Mechanistic Formulation of PWSCC Growth Rates of Ni-Base Alloys and Weld Metals

2011 ◽  
Author(s):  
Zhanpeng Lu ◽  
Tetsuo Shoji ◽  
He Xue ◽  
Chaoyang Fu
Keyword(s):  
Thermal Activation ◽  
Structural Integrity ◽  
Thermal Activation Energy ◽  
Evaluation Procedure ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Metals ◽  
Primary Water ◽  
Integrity Analysis

Several Ni-base alloys and their weld metals such as Alloy 600 and Alloy 82/182 suffered from stress corrosion cracking in pressurized water reactor primary water environments. Materials Reliability Program (MRP) proposed a CGR disposition curve in a report MRP 55 for PWSCC of thick-section Alloy 600 materials. This deterministic CGR equation has been adopted by Section XI Nonmandatory Appendix O of the ASME Boiler and Pressure Code for flaw evaluation. MRP also proposed a CGR disposition curve in MRP report 115 for PWSCC of Alloy 82/182/132 weld metals. In the same fashion, JSME and JNES also provided CGR disposition curves in the flaw evaluation procedure in structural integrity analysis. Stress intensity factor (K), temperature and thermal activation energy are included in both MRP 55 and MRP 115 reports. Both MRP 55 and MRP 115 are engineering-based rather than mechanism-based. The fundamental correlations such as crack growth rate vs. K are quantified based on the theoretical model and screened experimental data, which are compared to the reported disposition curves and used for improving the prediction.

Deterministic Formulation of the Effect of Stress Intensity Factor on PWSCC of Ni-Base Alloys and Weld Metals

2013 ◽  
Vol 135 (2) ◽  
Author(s):  
Zhanpeng Lu ◽  
Tetsuo Shoji ◽  
He Xue ◽  
Chaoyang Fu
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Thermal Activation ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Metals ◽  
Primary Water ◽  
Nickel Base

The fundamental correlations such as crack growth rate (CGR) versus K for primary water stress corrosion cracking (PWSCC) of nickel-base alloys in simulated pressurized water reactor environments are quantified with the theoretical model based on the combination of crack tip mechanics and oxidation kinetics. Materials reliability program (MRP) proposed a CGR disposition curve in a report MRP 55 for PWSCC of thick-section Alloy 600 materials. This deterministic CGR equation has been adopted by Section XI Nonmandatory Appendix O of the ASME Boiler and Pressure Code for flaw evaluation. MRP also proposed a CGR disposition curve in a report MRP 115 for PWSCC of Alloy 82/182/132 weld metals. Stress intensity factor (K), temperature and thermal activation energy are included in both MRP 55 and MRP 115 reports. Both MRP 55 and MRP 115 are engineering-based. The results of mechanism-based modeling are compared with the screened experimental data for typical PWSCC systems of nickel-base alloys and the consistence is observed.

Examinations of Oxidation and Sulfidation of Grain Boundaries in Alloy 600 Exposed to Simulated Pressurized Water Reactor Primary Water

2013 ◽  
Vol 19 (3) ◽  
pp. 676-687 ◽  
Author(s):  
D.K. Schreiber ◽  
M.J. Olszta ◽  
D.W. Saxey ◽  
K. Kruska ◽  
K.L. Moore ◽  
...  
Keyword(s):  
Electron Microscopy ◽  
Grain Boundary ◽  
Grain Boundaries ◽  
Atom Probe Tomography ◽  
Atom Probe ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Intergranular Attack ◽  
Pressurized Water ◽  
Primary Water

AbstractHigh-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325°C simulated pressurized water reactor primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy, and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.

Advanced Inlay System for Inlet/Outlet Nozzles of RV for Preventive Maintenance Against Alloy 600 PWSCC in Japanese PWR Plants

2013 ◽  
Author(s):  
Kazuhide Yamamoto ◽  
Masahiko Kizawa ◽  
Hiroki Kawazoe ◽  
Yuki Kobayashi ◽  
Ken Onishi ◽  
...  
Keyword(s):  
Preventive Maintenance ◽  
Heat Treatments ◽  
Residual Tensile Stress ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Nuclear Plants ◽  
Alloy 690 ◽  
Primary Water

Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes the specifications of the advanced INLAY system and introduces the maintenance activities which MHI has applied for three plants in Japan by March 2012.

