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

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

1978 ◽  
Vol 36 (1) ◽  
pp. 366-367
Author(s):  
D.O. Northwood ◽  
R.W. Gilbert ◽  
P.M. Kelly ◽  
P.K. Madden ◽  
D. Faulkner ◽  
...  
Keyword(s):  
Zirconium Alloys ◽  
Round Robin ◽  
Irradiation Damage ◽  
Exact Nature ◽  
Dislocation Loops ◽  
The Past ◽  
Irradiation Growth ◽  
Transmission Electron ◽  
International Round

Over the past few years there has been disagreement between laboratories on the exact nature of the damage in irradiated zirconium alloys. The main disagreement has centred on whether or not dislocation loops with c-component Burgers' vectors are formed during the irradiation. Since the presence of c-component loops was required in one of the current theories of irradiation growth and is considered in many other models, it was desirable to clear up this point and others relating to the nature of the damage such as loop size, loop concentration and the nature of the loop population, i.e. vacancy or interstitial. To this end a ‘round-robin’ series of transmission electron microscopy (TEM) examinations of neutron irradiated zirconium alloys was organized and the results are reported herein.The participants in the ‘round-robin’ included laboratories who had previously claimed to have seen evidence for c-component damage. The materials examined included zirconium and Zircaloy-2 irradiated at temperatures from 250-400°C, Table 1, the materials irradiated at 400°C providing samples with dislocation loops large enough to determine the interstitial/vacancy nature by inside/outside contrast techniques.

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.

Effects of Neutron Irradiation on the Mechanical and Electromagnetic Properties of Reactor Pressure Vessel Steels

2017 ◽  
Author(s):  
Li Chengliang ◽  
Shu Guogang ◽  
Chen Jun ◽  
Liu Yi ◽  
Liu Wei ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Power Plants ◽  
Hysteresis Loops ◽  
Reactor Pressure Vessel ◽  
Electromagnetic Properties ◽  
Irradiation Damage ◽  
Indentation Hardness ◽  
Rpv Steel ◽  
Reactor Pressure

The effect of neutron irradiation damage of reactor pressure vessel (RPV) steels is a main failure mode. Accelerated neutron irradiation experiments at 292 °C were conducted on RPV steels, followed by testing of the mechanical, electrical and magnetic properties for both the unirradiated and irradiated steels in a hot laboratory. The results showed that a significant increase in the strength, an obvious decrease in toughness, a corresponding increase in resistivity, and the clockwise turn of the hysteresis loops, resulting in a slight decrease in saturation magnetization when the RPV steel irradiation damage reached 0.0409 dpa; at the same time, the variation rate of the resistivity between the irradiated and unirradiated RPV steels shows good agreement with the variation rates of the mechanical properties parameters, such as nano-indentation hardness, ultimate tensile strength, yield strength at 0.2% offset, upper shelf energy and reference nil ductility transition temperature. Thus, as a complement to destructive mechanical testing, the resistivity variation can be used as a potentially non-destructive evaluation technique for the monitoring of the RPV steel irradiation damage of operational nuclear power plants.

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