Challenges for plasma-facing components in nuclear fusion

This paper summarizes the latest progress in the ITER blanket system design as it proceeds through its final design phase with the Final Design Review planned for Spring 2013. The blanket design is constrained by demanding and sometime conflicting design and interface requirements from the plasma and systems such as the vacuum vessel, in-vessel coils and blanket manifolds. This represents a major design challenge, which is highlighted in this paper with examples of design solutions to accommodate some of the key interface and integration requirements.

https://iopscience.iop.org/article/10.1088/0029-5515/54/3/033004/pdf

The ITER blanket system design challenge

As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce the risks of potential release during accidents, the inventory limit being set at 1 kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be) eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513 K for the FW and 623 K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure) co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

Wall protection strategies for DEMO plasma transients

As in many of today's tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW pane ...

http://iopscience.iop.org/article/10.1088/0741-3335/58/7/074005/pdf

Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

The ITER blanket design process is very challenging due to demanding design and interface requirements and constraints, including high heat fluxes from the plasma, large electromagnetic loads during off-normal events, sufficient contribution to the shielding of the vacuum vessel and superconducting coils, and tight interfacing space constraints with many key components. This paper highlights some of these challenges and the associated solutions developed as part of the final blanket design effort.

ITER blanket engineering challenges and solutions

Design evolution and integration of the ITER in-vessel components

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The ITER blanket system design challenge

ITER blanket engineering challenges and solutions

Design evolution and integration of the ITER in-vessel components