A Review of Alloy 600 Cracking in Operating Nuclear Plants Including Alloy 82 and 182 Weld Behavior

Bamford, W. H.; Hall, J. B.

doi:10.1115/icone12-49520

Cited by 22 publications

(18 citation statements)

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“…Incidences of stress corrosion cracking (SCC) of Alloy 182 weld in both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) have been reported [1][2][3][4][5]. In the recent decades, Alloy 52/152, which has a higher chromium content than Alloy 82/182, has been successfully used to repair the defected Control Rod Drive Mechanism (CRDM) and thermocouple penetration nozzles, Pressurizer (PZR) nozzles and hot leg nozzles, etc.…”

Section: Introductionmentioning

confidence: 99%

Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

Huang

Chiang

Jeng

et al. 2013

Journal of Nuclear Materials

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Section: Introductionmentioning

confidence: 99%

Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

Huang

Chiang

Jeng

et al. 2013

Journal of Nuclear Materials

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“…Eddy current testing (ECT) techniques to detect a defect, especially a stress corrosion cracking (SCC), on a reactor vessel (RV) and reactor internals have been developed as one of the surface inspection methods for nuclear power plants [1][2][3][4][5][6][7]. As a part of maintenance methods for the RV and reactor internals, laser peening and underwater laser beam welding techniques to prevent and repair from the SCC have been developed [8][9][10][11].…”

Section: Introductionmentioning

confidence: 99%

Eddy current testing system for bottom mounted instrumentation welds

Kobayashi

Ueno

Suganuma

et al. 2015

EPJ Nuclear Sci. Technol.

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Abstract. The capability of eddy current testing (ECT) for the bottom mounted instrumentation (BMI) weld area of reactor vessel in a pressurized water reactor was demonstrated by the developed ECT system and procedure. It is difficult to position and move the probe on the BMI weld area because the area has complexly curved surfaces. The space coordinates and the normal vectors at the scanning points were calculated as the scanning trajectory of probe based on the measured results of surface shape on the BMI mock-up. The multi-axis robot was used to move the probe on the mock-up. Each motion-axis position of the robot corresponding to each scanning point was calculated by the inverse kinematic algorithm. In the mock-up test, the probe was properly contacted with most of the weld surfaces. The artificial stress corrosion cracking of approximately 6 mm in length and the electrical-discharge machining slit of 0.5 mm in length, 1 mm in depth and 0.2 mm in width given on the weld surface were detected. From the probe output voltage, it was estimated that the average probe tilt angle on the surface under scanning was 2.6°.

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“…1 However, in laboratory tests in PWR coolant environments, the stress corrosion cracking (SCC) susceptibility of Alloy 182 is usually found to be greater than that of Alloy 600, while that of Alloy 82 is comparable to that of Alloy 600. This apparent inconsistency between field and laboratory experience is an issue that needs to be better understood.…”

Section: Introductionmentioning

confidence: 99%

“…1 Also, small-bore pipes and tubes have cracked earlier than larger components. 1 The PWSCC of Alloy 600 steam generator tubes in PWRs has been studied intensively. [2][3][4] In general, cracking occurs in regions of high residual stress due to cold work, such as the tube roll transition zone (RTZ), U-bends, tube denting locations, and plugs and sleeves.…”

Section: Introductionmentioning

confidence: 99%

Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

Alexandreanu¹,

Chopra²,

Shack³

2008

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In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290°C and 350°C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.iv This page is intentionally left blank. Foreword vThis report presents crack growth rate (CGR) data and the results of the corresponding fracture surface and metallographic examinations from cyclic loading and primary water stress-corrosion cracking (PWSCC) tests of two nickel-base Alloy 182 (A182) weldments. These weldments are typical of those used in vessel penetrations and piping butt welds in nuclear power plants. The effect of crack orientation with respect to dendrite orientation is the most significant variable investigated in this study. However, this report also compiles data from other laboratories that describe the effects of material composition, loading characteristics, and chemistry of the aqueous environment. The PWSCC growth rates described for the A182 specimens in the report are comparable to the CGR that characterize the performance of Alloy 600 (A600).This report is one of a series of reports documenting the results of CGR testing in vessel head penetration materials. This report focuses on the Alloy 82 (A82) and A182 weld metals. This report also presents results from tests of the A600 base metal. The researchers tested (1) a laboratory-fabricated, shieldedmetal-arc deposit of A182; (2) a weldment sample from the J-groove weld of a control rod drive mechanism nozzle from the Davis-Besse Nuclear Power Plant; and (3) a hot leg nozzle-to-safe end weld from the Virgil C. Summer Nuclear Station.

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A Review of Alloy 600 Cracking in Operating Nuclear Plants Including Alloy 82 and 182 Weld Behavior

Cited by 22 publications

References 0 publications

Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments

Eddy current testing system for bottom mounted instrumentation welds

Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

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