Characterisation of Neutron Irradiation Damage in Zirconium Alloys PDF Download
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Author: P. M. Kelly
Publisher:
ISBN: 9780642997654
Category : Zirconium alloys
Languages : en
Pages : 9
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Book Description
Author: P. M. Kelly
Publisher:
ISBN: 9780642997654
Category : Zirconium alloys
Languages : en
Pages : 9
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Book Description
Author: D. O. Northwood
Publisher: [Amsterdam] : North-Holland Publishing Company
ISBN:
Category :
Languages : en
Pages : 16
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Book Description
Author: Gerry D. Moan
Publisher: ASTM International
ISBN: 0803128959
Category : Nuclear fuel claddings
Languages : en
Pages : 891
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Book Description
Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.
Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907
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Book Description
Author: Frederic R. Shober
Publisher:
ISBN:
Category : Metals
Languages : en
Pages : 120
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Book Description
The effect of fast-neutron (>1 Mev) irradiation on the mechanical properties of structural metals and alloys was studied. Although the yield strengths and ultimate tensile strengths are increased su stantially for most materials, the ductility suffers severe decreases. This report presents these changes in properties of several structural metals for a number of neutron exposures within the 1.0 x 10 to the 18th power to 5.0 x 10 to the 21st power n/sq cm range. Data summarizing these effects on several classes of materials such as carbon steels, low-alloy steels, stainless steels, Zr-base alloys, ni-base alloys, Al-base alloys, and Ta are given. Additional data which show the influence f irradiation temperatures and of post-irradiation annealing on the radiation-induced property changes are also given and discussed. Increases as great as 175% in yield strength, 100% in ultimate strength, and decreases of 80% in total elongation are reported for fast-neutron exposures as great as 5 10 to the 21st power n/sq cm. (Author).
Author: CE. Coleman
Publisher:
ISBN:
Category : Electron microscopy
Languages : en
Pages : 19
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Book Description
Structures developed in zirconium alloys during irradiation creep have been characterized by transmission electron microscopy (TEM). Alloys investigated were annealed Zircaloy-2, cold-worked Zircaloy-2 and cold-worked Zr-2.5Nb pressure tube material. Thin films were taken from material deformed in the NRU, NRX and Pickering-3 reactors at temperatures of 530 to 600 K under stresses of 117 to 552 MPa giving strains in the range 0.14 to 8.8 percent. Stress-induced orientation of dislocation loops makes a negligible contribution to irradiation creep at all stresses. At the lower stresses (and hence strains), the size and distribution of the damage is unaffected by stress, being the same in the head and gage sections of creep specimens. At higher stresses (strains), there is much clearing of the damage by plastic deformation. The deformation however is very uneven, producing structures in different grains of the same specimen that can show no deformation, swaths cleared of irradiation damage, or dislocation tangles or cell formation. The relevance of these TEM observations to irradiation creep mechanisms is discussed.
Author: VN. Shishov
Publisher:
ISBN:
Category : Damage dose
Languages : en
Pages : 14
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Book Description
The irradiation-induced damage of zirconium alloys subjected to neutron irradiation up to dose levels of ~50 dpa was investigated. Specimens of unalloyed zirconium, Zr-1%Nb, Zr-2.5%Nb and Zr-1%Nb-1.3%Sn-0.4%Fe were irradiated in the BOR-60 reactor over the temperature range 320-420°C. The dose dependence of the irradiation growth strain increased sharply in zirconium and Zr-Nb irradiated at ~350°C at doses above ~10 dpa. In the case of Zr-1%Nb-1.3%Sn-0.4%Fe, it increased at doses of ~37 dpa. Upon increasing the irradiation temperature to 420°C, a sharp accelerated irradiation growth of the Zr-1%Nb alloy began shifting up to about 30 dpa. For the Zr- 1%Nb-1.3%Sn-0.4%Fe, no change of the irradiation growth rate was observed up to a dose of 55 dpa. The onset of increased irradiation growth in alloys correlates with the occurrence of c-component dislocation loops which coincides with a loss of coherence of finely-dispersed precipitates. Post-irradiation annealing experiments demonstrated that a delay in loop formation leads to displacement of the "break-away" beginning in the dose dependence of the irradiation growth in the direction of high doses. The a+c-type dislocation loops were also formed in Zr-1%Nb alloy at high doses, but their influence on the change of macroscopic properties was not observed.
Author: PCK Chow
Publisher:
ISBN:
Category : Corrosion
Languages : en
Pages : 34
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Book Description
X-ray diffraction (XRD) and analytical electron microscopy (AEM) have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components.
Author: J. H. Schemel
Publisher: ASTM International
ISBN: 9780803107571
Category : Science
Languages : en
Pages : 544
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Book Description
Author: X. Iltis
Publisher:
ISBN:
Category : Alloying element redistribution
Languages : en
Pages : 23
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Book Description
To understand the acceleration of the Zircaloy corrosion kinetics in PWR conditions, TEM microstructural characterizations of oxide layers grown in an autoclave or directly in-reactor have been performed. To separate the influence on the oxidation process of the irradiation damage in the alloy from the dynamic effect of neutron flux, oxide layers have also been grown in an autoclave on previously neutron-irradiated cladding. The comparative characterization of these oxide layers leads to the following results: the nucleation and growth process are observed to be similar on oxides formed in-autoclave and significantly different on oxides grown directly in-reactor, indicating that this process is essentially affected by neutron irradiation or, more generally, parameters specific to the reactor environment. Concerning grain growth phenomena, it appears that the high microstructural instability noticed in oxides formed in-reactor is also the consequence of parameters specific to the reactor environment such as neutron irradiation or the lithium concentration gradient. Finally, the iron distribution in the oxide is almost the direct image of the iron distribution in the metal.
