Published 1984 | Version v1
Book

Effect of radiation environment on SCC of austenitic materials

  • 1. General Electric Co., Vallecitos Nuclear Center, Pleasanton, CA 94566

Description

A program is in progress to study the effects of irradiation on the SCC behavior of AISI-304 stainless steel in high-temperature water. A series of Constant Extension Rate Tensile tests have been completed at 288lt. slashC in oxygenated water on annealed materials machined from irradiated reactor components. The tests confirmed that high levels of irradiation promote Intergranular Stress Corrosion Cracking in AISI-304 stainless steel, and that a threshold fluence exists for high plastic strain levels at about 5x10blt. slash n/cmb (E>1 MeV)

Additional details

Publishing Information

Publisher
National Association of Corrosion Engineers.
Imprint Place
Houston, TX (USA)
Imprint Title
Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
Journal Page Range
p. 451-461.

Conference

Title
Environmental degradation of materials in nuclear power systems-water reactors.
Dates
22-24 Aug 1983.
Place
Myrtle Beach, SC (USA).