G. Vayakis

Plasma diagnostics for INTER-FEAT

A comprehensive set of diagnostics is planned for ITER-FEAT. The design of the systems is a substantial technical challenge because of the combination of the harsh environment with the demanding measurement requirements. Through a combination of careful choice of technique, materials, and design, supported by dedicated research and development, an extensive diagnostic set has been developed. The designs are based on existing techniques as much as possible but in some cases novel approaches have to be adopted. In the article the requirements for measurements are outlined and representative diagnostic designs are presented. Key issues in the design are identified and areas requiring further development are highlighted.

Reflectometry on ITER

Reflectometry will be used on ITER to measure the density profile in the main plasma and divertor regions, and to measure the plasma position and shape in order to provide a standby reference for the magnetic diagnostics in long pulse discharges. The high temperatures of the ITER core and the resultant significant relativistic downshift of the second-harmonic electron cyclotron absorption imply that both low-field side O-mode and high-field side lower cut-off (X−l mode) systems are required to access the full plasma profile. A low-field side upper cut-off (X−u mode) system will also be required for measurements of the scrape-off layer. For measurements of the plasma position and shape, an O-mode system is optimum due to the large range of magnetic field along the plasma periphery and the wide range of possible plasma configurations achievable on ITER. A robust real-time calibration technique of the whole transmission line is required. It is likely that an accurate estimate of the position of the plasma will require the simultaneous use of signals from the profile reflectometer. For the divertor, profiles with peak densities in the range 1019–1022/m3 are to be measured with a target resolution of 3 mm. The large density range will necessitate the use of more than one system. Installing these reflectometers on ITER incurs additional difficulties such as the routing of the millimetre wave radiation around the complicated first wall and divertor structures and design of antennas able to operate through the first wall and blanket.

Technical issues associated with the control of steady-state tokamaks

Note: Proceedings of the 2nd IAEA Technical Committee Meeting on Steady-State Operation of Magnetic Fusion Devices - Plasma Control and Plasma Facing Components, Fukuoka, Japan, October 1999 Reference CRPP-CONF-2000-044 Record created on 2008-05-13, modified on 2017-05-12

Overview of the ITER Diagnostic System

The individual diagnostics which constitute the ITER Diagnostic System are outlined and the present state of development of the designs is summarised. The results of an assessment of the overall performance of the System are presented and the areas where the probable performance falls below the target specifications are identified. The design and RsD plans which are in place to address the shortcomings are outlined.

Irradiation Tests on ITER Diagnostic Components

Radiation effects on key components of diagnostic systems expected to be subjected to high neutron and gamma fluxes and fluences are being examined in irradiation tests to evaluate and establish an ITER-relevant database to support the design. A comprehensive irradiation database has been accumulated and permits conclusions to be drawn on the application of these components in ITER. The design studies on prototypical assemblies of diagnostic components are continuing based on the irradiation data bases, neutronics calculations for evaluating irradiation environment of diagnostic components and required specification of diagnostic systems. These studies will aid the recognition of detailed requirements of diagnostic systems leading to more specific irradiation tests on diagnostic components.

Role and Requirements for Plasma Measurements on ITER

Measurement of plasma and key first wall parameters will have three main roles on ITER. Some of the measurements will be used in real time to prevent the on-set of conditions which could potentially damage the first wall and other in-vessel components (machine protection); others will be used in real-time feedback control loops to control the value of key parameters at values required for specific plasma performance (plasma control); while others will be used to evaluate the plasma performance and to provide information on key phenomena which may limit ITER performance (physics studies). The measurements of some parameters may contribute to all three roles although the requirements on the measurements (accuracies, resolutions etc.) may be different depending on the role.

ITER physics program and implications for plasma measurements

Key objectives of the first ten years of ITER operation are the investigation of the physics of burning plasmas and the demonstration of long-pulse ignited plasma technologies. These include studies of plasma confinement and stability, divertor operation, disruption mitigation and control, noninductive current drive, and steady state operation under conditions when the plasma is heated predominantly by alpha particles. The ITER operational plan envisages two and a half years for commissioning and initial operation with hydrogen plasmas at up to 100 MW of auxiliary heating power when initial tests of divertor operation and evaluation of disruption effects will be made. In order to meet the operational and programmatic goals, it will be necessary to make a wide range of plasma measurements. In this article the preliminary operational plan and physics program are presented and the implications for plasma measurements are outlined.

Operation and control of ITER plasmas

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that facilitate operation and control of ignited and/or high?Q DT plasmas are presented. Control methods based upon straightforward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for ITER plasma control with fusion powers of up to 1.5?GW and burn durations of ? 1000?s. Examples of simulations of key plasma control functions, including plasma magnetic configuration control and fusion burn (power) control, are given. The prospects for the creation and control of steady state plasmas sustained by non-inductive current drive and bootstrap current are also discussed.

INTEGRATION OF VACUUM COUPLED DIAGNOSTICS

A special class of ITER diagnostics consists of those which extend the primary or tokamak vacuum outside the cryostat wall. This diagnostic set consists of the x-ray crystal spectrometers (XCS) [1], the vacuum ultra-violet spectrometers (VUV) [2], the neutral particle analysers (NPA) [3] and the low-field-side microwave reflectometers [4]. Two XCS systems are included, a high resolution five radial channel array system (XCS-A) designed to view specific spectral lines for temperature and rotation velocity profiles and a survey instrument package (XCS-S) to monitor a broad spectrum for identifying impurity concentrations. The vacuum vessel rough-out line is also located at this port.

ITER physics program and implications for plasma measurements (abstract)

Key objectives of the first ten years of ITER operation are the investigation of the physics of burning plasmas and the demonstration of long-pulse ignited plasma technologies. These include studies of plasma confinement and stability, divertor operation, disruption mitigation and control, noninductive current drive, and steady state operation under conditions when the plasma is heated predominantly by alpha particles. The ITER operational plan envisages two and a half years for commissioning and initial operation with hydrogen plasmas at up to 100 MW of auxiliary heating power when initial tests of divertor operation and evaluation of disruption effects will be made. In order to meet the operational and programmatic goals, it will be necessary to make a wide range of plasma measurements. In this article the preliminary operational plan and physics program are presented and the implications for plasma measurements are outlined.