V. Mukhovatov

Plasma Control Requirements and Concepts For ITER

Plasma control requirements for the International thermonuclear Experimental Reactor (ITER) are identified, and an overview of proposed ITER plasma control concepts is presented. ITER will operate with a burning deuterium-tritium plasma to produce 1.5 GW of fusion power for durations of 1000 s or more. Key plasma control requirements to achieve these objectives encompass (a) plasma scenario and sequencing: plasma initiation, current rampup, divertor formation, auxiliary heating, ignition and burn, deignition (fusion power shutdown), and current rampdown and termination; (b) plasma magnetics control: plasma current and shape (R 0 , a, κ, δ) versus time, plus control of critical plasma-to-first-wall clearance gaps, including ion-cyclotron coupling gap and divertor magnetic configuration, during the diverted heating/ignition/burn/deignition phase of the plasma scenario; (c) plasma kinetics and divertor control: core plasma density and/or fusion power, core impurity content and/or radiated power fraction; core profile control (auxiliary heating and/or current drive), and divertor control (pumping, in-divertor gas and/or impurity injection and magnetic configuration optimization for divertor performance); and (d) fast plasma shutdown: fusion power and current shutdown by means of impurity injection. Physics and hardware concepts are presented as to how these plasma control junctions will be implemented. Diagnostic measurements needed for plasma control are summarized. The relationship of plasma control to machine protection and public safety is also addressed.

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.

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.

RTO/RC ITER Plasma Performance: Inductive and Steady-State Operation

The plasma performance in two design options of the reduced-technical objectives/reduced cost (RTO/RC) ITER, i.e. IAM (intermediate aspect ratio machine) and LAM (low aspect ratio machine) is analysed. It is shown that Q = Pfus/Paux~10 can be obtained in both options at inductively driven ELMy H-mode operation. The operation domain in LAM is found to be marginally larger than that in IAM. The non-inductive operation with Q≈5 will be possible in both machines, provided a large amount of power with a high current drive efficiency is applied, or substantial improvement of the energy confinement time relative to the ELMy H-mode (HH = 1.2-1.4) is obtained. The required values of HH and βN are marginally smaller in IAM. The IAM-like machine, ITER-FEAT (fusion energy advanced tokamak), proposed for a detailed engineering design is discussed in brief.