Thermally activated dislocation creep model for primary water stress corrosion cracking of NiCrFe alloys

Conference · Sun Dec 31 04:00:00 EST 1995

OSTI ID:219245

Hall, Jr, M M

There is a growing awareness that awareness that environmentally assisted creep plays an important role in integranular stress corrosion cracking (IGSCC) of NiCrFe alloys in the primary coolant water environment of a pressurized water reactor (PWR). The expected creep mechanism is the thermally activated glide of dislocations. This mode of deformation is favored by the relatively low temperature of PWR operation combined with the large residual stresses that are most often identified as responsible for the SCC failure of plant components. Stress corrosion crack growth rate (CGR) equations that properly reflect the influence of this mechanism of crack tip deformation are required for accurate component life predictions. A phenomenological IGSCC-CGR model, which is based on an apriori assumption that the IGSCC-CGR is controlled by a low temperature dislocation creep mechanism, is developed in this report. Obstacles to dislocation creep include solute atoms such as carbon, which increase the lattice friction force, and forest dislocations, which can be introduced by cold prestrain. Dislocation creep also may be environmentally assisted due to hydrogen absorption at the crack tip. The IGSCC-CGR model developed here is based on an assumption that crack growth occurs by repeated fracture events occurring within an advancing crack-tip creep-fracture zone. Thermal activation parameters for stress corrosion cracking are obtained by fitting the CGR model to IGSCC-CGR data obtained on NiCrFe alloys, Alloy X-750 and Alloy 600. These IGSCC-CGR activation parameters are compared to activation parameters obtained from creep and stress relaxation tests. Recently reported CGR data, which exhibit an activation energy that depends on yield stress and the applied stress intensity factor, are used to benchmark the model. Finally, the effects of matrix carbon concentration, grain boundary carbides and absorbed hydrogen concentration are discussed within context of the model.

• Research Organization: Westinghouse Electric Corp., West Mifflin, PA (United States). Bettis Atomic Power Lab.
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC11-93PN38195
• OSTI ID: 219245
• Report Number(s): WAPD-T--3045; CONF-9505335--1; ON: DE96007598
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
CHROMIUM ALLOYS
CRACK PROPAGATION
CRACKS
CREEP
DEFORMATION
INCONEL 600
INCONEL X750
INTERGRANULAR CORROSION
IRON ALLOYS
MATHEMATICAL MODELS
NICKEL ALLOYS
PWR TYPE REACTORS
REACTOR COOLING SYSTEMS
STRESS CORROSION
STRESSES