Effect of Copper and Nickel on the Neutron Irradiation Damage in Iron Alloys

1994 ◽  
Vol 373 ◽  
Author(s):  
D.T. Hoelzer ◽  
F. Ebrahimi
Keyword(s):  
Neutron Irradiation ◽  
Dislocation Loop ◽  
Alloy Composition ◽  
Iron Alloys ◽  
Irradiation Damage ◽  
Dislocation Loops ◽  
Number Density ◽  
Tem Analysis ◽  
Vacancy Loops ◽  
Effect Of Copper

AbstractIron alloys containing copper and nickel were irradiated at 288°C to a fluence of 4.63 × 1019 neutrons/cm2. Neutron irradiation produced defects which were observable by TEM in all of the iron alloys studied. The TEM analysis of the defects showed them to be interstitial dislocation loops with a < 100 > and a/2 < 111 > Burgers vectors. The size, the number density, and the Burgers vector of dislocations were affected by the alloy composition. The addition ofcopper and nickel decreased the dislocation loop size and increased the fraction of a/2 < 111 > loops. No voids or vacancy loops were observed in the irradiated iron alloys. The results are discussed in terms of dislocation loop nucleation and growth.

Effects of dislocation loop morphology in titanium on diffraction contrast

1984 ◽  
Vol 42 ◽  
pp. 530-531
Author(s):  
C.D. Cann ◽  
M. Griffiths ◽  
R.C. Styles
Keyword(s):  
Neutron Irradiation ◽  
Dislocation Loop ◽  
Dislocation Loops ◽  
Fringe Contrast ◽  
Burgers Vector ◽  
Vacancy Loops ◽  
Diffraction Contrast ◽  
Hexagonal Metals

In hexagonal metals, such as Zr and Ti, both vacancy and interstitial loops may form (or develop) under neutron irradiation. These loops generally have a perfect Burgers vector, b, although faulted vacancy loops have been observed in Zr. Recently, Griffiths et al. have reported the observation of fringe contrast associated with dislocation loops in Ti. These loops had a b in the direction, but the presence of the fringe contrast was evidence that it differed from a perfect vector. An example of this fringe contrast is shown in Fig. 1. Griffiths et al. suggested two possibilities for the origin of this fringe contrast. The first was that this contrast resulted from impurities in the vicinity of these loops and the second was that the loops were faulted with . However, for both of these possibilities it was difficult to explain that the fringes were observed only under conditions of inside contrast and that fringed loops were observed in only certain grains.

Localized Deformation in Neutron-Irradiated Zirconium and Zircaloy-4

2007 ◽  
Vol 561-565 ◽  
pp. 1769-1772 ◽  
Author(s):  
Naoyuki Hashimoto ◽  
T.S. Byun
Keyword(s):  
Neutron Irradiation ◽  
Plastic Instability ◽  
Number Density ◽  
Localized Deformation ◽  
Tem Analysis ◽  
Hexagonal Close Packed ◽  
Channel Deformation ◽  
Transmission Electron ◽  
Hcp Materials ◽  
Higher Doses

The effects of neutron-irradiation near 80°C on the deformation behavior of hexagonal close packed (hcp) materials, zirconium and zircalloy-4, were investigated by transmission electron microscopy (TEM). Particular emphasis is placed on the deformation microstructure responsible for the changes in mechanical behavior. Neutron irradiation at low temperature up to 1 dpa induced a high number density of defect clusters, which resulted in irradiation-induced hardening. Dislocation channel deformation is observed for doses greater than 0.1 dpa, and is coincident with prompt plastic instability at yield. TEM analysis suggests that the loss of work hardening capacity in irradiated zirconium and zircaloy-4 at higher doses is mainly due to dislocation channels that are formed under a high local resolved shear stress, leading to the observed localized deformation.

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.

