Preliminary tritium safety analysis on China DFLL-TBM for ITER

被引：6

作者：

Song, Yong ^{[1
]}

Huang, Qunying ^{[1
,2
]}

Ni, Muyi ^{[2
]}

Wang, Yongliang ^{[3
]}

机构：

[1] Chinese Acad Sci, Inst Plasma Phys, Hefei 230031, Anhui, Peoples R China

[2] Univ Sci & Technol China, Sch Nucl Sci & Technol, Hefei 230027, Anhui, Peoples R China

[3] Sichuan Univ, Coll Phys Sci & Technol, Chengdu 610064, Peoples R China

来源：

FUSION ENGINEERING AND DESIGN | 2009年 / 84卷 / 12期

关键词：

Tritium; Safety analysis; Test blanket module; Liquid LiPb; TEST BLANKET MODULE; CONCEPTUAL DESIGN; HYDROGEN ISOTOPES; FUSION; REACTOR; PERMEATION; BERYLLIUM; PROGRESS; STRATEGY; STEEL;

D O I：

10.1016/j.fusengdes.2009.01.080

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：

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM. (C) 2009 Elsevier B.V. All rights reserved.

引用

页码：2114 / 2117

页数：4

共 23 条

[1] Post-irradiation studies of beryllium reflector of fission reactor - Examination of gas release, swelling and structure of beryllium under annealing [J].

Andreev, DV ;

Bespalov, VN ;

Birjukov, AJ ;

Gurovich, BA ;

Platonov, PA .

JOURNAL OF NUCLEAR MATERIALS, 1996, 233 :880-885

[2] Modelling of tritium permeation through beryllium as plasma facing material [J].

Berardinucci, L .

JOURNAL OF NUCLEAR MATERIALS, 1998, 258 :777-781

[3]

BERARDINUCI L, 1996, FUS TECHNOL, P1427

[4] Tritium control modelling for a helium cooled lithium-lead blanket of a fusion power reactor [J].

Farabolini, W ;

Ciampichetti, A ;

Dabbene, F ;

Fütterer, MA ;

Giancarli, L ;

Laffont, G ;

Puma, AL ;

Raboin, S ;

Poitevin, Y ;

Ricapito, I ;

Sardain, P .

FUSION ENGINEERING AND DESIGN, 2006, 81 (1-7) :753-762

[5] TRITIUM PERMEATION THROUGH HELIUM-HEATED STEAM-GENERATORS OF CERAMIC BREEDER BLANKETS FOR DEMO [J].

FUTTERER, MA ;

RAEPSAET, X ;

PROUST, E .

FUSION ENGINEERING AND DESIGN, 1995, 29 :225-232

[6] Progress in development of China Low Activation Martensitic Steel for fusion application [J].

Huang, Q. ;

Li, C. ;

Li, Y. ;

Chen, M. ;

Zhang, M. ;

Peng, L. ;

Zhu, Z. ;

Song, Y. ;

Gao, S. .

JOURNAL OF NUCLEAR MATERIALS, 2007, 367 :142-146

[7]

[黄群英 HUANG Qunying], 2008, [核科学与工程, Chinese Journal of Nuclear Science and Engineering], V28, P256

[8]

*IAEA, 2005, SAF TECHN BAS ITER F, pCH5

[9]

*ITER, 2004, GEN SIT SAF REP GSSR, V3

[10] Safety assessment of two advanced ferritic steel molten salt blanket design concepts [J].

Merrill, BJ ;

Sawan, M ;

Wong, CPC ;

Nygren, RE ;

Cadwallader, LC ;

Malang, S ;

Sze, DK .

FUSION ENGINEERING AND DESIGN, 2004, 72 (1-3) :277-306

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