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

CORROSION ◽  
10.5006/0935 ◽  
2013 ◽  
Vol 69 (11) ◽  
pp. 1051-1059 ◽  
Author(s):  
Seong Sik Hwang ◽  
Yun Soo Lim ◽  
Sung Woo Kim
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Primary Water

The effect of prior deformation on stress corrosion cracking growth rates of Alloy 600 materials in a simulated pressurized water reactor primary water

2008 ◽  
Vol 50 (3) ◽  
pp. 835-846 ◽  
Author(s):  
Seiya Yamazaki ◽  
Zhanpeng Lu ◽  
Yuzuru Ito ◽  
Yoichi Takeda ◽  
Tetsuo Shoji
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Growth Rates ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Prior Deformation ◽  
Water Reactor ◽  
Primary Water

Stress Corrosion Cracking Behavior of Alloy 182 Weld in Pressurized Water Reactor Primary Water Environment at 325ºC

CORROSION ◽  
2011 ◽  
Vol 67 (8) ◽  
pp. 085004-1-085004-9 ◽  
Author(s):  
L.I.L. Lima ◽  
M.M.A.M. Schvartzman ◽  
C.A. Figueiredo ◽  
A.Q. Bracarense
Keyword(s):  
Stress Corrosion ◽  
Oxide Layer ◽  
Stress Corrosion Cracking ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Primary Water

Abstract The weld used to connect two different metals is known as a dissimilar metal weld (DMW). In nuclear power plants, this weld is used to join stainless steel to low-alloy steel components in the nuclear pressurized water reactor (PWR). The most common weld metal is Alloy 182 (UNS W86182). Originally selected for its high corrosion resistance, it exhibited, after a long operation period, susceptibility to stress corrosion cracking (SCC) in PWR. The goal of this work was to study the electrochemical corrosion behavior and SCC susceptibility of Alloy 182 weld in PWR primary water containing 25 cm3 and 50 cm3 H2/kg H2O at standard temperature and pressure (STP). For this purpose, slow strain rate tensile (SSRT) tests and potentiodynamic polarization measurements were carried out. Scanning electron microscopy (SEM) with energy-dispersive spectrometry (EDS) was used to evaluate fracture morphology and determine the oxide layer chemical composition and morphology. The results indicated that at 325°C Alloy 182 weld is more susceptible to SCC at 25 cm3 (STP) H2/kg H2O and the increase of dissolved hydrogen decreased the crystal size of the oxide layer.

Effect of shot peening on primary water stress corrosion cracking of Alloy 600 steam generator tubes in an operating PWR plant

2004 ◽  
Vol 227 (2) ◽  
pp. 155-160 ◽  
Author(s):  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Deok Hyun Lee ◽  
Myung Ho Song ◽  
Seon Jin Kim ◽  
...  
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Shot Peening ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Primary Water ◽  
Steam Generator Tubes

Effects of Dissolved Hydrogen on the Primary Water Stress Corrosion Cracking Behavior of Alloy 600 at 325 °C

2004 ◽  
Vol 261-263 ◽  
pp. 943-948 ◽  
Author(s):  
Q.J. Peng ◽  
Tetsuo Shoji
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Experimental Results ◽  
Alloy 600 ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Primary Water ◽  
Dissolved Hydrogen

Primary water stress corrosion cracking (PWSCC) of Alloy 600 has been a great concern to the nuclear power industry. Reliable PWSCC growth rate data, especially at temperatures in the range of 290-330°C, of the alloy are required in order to evaluate the lifetime of power plant components. In this study, three tests were carried out in simulated pressurized water reactor (PWR) primary water at 325°C at different dissolved hydrogen (DH) concentrations using standard one-inch compact tension (1T-CT) specimens. The initiation and growth of cracks as well as insights into the different PWSCC mechanisms proposed in the literature were discussed. The experimental results show that the detrimental effects of hydrogen on crack initiation and growth reached a maximum at a certain level of DH in water. The experimental results were explained in terms of changes in the stability of the surface oxide films under different DH levels. The experimental results also support the assumption that hydrogen absorption as a result of cathodic reactions within the metal plays a fundamental role in PWSCC.

Precursor Evolution and Stress Corrosion Cracking Initiation of Cold-Worked Alloy 690 in Simulated Pressurized Water Reactor Primary Water

CORROSION ◽  
10.5006/2470 ◽  
2017 ◽  
Vol 73 (10) ◽  
pp. 1224-1236 ◽  
Author(s):  
Ziqing Zhai ◽  
Mychailo Toloczko ◽  
Karen Kruska ◽  
Stephen Bruemmer
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Alloy 690 ◽  
Primary Water ◽  
Cold Worked
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