The Stress Corrosion Cracking Behavior of Alloys 690 and 152 Weld in a PWR Environment

2008 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Water Environment ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Alloy 690 ◽  
Primary Water ◽  
Cold Rolled ◽  
Water Reactors

Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10−11 m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25–870°C.

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.

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.

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