Published October 2009 | Version v1
Journal article Metadata-only

Corrosion behaviors of candidate materials for supercritical-cooled water reactor

Creators

  • 1. School of Nuclear Science and Engineering, Shanghai Jiaotong Univ., Shanghai (China)

Description

The corrosion behaviors of ferritic-martensitic steel P92, austenitic stainless steel type 316L and nickel-based alloy 690 have been investigated in supercritical water (SCW) at 600 degree C and 23 MPa. After exposed to SCW for 625 h, the weight gain of alloy 690, 316L and P92 are 0.001 02 g/(m2·h), 0.060 6 g/(m2·h) and 0.101 27g/(m2·h), respectively. SEM observation shows that a three-layer oxide film is formed on P92, a single-layer oxide film is formed on 316L and a very thin oxide film with some pitting is formed on alloy 690. (authors)

Additional details

Publishing Information

Journal Title
Nuclear Power Engineering
Journal Volume
30
Journal Issue
5
Journal Page Range
p. 62-66
ISSN
0258-0926

INIS

Country of Publication
China
Country of Input or Organization
China
INIS RN
42073347
Subject category
S36: MATERIALS SCIENCE;
Descriptors DEI
AUSTENITIC STEELS; CORROSION; FERRITIC STEELS; FILMS; GAIN; LAYERS; MARTENSITIC STEELS; MATERIALS; MAXIMUM PERMISSIBLE ACTIVITY; NICKEL BASE ALLOYS; OXIDES; PRESSURE RANGE MEGA PA; PWR TYPE REACTORS; SCANNING ELECTRON MICROSCOPY; SUPERCRITICAL STATE; WATER; WEIGHT
Descriptors DEC
ALLOYS; AMPLIFICATION; CARBON ADDITIONS; CHALCOGENIDES; CHEMICAL REACTIONS; ELECTRON MICROSCOPY; ENRICHED URANIUM REACTORS; HYDROGEN COMPOUNDS; IRON ALLOYS; IRON BASE ALLOYS; MICROSCOPY; NICKEL ALLOYS; OXYGEN COMPOUNDS; POWER REACTORS; PRESSURE RANGE; REACTORS; SAFETY STANDARDS; STANDARDS; STEELS; THERMAL REACTORS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Notes
6 figs., 1 tab., 11 refs.