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.

Compared Modeling Study of Primary Water Stress Corrosion Cracking at Dissimilar Weld of Alloy 182 of Pressurized Water Nuclear Reactor According to Hydrogen Concentration

2015 ◽  
Vol 10 (4) ◽  
pp. 641-646
Author(s):  
Omar F. Aly ◽  
◽  
Miguel M. Neto ◽  
Mônica M. A. M. Schvartzman ◽  
Luciana I. L. Lima ◽  
...  
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Slow Strain Rate Test ◽  
Pressurized Water ◽  
Dissimilar Weld ◽  
Primary Water ◽  
Dissolved Hydrogen ◽  
Four Levels

One of the main failure mechanisms of pressurized water reactors (PWR) is primary water stress corrosion cracking (PWSCC), which occurs in alloy 600 (75Ni-15Cr-9Fe) and weld metals such as alloy 182 (70Ni-14Cr-9Fe), and alloy 82 (73Ni-19Cr-2Fe). Corrosion cracking is due, for example, in reactor nozzles welded dissimilarly with alloys 182/82 between ASTM A-508 G3 steel and AISI316L stainless steel. Corrosion cracks can cause problems reducing nuclear installations safety and reliability. Hydrogen dissolved into primary water to prevent radiolysis, also may enhance PWSCC growth. This article begins from a study by Lima et al. (2011) based on experimental data from the CDTN-Brazilian Nuclear Technology Development Center, and related to a slow strain rate test (SSRT). This was prepared and used for testing welds in the laboratory, similar to the dissimilar weld in pressurizer relief nozzles operating at the Brazilian Angra Unit 1 nuclear power plant. It was simulated for tests, primary water at 325°C and 12.5 MPa containing four levels of dissolved hydrogen. Our objective in this article is to clarify, and discuss adequate modeling based on the SSRT experimental results, and to compare them with those from another database and modeling, of the PWSCC growth rate based on levels of dissolved hydrogen.

Primary water stress corrosion cracking behavior of an Alloy 600/182 weld

2015 ◽  
Vol 100 ◽  
pp. 12-22 ◽  
Author(s):  
Yun Soo Lim ◽  
Seong Sik Hwang ◽  
Sung Woo Kim ◽  
Hong Pyo Kim
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Cracking Behavior ◽  
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.

Evaluation and Repair of Primary Water Stress Corrosion Cracking in Alloy 600/182 Control Rod Drive Mechanism Nozzles

2002 ◽  
Author(s):  
Charles R. Frye ◽  
Melvin L. Arey ◽  
Michael R. Robinson ◽  
David E. Whitaker
Keyword(s):  
Water Stress ◽  
Boric Acid ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Base Material ◽  
Corrosion Cracking ◽  
Alloy 600 ◽  
Control Rod ◽  
Drive Mechanism ◽  
Primary Water

In February 2001, a routine visual inspection of the reactor vessel head of Oconee Nuclear Station Unit 3 identified boric acid crystals at nine of sixty-nine locations where control rod drive mechanism housings (CRDM nozzles) penetrate the head. The boric acid deposits resulted from primary coolant leaking from cracks in the nozzle attachment weld and from through-thickness cracks in the nozzle wall. A general overview of the inspection and repair process is presented and results of the metallurgical analysis are discussed in more detail. The analysis confirmed that primary water stress corrosion cracking (PWSCC) is the mechanism of failure of both the Alloy 182 weld filler material and the alloy 600 wrought base material.

Benchmarking Study on Probabilistic Fracture Mechanics Analysis Codes xLPR and PASCAL-SP Considering Primary Water Stress Corrosion Cracking

2021 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Water Stress ◽  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Probabilistic Fracture Mechanics ◽  
Primary Water ◽  
Failure Probabilities

Abstract Probabilistic fracture mechanics (PFM) is expected as a more rational methodology for the structural integrity assessments of nuclear power components because it can consider the inherent probabilistic distributions of various influencing factors and quantitatively evaluate the failure probabilities of the components. The Japan Atomic Energy Agency (JAEA) has developed a PFM analysis code, PASCAL-SP, to evaluate the failure probabilities of piping caused by aging degradation mechanisms, such as fatigue and stress corrosion cracking in the environments of both pressurized water and boiling water reactors. To improve confidence in the analysis results obtained from PASCAL-SP, a benchmarking study was conducted together with the PFM analysis code, xLPR, which was developed jointly by the U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute. The benchmarking study was composed of deterministic and probabilistic analyses related to primary water stress corrosion cracking in a dissimilar metal weld joint in a pressurized water reactor surge line. The analyses were conducted independently by NRC staff and JAEA using their own codes and under common analysis conditions. In the present paper, the analysis conditions for the deterministic and probabilistic analyses are described in detail, and the analysis results obtained from the xLPR and PASCAL-SP codes are presented. It was confirmed that the analysis results obtained from the two codes were in good agreement.

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