Mechanical and Corrosion Characteristics of Zr + 2.5% Nb Zirconium Reactor Alloy

被引:0
作者
A. I. Stukalov
V. M. Gritsina
T. P. Chernyaeva
D. A. Baturevich
机构
[1] National Scientific Center “Kharkov Physicotechnical Institute,
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来源
Materials Science | 2000年 / 36卷
关键词
Zirconium; Corrosion Resistance; Transverse Direction; Nuclear Power Plant; Corrosion Test;
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摘要
The results of structural and phase hardening of pipes made of Zr + 2.5% Nb alloy show that ultrahigh-frequency thermal treatment of pipes (fast heating to the temperature of existence of the β-phase followed by sharp cooling and annealing in the high-temperature range of the α-phase) destroys the texture and forms a fine-grained structure (the grain diameter is about 1 μm) with numerous transitional twins and a high density of precipitations of the secondary β-niobium phase (∼  1016 cm−3). In this state, the alloy is rather strong and plastic (at room temperature, σu ≈ 650 MPa, σ0.2 ≈ 550 MPa, and δ ∼ 20% both in the longitudinal and transverse directions). The efficiency of hardening by ultrahigh-frequency thermal treatment is not reduced with increase in the temperature of testing up to 500°C. Corrosion tests of channel pipes made of Zr + 2.5% Nb alloy subjected to ultrahigh-frequency thermal treatment in water containing various amounts of oxygen (from 0.1–0.3 to 600 mg/kg) at temperatures of 285–350°C for 700–6600 h under static conditions and in reactor water of the Ignalina Nuclear Power Plant for ∼ 5000 h under dynamic conditions showed that the corrosion resistance of this alloy is on a par with the corrosion resistance of the material of assembly channels of high-power channel reactors subjected to a standard treatment.
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页码:669 / 674
页数:5
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