Deoxidation Process of Oxidized Zirconium Alloy

2020 ◽  
Vol 993 ◽  
pp. 22-28
Author(s):  
Jun Song Zhang ◽  
Chong Sheng Long ◽  
Jing Jing Liao ◽  
Tian Guo Wei ◽  
Zhong Bo Yang
Keyword(s):  
Oxide Film ◽  
Zirconium Alloy ◽  
Zirconium Alloys ◽  
Oxide Thickness ◽  
Sem Analysis ◽  
Corrosion Mechanism ◽  
Ion Transfer ◽  
Deoxidation Process ◽  
Oxidized Zirconium ◽  
Pure Steam

When zirconium alloy is corroded, an oxide film is formed on the surface, which hinders the ion transfer during the corrosion process. Therefore, the analysis of the oxide film is an important part of the research on the corrosion resistance of zirconium alloys. In this paper, two kinds of Zr-Sn-Nb alloys were corroded in 400 °C/10.3 MPa pure steam and 500 °C/10.3 MPa pure steam in autoclave to obtain samples with oxide thickness of 14 um and 18 um respectively. Then they were annealed at 800 °C at a pressure of 10-4 Pa for 18 h. XRD and WDS studies were used to analyze the structure and oxygen content of the oxide film after annealing. The results indicate that the oxide films of alloys change from zirconium dioxide to zirconium after annealing. The oxygen diffuses into the substrate and its content decreases continuously with increasing diffusion distance. Combined with the SEM analysis of cross-section samples, it is found that the annealed samples are composed of several layers. An oxygen-saturated zirconium layer, a transitional layer with micro-cracks, an oxygen-dissolved α-Zr layer and a β-Zr layer are identified. Based on these results, the mechanism of the ion transfer in the oxide film during annealing is analyzed deeply. It is proposed that space charges in the oxide film have a major impact on deoxidation kinetics. This study provides a new research method for the corrosion mechanism of zirconium alloys.

Control of hydrogen absorption by nickel films obtained upon magnetic spraying of zirconium alloy using the thermoEMF method

2020 ◽  
Vol 86 (8) ◽  
pp. 32-37
Author(s):  
V. V. Larionov ◽  
Xu Shupeng ◽  
V. N. Kudiyarov
Keyword(s):  
Magnetron Sputtering ◽  
Hydrogen Concentration ◽  
Hydrogen Absorption ◽  
Zirconium Alloy ◽  
Hydrogen Adsorption ◽  
Zirconium Alloys ◽  
Nickel Film ◽  
Protective Film ◽  
Nickel Films ◽  
Hydrogen Penetration

Nickel films formed on the surface of zirconium alloys are often used to protect materials against hydrogen penetration. Hydrogen adsorption on nickel is faster since the latter actively interacts with hydrogen, oxidizes and forms a protective film. The goal of the study is to develop a method providing control of hydrogen absorption by nickel films during vacuum-magnetron sputtering and hydrogenation via measuring thermoEMF. Zirconium alloy E110 was saturated from the gas phase with hydrogen at a temperature of 350°C and a pressure of 2 atm. A specialized Rainbow Spectrum unit was used for coating. It is shown that a nickel film present on the surface significantly affects the hydrogen penetration into the alloy. A coating with a thickness of more than 2 μm deposited by magnetron sputtering on the surface of a zirconium alloy with 1% Nb, almost completely protects the alloy against hydrogen penetration. The magnitude of thermoemf depends on the hydrogen concentration in the zirconium alloy and film thickness. An analysis of the hysteresis width of the thermoEMF temperature loop and a method for determining the effective activation energy of the conductivity of a hydrogenated material coated with a nickel film are presented. The results of the study can be used in assessing the hydrogen concentration and, hence, corrosion protection of the material.

Determination of the Hydrogen Concentration in Zr-2.5Nb Alloy by a Resonant Ultrasound Spectroscopy

2006 ◽  
Vol 321-323 ◽  
pp. 1576-1579
Author(s):  
Yong Moo Cheong ◽  
Young Suk Kim
Keyword(s):  
Hydrogen Concentration ◽  
Zirconium Alloy ◽  
Nuclear Reactor ◽  
Zirconium Alloys ◽  
Resonant Ultrasound Spectroscopy ◽  
Heavy Water Reactor ◽  
Resonant Ultrasound ◽  
Non Destructive

Zirconium alloys are used for many applications in nuclear components, such as the pressure tube material in a pressurized heavy water reactor, nuclear fuel cladding, etc. One of the problems during the operation of a nuclear reactor is the degradation of the zirconium alloys, which is due to an increase of the hydrogen content in the zirconium alloy. Therefore a non-destructive determination of the hydrogen concentration in zirconium alloy is one of the important issues that need to be addressed. The resonant ultrasound spectroscopy (RUS) technique is evaluated for a characterization of the hydrogen concentration in Zr-2.5Nb alloy. Referring to the terminal solid solubility for dissolution (TSSD) of Zr-2.5Nb alloy, the plot of the mechanical damping coefficient (Q-1) versus the temperature or the deviation of the resonant frequency for the temperature (df/dT) versus the temperature was correlated for the hydrogen concentration in Zr-2.5Nb alloy. It was found that the temperature at an abrupt change of the slope can be correlated with the hydrogen concentration of the Zr-2.5Nb alloy.

