[Table of Contents]

Plasma and Fusion Research

Volume 7, 2405016 (2012)

Regular Articles

Analysis on Tritium Management in FLiBe Blanket for LHD-Type Helical Reactor FFHR2
Yong SONG1), Akio SAGARA2), Takeo MUROGA2), Qunying HUANG1,3), Muyi NI3) and Yican WU1,3)
Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031, China
Reactor Engineering Research Center, National Institute of Fusion Sciences, Oroshi-cho, Toki 509-5292, Japan
School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027, China
(Received 2 December 2011 / Accepted 2 February 2012 / Published 15 March 2012)


In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R&D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc..


tritium, permeation, extraction, FLiBe

DOI: 10.1585/pfr.7.2405016


  • [1] A. Sagara, S. Imagawa, O. Mitarai et al., Nucl. Fusion 45, 258 (2005).
  • [2] S. Fukada, A. Morisaki, A. Sagara and T. Terai, Fusion Eng. Des. 81, 477 (2006).
  • [3] G. Gervasini and F. Reiter, J. Nucl. Mater. 168 (3), 304 (1989).
  • [4] W. Farabolini, A. Ciampichetti, F. Dabbene, M.A. Fütterer, L. Giancarli et al., Fusion Eng. Des. 81, 753 (2006).
  • [5] T. Tanaka, A. Sagara, T. Muroga et al., Nucl. Fusion 48, 035005 (2008).
  • [6] S. Fukada, A design for recovery of tritium from Flibe loop in FFHR-2, Fusion Power Plants and Related Advanced Technologies, Feb. 5, 2007.
  • [7] A. Sagara, O. Mitarai, T. Tanaka, S. Imagawa, Y. Kozaki, M. Kobayashi et al., Fusion Eng. Des. 83, 1690 (2008).
  • [8] B. Merrill and S. Malang, Tritium Permeation and Extraction Issues for the ARIES-CS, ARIES Meeting, University of Wisconsin, June 15th, 2005.
  • [9] Y. Song, Q. Huang, Y. Wang and M. Ni, Fusion Eng. Des. 84, 1779 (2009).
  • [10] C.P.C. Wong, M. Abdou, J. Blanchard, P. Calderoni, D.P. Carosella, M. Dagher et al., Design Description Document for the U.S. Dual Coolant Pb-17Li (DCLL) Test Blanket Module, TBWG Meeting, Nov 15, 2005.
  • [11] E. Serra, G. Benamati and O.V. Ogorodnikova, J. Nucl. Mater. 255, 105 (1998).
  • [12] R.A. Causey and W.R. Eampler, J. Nucl. Mater. 220-222, 823 (1995).
  • [13] P.X. Wang and J.S. Song, Helium in Material and the Permeation of Tritium (National defence industrial publication company of China, 2002).
  • [14] N. Baluc, K. Abe, J.L. Boutard et al., Nucl. Fusion 47, S696 (2007).
  • [15] H. Yoshida, O. Kveton, J. Koonce, D. Holland and R. Haange, Fusion Eng. Des. 39-40, 875 (1998).

This paper may be cited as follows:

Yong SONG, Akio SAGARA, Takeo MUROGA, Qunying HUANG, Muyi NI and Yican WU, Plasma Fusion Res. 7, 2405016 (2012).