High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

被引:10
|
作者
Raffray, AR
Chiocchio, S
Ioki, K
Krassovski, D
Kubik, D
Tivey, R
机构
[1] Max Planck Inst Plasma Phys, ITER JWS, D-85748 Garching, Germany
[2] NIKIET Entek, Dept 310, Moscow, Russia
[3] McDonnell Douglas Corp, Aerosp, St Louis, MO 63166 USA
关键词
D O I
10.1016/S0920-3796(98)00225-7
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
Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest ht:at load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application. of results from on-going high heat flux R&D and a brief description of future R&D effort to address remaining issues are also included. (C) 1998 Elsevier Science S.A. All rights reserved.
引用
收藏
页码:323 / 331
页数:9
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