Author: Atomic Energy of Canada LimitedPublisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Proposal to Study the Stability of Zircaloy-clad U02 Fuel Elements During Long-term Storage
Author: Atomic Energy of Canada LimitedPublisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
A Proposal to Study the Stability of Zircaloy-clad UO↓2 Fuel Elements During Long-term Storage
Author: A. S. BainPublisher: Chalk River, Ont. : Fuel Engineering Branch, Chalk River Nuclear Laboratories
ISBN:
Category :
Languages : en
Pages : 13
Book Description
A proposal to study the effect of power ramp conditions on the dimensional stability of zircaloy-clad uo2 fuel elements
Author: P. J. FehrenbachPublisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
A proposal for the irradiation of zircaloy-4 clad, uo2 fuel elements at post-dryout sheath temperatures in the 625-675 degrees c range in the x-4 loop
Author: V. J. LangmanPublisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Interdiffusion in Zircaloy-2 Clad U-2 W/o Zr Fuel Materials and Its Effect Upon Corrosion Behavior
Author: A. L. GearyPublisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 28
Book Description
A proposal to irradiate zircaloy clad tho2-uo2 fuel elements in the x-3 loop at average linear heat outputs of 70 kw/m and to burnups of 500 mwh/kg u
Author: R. D. MacdonaldPublisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
An Assessment of the Long-term Storage of Zircaloy Fuel Rods in Water
Author: Zuhair A. MunirPublisher:
ISBN:
Category : Nuclear fuel claddings
Languages : en
Pages : 36
Book Description
Assembly and testing of the fuel elements for the prolonged irradiation of zircaloy-2 clad uo2
Author: M. B. WatsonPublisher:
ISBN:
Category :
Languages : en
Pages : 0
Book Description
Energy Research Abstracts
Author: Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 816
Book Description
CORROSION BEHAVIOR OF DEFECTED FUEL ELEMENTS WITH U-2 W/o Zr CORE CLAD WITH ZIRCALOY-2
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
The aqueous corrosion behavior of defected fuel elements having a U-2 wt. % Zr core with Zircaloy-2 cladding has been studied. A standard diffusion heat treatment for 7 hours at 880 C, followed by moderately rapid cooling, was established to overcome the effect of subtle defects in 15-mil Zircaloy cladding. The development of this heat treatment was carried out primarily with small rod specimens, which were also used to obtain data on aurxiliary effects of thw diffusion heat treatment. The heat treatment has been applied to full diameter tubes. One of these tubes was tested with subtle defects and various sections have been observed following the insertion of gross defects. These observations have shown a difference in the rate of hydrogen evolution as a result of heat treatment. In addition, the heat treatment changes the nature of failure in both full diameter tube sections and small rods. The presence of 2 wt. % Zr in the core reduces the corrosion rate considerably. Comparative quantitative data are presented for uranium alloys with various zirconium contents up to 15 wt. %. (auth).
