Overview of ITER TBM program objectives and management

Abstract

SUMMARYOne of the most important ITER missions is to test tritium breeding blanket (TBB) concepts that would result in tritium self-sufficiency, an extraction of high grade heat and net electricity production in future fusion reactors. Knowledge of the TBB performance characteristics in the integrated fusion nuclear environment is essential to designing the TBB of the DEMO and the fusion reactor. Three equatorial ports in ITER are allotted for testing test blanket modules (TBMs), which are mock-ups of the TBB, allowing up to six TBMs to be simultaneously installed and tested -two TBMs per each port. Conceptual designs for all TBM systems are now completed, and the important next milestones are the achievement of the preliminary and the final design by 2025. The TBM systems will be installed during Assembly Phase III, which is planned to start in 2030, and tested during the ITER Pre-Fusion Power Operation 2, which is planned to start in 2032. This paper describes the objectives and management aspects of the TBM Program.

Development and experiments of LiPb coolant technologies for fusion blanket in China

“…The liquid metal blanket concept is a promising blanket design for fusion reactor including international thermonuclear experimental reactor (ITER) and fusion demonstration (DEMO) power plant 1‐3 . The lithium lead (LiPb) eutectic is considered as one of the most promising candidate tritium breeder, neutron multiplier and coolant due to its favorable properties such as high tritium breeding rate, high thermal conductivity and low chemical activity etc 4‐9 .…”

Analysis on stress and deformation of large-scale concentric recuperator for high-temperature PbLi loop

“…Aimed at the use of fusion energy, many nuclear fusion science facilities have been constructed in worldwide. Especially, the most ambitious fusion device, International Thermonuclear Experimental Reactor (ITER), is currently underway . The blanket is one of the most important components for energy transformation and extraction in fusion reactor, and liquid lead lithium (PbLi) blanket is proposed as one of the most promising blanket designs .…”

An experimental investigation of bonding strengths for plasma facing components

“…1,2 Normally, it requires reliable compatibility with the plasma, good resistance to high heat load, high flux low-energy ion, and neutral particle irradiation and excellent heat transfer property. 1,2 Normally, it requires reliable compatibility with the plasma, good resistance to high heat load, high flux low-energy ion, and neutral particle irradiation and excellent heat transfer property.…”

“…One important issue in International Thermonuclear Experimental Reactor (ITER) is to select a proper material for the plasma facing components (FPC) due to its severe work environment. 1,2 Normally, it requires reliable compatibility with the plasma, good resistance to high heat load, high flux low-energy ion, and neutral particle irradiation and excellent heat transfer property. At this moment, the existing material is hard to completely satisfy such requirements, and the typical way is to integrate plasma facing material (PFM) with the heat sink material with high thermal conductivity.…”