Corrosion Model for Zirconium-Niobium Alloys in Pressurized Water Reactors

被引：6
|
作者：
Likhanskii, V. V. [1 ]
Evdokimov, I. A. [1 ]
Aliev, T. N. [1 ]
Kon'kov, V. F. [2 ]
Markelov, V. A. [2 ]
Novikov, V. V. [2 ]
Khokhunova, T. N. [2 ]
机构：
[1] Troitsk Inst Innovat & Fus Res GNTs RF TRINITI, State Sci Ctr Russian Federat, Troitsk, Russia
[2] Bochvar All Russia Res Inst Inorgan Mat VNIINM, Moscow, Russia
来源：
ATOMIC ENERGY | 2014年 / 116卷 / 03期
关键词：
Oxide Film; Fuel Element; Heat Flux Density; Fuel Assembly; Engineering Model;
D O I：
10.1007/s10512-014-9839-7
中图分类号：
TL [原子能技术]; O571 [原子核物理学];
学科分类号：
0827 ; 082701 ;
摘要：
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.
引用
收藏
页码：186 / 193
页数：8
相关论文
共 50 条
12345