniques have become increasingly sophisticated, a reactor scale facility will involve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficient removal of tritium incorporated into co-deposited layers on plasma facing components and their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized. Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactor scale plasmas will rely on the same principles, but will face additional challenges, exemplified by the ITER/FDR design. The absolute positioning of outermost flux surface and divertor strike points will have to be precise and reliable in view of the high heat fluxes at the separatrix. Long pulses will require minimal control actions, to reduce accumulation of AC losses in superconducting PF and TF coils. To this end, more complex feedback controllers are envisaged, and the experimental validation of the plasma equilibrium response models on which such controllers are designed is encouraging. Present simulation codes provide an adequate platform on which equilibrium response techniques can be validated. Burning plasmas require kinetic control in addition to traditional magnetic shape and position control. Kinetic control refers to measures controlling density, rotation and temperature in the plasma core as well as in plasma periphery and divertor. The planned diagnostics (Chapter?7) serve as sensors for kinetic control, while gas and pellet fuelling, auxiliary power and angular momentum input, impurity injection, and non-inductive current drive constitute the control actuators. For example, in an ignited plasma, core density controls fusion power output. Kinetic control algorithms vary according to the plasma state, e.g. H- or L-mode. Generally, present facilities have demonstrated the kinetic control methods required for a reactor scale device. Plasma initiation - breakdown, burnthrough and initial current ramp - in reactor scale tokamaks will not involve physics differing from that found in present day devices. For ITER, the induced electric field in the chamber will be ~0.3V??m-1?- comparable to that required by breakdown theory but somewhat smaller than in present devices. Thus, a start-up 3MW electron cyclotron heating system will be employed to assure burnthrough. Simulations show that plasma current ramp up and termination in a reactor scale device can follow procedures developed to avoid disruption in present devices. In particular, simulations remain in the stable area of the li-q plane. For design purposes, the resistive V?s consumed during initiation is found, by experiments, to follow the Ejima expression, 0.45?0 RIp. Advanced tokamak control has two distinct goals. First, control of density, auxiliary power, and inductive current ramping to attain reverse shear q profiles and internal transport barriers, which persist until dissipated by magnetic flux diffusion. Such internal transport barriers can lead to transient ignition. Second, combined use poloidal field shape control with non-inductive current drive and NBI angular momentum injection to create and control steady state, high bootstrap fraction, reverse shear discharges. Active n = 1 magnetic feedback and/or driven rotation will be required to suppress resistive wall modes for steady state plasmas that must operate in the wall stabilized regime for reactor levels of ? ? 0.03.

http://iopscience.iop.org/article/10.1088/0029-5515/39/12/308/pdf

Chapter?8: Plasma operation and control

Hydrogen and helium recycling in tokamaks with carbon walls

The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-6O and DIE-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables.

https://iopscience.iop.org/article/10.1088/0029-5515/28/4/007/pdf

Plasma diagnostics on large tokamaks

Graphite as plasma facing material

Transport simulations of ohmically heated TFTR experiments with recently developed microinstability based, profile consistent models for the anomalous thermal diffusivities, χe and χi, give good agreement with experimental data. The steady state temperature profiles and the total energy confinement times, τE, were found to agree for each of the Ohmic TFTR experiments simulated, including three high radiation cases and two plasmas fuelled by pellet injection. Both collisional and collisionless models are tested. The trapped electron drift wave microinstability model results are consistent with the thermal confinement of large plasma Ohmic experiments on TFTR. It is also found that transport due to the profile consistent model based on toroidal ion temperature gradient (ηi) mode transport can cause saturation in τE at the highest densities comparable to that observed on TFTR and equivalent to a neoclassical anomaly factor of three. Predictions based on stabilized ηi mode driven ion transport are found to be in agreement with the enhanced global energy confinement times for pellet fuelled plasmas.

http://iopscience.iop.org/0029-5515/27/12/003/pdf/0029-5515_27_12_003.pdf

Team Tftr Team

Papers

Plasma-material interactions in TFTR