China Fusion Engineering Test Reactor (CFETR) is a tokamak reactor; one design option under the consideration of the China National Integration Design Group employs superconducting magnets. The fusion power is at the range of 50-200 MW and the duty cycle (or burning time) was envisioned as 30%-50%. The plasma current will be 10 MA and the major and minor radii are 5.7 and 1.6 m, respectively. The concept engineering design including the magnet system, vacuum vessel (VV) system, and maintenance method has been carried out in the past years. The toroidal magnetic field strength at R0 is 5 T and the maximum flux swing provided by central solenoid winding will be 160 VS. The main design work, including the electromagnetic analysis of the magnet system, has already been carried out. This paper mainly probes into the VV design and optimization based on three types of maintenance ports. Furthermore, the maintenance method, counted as one of the most important design studies, is presented specifically in this paper. It includes the design of the maintenance ports and the remote handling system design, and so on. The next design stage will mainly include mechanical design, conductor stability, different types of divertor system design, and so on.

Concept Design of CFETR Tokamak Machine

Finalizing the ITER divertor design: The key role of SOLPS modeling

In spite of being highly suited for advanced plasma performance operation of tokamaks, as demonstrated over at least two decades of fusion plasma research, carbon is not currently considered as an integrating element of the plasma-facing components (PFCs) for the active phase of ITER. The main reason preventing its use under the very challenging scenarios foreseen in this phase, with edge-localized modes delivering several tens of MW m−2 to the divertor target every second or less, is the existing concern about reaching the tritium inventory value of 1000 g used in safety assessments in a time shorter than the projected lifetime of the divertor materials eroded by the plasma, set at 3000 shots. Although several mechanisms of tritium trapping in carbon components have been identified, co-deposition of the carbon radicals arising from chemically eroded chlorofluorocarbons in remote areas appears to play a dominant role. Several possible ways to keep control of the tritium build-up during the full operation of ITER have been put forward, mostly based on the periodic removal of the co-deposits by chemical (thermo-oxidation, plasma chemistry) or physical (laser, flash lamps) methods. In this work, we review the techniques for the inhibition and removal of tritium-rich co-deposits based on the strong chemical reactivity of some N-bearing molecules with carbon. The integration of these techniques into a possible scheme for tritium inventory control in the active phase of ITER under carbon-based PFCs with minimum down-time is discussed and the existing caveats are addressed.

http://iopscience.iop.org/article/10.1088/0963-0252/22/3/033001/pdf

Tritium inventory control during ITER operation under carbon plasma-facing components by nitrogen-based plasma chemistry: a review

The ITER divertor design is the culmination of years of physics and engineering effort, building confidence that this critical component will satisfy the requirements and meet the challenge of burning plasma operation. With 54 cassette assemblies, each weighing ~9 tonnes, nearly 3900 actively cooled high heat flux elements rated to steady-state surface power flux densities of 10 MW m−2 and a total of ~60 000 carbon fibre composite monoblocks and ~260 000 tungsten monoblocks/flat tiles, the ITER divertor will be the largest and most advanced of its kind ever constructed. Both the ITER Design Review and subsequent follow-up activities have led to a number of modifications to the device, including the divertor design, significantly improving ITER's operational flexibility. This paper outlines the salient features of the final divertor design, with emphasis on the physics rationale that has defined the design choices and on the performance of the resulting configuration.

https://iopscience.iop.org/article/10.1088/0031-8949/2009/T138/014001/pdf

Status and physics basis of the ITER divertor

The EAST research program aims to demonstrate steady-state long-pulse advanced high-performance H-mode operations with ITER-like poloidal configuration and RF-dominated heating schemes. Since the 2014 IAEA FEC, EAST has been upgraded with all ITER-relevant auxiliary heating and current drive systems, enabling the investigation of plasma profile control by the coupling/integration of various auxiliary heating combinations. Fully non-inductive steady-state H-mode plasma (H 98,y2 > 1.1) was extended over 60 s for the first time with sole RF heating plus good power coupling and impurity and particle control. By means of the 4.6 GHz and 2.45 GHz LHCD systems, H-mode can be obtained and maintained at relatively high density, even up to n e ~ 4.5 × 1019 m−3, where a current drive effect is still observed. Significant progress has been achieved on EAST, including: (i) demonstration of a steady-state scenario (fully non-inductive with V loop ~ 0.0 V at high β P ~ 1.8 and high-performance in upper single-null (e ~ 1.6) configuration with the tungsten divertor; (ii) discovery of a stationary H-mode regime with no/small ELM using 4.6 GHz LHCD, and; (iii) achievement of ELM suppression in slowly rotating H-mode plasma with n = 1 and 2 RMP compatible with long-pulse operations. The new advances in scenario development provide an integrated solution in achieving long-pulse steady-state operations on EAST.

https://iopscience.iop.org/article/10.1088/1741-4326/aa7861/pdf

Overview of EAST experiments on the development of high-performance steady-state scenario

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

/pdf/principal-physics-developments-evaluated-in-the-iter-design-35npzpn1e1.pdf

J. J. Cordier

Papers

Principal Physics Developments Evaluated in the ITER Design Review