Formation and Deposition of Iron Oxides on Stainless Steel and Carbon Steel in Conditions of Secondary Circuits of Pressurized Water Reactors

CORROSION ◽  
2011 ◽  
Vol 67 (1) ◽  
pp. 015003-1-015003-10 ◽  
Author(s):  
S. Delaunay ◽  
C. Mansour ◽  
E.-M. Pavageau ◽  
G. Cote ◽  
G. Lefèvre ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Carbon Steel ◽  
Iron Oxides ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Water Reactors

scholarly journals An Experimental Programme optimized with Uncertainty Propagation: PETALE in the CROCUS Reactor

2019 ◽  
Vol 211 ◽  
pp. 03003 ◽  
Author(s):  
Vincent Lamirand ◽  
Axel Laureau ◽  
Dimitri Rochman ◽  
Gregory Perret ◽  
Adrien Gruel ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Energy Range ◽  
Present Article ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Experimental Programme ◽  
Iron Nickel ◽  
Water Reactors

The PETALE experimental programme in the CROCUS reactor at EPFL intends to contribute to the validation and improvement of neutron nuclear data in the MeV energy range for stainless steel, particularly in the prospect of heavy reflector elements of pressurized water reactors. It mainly consists of several transmission experiments: first, through metallic sheets of nuclear-grade stainless steel interleaved with dosimeter foils, and, successively, through its elemental components of interest – iron, nickel, and chromium. The present article describes the study for optimizing the response of the dosimetry experiments to the nuclear data of interest.

Method for Predicting Mixing Tee Thermal Fatigue in Carbon Steel Based on Austenitic Stainless Steel Operating Experience

2009 ◽  
Author(s):  
Robert O. McGill ◽  
David O. Harris ◽  
Ken Wolfe
Keyword(s):  
Stainless Steel ◽  
Carbon Steel ◽  
Austenitic Stainless Steel ◽  
Thermal Fatigue ◽  
Operating Experience ◽  
The United States ◽  
Pressurized Water ◽  
Short Period ◽  
Water Reactors ◽  
Steel Piping

Over the past ten years, a significant amount of research has been conducted regarding mixing tee thermal fatigue in pressurized water reactors prompted by the leakage event at Civaux Unit 1 in 1998. The plant experienced a leak in the reactor residual heat removal system piping after a short period of operation during plant start-up. An evaluation as to the cause of the leakage concluded that mixing of hot and cold fluid upstream from the failed austenitic stainless steel elbow resulted in thermal fatigue cracking. Recently, an assessment of susceptibility to this thermal fatigue mechanism in boiling water reactors in the United States was completed. The piping systems in these reactors where the potential for thermal mixing exists are predominantly constructed of carbon steel. Thus, an analytical model was developed for predicting mixing tee thermal fatigue in carbon steel piping based on the austenitic stainless steel piping operating experience at Civaux. This paper describes how the model was developed and presents some general findings.

INCONEL FILLER METAL 52

Alloy Digest ◽  
1992 ◽  
Vol 41 (9) ◽  
Keyword(s):  
Corrosion Resistance ◽  
Arc Welding ◽  
Filler Metal ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
High Chromium ◽  
Inconel Alloy ◽  
Gas Tungsten Arc ◽  
Alloy 690 ◽  
Water Reactors

Abstract INCONEL FILLER METAL 52 is a high chromium filler metal for gas-metal-arc and gas-tungsten-arc welding of Inconel Alloy 690 (See Alloy Digest Ni-266, March 1981). Higher chromium is beneficial in resisting stress-corrosion cracking in high purity water for pressurized water reactors and for resistance to oxidizing acids. This datasheet provides information on composition and tensile properties. It also includes information on corrosion resistance as well as joining. Filing Code: Ni-412. Producer or source: Inco Alloys International Inc..

JESSOP-SAVILLE ZIRCONIUM ALLOY

Alloy Digest ◽  
1965 ◽  
Vol 14 (3) ◽  
Keyword(s):  
Corrosion Resistance ◽  
Zirconium Alloy ◽  
Heat Treating ◽  
Pressurized Water Reactors ◽  
Neutron Absorption ◽  
High Melting Point ◽  
Absorption Properties ◽  
Pressurized Water ◽  
Excellent Corrosion Resistance ◽  
Water Reactors

Abstract JESSOP-SAVILLE ZIRCONIUM Alloy has a high melting point and possesses excellent corrosion resistance coupled with low neutron absorption properties. It is equivalent to ZIRCALOY 2. It is recommended for pressurized water reactors. This datasheet provides information on composition, physical properties, elasticity, and tensile properties as well as creep. It also includes information on corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: Zr-2. Producer or source: Jessop-Saville Ltd, Brightside Works.

Sign in / Sign up

Export Citation Format

Share Document