Special Issue "Zirconium Alloys"

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

Deadline for manuscript submissions: closed (30 April 2018)

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 (7 papers)

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Research

Open AccessArticle Effects of Cold Swaging on Mechanical Properties and Magnetic Susceptibility of the Zr–1Mo Alloy
Metals 2018, 8(6), 454; https://doi.org/10.3390/met8060454
Received: 20 April 2018 / Revised: 23 May 2018 / Accepted: 12 June 2018 / Published: 13 June 2018
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Abstract
Zr alloy is expected to decrease the artifact volume of magnetic resonance imaging (MRI) due to its relatively small magnetic susceptibility. To improve the mechanical properties of a Zr–1mass%Mo alloy that yielded a reduced artifact volume during MRI, the alloy was melted, hot-forged,
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Zr alloy is expected to decrease the artifact volume of magnetic resonance imaging (MRI) due to its relatively small magnetic susceptibility. To improve the mechanical properties of a Zr–1mass%Mo alloy that yielded a reduced artifact volume during MRI, the alloy was melted, hot-forged, and cold-swaged with area reduction ratios of 30%, 50%, 60%, 70%, and 84%. The effects of cold swaging on the microstructure, mechanical properties, and magnetic susceptibility of the alloy were investigated. Before cold swaging, the microstructure consisted of laminated and layered α- and β-phases; however, after cold swaging, the α- and β-phases were bent and distorted, and the α-phase became oriented along the {10 1¯ 0} plane. The ultimate tensile strength and elongation to fracture of the Zr–1Mo alloy after cold swaging with an 84% area reduction were 1001 MPa and 10.7%, respectively. The alloy only experienced work-hardening when subjected to large deformations. On the other hand, the change in magnetic susceptibility with cold-swaging was small, from 13.85 × 10−9 to 14.87 × 10−9 m3·kg−1. Thus, a good balance of mechanical properties and low magnetic susceptibility in the Zr–1Mo alloy was obtained by cold swaging. Therefore, this alloy is suitable for utilization in medical devices and is expected to decrease the artifact volume. Full article
(This article belongs to the Special Issue Zirconium Alloys)
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Open AccessArticle Phase Diagram of near Equiatomic Zr-Pd Alloy
Metals 2018, 8(5), 366; https://doi.org/10.3390/met8050366
Received: 28 April 2018 / Revised: 16 May 2018 / Accepted: 18 May 2018 / Published: 21 May 2018
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Abstract
The exact eutectoid and peritectoid temperatures in near equiatomic Zr-Pd compositions have been determined by using the diffusion couple method and microstructure analysis. The crystal structure of Zr13Pd12 compound were estimated to be orthorhombic with a = 1.78 nm, b
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The exact eutectoid and peritectoid temperatures in near equiatomic Zr-Pd compositions have been determined by using the diffusion couple method and microstructure analysis. The crystal structure of Zr13Pd12 compound were estimated to be orthorhombic with a = 1.78 nm, b = 0.80 nm and c = 1.00 nm from the electron diffraction experiments. The Zr13Pd12 compound is formed at 1100 ± 2 K with a peritectoid reaction between Zr2Pd and ZrPd compounds. The ZrPd compound transforms to Zr13Pd12 and Zr9Pd11 compounds by a eutectoid reaction at 1028 ± 4 K. Based on these results, the phase diagram of near equiatomic Zr-Pd binary system is reconstructed. Full article
(This article belongs to the Special Issue Zirconium Alloys)
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Open AccessArticle Thermophysical and Mechanical Analyses of UO2-36.4vol % BeO Fuel Pellets with Zircaloy, SiC, and FeCrAl Claddings
Metals 2018, 8(1), 65; https://doi.org/10.3390/met8010065
Received: 31 October 2017 / Revised: 14 January 2018 / Accepted: 15 January 2018 / Published: 18 January 2018
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Abstract
The thermophysical performance and solid mechanics behavior of UO2-36.4vol % BeO fuel pellets cladded with Zircaloy, SiC, and FeCrAl, and Zircaloy cladding materials coated with SiC and FeCrAl, are investigated based on simulation results obtained by the CAMPUS code. In addition,
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The thermophysical performance and solid mechanics behavior of UO2-36.4vol % BeO fuel pellets cladded with Zircaloy, SiC, and FeCrAl, and Zircaloy cladding materials coated with SiC and FeCrAl, are investigated based on simulation results obtained by the CAMPUS code. In addition, the effect of coating thickness (0.5, 1 and 1.5 mm) on fuel performance and mechanical interaction is discussed. The modeling results show that Zircaloy claddings are more effective in decreasing fuel centerline temperature and fission gas release than other kinds of cladding material because of the smaller gap between cladding and fuel at the same burnup. SiC claddings and SiC-coated Zircaloy claddings possess smaller plenum pressure than other kinds of cladding. SiC claddings contribute more to fuel radial displacement but less to fuel axial displacement. FeCrAl claddings exhibit very different radial and axial displacements in different axial positions. FeCrAl-coated Zircaloy claddings have a lower fuel centerline temperature than Zircaloy claddings at burnup below 850 MWh/kg U, but a higher fuel centerline temperature at higher burnup. The gap between FeCrAl-coated Zircaloy claddings and fuel pellets closes earlier than that of Zircaloy claddings. SiC-coated claddings increase fuel radial and axial displacements, and cladding axial displacements of inner and outer cladding surfaces. Full article
