Flaw Evaluation for PWR and BWR Component Weld Joints Using Advanced FEA Modeling Techniques

2009 ◽  
Author(s):  
T. Hayashi ◽  
S. F. Hankinson ◽  
T. Saito ◽  
C. K. Ng ◽  
W. H. Bamford
Keyword(s):  
Crack Propagation ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Water Reactor ◽  
Crack Shape ◽  
Weld Joints ◽  
Flaw Shape

Primary Water Stress Corrosion Cracking (PWSCC) of Pressurized Water Reactor (PWR) primary loop piping/nozzle Dissimilar Metal Weld (DMW) joints and Inter Granular Stress Corrosion Cracking (IGSCC) of Boiling Water Reactor (BWR) weld joints is an ongoing issue in the nuclear power industry. Recent field experiences with PWSCC of various DMW joints in US plants led to the development and application of an Advanced Finite Element Analyses (AFEA) methodology that permits crack propagation with a natural flaw shape. Crack growth and fracture evaluations for both PWR and BWR components are generally performed based on a conservative, idealized crack shape model, e.g. semi-ellipse, rectangle, etc., depending on the geometry of the crack and the component. Conventional evaluation methodologies and/or assumptions of this kind, in some cases may provide excessive conservatisms. The use of natural flaw shape development with crack propagation might provide a more realistic assessment of crack growth and structural integrity. The prime purpose of this study is to demonstrate the conservatism/margins in the conventional “idealized crack shape” methodology. A comparison study of crack growth behavior between the applications of the idealized and natural crack shape methodologies has been performed in order to assess the level of conservatism/margins in the conventional crack growth evaluation methodology and the possible impacts on the structural integrity evaluation for both PWR and BWR components. Comparison studies on the impacts of the differences in crack growth law and loading condition used for crack growth evaluations have been performed as well.

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Technical Basis for Revisions to ASME Section XI Acceptance Standards for Surface Flaws in Piping in Stress Corrosion Cracking Susceptible Materials

2009 ◽  
Author(s):  
Douglas A. Scarth ◽  
Katsumasa Miyazaki ◽  
Kunio Hasegawa ◽  
Warren H. Bamford
Keyword(s):  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Pressurized Water Reactor ◽  
Planar Surface ◽  
Pressurized Water ◽  
Water Reactor ◽  
Crack Growth Mechanism ◽  
Technical Basis

Acceptance Standards for flaws in piping are provided in Section XI of the ASME B&PV Code to permit acceptance of relatively small flaws without the need to perform an analytical evaluation. The Acceptance Standards are based on maintaining large margins against failure, and are based on the assumption that flaw growth will be insignificant. The assumption of a small amount of flaw growth is justified when fatigue crack growth is the only crack growth mechanism. However, when stress corrosion cracking is operative, flaw growth could be significant. This conclusion was illustrated by comparison of the crack growth results due to fatigue and stress corrosion cracking in Pressurized Water Reactor (PWR), and Boiling Water Reactor (BWR), coolant environments. For this reason, IWB-3514 of Section XI prohibits use of the Acceptance Standards for planar surface-connected flaws that are detected in piping materials that are susceptible to stress corrosion cracking and are in reactor coolant environments. As part of a recent Code revision to include new Acceptance Standards tables for flaws in piping, restrictions on use of the Acceptance Standards of IWB-3514 have been refined and clarified. The recent Code revision now specifies different restrictions and requirements for use of the Acceptance Standards for such planar surface-connected flaws detected by preservice and inservice examination. In addition, similar restrictions have been imposed on use of the new Acceptance Standards for such planar surface-connected flaws in Class 2 piping in IWC-3514 of Section XI. The technical basis for the restrictions and requirements for use of the Acceptance Standards for planar surface-connected flaws in piping materials that are susceptible to stress corrosion cracking is provided in this paper.

scholarly journals Fracture, Fatigue, and Structural Integrity of Metallic Materials

Metals ◽  
2019 ◽  
Vol 9 (8) ◽  
pp. 913
Author(s):  
Sergio Cicero ◽  
José Alberto Álvarez
Keyword(s):  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Creep Crack Growth ◽  
Corrosion Cracking ◽  
Metallic Materials ◽  
Creep Crack ◽  
Points Of View ◽  
Open Issues

Fracture, fatigue, and other subcritical processes, such as creep crack growth or stress corrosion cracking, present numerous open issues from both scientific and industrial points of view [...]

