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

A small relaxation of the axisymmetric magnetic field of a tokamak into a non-axisymmetric three-dimensional (3D) configuration can be effective to control magnetohydrodynamic instabilities, such as edge-localized modes. However, a major challenge to the concept of 3D tokamaks is that there are virtually unlimited possible choices for a 3D magnetic field, and most of them will only destabilize or degrade plasmas by symmetry breaking. Here, we demonstrate the phase-space visualization of the full 3D field-operating windows of a tokamak, which allows us to predict which configurations will maintain high confinement without magnetohydrodynamic instabilities in an entire region of plasmas. We test our approach at the Korean Superconducting Tokamak Advanced Research (KSTAR) facility, whose 3D coils with many degrees of freedom in the coil space make it unique for this purpose. Our experiments show that only a small subset of coil configurations can accomplish edge-localized mode suppression without terminating the discharge with core magnetohydrodynamic instabilities, as predicted by the perturbative 3D expansion of plasma equilibrium and the optimizing principle of local resonance. The prediction provided excellent guidance, implying that our method can substantially improve the efficiency and fidelity of the 3D optimization process in tokamaks. A theoretical and numerical approach, validated by experiments at the KSTAR facility, shows how magnetohydrodynamic instabilities in tokamak plasmas can be efficiently controlled by a small relaxation of the confining field into a 3D configuration.

3D field phase-space control in tokamak plasmas

An electron-beam ion trap (EBIT) is used to measure extreme ultraviolet spectra between 10 and 25 nm from highly charged ions of tungsten with an open $3d$ shell (W xlviii through W lvi). We found that almost all strong lines are due to the forbidden magnetic-dipole ($M1$) transitions within $3{d}^{n}$ ground configurations. A total of 37 previously unknown spectral lines are identified using detailed collisional-radiative (CR) modeling of the EBIT spectra. A level-merging scheme for compactification of rate equations is described. The CR simulations for Maxwellian plasmas show that several line ratios involving these $M1$ lines can be used to reliably diagnose temperature and density in hot fusion devices.

Spectroscopy of diagnostically-important magnetic-dipole lines in highly-charged 3d n ions of tungsten

Nuclear analysis of the ITER torus cryopumps

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

Principal Physics Developments Evaluated in the ITER Design Review

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Nuclear analysis of the ITER torus cryopumps

Principal Physics Developments Evaluated in the ITER Design Review