Do All RPV Head Penetration Leaks Have the Potential to Cause Head Wastage?

2004 ◽  
Author(s):  
Chandra M. Roy ◽  
John R. Fessler ◽  
Jude R. Foulds ◽  
Ronald M. Latanision ◽  
David E. Taylor
Keyword(s):  
Water Stress ◽  
Pressure Vessel ◽  
Recent Experience ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Head Penetration ◽  
Failure Risk ◽  
Primary Water ◽  
Reactor Pressure

The identification of the PWSCC (Primary Water Stress Corrosion Cracking) mechanism responsible for leakage from an Alloy 600 nozzle tube of a PWR RPV (pressurized water reactor reactor pressure vessel) head more than a decade ago led to a significant body of research into understanding the phenomenon and to development of bases for safely managing primary pressure boundary integrity. However, the relatively recent experience at Davis-Besse, wherein penetration leakage resulted in significant vessel head material wastage, led to the heretofore unconsidered issue of vessel failure risk due to head rupture. This paper addresses, in preliminary fashion, one key input to determining the risk associated with head material wastage and potential rupture — the local environmental and fluid conditions associated with a range of leak paths. The results indicate a need for rigorous prediction of fluid conditions for a range of leak situations to help establish criteria for addressing penetration leaks.

Stress Corrosion Cracking Growth of Alloy 800NG in Pressurized Water Reactor Primary Water

CORROSION ◽  
10.5006/2572 ◽  
2017 ◽  
Vol 74 (1) ◽  
pp. 24-36 ◽  
Author(s):  
Koji Arioka ◽  
Roger W. Staehle ◽  
Robert L. Tapping ◽  
Takuyo Yamada ◽  
Tomoki Miyamoto
Keyword(s):  
Beneficial Effect ◽  
Chromium Concentration ◽  
Alloy 600 ◽  
Cavity Formation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Chromium Addition ◽  
Primary Water

The primary purpose of this research is to examine the stress corrosion cracking (SCC) resistance of Alloy 800NG in pressurized water reactor (PWR) primary water and pressurized heavy water reactor (PHWR) primary water. Rates of SCC growth of 20% cold-worked (CW) Alloy 800NG measured over the temperature range between 270°C and 360°C were compared with previously reported results for 20% CW Alloy TT690 and 20% CW Alloy 600 in order to consider which material is the most SCC resistant among materials presently being used for steam generator (SG) tubing worldwide. The secondary purpose is to examine the effect of chromium addition on SCC growth in PWR primary water of a series of alloys based on the Alloy 800 composition. SCC growth measurements were performed in PWR primary water over the chromium concentration range from 16% to 27% to obtain fundamental knowledge useful for considering a future alternative SCC-resistant material for SG tubing in extended life PWRs and PHWRs. The third objective is to examine the rate of cavity formation of 20% CW Alloy 800NG to obtain basic knowledge of one possible mechanism for SCC initiation after long-term operation. Measured rates of cavity formation in 20% CW Alloy 800NG were compared with previously reported results of 20% CW Alloy TT690 to compare the rate of SCC initiation caused by cavity formation. Four important patterns were observed. First, excellent SCC growth resistance was observed for 20% CW Alloy 800NG compared to 20% CW Alloy TT690 at 320°C, 340°C, and 360°C. Second, an inverse temperature dependence on SCC growth was observed in Alloy 800NG. The rate of SCC growth increased with decreasing temperature which was completely different from the trend for Alloy 600. Third, a significant beneficial effect by chromium addition in 800 series alloys on SCC growth resistance was observed in PWR primary water in the operating temperature range of PWRs and PHWRs. The rate of SCC growth decreased with increasing chromium concentration in the chromium concentration range between 16% and 27% chromium at 270°C, 290°C, and 320°C. However, no beneficial effect of chromium addition in these alloys was observed at 340°C and 360°C. Finally, a more than 10 times slower rate of cavity formation was observed in 20% CW Alloy 800NG than for 20% CW Alloy TT690. Results suggested that because of cavity formation, a more than 10-fold faster crack initiation occurred in Alloy TT690 than in Alloy 800NG. Further, carbide coverage and grain size significantly affected the rate of cavity formation. Detailed and comprehensive studies of long-term SCC initiation are necessary to ensure the future reliability of life-extended PWRs and PHWRs.

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