Modeling the Interactions of Hydrogen, Stress and Dissolution in Near-Neutral pH Stress Corrosion Cracking of Pipelines

2006 ◽  
Author(s):  
Frank Y. Cheng
Keyword(s):  
Stress Corrosion ◽  
Dissolution Rate ◽  
Crack Growth ◽  
Anodic Dissolution ◽  
Hydrogen Concentration ◽  
Stress Corrosion Cracking ◽  
Crack Tip ◽  
Corrosion Cracking ◽  
Neutral Ph ◽  
Ph Stress

A thermodynamic model was developed to determine the interactions of hydrogen, stress and anodic dissolution at the crack-tip during near-neutral pH stress corrosion cracking in pipelines. By analyzing the free-energy of the steel in the presence and absence of hydrogen and stress, it is demonstrated that a synergism of hydrogen and stress promotes the cracking of the steel. The enhanced hydrogen concentration in the stressed steel significantly accelerates the crack growth. The quantitative prediction of the crack growth rate in near-neutral pH environment is based on the determination of the effect of hydrogen on the anodic dissolution rate in the absence of stress, the effect of stress on the anodic dissolution rate in the absence of hydrogen, the synergistic effect of hydrogen and stress on the anodic dissolution rate at the crack-tip and the effect of the variation of hydrogen concentration on the anodic dissolution rate.

Mechanical Stress Improvement Process (MSIP) Used to Prevent and Mitigate Primary Water Stress Corrosion Cracking (PWSCC) in Reactor Vessel Piping at V.C. Summer

2003 ◽  
Author(s):  
E. A. Ray ◽  
K. Weir ◽  
C. Rice ◽  
T. Damico
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Mechanical Stress ◽  
Stress Corrosion Cracking ◽  
Reactor Vessel ◽  
Corrosion Cracking ◽  
The Other ◽  
Water Reactor ◽  
Improvement Process ◽  
Primary Water

During the October 2000 refueling outage at the V.C. Summer Nuclear Station, a leak was discovered in one of the three reactor vessel hot leg nozzle to pipe weld connections. The root cause of this leak was determined to be extensive weld repairs causing high tensile stresses throughout the pipe weld; leading to primary water stress corrosion cracking (PWSCC) of the Alloy 82/182 (Inconel). This nozzle was repaired and V.C. Summer began investigating other mitigative or repair techniques on the other nozzles. During the next refueling outage V.C. Summer took mitigative actions by applying the patented Mechanical Stress Improvement Process (MSIP) to the other hot legs. MSIP contracts the pipe on one side of the weldment, placing the inner region of the weld into compression. This is an effective means to prevent and mitigate PWSCC. Analyses were performed to determine the redistribution of residual stresses, amount of strain in the region of application, reactor coolant piping loads and stresses, and effect on equipment supports. In May 2002, using a newly designed 34-inch clamp, MSIP was successfully applied to the two hot-leg nozzle weldments. The pre- and post-MSIP NDE results were highly favorable. MSIP has been used extensively on piping in boiling water reactor (BWR) plants to successfully prevent and mitigate SCC. This includes Reactor Vessel nozzle piping over 30-inch diameter with 2.3-inch wall thickness similar in both size and materials to piping in pressurized water reactor (PWR) plants such as V.C. Summer. The application of MSIP at V.C. Summer was successfully completed and showed the process to be predictable with no significant changes in the overall operation of the plant. The pre- and post-nondestructive examination of the reactor vessel nozzle weldment showed no detrimental effects on the weldment due to the MSIP.

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.

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