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Title: Tests with Inconel 600 to obtain quantitative stress-corrosion cracking data for evaluating service performance. [PWR]

Conference ·
OSTI ID:6628803

Inconel 600 tubes in pressurized water reactor (PWR) steam generators form a pressure boundary between radioactive primary water and secondary water which is converted to steam and used for generating electricity. Under operating conditions the performance of alloy 600 has been good, but with some occasional small leaks resulting from stress corrosion cracking (SCC), related to the presence of unusually high residual or operating stresses. The suspected high stresses can result from either the deformation of tubes during manufacture, or distortion during abnormal conditions such as denting. The present experimental program addresses two specific conditions, i.e., (1) where deformation occurs but is no longer active, such as when denting is stopped and (2) where plastic deformation of the metal continues, as would occur during denting. Laboratory media consist of pure water as well as solutions to simulate environments that would apply in service; tubing from actual production is used in carrying out these tests. The environments include both normal and off chemistries for primary and secondary water. The results reported here were obtained in several different tests. The main ones are (1) split tube reverse U-bends, (2) constant extension rate tests (CERT), and (3) constant load. The temperature range covered is 290 to 365/sup 0/C.

Research Organization:
Brookhaven National Lab., Upton, NY (USA)
DOE Contract Number:
AC02-76CH00016
OSTI ID:
6628803
Report Number(s):
BNL-NUREG-31814; CONF-821037-39; ON: DE83002102
Resource Relation:
Conference: 10. water reactor safety research information conference, Gaithersburg, MD, USA, 12 Oct 1982; Other Information: Portions of document are illegible
Country of Publication:
United States
Language:
English