Overview of the 10 MW high temperature gas cooled reactor-test module project

被引：93

作者：

Xu, YH ^{[1
]}

Zuo, KF ^{[1
]}

机构：

[1] Tsing Hua Univ, Inst Nucl Energy Technol, Beijing 100084, Peoples R China

来源：

NUCLEAR ENGINEERING AND DESIGN | 2002年 / 218卷 / 1-3期

关键词：

Graphite - High temperature reactors - Reactor cores - Safety factor;

D O I：

10.1016/S0029-5493(02)00181-4

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：

This paper discusses the historical development of the high temperature gas cooled reactor (HTGR) in China. China's development strategy of the HTGR will be explained in this text. The aim, design, construction and commissioning of the 10 MW HTGR-test module (HTR-10) will be explained herein. The engineering experiments, which were developed for the HTR-10, will also be introduced. The experience leading to an accumulation of knowledge during the development of China's HTGR will be summarized in this article. (C) 2002 Elsevier Science B.V. All rights reserved.

引用

页码：13 / 23

页数：11

共 8 条

[1]

*INET KFA, 1988, HIGH TEMP REACT HEAV

[2] TECHNICAL DESIGN-FEATURES AND ESSENTIAL SAFETY-RELATED PROPERTIES OF THE HTR-MODULE [J].

LOHNERT, GH .

NUCLEAR ENGINEERING AND DESIGN, 1990, 121 (02) :259-275

[3] STATUS OF DESIGN OF THE HTR TEST MODULE CHINA [J].

STEINWARZ, W ;

XU, YH .

NUCLEAR ENGINEERING AND DESIGN, 1990, 121 (02) :317-324

[4]

Sun Y. L., 2000, WORKSH SAF LIC ASP M

[5]

WANG DZ, 1982, UNTERSUCHUNG HOCHTEM

[6]

Xu Y. H., 1997, NEA WORKSH HIGH TEMP

[7]

ZHAO RK, 2001, PROJECT PROGR ENERGY, P121

[8]

ZHONG DX, 1985, 198511 INET

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