High-fluence irradiation growth of zirconium alloys at 644 to 725 k PDF Download
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Author: R. P. Tucker
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author: R. P. Tucker
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author: RB. Adamson
Publisher:
ISBN:
Category : Alloys
Languages : en
Pages : 23
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Book Description
Irradiation growth behavior of zirconium, Zircaloy-2 and Zircaloy-4,Zr-2.5Nb, and Zr-3.5Sn-0.8Mo-0.8Nb (EXCEL) was studied on specimens irradiated in the Experimental Breeder Reactor II (EBR-II) to fluences of 1.2 to 16.9 x 1025 neutrons (n).m-2 (E > 1 MeV) in the temperature range 644 to 725 K. In Zircaloy, growth and growth rate were observed to increase continuously with fluence up to 16.9 x 1025 n.m-2 with no indication of saturation in either recrystallized or cold-worked materials. Positive growth strains of 1.5% and negative strains of approximately 2% to 2.5% were observed in both recrystallized and cold-worked Zircaloy. The formation of both a-type loops and c component dislocations is recrystallized Zircaloy under irradiation appears to be the basis in this material for growth strains similar in magnitude to those in cold-worked Zircaloy. Alloy additions to zirconium can increase growth by as much as an order of magnitude for a given texture at the higher irradiation temperatures and fluences. A sharp change to increasing growth rate with temperature occurs in Zircaloy at ~670 K, with a similar trend indicated for the other alloys. Although growth in all these alloys is a strong function of crystallographic texture, an exact (1-3f) type of dependence is not always apparent. In Zr-2.5Nb the dependence of growth on texture appears to be masked by the precipitation of betaniobium, with a transition to a well-defined texture dependence being a function of fluence and temperature. Significant differences in growth behavior were observed in nominally similar Zircaloys, apparently due to minor microstructural or chemical differences.
Author:
Publisher: ASTM International
ISBN:
Category :
Languages : en
Pages : 849
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Author: D. G. Franklin
Publisher: ASTM International
ISBN: 9780803102705
Category : Science
Languages : en
Pages : 866
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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: 0803124996
Category : Microstructure
Languages : en
Pages : 953
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Author:
Publisher:
ISBN:
Category : Bioengineering
Languages : en
Pages : 872
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Author: N. H. Packan
Publisher: ASTM International
ISBN: 0803112661
Category : Materials
Languages : en
Pages : 679
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Annotation Effects of Radiation on Materials: Fourteenth International Symposium was presented at Andover, MA, June 1988. The symposium was sponsored by ASTM Committee E-10 on Nuclear Technology and Applications. The papers from the first three days of the symposium appear in the two volumes of this publication. Volume I encompasses radiation damage- induced microstructures; point defect, solute, and gas atom effects; atomic-level measurement techniques; and applications of theory. Volume II includes mechanical behavior, all papers dealing with pressure-vessel steels, breeder reactor components, dosimetry, and nuclear fuels. The fourth day of the symposium was devoted to the single topic of reduced-activation materials (see TK9204). The two volumes are separately sold at $127 and $128 respectively; each is independently indexed. Annotation copyrighted by Book News, Inc., Portland, OR.
Author: RB. Adamson
Publisher:
ISBN:
Category : Annealed
Languages : en
Pages : 27
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Irradiation growth behavior of Zircaloy-2 and -4 was studied on specimens irradiated in the Experimental Breeder Reactor II to fluences of 1.4 to 6.3 x 1025 neutrons (n).m-2 (E > 1 MeV) in the temperature range 644 to 723 K. Measurements in the three principal directions on annealed and cold-worked/stress-relieved Zircaloy-2 slab materials provided evidence that growth is a constant-volume process up to about 680 K. The growth strains were shown to be determined by the crystallographic texture, that is, proportional to (1-3(1-3fdc)), where), where fdc is the fraction of basal poles, is the fraction of basal poles, fc, in the direction d. The growth strains for annealed and cold-worked Zircaloy were large relative to previously reported data, were similar in magnitude, were strongly dependent on irradiation temperature, and varied linearly with fluence over the range investigated. Transmission electron microscopy on annealed Zircaloy-4 specimens revealed a few small voids and larger cavities, a grain boundary second phase, and dislocation loops, tangles, and arrays. The high growth strains in annealed Zircaloy appear to be governed by dislocation arrays formed during irradiation. This implies a change in growth mechanism from that pertaining at lower temperatures in annealed material. The data suggest a transition from saturating steady-state growth at lower temperatures to increasing and eventually high steady-state rates under the conditions of these tests.
Author: M. Griffiths
Publisher:
ISBN:
Category : Anisotropic diffusion
Languages : en
Pages : 20
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Book Description
Sponge zirconium and Zr-2.5 wt% Nb, Zircaloy, or Excel alloys all exhibit accelerated irradiation growth compared with high-purity crystal-bar zirconium for irradiation temperatures between 550 to 710 K and fluences between 0.1 to 10 x 1025 n • m-2 (E > 1 MeV). There is generally an incubation period or fluence before the onset of accelerated or "breakaway" growth, which is dependent on the particular material being irradiated, its metallurgical condition before irradiation, and the irradiation temperature.
