Supercritical water-cooled reactor materials - Summary of research and open issues

被引：61

作者：

Guzonas, D. ^{[1
]}

Novotny, R. ^{[2
]}

机构：

[1] Atom Energy Canada Ltd, Chalk River Labs, Chalk River, ON K0J 1P0, Canada

[2] JRC IET, NL-1755 LE Petten, Netherlands

来源：

PROGRESS IN NUCLEAR ENERGY | 2014年 / 77卷

关键词：

SCWR; Gen IV; Water chemistry; Corrosion; Stress corrosion cracking; Cladding materials; EXPERIENCE PROVIDING BASES; STRESS-CORROSION CRACKING; STAINLESS-STEEL; PREDICTING CORROSION; HIGH-TEMPERATURE; ALLOYS; PRESSURE; BEHAVIOR; CHALLENGES; CHEMISTRY;

D O I：

10.1016/j.pnucene.2014.02.008

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 Supercritical Water Reactor (SCWR) is one of the six reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). Research on materials and chemistry for supercritical water-cooled reactors dates back to the 1960s when a number of reactor concepts using water at supercritical temperatures but sub-critical pressures (nuclear steam) were studied. There is also significant experience available from the operation of supercritical fossil-fired power plants. In this paper, the materials requirements of the various SCWR concepts are introduced, with a focus on the European Union pressure vessel concept and the Canadian pressure tube concept. The current understanding of the key materials degradation issues is reviewed, and knowledge gaps identified. (C) 2014 Elsevier Ltd. All rights reserved.

引用

页码：361 / 372

页数：12

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