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2 pages, 144 KiB

Open AccessEditorial

Improvement and Application of Zirconium Alloys

by Wen Qin

Metals 2018, 8(10), 794; https://doi.org/10.3390/met8100794 - 04 Oct 2018

Cited by 3 | Viewed by 2440

Abstract

n/a Full article

(This article belongs to the Special Issue Zirconium Alloys)

Research

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10 pages, 3081 KiB

Open AccessArticle

Effects of Cold Swaging on Mechanical Properties and Magnetic Susceptibility of the Zr–1Mo Alloy

by Maki Ashida, Masahiro Morita, Yusuke Tsutsumi, Naoyuki Nomura, Hisashi Doi, Peng Chen and Takao Hanawa

Metals 2018, 8(6), 454; https://doi.org/10.3390/met8060454 - 13 Jun 2018

Cited by 11 | Viewed by 3793

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, [...] Read more.

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

\bar{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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9 pages, 11323 KiB

Open AccessArticle

Phase Diagram of near Equiatomic Zr-Pd Alloy

by Mitsuhiro Matsuda, Tomohiro Nishiura, Takateru Yamamuro and Minoru Nishida

Metals 2018, 8(5), 366; https://doi.org/10.3390/met8050366 - 21 May 2018

Cited by 4 | Viewed by 3986

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 [...] Read more.

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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14 pages, 2146 KiB

Open AccessArticle

Thermophysical and Mechanical Analyses of UO₂-36.4vol % BeO Fuel Pellets with Zircaloy, SiC, and FeCrAl Claddings

by Wei Zhou and Wenzhong Zhou

Metals 2018, 8(1), 65; https://doi.org/10.3390/met8010065 - 18 Jan 2018

Cited by 7 | Viewed by 4324

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, [...] Read more.

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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16 pages, 27095 KiB

Open AccessArticle

Oxidation Behavior of Zr–1Nb Corroded in Air at 400 °C after Plasma Immersion Titanium Implantation

by Aleksei Obrosov, Alina N. Sutygina, Anton Manakhov, Sebastian Bolz, Sabine Weiß and Egor B. Kashkarov

Metals 2018, 8(1), 27; https://doi.org/10.3390/met8010027 - 02 Jan 2018

Cited by 11 | Viewed by 5880

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 [...] Read more.

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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4904 KiB

Open AccessArticle

Fabrication and Characterization of a Low Magnetic Zr-1Mo Alloy by Powder Bed Fusion Using a Fiber Laser

by Xiaohao Sun, Weiwei Zhou, Keiko Kikuchi, Naoyuki Nomura, Akira Kawasaki, Hisashi Doi, Yusuke Tsutsumi and Takao Hanawa

Metals 2017, 7(11), 501; https://doi.org/10.3390/met7110501 - 13 Nov 2017

Cited by 19 | Viewed by 5694

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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10186 KiB

Open AccessFeature PaperArticle

Precipitate Stability in a Zr–2.5Nb–0.5Cu Alloy under Heavy Ion Irradiation

by Qingshan Dong, Zhongwen Yao, Qiang Wang, Hongbing Yu, Mark A. Kirk and Mark R. Daymond

Metals 2017, 7(8), 287; https://doi.org/10.3390/met7080287 - 27 Jul 2017

Cited by 17 | Viewed by 4488

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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4440 KiB

Open AccessArticle

New Low-Sn Zr Cladding Alloys with Excellent Autoclave Corrosion Resistance and High Strength

by Ruiqian Zhang, Beibei Jiang, Chang Pang, Xun Dai, Yongduo Sun, Wei Liao, Qing Wang and Chuang Dong

Metals 2017, 7(4), 144; https://doi.org/10.3390/met7040144 - 19 Apr 2017

Cited by 6 | Viewed by 4999

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 σ_UTS = 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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