Primary water stress corrosion cracking inspection ranking scheme for alloy 600 components

Hoang, Phuong H.; Gangadharan, A. C.; Ramalingam, Suresh C

doi:10.1016/s0029-5493(97)00346-4

Cited by 10 publications

(6 citation statements)

References 1 publication

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“…[96,100] Bruemmer and Was [78] reported that primary water SCC has not been observed at temperatures below 523 K (250°C); it is most significant in the hottest part of the system, where the temperature exceeds 573 K (300°C). This agrees with Hoang et al, [88] who believe that temperature is the most influential variable to PWSCC growth rate of Alloy 600. Many experimental test results indicate that IGSCC is a thermally activated process and that the CGR is strongly dependent on temperature.…”

Section: Effect Of Cold-worksupporting

confidence: 92%

“…[46,79,[84][85][86][87] Intragranular carbides have not been shown to have a beneficial effect. [88] The influence of special GBs (e.g., CSL boundaries, low, and HABs) together with C in solution and GB carbides was also investigated. [79] It was found that higher C content in solution increases IGSCC resistance and GB carbides further improve resistance to SCC on top of already present benefits.…”

Section: Effect Of Cold-workmentioning

confidence: 99%

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SCC in PWRs: Learning from a Bottom-Up Approach

Lozano-Pérez

Dohr

Meisnar

et al. 2014

Metallurgical and Materials Transactions E

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Stress corrosion cracking (SCC) of steels and Ni-base alloys in the pressurized water reactor (PWR) primary circuit has been a cause for reactor outages for many decades. Although the nuclear industry has made a considerable research effort to understand and predict SCC in this system, this enterprise is complicated by the sheer number of interdependent variables that have a major influence on the degradation behavior. SCC is highly time-dependant and often only sets in after many years. Therefore, autoclave testing, even with accelerating conditions, is an expensive and time-consuming endeavor. However, the results collected by a great number of research groups over many years have identified the most important parameters and, in many cases, how they influence the degradation behavior. The community has been constantly working on the development of a general theory that encompasses the underlying mechanisms and is capable of describing and predicting the observed degradation behavior. In the last two decades, the focus of research has shifted from the traditional approach of autoclave testing (for susceptibility and crack growth rates) to high-resolution microscopy and chemical analysis. The newly available techniques are providing data on chemical and structural changes locally at the crack tip where SCC occurs. The techniques covered in this review can provide very high chemical sensitivity at atomic resolution, which is ultimately needed in the quest for a generalized theory.

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Section: Effect Of Cold-worksupporting

confidence: 92%

Section: Effect Of Cold-workmentioning

confidence: 99%

SCC in PWRs: Learning from a Bottom-Up Approach

Lozano-Pérez

Dohr

Meisnar

et al. 2014

Metallurgical and Materials Transactions E

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“…304不锈钢在超临界水中裂纹扩展速率和裂纹深度均随着温度升高而增加 [54] . SCC裂纹扩展速率与温度的关系可用Arrhenius方程 [55] 表示: 通过实验分析得到了SCC扩展速率与应力强度因子之间的关系. Andresen和Ford [58] 基于实验和理论研究提出材料应力腐蚀扩展速率的For-Andresen模型如下: Nuclear energy is an important option to ensure the national energy safety.…”

Section: 展速率影响显著 200°c时水环境裂纹扩展速率是25°cunclassified

Key problems on the mechanical behavior of nuclear materials and structures of pressured water reactors

Wang¹,

Zhang²,

Ma³

et al. 2019

Sci. Sin.-Phys. Mech. Astron.

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“…The activation energy for crack initiation of Alloy 600 was suggested by others. [20][21][22][23][24][25] For a susceptibility prediction in this study, the value of 210 KJ/mole was selected for crack initiation of Alloy 600 with reasonable consensus from six studies. 20-25…”

Section: Comparison With Alloy 690 Nozzle At Hot Leg Sidementioning

confidence: 99%

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

Hwang

Lim²,

Kim³

2013

Corrosion

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Steam generator (SG) drain nozzles on the cold leg side in a pressurized water reactor (PWR) have exhibited stress corrosion cracking (SCC). The nozzles were made of forged and annealed Alloy 600 (UNS N06600) base material containing welds with Alloys 82 (UNS N06082) and 182 (UNS W86182). Carbides were distributed at the grain boundaries, but no grain boundary Cr depletion was observed in the base metal. Circumferential cracks developed at the root of a J weld inside a thick wall pipe. Axial cracks were distributed with a length around 6 mm to 10 mm long. All cracks were initiated from the inner surface of the pipe. Some more shallow cracks were also observed, but they were not detected by the field non-destructive examination using eddy current test (ECT) and ultrasound test (UT). Two out of the 12 cracks penetrated through the pipe wall, and therefore the primary water was leaking during the operation. The high tensile stress region calculated by a finite element modeling (FEM) model coincided with the crack locations. The highest tensile stress region was at the root of the J weld, as judged from metallography and residual stress analyses. Residual tensile stresses from welding are considered to be the main contributor to the overall stress.

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Primary water stress corrosion cracking inspection ranking scheme for alloy 600 components

Cited by 10 publications

References 1 publication

SCC in PWRs: Learning from a Bottom-Up Approach

SCC in PWRs: Learning from a Bottom-Up Approach

Key problems on the mechanical behavior of nuclear materials and structures of pressured water reactors

Primary Water Stress Corrosion Cracking Analysis in Alloy 600 Steam Generator Nozzle of a Pressurized Water Reactor

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