Primary water stress corrosion cracking in alloy 600 and intergranular crack growth rate

2010 ◽  
Vol 16 (2) ◽  
pp. 267-271 ◽  
Author(s):  
Y. C. Jung ◽  
H. S. Chung ◽  
K. S. Lee ◽  
N. H. Heo
Keyword(s):  
Growth Rate ◽  
Water Stress ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Intergranular Crack ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Primary Water

Weld Residual Stresses and Primary Water Stress Corrosion Cracking in Bimetal Nuclear Pipe Welds

2007 ◽  
Author(s):  
Frederick W. Brust ◽  
Paul M. Scott
Keyword(s):  
Water Stress ◽  
Residual Stresses ◽  
Weld Metal ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Pressure Vessel ◽  
Large Diameter ◽  
Corrosion Cracking ◽  
Primary Water

There have been incidents recently where cracking has been observed in the bi-metallic welds that join the hot leg to the reactor pressure vessel nozzle. The hot leg pipes are typically large diameter, thick wall pipes. Typically, an inconel weld metal is used to join the ferritic pressure vessel steel to the stainless steel pipe. The cracking, mainly confined to the inconel weld metal, is caused by corrosion mechanisms. Tensile weld residual stresses, in addition to service loads, contribute to PWSCC (Primary Water Stress Corrosion Cracking) crack growth. In addition to the large diameter hot leg pipe, cracking in other piping components of different sizes has been observed. For instance, surge lines and spray line cracking has been observed that has been attributed to this degradation mechanism. Here we present some models which are used to predict the PWSCC behavior in nuclear piping. This includes weld model solutions of bimetal pipe welds along with an example calculation of PWSCC crack growth in a hot leg. Risk based considerations are also discussed.

scholarly journals The Application of Reliability-Based Design Factors in Stress Corrosion Cracking Evaluations

2002 ◽  
Author(s):  
Edward Friedman
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Crack Initiation ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Initiation Time ◽  
Reliability Based Design ◽  
Design Factors

First-order reliability methodology (FORM) is used to develop reliability-based design factors for deterministic analyses of stress corrosion cracking. The basic elements of FORM as applied to structural reliability problems are reviewed and then employed specifically to stress corrosion cracking evaluations. Failure due to stress corrosion cracking is defined as crack initiation followed by crack growth to a critical depth. The stress corrosion cracking process is thus represented in terms of a crack initiation time model and a crack growth rate model, with the crack growth rate integrated from the initiation time to the time at which the crack grows to its critical depth. Both models are described by log-normal statistical distribution functions. A procedure is developed to evaluate design factors that are applied to the mean values of the crack initiation time and the crack growth rate for specified temperature and stress conditions. The design factors, which depend on the standard deviations of the statistical distributions, are related to a target reliability, which is inversely related to an acceptable probability of failure. The design factors are not fixed, but are evaluated on a case-to-case basis for each application. The use of these design factors in a deterministic analysis assures that the target reliability will be attained and the corresponding acceptable probability of failure will not be exceeded. An example problem illustrates use of this procedure.

Crack Growth Rate Testing and Large Plate Demonstration Under Chloride-Induced Stress Corrosion Cracking Conditions in Stainless Steel Canisters for Storage of Spent Nuclear Fuel

2019 ◽  
Author(s):  
Poh-Sang Lam ◽  
Andrew J. Duncan ◽  
Lisa N. Ward ◽  
Robert L. Sindelar ◽  
Yun-Jae Kim ◽  
...  
Keyword(s):  
Growth Rate ◽  
Stainless Steel ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Compact Tension ◽  
Corrosion Cracking

Abstract Stress corrosion cracking may occur when chloride-bearing salts deposit and deliquesce on the external surface of stainless steel spent nuclear fuel storage canisters at weld regions with high residual stresses. Although it has not yet been observed, this phenomenon leads to a confinement concern for these canisters due to its potential for radioactive materials breaching through the containment system boundary provided by the canister wall during extended storage. The tests for crack growth rate have been conducted on bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The test and characterization protocols are performed to provide bounding conditions in which cracking will occur. The results after 2- and 6-month exposure are examined in relation to previous studies in condensed brine and compared with other experimental data in the open literature. The knowledge gained from bolt-load compact tension testing is being applied to a large plate cut from a mockup commercial spent nuclear fuel canister to demonstrate the crack growth behavior induced from starter cracks machined in regions where the welding residual stress is expected. All these tests are conducted to support the technical basis for ASME Boiler and Pressure Vessel Section XI Code Case N-860.

Evaluation and Repair of Primary Water Stress Corrosion Cracking in Alloy 600/182 Control Rod Drive Mechanism Nozzles

2002 ◽  
Author(s):  
Charles R. Frye ◽  
Melvin L. Arey ◽  
Michael R. Robinson ◽  
David E. Whitaker
Keyword(s):  
Water Stress ◽  
Boric Acid ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Base Material ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Control Rod ◽  
Drive Mechanism ◽  
Primary Water

In February 2001, a routine visual inspection of the reactor vessel head of Oconee Nuclear Station Unit 3 identified boric acid crystals at nine of sixty-nine locations where control rod drive mechanism housings (CRDM nozzles) penetrate the head. The boric acid deposits resulted from primary coolant leaking from cracks in the nozzle attachment weld and from through-thickness cracks in the nozzle wall. A general overview of the inspection and repair process is presented and results of the metallurgical analysis are discussed in more detail. The analysis confirmed that primary water stress corrosion cracking (PWSCC) is the mechanism of failure of both the Alloy 182 weld filler material and the alloy 600 wrought base material.

Sign in / Sign up

Export Citation Format

Share Document