Stress corrosion cracking (SCC) tests were conducted under the supercritical-water cooled power reactor (SCPR) coolant conditions, which are predicted to consist of wide temperature range, extremely high pressure, and high dosage of neutron irradiation. This paper describes the results of SCC tests for austenitic stainless steels and a Ni base alloy. The susceptibility to intergranular stress corrosion cracking (IGSCC) was evaluated by slow strain rate tests (SSRTs), which were performed using a supercritical-water (SCW) test loop. The results for sensitized Type 304 SS indicated that the SCC susceptibility decreased with increasing temperature. The upper limit temperature of SCC susceptibility for sensitized type 304 SS existed at around 400°C, above which IGSCC did not occur in oxygenated supercritical water. No IGSCC of Type 316L SS and Alloy 600 was exhibited in fracture surface under SCPR conditions, even though small cracks were observed in side surface.

Subject
Scanning electron microscopy, Materials, Intergranular stress corrosion cracking, Metal surfaces, Reactors, Fracture surfaces, Stress strain curves, Irradiation assisted stress corrosion cracking, Corrosion resistance, Fracture, Alloys, Cracks, Stress corrosion cracking
Keywords:
Supercritical-water cooled power reactor, Type 304 SS, Type316LSS, Alloy600, IGSCC, SCC susceptibility
© 2004 Association for Materials Protection and Performance (AMPP). All rights reserved. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means (electronic, mechanical, photocopying, recording, or otherwise) without the prior written permission of AMPP. Positions and opinions advanced in this work are those of the author(s) and not necessarily those of AMPP. Responsibility for the content of the work lies solely with the author(s).
2004
Association for Materials Protection and Performance (AMPP)
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