Effect of Self-Interstitial Diffusion Anisotropy in Electron-Irradiated Zirconium. A Cluster Dynamics Modeling

2005 ◽  
Vol 237-240 ◽  
pp. 659-664
Author(s):  
Frédéric Christien ◽  
Alain Barbu
Keyword(s):  
Point Defects ◽  
Thin Foil ◽  
Elastic Interaction ◽  
Dislocation Loops ◽  
Considerable Influence ◽  
Dynamics Modeling ◽  
Dynamics Model ◽  
Diffusion Anisotropy ◽  
Vacancy Loops ◽  
Transmission Electron

Irradiation of metals leads to the formation of point-defects (vacancies and selfinterstitials) that usually agglomerate in the form of dislocation loops. Due to the elastic interaction between SIA (self-interstitial atoms) and dislocations, the loops absorb in most cases more SIA than vacancies. That is why the loops observed by transmission electron microscopy are almost always interstitial in nature. Nevertheless, vacancy loops have been observed in zirconium following electron or neutron irradiation (see for example [1]). Some authors proposed that this unexpected behavior could be accounted for by SIA diffusion anisotropy [2]. Following the approach proposed by Woo [2], the cluster dynamics model presented in [3] that describes point defect agglomeration was extended to the case where SIA diffusion is anisotropic. The model was then applied to the loop microstructure evolution of a zirconium thin foil irradiated with electrons in a high-voltage microscope. The main result is that, due to anisotropic SIA diffusion, the crystallographic orientation of the foil has considerable influence on the nature (vacancy or interstitial) of the loops that form during irradiation.

Effect of Neutron Irradiation on the Mechanical Properties of an A508 CL2 and 15Kh2NMFA Irradiated in the NOMAD_3 Rig in the BR2 Cooling Water

2021 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Cooling Water ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Significant Loss ◽  
Irradiation Temperature ◽  
Irradiation Damage ◽  
Transition Curves ◽  
Reactor Pressure

Abstract The typical operating temperatures of a nuclear reactor pressure vessel in a PWR are between 290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 260°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing happens below approximately 230°C. The NOMAD_3 irradiation consisted in total of 24 Charpy sized samples from an A508 Cl.2 forging and a 15Kh2NMFA material. They were irradiated to three various fluences between 1.55 and 7.90 × 1019 n/cm2 (E &gt; 1MeV) at approximately 100°C. The hardening of the A508 Cl.2 was between 260 and 400 MPa which was much higher than the NOMAD_0 properties which were irradiated at approximately 280°C. The tensile tests of irradiated materials are all characterized by a significant loss of work hardening capacity leading to plastic flow localization promptly after the yield strength is reached. This affects also the shape of the Charpy impact transition curves. The radiation embrittlement derived from Charpy impact tests, ΔT41J, is up to 156°C for the highest fluence. For this irradiation, the embrittlement to hardening ratio was also around 0.43 +/−0.2°C/MPa as it was found in the previous campaign NOMAD_0. This paper discusses the tensile, hardness and impact properties of the NOMAD_3 irradiation campaign. It is compared to the NOMAD_0 with respect to effect of irradiation temperature and annealing recovery.

Defects In Erbium/Oxygen Implanted Silicon

1996 ◽  
Vol 442 ◽  
Author(s):  
X. Duan ◽  
J. Palm ◽  
B. Zheng ◽  
M. Morse ◽  
J. Michel ◽  
...  
Keyword(s):  
Habit Plane ◽  
Dislocation Loop ◽  
Systematic Study ◽  
Sharp Contrast ◽  
Dislocation Loops ◽  
Different Temperatures ◽  
Free Material ◽  
Habit Planes

AbstractA systematic study of defects in the Er/O implanted silicon was conducted using TEM, HRTEM and SIMS. Defect-free material was obtained after the annealing of 400 keV Er+ implanted (100)Si. In sharp contrast, several forms of secondary defects consisting of dislocations, dislocation loops and precipitates were induced upon annealing at different temperatures in the 4.5 MeV implanted (100)Si sample. The isothermal evolution of the defects and reactions between dopants and defects were studied. Oxygen tends to segregate into the dislocation loop zones, where platelet precipitates with habit planes of {111} were found. Following dissociation of oxygen and erbium, plate-like Er precipitates were generated, which are most likely ErSi2 with a habit plane of {111}

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