Published February 1987 | Version v1
Report

BWR pipe crack remedies evaluation

  • 1. Argonne National Lab., IL

Description

This paper presents results on (a) the influence of simulated BWR environments on the stress-corrosion-cracking (SCC) susceptibility of Types 304, 316NG< and 347 stainless steel (SS), (b) fracture-mechanics crack growth rate measurements on these materials and weld overlay specimens in different environments, and (c) residual stress measurements and metallographic evaluations of conventional pipe weldments treated by a mechanical-stress-improvement process (MSIP) as well as those produced by a narrow-gap welding procedure. Crack initiation studies on Types 304 and 316NG SS under crevice and non-crevice conditions in 2890C water containing 0.25 ppm dissolved oxygen with low sulfate concentrations indicate that SCC initiates at very low strains (<3%) in the nuclear grade material. Crack growth measurements on fracture-mechanics-type specimens, under low-frequency cyclic loading, show that the Type 316NG steel cracks at a somewhat lower rate (∼40%) than sensitized Type 304 SS in an impurity environment with 0.25 ppm dissolved oxygen; however, the latter material stops cracking when sulfate is removed from the water. Crack growth in both materials ceases under simulated hydrogen-water chemistry conditions (<5 ppb oxygen) even with 100 ppb sulfate present in the water. An unexpected result was obtained in the test on a weld overlay specimen in the impurity environment, viz., the crack grew to the overlay interface at a nominal rate, branched at 900 in both directions, and then grew at a high rate (parallel to the nominal applied load). Residual stress measurements on MSIP-treated weldments and those produced by a narrow-gap welding procedure indicate that these techniques produce compressive stresses over most of the inner surface near the weld and heat-affected zones

Additional details

Publishing Information

Imprint Title
Proceedings of the US Nuclear Regulatory Commission fourteenth water reactor safety information meeting. Volume 2. Research in nondestructive evaluation, environmental effects in primary system components, pressure vessel research, irradiation effects on RPV steels, degraded piping research
Journal Page Range
p. 101-118.
Report number
NUREG/CP--0082-Vol.2

Conference

Title
14. water reactor safety information meeting.
Dates
27-31 Oct 1986.
Place
Gaithersburg, MD (USA).