Characterization of Hydrides and the α-Zr Matrix in Zirconium Alloys: Effects of Stresses, Microstructure, and Neutron Irradiation on Hydride Texture, Terminal Solid Solubility, and Dislocation Structure

2021 ◽  
pp. 786-811
Author(s):  
Pablo Vizcaino ◽  
Alejandra V. Flores ◽  
Miguel A. Vicente Alvarez ◽  
Javier R. Santisteban ◽  
Gladys Domizzi ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Dislocation Structure ◽  
Solid Solubility ◽  
Zirconium Alloys ◽  
Terminal Solid Solubility

scholarly journals Influence of kinetic effects on terminal solid solubility of hydrogen in zirconium alloys

2018 ◽  
Vol 510 ◽  
pp. 277-281 ◽  
Author(s):  
Peter Kaufholz ◽  
Maik Stuke ◽  
Felix Boldt ◽  
Marc Péridis
Keyword(s):  
Solid Solubility ◽  
Zirconium Alloys ◽  
Terminal Solid Solubility ◽  
Kinetic Effects

Critical Temperatures for Initiating and Arresting Delayed Hydride Cracking in a Zr-2.5Nb Pressure Tube

2005 ◽  
Vol 297-300 ◽  
pp. 1685-1690
Author(s):  
Young Suk Kim ◽  
Kyung Soo Im ◽  
Yong Moo Cheong
Keyword(s):  
Hydrogen Concentration ◽  
Solid Solubility ◽  
Driving Force ◽  
Stress Gradient ◽  
Zirconium Alloys ◽  
Concentration Limit ◽  
Critical Temperatures ◽  
Terminal Solid Solubility ◽  
Delayed Hydride Cracking ◽  
Hydride Cracking

The hydrogen concentration limit and critical temperatures for a delayed hydride cracking (DHC) in zirconium alloys have been reanalyzed using Kim’s DHC model that a driving force for DHC is not the stress gradient but the supersaturated hydrogen concentration or ∆C arising from a hysteresis of the terminal solid solubility on a heating and on a cooling. The DHC initiation occurs generally at the temperatures corresponding to the terminal solid solubility for precipititation (TSSP), demonstrating that the supercooling from the terminal solid solubility for dissolution (TSSD) is required to initiate the DHC. The DHC arrest temperatures correspond to the temperatures where the ∆C is reduced to zero. Therefore, we conclude that the ∆C is the driving force for the DHC and that the Kim’s DHC model is feasible.

scholarly journals Irradiation Hardening and Microstructure Characterization of Zr -1% Nb During Low Dose Neutron Irradiation

2021 ◽  
Vol 10 (2) ◽  
pp. 63-72
Author(s):  
Carolina Vazquez ◽  
Eugenia Zelaya ◽  
Ana Maria Fortis ◽  
Patricia B. Bozzano
Keyword(s):  
Neutron Irradiation ◽  
Low Dose ◽  
Zirconium Alloys ◽  
Life Time ◽  
Tensile Tests ◽  
Crystal Defects ◽  
Second Phase ◽  
Hydrogen Damage ◽  
Transmission Electron

Due to low neutron absorption cross section, high mechanical strength, high thermal conductivity and good corrosion resistance in water and steam, Zirconium alloys are widely used as fuel cladding material in nuclear reactors. During life-time of a reactor the microstructure of this alloy is affected due to, among other factors, radiation damage and hydrogen damage. In this work mechanical properties changes on neutron irradiated Zr-1wt.% Nb at low temperatures (< 100 °C) and low dose (3.5 ´ 1023 n m-2 (E > 1 MeV)) were correlated with hydrides and crystal defects evolution during irradiation. To achieve this propose, tensile tests of: 1) Non-hydrided and non-irradiated material, 2) Hydrided and non-irradiated material and 3) Hydrided and irradiated material were performed at 25 ºC and 300 ºC. Different phases, hydrides and second phase precipitates were characterized by transmission electron microscopy (TEM) techniques. For the hydrided and irradiated material, the ductility decreased sharply with respect to the hydrided and non-irradiated material, among other factors, due to the change in the microstructure produced mainly by neutron irradiation. Even if the presence of the hydride ζ (zeta) was observed, both in the irradiated and non-irradiated material, tensile tests showed that ζ-hydrides did not affect ductility, since hydrided samples are more ductile than non-hydrided samples.

Characterization of neutron irradiation damage in zirconium alloys — an international “round-robin” experiment

1979 ◽  
Vol 79 (2) ◽  
pp. 379-394 ◽  
Author(s):  
D.O. Northwood ◽  
R.W Gilbert ◽  
L.E. Bahen ◽  
P.M. Kelly ◽  
R.G. Blake ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Zirconium Alloys ◽  
Round Robin ◽  
Irradiation Damage ◽  
International Round

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.

Effect of Irradiation on Terminal Solid Solubility of Hydrogen in Zr-2.5Nb

2018 ◽  
pp. 1136-1166
Author(s):  
Heidi M. Nordin ◽  
Vicky Hilton ◽  
Andrew W. Buyers ◽  
Christopher E. Coleman ◽  
Glenn A. McRae
Keyword(s):  
Solid Solubility ◽  
Terminal Solid Solubility
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