MANAGEMENT OF TRITIUM IN ITER WASTE

被引:11
作者
Rosanvallon, S. [1 ]
Benchikhoune, M. [1 ]
Ciattaglia, S. [1 ]
Uzan, J. Elbez [1 ]
Gastaldi, O. [2 ]
Na, B. C. [1 ]
Taylor, N. [1 ]
机构
[1] ITER Org, CS 90 046, F-13067 St Paul Les Durance, France
[2] Assoc Euratom CEA, DEN DTN STPA, F-13108 St Paul Les Durance, France
关键词
D O I
10.13182/FST11-A12553
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
ITER will use tritium as fuel. Procedures and processes are thus put in place in order to recover the tritium that is not used in the fusion reaction, including from waste and effluents. The tritium thus recovered can be re-injected into the fuel cycle. Moreover, tritium content and thus outgassing may be a safety concern, because of the potential for releases to the environment, both from the facility and from the final disposal (subjected to stringent acceptance criteria in the current waste final disposal). The aim of this paper is to present the measures considered to deal with the specific case of tritium in the liquid and solid waste that will arise from ITER operation and decommissioning. It concerns the processes that are considered from the waste production to its final disposal and in particular: the tritium removal stages (in-situ divertor baking at 350 C and tritium removal from solid waste and liquid and gaseous effluents), the removal of dust contamination (dust containing tritium produced by plasma-wall interaction and by the maintenance/ refurbishment processes) and the measures to enable safe processing and storage of the waste (wall-liner in the hot cell facility to limit concrete contamination and interim storage enabling tritium decay for waste that could not be directly accepted in the host-country final disposal facilities).
引用
收藏
页码:855 / 860
页数:6
相关论文
共 10 条
  • [1] Gaël B, 2008, FUSION SCI TECHNOL, V54, P205
  • [2] Optimisation of operating parameters for plasma facing components detritiation
    Gastaldi, O.
    Ghirelli, N.
    [J]. FUSION ENGINEERING AND DESIGN, 2009, 84 (2-6) : 815 - 820
  • [3] Pre-study of a detritiation process for plasma facing components in the ITER hot cells
    Ghirelli, N.
    Blet, V.
    Gastaldi, O.
    [J]. FUSION ENGINEERING AND DESIGN, 2009, 84 (2-6) : 821 - 824
  • [4] Fuel cycle design for ITER and its extrapolation to DEMO
    Konishi, Satoshi
    Glugla, Manfred
    Hayashi, Takumi
    [J]. FUSION ENGINEERING AND DESIGN, 2008, 83 (7-9) : 954 - 958
  • [5] Detritiation studies for JET decommissioning
    Perevezentsev, A. N.
    Bell, A. C.
    Williams, J.
    Brennan, P. D.
    [J]. FUSION ENGINEERING AND DESIGN, 2008, 83 (10-12) : 1364 - 1367
  • [6] ITER waste management
    Rosanvallon, S.
    Na, B. C.
    Benchikhoune, M.
    Uzan, J. Elbez
    Gastaldi, O.
    Taylor, N.
    Rodriguez, L.
    [J]. FUSION ENGINEERING AND DESIGN, 2010, 85 (10-12) : 1788 - 1791
  • [7] Recent analysis of key plasma wall interactions issues for ITER
    Roth, Joachim
    Tsitrone, E.
    Loarte, A.
    Loarer, Th.
    Counsell, G.
    Neu, R.
    Philipps, V.
    Brezinsek, S.
    Lehnen, M.
    Coad, P.
    Grisolia, Ch.
    Schmid, K.
    Krieger, K.
    Kallenbach, A.
    Lipschultz, B.
    Doerner, R.
    Causey, R.
    Alimov, V.
    Shu, W.
    Ogorodnikova, O.
    Kirschner, A.
    Federici, G.
    Kukushkin, A.
    [J]. JOURNAL OF NUCLEAR MATERIALS, 2009, 390-91 : 1 - 9
  • [8] A review of dust in fusion devices: Implications for safety and operational performance
    Sharpe, JP
    Petti, DA
    Bartels, HW
    [J]. FUSION ENGINEERING AND DESIGN, 2002, 63-64 : 153 - 163
  • [9] SUGIYAMA K., 2010, 19 PSI C SAN DIEG US
  • [10] The ITER remote maintenance system
    Tesini, A.
    Palmer, J.
    [J]. FUSION ENGINEERING AND DESIGN, 2008, 83 (7-9) : 810 - 816