Examinations of the Corrosion Mechanism of Zirconium Alloys

2009 ◽  
pp. 615-615-29 ◽  
Author(s):  
HJ Beie ◽  
A Mitwalsky ◽  
F Garzarolli ◽  
H Ruhmann ◽  
HJ Sell
Keyword(s):  
Zirconium Alloys ◽  
Corrosion Mechanism

scholarly journals Hydride Rim Formation in E110 Zirconium Alloy during Gas-Phase Hydrogenation

Metals ◽  
2020 ◽  
Vol 10 (2) ◽  
pp. 247
Author(s):  
Viktor Kudiiarov ◽  
Ivan Sakvin ◽  
Maxim Syrtanov ◽  
Inga Slesarenko ◽  
Andrey Lider
Keyword(s):  
Mechanical Properties ◽  
Gas Phase ◽  
Zirconium Alloy ◽  
Zirconium Alloys ◽  
Subsequent Formation ◽  
Delayed Hydride Cracking ◽  
Hydrogen Penetration ◽  
Significant Change ◽  
Hydride Cracking

The work is devoted to the study of the laws of the formation of a hydride rim in E110 zirconium alloy claddings during gas-phase hydrogenation. The problem of hydrogen penetration and accumulation and the subsequent formation of hydrides in the volume of zirconium cladding tubes of water-cooled power reactors remain relevant. The formation of brittle hydrides in a zirconium matrix firstly, leads to a significant change in the mechanical properties, and secondly, can cause the destruction of the claddings by the mechanism of delayed hydride cracking. The degree of the hydride’s effect on the mechanical properties of zirconium cladding is mainly determined by the features of the hydride’s distribution and orientation. The problem of hydride rim formation in zirconium alloys with niobium is quite new and poorly studied. Therefore, the study of hydride rim formation in Russian zirconium alloy is important and necessary for predicting the behavior of claddings during the formation of the hydride rim.

Phase transformation of oxide film in zirconium alloy in high temperature hydrogenated water

2015 ◽  
Vol 99 ◽  
pp. 134-144 ◽  
Author(s):  
Taeho Kim ◽  
Jongjin Kim ◽  
Kyoung Joon Choi ◽  
Seung Chang Yoo ◽  
Seunghyun Kim ◽  
...  
Keyword(s):  
Phase Transformation ◽  
High Temperature ◽  
Oxide Film ◽  
Zirconium Alloy

The use of oxidized zirconium alloy in knee arthroplasty

2012 ◽  
Vol 9 (4) ◽  
pp. 409-421 ◽  
Author(s):  
Thomas J Heyse ◽  
Steven B Haas ◽  
Turgay Efe
Keyword(s):  
Knee Arthroplasty ◽  
Zirconium Alloy ◽  
Oxidized Zirconium

General Corrosion of Chromium-Coated Zirconium- and Titanium-Based Alloys in Supercritical Water at 500 °C

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
K. Khumsa-Ang ◽  
M. Edwards ◽  
S. Rousseau
Keyword(s):  
Supercritical Water ◽  
Fuel Cycle ◽  
Neutron Cross Section ◽  
Zirconium Alloys ◽  
Oxide Thickness ◽  
General Corrosion ◽  
Coating Materials ◽  
Grade 5 ◽  
Oxide Phases ◽  
Enriched Uranium

Abstract The 300 MWel small Canadian supercritical water-cooled reactor (SCWR), which is a scaled-down version of the original 1200 MWel concept, has a smaller core, uses low enriched uranium fuel instead of a plutonium–thorium fuel, and features a lower (maximum) cladding temperature of 500 °C. The lower cladding temperature may permit the use of different alloys, including zirconium alloys, which had been ruled out as candidates for the Canadian SCWR, whose cladding temperature may reach 850 °C. The potential to use zirconium alloys is exciting because they have a low neutron cross section, which in turn means that fewer neutrons are lost, and the fuel can be used more efficiently. One advantage, for example,, is that the fuel cycle can be lengthened. In this paper, we report on the results of corrosion experiments used to screen zirconium- and titanium-based alloys as well as corrosion-resistant coating materials such as Cr and Al as potential candidates for fuel cladding in the small Canadian SCWR. These experiments were conducted in a refreshed autoclave in deaerated supercritical water at 500 °C and 23.5 MPa. After exposure, the weight gain was measured, and the oxide thickness and the oxide phases were examined. Of all materials, the coated and uncoated Ti-grade 2 and Ti-grade 5 alloys met our screening qualification criteria, however, Al/Cr-coated zirconium coupons showed notable improvement and will be explored further in future testing.

Corrosion characteristics of zirconium alloy with a high temperature pre-formed oxide film

2005 ◽  
Vol 388 (2) ◽  
pp. 279-283 ◽  
Author(s):  
Hyun Seon Hong ◽  
Yongseung Yun ◽  
Kyung Sub Lee
Keyword(s):  
High Temperature ◽  
Oxide Film ◽  
Zirconium Alloy
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