(This article belongs to the Special Issue Zirconium Alloys)
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Open AccessArticle Oxidation Behavior of Zr–1Nb Corroded in Air at 400 °C after Plasma Immersion Titanium Implantation
Metals 2018, 8(1), 27; https://doi.org/10.3390/met8010027
Received: 4 October 2017 / Revised: 24 December 2017 / Accepted: 30 December 2017 / Published: 2 January 2018
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Abstract
In this paper, the influence of plasma immersion titanium implantation into the zirconium alloy Zr-1Nb on the oxidation behavior at 400 °C for 5, 24, 72, and 240 h in air under normal atmospheric pressure (101.3 kPa) was shown. The influence of implantation
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In this paper, the influence of plasma immersion titanium implantation into the zirconium alloy Zr-1Nb on the oxidation behavior at 400 °C for 5, 24, 72, and 240 h in air under normal atmospheric pressure (101.3 kPa) was shown. The influence of implantation on the protective properties of the modified layer was shown. The valence of the oxides before and after implantation was analyzed by means of X-ray photoelectron spectroscopy (XPS). Grazing incidence X-ray diffraction (GIXRD) was carried out to examine the phase composition after titanium ion implantation and oxidation. Differential scanning calorimetry (DSC) revealed that titanium implantation exhibited effects of stabilizing the β phase. The formation of the t-ZrO2 and m-ZrO2 was observed during the oxidation of the as-received and modified Zr-1Nb. The measurement of weight gain showed an improvement in oxidation resistance of Ti implanted Zr-1Nb at the oxidation up to 24 h when compared with that of the as-received Zr-1Nb. However, at longer oxidation cycle the oxidation rate of Ti-implanted zirconium alloy is the same with the as-received alloy, which attributed to the layer thickness. Nevertheless, the corrosion of the Ti-implanted alloy is more uniform, while a local corrosion and cracks was detected on the surface of the as-received alloy. Full article
(This article belongs to the Special Issue Zirconium Alloys)
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Open AccessArticle Fabrication and Characterization of a Low Magnetic Zr-1Mo Alloy by Powder Bed Fusion Using a Fiber Laser
Metals 2017, 7(11), 501; https://doi.org/10.3390/met7110501
Received: 19 October 2017 / Revised: 4 November 2017 / Accepted: 8 November 2017 / Published: 13 November 2017
Cited by 1 | PDF Full-text (4904 KB) | HTML Full-text | XML Full-text
Abstract
A low magnetic Zr-1Mo alloy was fabricated by a powder bed fusion (PBF) process using a fiber laser. The microstructure, surface morphology, and pore distribution of the as-built Zr-1Mo alloy were observed. Its magnetic susceptibility and Vickers hardness were evaluated by magnetic susceptibility
[...] Read more.
A low magnetic Zr-1Mo alloy was fabricated by a powder bed fusion (PBF) process using a fiber laser. The microstructure, surface morphology, and pore distribution of the as-built Zr-1Mo alloy were observed. Its magnetic susceptibility and Vickers hardness were evaluated by magnetic susceptibility balance and a microindentation tester, respectively. The as-built Zr-1Mo alloy mainly consisted of an α′ phase with an acicular structure. From the processing maps of the surface morphology and pore distribution, open pores on the top surface due to the lack of fusion corresponded to grid-like distributed pores, and large pores corresponded to balling particles on the top surface. The Vickers hardness was influenced by the oxygen and nitrogen contents rather than the porosity. The magnetic susceptibilities of the as-built Zr-1Mo alloy still were one-third those of Ti-6Al-4V and Ti-6Al-7Nb, thus PBF can be applicable to the fabrication process for the low magnetic Zr-1Mo alloy. Full article
(This article belongs to the Special Issue Zirconium Alloys)
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Open AccessFeature PaperArticle Precipitate Stability in a Zr–2.5Nb–0.5Cu Alloy under Heavy Ion Irradiation
Metals 2017, 7(8), 287; https://doi.org/10.3390/met7080287
Received: 4 July 2017 / Revised: 21 July 2017 / Accepted: 24 July 2017 / Published: 27 July 2017
Cited by 1 | PDF Full-text (10186 KB) | HTML Full-text | XML Full-text
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
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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; https://doi.org/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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