Irradiation data and post-irradiation examination of zirconium alloy clad fuel elements inadvertently operated at high sheath temperatures PDF Download
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Author: E. Proudfoot
Publisher:
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Category :
Languages : en
Pages : 0
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Author: E. Proudfoot
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 850
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Author: R. H. Hu
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Category :
Languages : en
Pages : 0
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Author: Richard E. Schreiber
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Category : Aluminum alloys
Languages : en
Pages : 112
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Author: V. J. Langman
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Category :
Languages : en
Pages : 0
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Author: Gerald E. Lamale
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Category : Hydrides
Languages : en
Pages : 46
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Enriched ZrH/sub 1.65/ -2 wt.% uranium specimens were irradiated under 1 atm. of hydrogen at center-line temperatures between 900 to 1400 deg F to uranium burnups of between 13 and 25 at.%. Specially designed irradiation capsules were used to provide the conditions of temperature and hydrogen atmosphere which were required. Each capsule was instrumented with five thermocouples so that ample temperature data could be obtained during the irradiation. Almost negligible density changes were produced in the material by the irradiation. Changes in length and diameter were of a degree which could fall within experimental error in measurement. Metallographic examination showed no change in microstructure which could be attributed to the effect of irradiation. (auth).
Author: B. Bourdiliau
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Category : Deformation mechanisms
Languages : en
Pages : 25
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Book Description
Zirconium alloys are commonly used in pressurized water reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the spent nuclear fuel assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on a nuclear power plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep, a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1 % Nb and stress-relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400 and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, transmission electron microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the as-irradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The transmission electron microscopy examinations, conducted on recrystallized Zr-1 % Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channeling was observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may also occur during dry transportation or at the beginning of dry storage.
Author: R. H. Hu
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Category :
Languages : en
Pages : 0
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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 76
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Book Description
This report discusses Zircaloy-2 clad uranium and uranium-2 weight percent zirconium fuel tubes which were irradiated to 3200 MWD/T in a high temperature water cooled loop. The outer clad of one tube split due to swelling of the uranium. Postirradiation examination of the fuel cores included metallography, electron microscopy, density determinations, dimensional measurements, and radiochemical burn-up analysis.
Author: Suresh Yagnik
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 48
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Book Description
In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components--especially at high exposures--cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multiyear (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2 x 1026 m-2 (E > 1 MeV)--equivalent to maximum of approximately 37 dpa exposure--followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320 ± 10°C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM).
