Overview of recent ITER TBM Port Plug R&D activities

The objective of ITER project is the demonstration of the scientific and technological feasibility of fusion energy. Tritium fuel is required for the fusion reaction in DEMOnstration reactors and future commercial reactors but it is not available in nature because of its instability. To investigate tritium breeding self-sufficiency, two equatorial ports in ITER are dedicated to operating test blanket modules (TBMs) using the TBM Port Plug (TBM-PP). The TBM-PP design should be investigated with the physical mock-ups to validate the leak tightness in the primary vacuum boundary and to ensure TBM replacement and refurbishment compatible with Remote Handling (RH) operations in the ITER Hot Cell Facility during the ITER lifetime. This paper summarizes the achievements on the two most relevant R & D topics as described herewith. Several TBM-PP components (i.e., back parts of the TBMFrame, the Dummy-TBMs, the TBM-Shields, and the TBM feeding pipes) are part of the vacuum boundary implying that these components are Vacuum Quality Classification (VQC) 1A. To test and develop high vacuum sealing for the TBM-Frame flange and the TBM-Set or Dummy-TBM flange, the ITER Large Seal Test Rig (LSTR), and metallic gasket seals for the TBM application have been manufactured and installed in the ITER site. The helium leak tests were performed at different operating temperatures to investigate the vacuum performance of metallic gasket seals with VQC 1A. The TBM-PP design shall guarantee the feasibility of a rapid replacement and refurbishment compatible with ITER Remote Handling (RH) operations by using RH tools in the Hot Cell Facility during the ITER's lifetime. Therefore, an experimental program has been performed to demonstrate the feasibility of the critical RH refurbishment tasks including the insertion and removal of TBM-Set/Dummy-TBM and the RH bolting.

Automated summary

The objective of the ITER project is to demonstrate the scientific and technological feasibility of fusion energy. To investigate the self-sufficiency of tritium fuel, test blanket modules are used in ITER. The design of the test blanket modules must be validated for leak tightness and compatible with remote handling operations. Experimental programs have been performed to demonstrate the feasibility of critical remote handling refurbishment tasks.