An Overview of Microstructural and Experimental Factors That Affect the Irradiation Growth Behavior of Zirconium Alloys

2008 ◽  
pp. 49-49-37 ◽  
Author(s):  
V Fidleris ◽  
RP Tucker ◽  
RB Adamson
Keyword(s):  
Zirconium Alloys ◽  
Growth Behavior ◽  
Irradiation Growth

scholarly journals Homogenization and Growth Behavior of Second-Phase Particles in a Deformed Zr–Sn–Nb–Fe–Cu–Si–O Alloy

Metals ◽  
2018 ◽  
Vol 8 (10) ◽  
pp. 759 ◽  
Author(s):  
Liang-Yu Chen ◽  
Peng Sang ◽  
Lina Zhang ◽  
Dongpo Song ◽  
Yan-Qiu Chu ◽  
...  
Keyword(s):  
Ostwald Ripening ◽  
Short Circuit ◽  
Zirconium Alloys ◽  
Lattice Diffusion ◽  
Growth Behavior ◽  
Homogeneous Distribution ◽  
Second Phase ◽  
Deformation Degree ◽  
Linear Distribution ◽  
Second Phase Particles

Homogeneous distribution of fine second-phase particles (SPPs) fabricated by cycles of deformation and annealing in zirconium alloys is a critical consideration for the corrosion resistance of fuel claddings. Different deformation degrees of zirconium alloys would result in distinctive microstructures, leading to a distinct growth of SPPs during subsequent annealing. Unfortunately, the homogenization and growth behavior of SPPs in deformed zirconium alloys have not been well studied. In this work, a β-quenched Zr–Sn–Nb–Fe–Cu–Si–O alloy was rolled and annealed at 580 °C or 680 °C. The morphologies, distributions, and sizes of SPPs resulting from the different processing procedures were investigated. A linear distribution of SPPs is found in the β-quenched sample. Afterward, SPPs grow and are randomly distributed during heat treatment as the deformation degree or annealing time (or temperature) increases. The homogenization and growth of SPPs are attributed to the Ostwald ripening mechanism that is governed by lattice diffusion and short-circuit diffusion. The sample with a higher deformation degree is speculated to have a larger number of defects that provide more shortcuts for the mass transfer of SPPs, thereby facilitating a homogeneous distribution of fine SPPs during annealing.

Test Reactor Irradiation Fuel Assembly Development in China

2017 ◽  
Author(s):  
Jie Ding ◽  
Yixiong Zheng ◽  
Yang Ding ◽  
Song Liu ◽  
Libing Zhu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Zirconium Alloys ◽  
Reactor Irradiation ◽  
Fuel Rods ◽  
Post Irradiation ◽  
Irradiation Growth ◽  
Corrosion Creep ◽  
Test Reactor ◽  
Irradiation Performance ◽  
Test Fuel

During the development of zirconium alloys, the irradiation in the test reactor is a critical step to comparison the irradiation properties of candidate alloys, such as corrosion, creep and irradiation growth. In this paper, a small scaled fuel assembly for test reactor irradiation is designed, which meets the needs of new zirconium alloys development. The irradiation fuel assembly (IFA) can be easily disassembled, and the test fuel rods or irradiation specimen can be easily replaced, which makes it possible to do the further post-irradiation examination in the hot cell to obtain the irradiation performance data. Now the IFA has finish fabrication and the test reactor irradiation program is planned to launch in 2017.

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.

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