Zirconium Moderator Cladding Irradiated in the Sodium Reactor Experiment PDF Download
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Author: J. J. Gill
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 42
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Author: J. J. Gill
Publisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 42
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Book Description
Author: J. H. Schemel
Publisher: ASTM International
ISBN: 9780803106017
Category : Business & Economics
Languages : en
Pages : 656
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Author: ASTM Committee B-10 on Reactive and Refractory Metals and Alloys
Publisher: ASTM International
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 694
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Author: J. A. Leppard
Publisher:
ISBN:
Category : Reactor moderators
Languages : en
Pages : 40
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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 1
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Author:
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 700
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Author: Holger Wiese
Publisher:
ISBN:
Category : High burnup
Languages : en
Pages : 34
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Book Description
The objective of this paper is to summarize the results of the latest observations performed at Paul Scherrer Institut on irradiated fuel claddings, to characterize their corrosion and hydrogen-uptake behavior. Two categories of studies have been performed. (1) A series of destructive tests were achieved on the fuel rods irradiated in a boiling-water reactor (BWR), including hydrogen concentration by hot-gas extraction. These results provided the hydrogen content of the cladding at different stages of irradiation, at different elevations along the rod. (2) Another series of examinations using a correlative microscopy method, i.e., using different techniques, including transmission electron microscopy (TEM), electron probe microanalysis (EPMA), and secondary ion mass spectrometry (SIMS), on the same material and in the same region of the metal-oxide interface have provided useful data regarding the oxide layer combining the signals from oxides and from hydrides. Furthermore, the effect of the type of alloying element has been examined for in-reactor oxidation. These studies are subsequently combined with the findings from out-of-pile studies, using techniques, such as neutron radiography, to confirm the in-reactor observations. Results have shown that: (i) the hydrogen pickup fraction varies at different conditions and could even decrease as the oxide thickness increases; (ii) the distribution of hydrogen in the cladding is usually inhomogeneous; (iii) the most determining parameter for hydrogen uptake seems to be the microstructure of the oxide, and the nature of the alloying element will influence to a certain extent this parameter; (iv) furthermore, the stress in the oxide layer can modify the crack distribution in the latter, cracks will in turn shorten the route for the hydrogen to access the metal. These results will be discussed as a contribution to the available knowledge about hydrogen uptake and will provide a global support for the models of the uptake phenomenon.
Author: S. Venugopal
Publisher:
ISBN:
Category : Accelerated irradiation test
Languages : en
Pages : 16
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Book Description
Fast Breeder Test Reactor (FBTR) at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, India is a sodium cooled fast reactor with a maximum neutron flux of about 3 x 1015 n/cm2/s. The neutron spectrum in the FBTR is hard and the damage rate attained in structural specimens is high. Compact pressurized capsules of zirconium alloys have been developed and subjected to irradiation in the FBTR to a fluence level of 1.1 x 1021 n/cm2 (E >1 MeV) at temperatures of 306 to 319°C to determine the irradiation creep rate of indigenously developed zirconium alloys (Zircaloy-2 and Zr-2.5%Nb alloy). To assess the changes in mechanical properties of FBTR grid plate material (modified type 316 stainless steel) due to prolonged low dose exposure, an accelerated irradiation test with dose levels up to 2.6 dpa (at 350°C) has also been carried out using miniature tensile test and disk specimens. Postirradiation examination (PIE) measurements carried out in the hot cells determined the creep rate of zirconium alloys, and indicated that the grid plate material has hardened but has still enough residual ductility. This paper presents salient features of the design and implementation of these irradiation experiments in FBTR and the results obtained during PIE.
Author:
Publisher:
ISBN:
Category : Nuclear engineering
Languages : en
Pages : 1244
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Author:
Publisher:
ISBN:
Category : Science
Languages : en
Pages : 1116
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