Corrosion Model for Zirconium-Niobium Alloys in Pressurized Water Reactors

• Published in: Atomic energy (New York, N.Y.) Vol. 116; no. 3; pp. 186 - 193
• Main Authors: Likhanskii, V. V., Evdokimov, I. A., Aliev, T. N., Kon’kov, V. F., Markelov, V. A., Novikov, V. V., Khokhunova, T. N.
• Format: Journal Article
• Language: English
• Published: Boston Springer US 01.07.2014; Springer; Springer Nature B.V
• Subjects: Alloys; Analysis; Autoclaves; Building materials; Corrosion; Corrosion and anti-corrosives; Engineering; Experiments; Hadrons; Heat; Heavy Ions; Hydrogen; Lithium; Mathematical models; Neutron irradiation; Nuclear Chemistry; Nuclear Energy; Nuclear fuel elements; Nuclear fuels; Nuclear Physics; Nuclear reactors; Oxidation; Physics; Physics and Astronomy; Pressurized water reactors; Reactors; Specialty metals industry; Studies; Zirconium; Zirconium alloys; Fuel Element; Fuel Assembly; Engineering Model; Oxide Film; Heat Flux Density
• ISSN: 1063-4258; 1573-8205
• DOI: 10.1007/s10512-014-9839-7

An engineering model of corrosion of zirconium-niobium alloys is described. It takes account of the alloying composition, the content of lithium and boron in the coolant, the heat flux on the surface of fuel elements and the intensity of the neutron irradiation. The parametric dependences used in the model are based on the results of tests performed in autoclaves and research reactors. The results of verification of the model on data from post-reactor studies of PWR and VVER fuel assemblies operating in nominal regimes are presented.