Special Issue "Zirconium Alloys"

A special issue of Metals (ISSN 2075-4701).

Deadline for manuscript submissions: 31 October 2017

Special Issue Editor

Guest Editor
Dr. Wen Qin

Department of Mechanical Engineering, University of Saskatchewan, Saskatoon, Canada
Website | E-Mail

Special Issue Information

Dear Colleagues,

Zirconium alloys are used as the primary structural components in the nuclear power industry, because of their low thermal neutron absorption cross section, high corrosion resistance, good ductility, and satisfactory strength. In general, several interconnected problems are given special attention: (1) Corrosion. Corrosion of zirconium alloys is one of the main factors in the degradation of zirconium alloys during service. The surface reaction between the zirconium cladding and the coolant water in nuclear reactors leads to the oxidation of the cladding and the release of hydrogen. The pick-up of hydrogen in zirconium alloys induces the embrittlement of the tubes due to hydride formation. Therefore, there is great incentive to minimize the amount of oxidation and hydriding that materials experience in reactors. (2) Coating. The coating can provide a protective layer for the zirconium alloy that can reduce oxidation and hydrogen pick-up. The protective coating prepared on the zirconium cladding tube can effectively increase safety margins of nuclear plants. In general, the mechanical properties, the adherence between the coating and the zirconium substrate, the stability under irradiation and self-healing are the main problems that need to be taken into account. (3) Irradiation-induced damage. During service in nuclear reactors, zirconium alloys are exposed to neutron irradiation. Neutron irradiation can affect microstructural evolution, and the mechanical and corrosion properties of zirconium alloys. Irradiation-induced damage is one of the most important factors affecting the lifetime of zirconium alloy components in reactors.

A new generation of reactors will offer higher fuel burn-up, higher efficiency and excellent safety of operation. The performance and high efficiency of these advanced reactors are linked to more severe service environments. New zirconium alloys with improved resistance to the environment of high temperature, high pressure, high corrosion and high radiation field are necessary.

Dr. Wen Qin
Guest Editor

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Keywords

  • Zirconium alloys
  • Mechanical properties
  • Corrosion
  • Oxidation
  • Hydrogen embrittlement
  • Hydride
  • Coating
  • Irradiation
  • Defects

Published Papers (2 papers)

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Research

Open AccessFeature PaperArticle Precipitate Stability in a Zr–2.5Nb–0.5Cu Alloy under Heavy Ion Irradiation
Metals 2017, 7(8), 287; doi:10.3390/met7080287
Received: 4 July 2017 / Revised: 21 July 2017 / Accepted: 24 July 2017 / Published: 27 July 2017
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Abstract
The stability of precipitates in Zr–2.5Nb–0.5Cu alloy under heavy ion irradiation from 100 °C to 500 °C was investigated by quantitative Chemi-STEM EDS analysis. Irradiation results in the crystalline to amorphous transformation of Zr2Cu between 200 °C and 300 °C, but
[...] Read more.
The stability of precipitates in Zr–2.5Nb–0.5Cu alloy under heavy ion irradiation from 100 °C to 500 °C was investigated by quantitative Chemi-STEM EDS analysis. Irradiation results in the crystalline to amorphous transformation of Zr2Cu between 200 °C and 300 °C, but the β–Nb remains crystalline at all temperatures. The precipitates are found to be more stable in starting structures with multiple boundaries than in coarse grain structures. There is an apparent increase of the precipitate size and a redistribution of the alloying element in certain starting microstructures, while a similar size change or alloying element redistribution is not detected or only detected at a much higher temperature in other starting microstructures after irradiation. Full article
(This article belongs to the Special Issue Zirconium Alloys)
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Open AccessArticle New Low-Sn Zr Cladding Alloys with Excellent Autoclave Corrosion Resistance and High Strength
Metals 2017, 7(4), 144; doi:10.3390/met7040144
Received: 21 March 2017 / Revised: 12 April 2017 / Accepted: 17 April 2017 / Published: 19 April 2017
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Abstract
It is expected that low-Sn Zr alloys are a good candidate to improve the corrosion resistance of Zr cladding alloys in nuclear reactors, presenting excellent corrosion resistance and high strength. The present work developed a new alloy series of Zr-0.25Sn-0.36Fe-0.11Cr-xNb (
[...] Read more.
It is expected that low-Sn Zr alloys are a good candidate to improve the corrosion resistance of Zr cladding alloys in nuclear reactors, presenting excellent corrosion resistance and high strength. The present work developed a new alloy series of Zr-0.25Sn-0.36Fe-0.11Cr-xNb (x = 0.4~1.2 wt %) to investigate the effect of Nb on autoclave corrosion resistance. Alloy ingots were prepared by non-consumable arc-melting, solid-solutioned, and then rolled into thin plates with a thickness of 0.7 mm. It was found that the designed low-Sn Zr alloys exhibit excellent corrosion resistances in three out of pile autoclave environments (distilled water at 633 K/18.6 MPa, 70 ppm LiOH solution at 633 K/18.6 MPa, and superheated water steam at 673 K/10.3 MPa), as demonstrated by the fact of the Zr-0.25Sn-0.36Fe-0.11Cr-0.6Nb alloy shows a corrosion weight gain ΔG = 46.3 mg/dm2 and a tensile strength of σUTS = 461 MPa following 100 days of exposure in water steam. The strength of the low-Sn Zr alloy with a higher Nb content (x = 1.2 wt %) is enhanced up to 499 MPa, comparable to that of the reference high-Sn N36 alloy (Zr-1.0Sn-1.0Nb-0.25Fe, wt %). Although the strength improvement is at a slight expense of corrosion resistance with the increase of Nb, the corrosion resistance of the high-Nb alloy with x = 1.2 (ΔG = 90.4 mg/dm2 for 100-day exposure in the water steam) is still better than that of N36 (ΔG = 103.4 mg/dm2). Full article
(This article belongs to the Special Issue Zirconium Alloys)
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