Ph. Chappuis

The ITER blanket system design challenge

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.

Possible divertor solutions for a fusion reactor.

Safe operation of a fusion reactor divertor requires the exhaust of heat fluxes ranging from 5 to 30 MW m−2. The different technical solutions which are proposed rely on various water cooled copper heat sink designs. The protective armour may be Tungsten, Beryllium or Carbon depending on the plasma interaction. Optimisation is achieved by an overall comparison stressing the compliance to industrial defects. Reliable designs have been proposed and tested on large elements with carbon tiles for several thousand cycles, at power levels up to 10 MW m−2.

ITER blanket engineering challenges and solutions

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.

Progress on performance assessment of ITER enhanced heat flux first wall technology after neutron irradiation

ITER first wall (FW) panels are irradiated by energetic neutrons during the nuclear phase. Thus, an irradiation and high heat flux testing programme is undertaken by the ITER organization in order to evaluate the effects of neutron irradiation on the performance of enhanced heat flux (EHF) FW components. The test campaign includes neutron irradiation (up to 0.6–0.8 dpa at 200 °C–250 °C) of mock-ups that are representative of the final EHF FW panel design, followed by thermal fatigue tests (up to 4.7 MW m−2). Mock-ups were manufactured by the same manufacturing process as proposed for the series production. After a pre-irradiation thermal screening, eight mock-ups will be selected for the irradiation campaigns. This paper reports the preparatory work of HHF tests and neutron irradiation, assessment results as well as a brief description of mock-up manufacturing and inspection routes.

Assembly and remote handling of ITER plasma facing components

The assembly and remote maintenance of the plasma-facing components of ITER presents major challenges for the construction, operation and maintenance of ITER. The main plasma-facing components, the divertor and the blankets, are large components, with exceptionally demanding alignment requirements, and require complex support systems to sustain loads under normal and off-normal operational conditions. The plasma-facing components are designed to be maintained using special remote handling (RH) tools. These tools and the associated RH systems and processes are currently undergoing detailed design and certain key processes have been subject to extensive R&D / prototyping activities.