scholarly journals Effect of oxide film on ECT detectability of surface IGSCC in laboratory-degraded alloy 600 steam generator tubing

2019 ◽  
Vol 51 (5) ◽  
pp. 1381-1389
Author(s):  
Tae Hyun Lee ◽  
Kyung Ha Ryu ◽  
Hong Deok Kim ◽  
Il Soon Hwang ◽  
Ji Hyun Kim ◽  
...  
Keyword(s):  
Oxide Film ◽  
Steam Generator ◽  
Alloy 600 ◽  
Steam Generator Tubing

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

2004 ◽  
Author(s):  
Warren Bamford ◽  
John Hall
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Steam Generator ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Control Rod ◽  
Nuclear Plants ◽  
Steam Generator Tubing ◽  
Long Time

Service induced cracking in Alloy 600 has been known for a long time, having been first observed in the 1980’s in steam generator tubing and small bore piping, and later, in 1991, in reactor vessel control rod drive mechanism (CRDM) head penetrations. Other than steam generator tubing, which cracked within a few years of operation, the first Alloy 600 cracking was in base metal of Combustion Engineering small bore piping, followed closely by CE pressurizer heater sleeves. The first reactor vessel CRDM penetrations (base metal) to crack were in France, US plants found CRDM cracking several years later. Three plants have discovered weld metal cracking at the outlet nozzle to pipe weld region. This was the first known weld metal cracking. This paper will chronicle the development of service-induced cracking in these components, and compare the behavior of welds as opposed to base metal, from the standpoint of time to crack initiation, growth rate of cracks, and their impact on structural integrity. In addition, a discussion of potential future trends will be provided.

IGSCC Behavior of Alloy 600 Steam Generator Tubing in Water or Steam Tests Above 360 C

CORROSION ◽  
1987 ◽  
Vol 43 (12) ◽  
pp. 727-734 ◽  
Author(s):  
G. Economy ◽  
R. J. Jacko ◽  
F. W. Pement
Keyword(s):  
Steam Generator ◽  
Alloy 600 ◽  
Steam Generator Tubing

Effect of Boric Acid on Pit Growth in Alloy 600 Steam Generator Tubing

CORROSION ◽  
1992 ◽  
Vol 48 (2) ◽  
pp. 103-113
Author(s):  
G. S. Was ◽  
D. Choi
Keyword(s):  
Boric Acid ◽  
Steam Generator ◽  
Alloy 600 ◽  
Steam Generator Tubing

Microstructure and Microdeformation Effects on IGSCC of Alloy 600 Steam Generator Tubing

CORROSION ◽  
1988 ◽  
Vol 44 (11) ◽  
pp. 782-788 ◽  
Author(s):  
S. M. Bruemmer ◽  
L. A. Charlot ◽  
C. H. Henager
Keyword(s):  
Steam Generator ◽  
Alloy 600 ◽  
Steam Generator Tubing

Pitting of steam-generator tubing alloys in solutions containing thiosulfate and sulfate or chloride

2015 ◽  
Vol 180 ◽  
pp. 233-249 ◽  
Author(s):  
William Zhang ◽  
Anatolie G. Carcea ◽  
Roger C. Newman
Keyword(s):  
Steam Generator ◽  
Concentration Ratio ◽  
Alloy 600 ◽  
Alloy Component ◽  
Sulfate Solutions ◽  
Steam Generator Tubing ◽  
Thiosulfate Concentration

The pitting of nuclear steam generator tubing alloys 600, 690 and 800 was studied at 60 °C using dilute thiosulfate solutions containing excess sulfate or (for Alloy 600) chloride. A potentiostatic scratch method was used. In sulfate solutions, all alloys pitted at low potentials, reflecting their lack of protective Mo. The alloys demonstrated the most severe pitting at a sulfate : thiosulfate concentration ratio of ∼40. Alloy 600 pitted worst at a chloride : thiosulfate ratio of ∼2000. The results are interpreted through the mutual electromigration of differently charged anions into a pit nucleus, and differences in the major alloy component.

Evaluation of the Creep Rupture Behavior of Alloy 690 Steam Generator Tubes Considering the Pressure Ramp Rate

2021 ◽  
Author(s):  
Jongmin Kim ◽  
Min-Chul Kim ◽  
Joonyeop Kwon
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Parameter Method ◽  
Alloy 600 ◽  
Steam Generator Tube ◽  
Failure Pressure ◽  
Alloy 690 ◽  
Rupture Model ◽  
Temperature And Pressure ◽  
Steam Generator Tubes

Abstract The materials used previously for steam generator tubes around the world have been replaced and will be replaced by Alloy 690 given its improved corrosion resistance relative to that of Alloy 600. However, studies of the high- temperature creep and creep-rupture characteristics of steam generator tubes made of Alloy 690 are insufficient compared to those focusing on Alloy 600. In this study, several creep tests were conducted using half tube shape specimens of the Alloy 690 material at temperatures ranging from 650 to 850C and stresses in the range of 30 to 350 MPa, with failure times to creep rupture ranging from 3 to 870 hours. Based on the creep test results, creep life predictions were then made using the well-known Larson Miller Parameter method. Steam generator tube rupture tests were also conducted under the conditions of a constant temperature and pressure ramp using steam generator tube specimens. The rupture test equipment was designed and manufactured to simulate the transient state (rapid temperature and pressure changes) in the event of a severe accident condition. After the rupture test, the damage to the steam generator tubes was predicted using a creep rupture model and a flow stress model. A modified creep rupture model for Alloy 690 steam generator tube material is proposed based on the experimental results. A correction factor of 1.7 in the modified creep rupture model was derived for the Alloy 690 material. The predicted failure pressure was in good agreement with the experimental failure pressure.

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