Estimation of Tritium Release from LLCB TBM and Its Ancillary Systems and Tritium Management in Different Locations of ITER. (3rd April 2017)

• Record Type: Journal Article
• Authors: Brahmbhatt, Priyanka; Sircar, Amit; Patel, Rudreksh; RajendraKumar, E.; Mohan, Sadhana; Bhanja, Kalyan
• Is Part Of: Fusion science and technology. Volume 71:Number 3(2017)
• Page Start: 391
• Page End: 396
• Publication Date: 2017-04-03
• Subjects: Tritium permeation -- tritium mass balance -- LLCB TBM; Fusion reactors -- Periodicals; Nuclear fusion -- Periodicals; Fusion reactors; Nuclear fusion; Periodicals; 621.48405
• DOI: 10.1080/15361055.2017.1289580
• Languages: English
• ISSNs: 1536-1055

Abstract

Abstract: The Indian Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is to be installed in one half of equatorial port #2 for testing in ITER machine. Liquid Pb-Li and solid Li2 TiO3 are the tritium breeder materials in LLCB TBM. Tritium permeates through structural materials in particular at higher temperatures, which is a major operational and safety concern. Therefore, tritium flows in different locations of ITER Tokamak complex have been estimated. Tritium transport from LLCB TBM and its ancillary systems into process rooms has been studied and analyzed in this work. A steady state diffusion limited permeation model neglecting surface effects has been used for the analysis. Tritium permeation to the Vacuum Vessel, Pipe Forest Area, Port Cell, Pipe Chase Area, Tokamak Cooling Water System Vault Annex (TCWS-VA) and Tritium Process Room in L-2 level has been estimated. The requirement to be fulfilled in each equatorial port cell is that the tritium concentration in the port cell during maintenance operations should be below the admissible limit for human access (regulatory maximum allowable value < 1 DAC = 3.4 × 10 5 Bq/m 3, Derived Air concentration). The presence of the Detritiation System (DS) in the Port cell has to be taken into account.This admissible limit for human access has to be reached in a sufficiently short time (target = 12 h) after plasma shutdown. Additional release during maintenance and radiological zoning with recommended <10 µSv/h need to be considered. Management of concentration of permeated tritium in different locations considering above requirement has also been discussed in this paper.