Effects of dissolved oxygen on corrosion fatigue cracking of Alloy 690(TT) in pressurized water reactor environments

2015 ◽  
Vol 74 ◽  
pp. 65-70 ◽  
Author(s):  
J. Xiao ◽  
S.Y. Qiu ◽  
Y. Chen ◽  
Z.X. Lin ◽  
Q. Xu ◽  
...  
Keyword(s):  
Dissolved Oxygen ◽  
Corrosion Fatigue ◽  
Fatigue Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Alloy 690

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

2004 ◽  
Author(s):  
J. D. Keller ◽  
A. J. Bilanin ◽  
S. T. Rosinski
Keyword(s):  
Thermal Cycling ◽  
Development Program ◽  
Fatigue Cracking ◽  
Evaluation Methodology ◽  
Experimental Investigations ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

The response of alloy 690 tubing in a pressurized water reactor environment

2007 ◽  
Vol 28 (2) ◽  
pp. 373-379 ◽  
Author(s):  
B.A. Young ◽  
Xiaosheng Gao ◽  
T.S. Srivatsan ◽  
Peter J. King
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Alloy 690

Corrosion Mechanisms of Ni-Base Alloys in Pressurized Water Reactor Primary Conditions

2008 ◽  
Vol 595-598 ◽  
pp. 529-537 ◽  
Author(s):  
Loïc Marchetti ◽  
Stéphane Perrin ◽  
Olivier Raquet ◽  
Michèle Pijolat
Keyword(s):  
Film Formation ◽  
Oxidation Mechanism ◽  
Passive Layer ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Boundary Diffusion Coefficient ◽  
Alloy 690 ◽  
Corrosion Mechanisms

Oxidation mechanism of Alloy 690 has been investigated in Pressurised Water Reactor (PWR) primary coolant conditions (325°C, aqueous hydrogenated media). Experiments performed with gold marker and RBS technique reveal that the passive film formation is the consequence of an anionic mechanism. This result is confirmed by experiments achieved with two sequences of corrosion in a H2 16O media and in a mixed H2 16O/ H2 18O media. The localisation of 18O by SIMS analysis in the thin passive layer underlines an oxidation mechanism due to oxygen diffusion by short circuits (like grain boundaries) in the oxide scale. Moreover grain boundary diffusion coefficient in chromite like oxide was estimated to be in the range 2 10-18 – 1 10-17 cm2.s-1 and compared to values extrapolated from higher temperature.

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