Actions for Mechanistic Modeling of Zircaloy Deformation and Fracture in Fuel Element Analysis
Mechanistic Modeling of Zircaloy Deformation and Fracture in Fuel Element Analysis / FA. Nichols
- Conference Author
- Zirconium in the Nuclear Industry (1985 : Strasbourg, France)
- Physical Description
- 1 online resource (18 pages) : illustrations, figures, tables
- Additional Creators
- Nichols, FA., American Society for Testing and Materials, and ASTM International
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- Restrictions on Access
- Subscription required for access to full text.
License restrictions may limit access. - Summary
- A review is given of the comprehensive model developed in the 1960s at the Bettis Atomic Power Laboratory to explain the creep of Zircaloy during neutron irradiation and applied to fuel element analysis and design. The in-pile softening observed at low stresses was hypothesized to be due to a combination of the growth-directed Roberts-Cottrell yielding creep originally proposed for ?-uranium and the formation of point defect loops preferentially on certain planes in response to the applied stress, with the second process being of relatively greater importance. The in-pile hardening observed at high stresses (or strain-rates) was proposed to be due to the cutting by dislocations of radiation-produced obstacles. In this stress (strain-rate) region, in-pile behavior was proposed to be identical to post-irradiation behavior. At intermediate stresses (strain-rates) a mechanism of radiation-enhanced climb around obstacles was suggested as being rate-controlling. As the stress is decreased, the climb process becomes easier, and the rate was then predicted to be controlled by glide at a flow-stress characteristic of unirradiated, annealed material, where radiation-enhanced diffusion enabled climbing around the normal strain-hardening obstacles. At still lower stresses, this glide process became negligibly slow compared with the growth-connected creep mechanism that was presumed to operate independently. The overall scheme was shown to be in good agreement with all the in-pile data then available and implemented into the computer analysis of fuel element behavior.
- Dates of Publication and/or Sequential Designation
- Volume 1987, Issue 939 (January 1987)
- Subject(s)
- Zircaloy-2.
- Texture
- Fracture
- Damage rate
- Strain rate
- Temperature
- Modeling
- Nuclear fuel cladding
- Irradiation growth
- Zirconium
- Strain
- Deformation
- Neutron irradiation
- Irradiation creep
- Pressurized water reactor
- Stress
- Mechanical properties
- Zirconium alloys—Congresses
- Zircaloy-2—Congresses
- Nuclear fuel claddings—Congresses
- Other Subject(s)
- ISBN
- 9780803150041 (e-ISBN)
9780803109353
0803109350 - Digital File Characteristics
- text file PDF
- Bibliography Note
- Includes bibliographical references 84.
- Other Forms
- Also available online via the World Wide Web. Tables of contents and abstracts freely available; full-text articles available by subscription.
Full text article also available for purchase.
Also available in PDF edition. - Reproduction Note
- Electronic reproduction. W. Conshohocken, Pa. : ASTM International, 1987. Mode of access: World Wide Web. System requirements: Web browser. Access may be restricted to users at subscribing institutions.
- Technical Details
- Mode of access: World Wide Web.
- Source of Acquisition
- ASTM International PDF Purchase price USD25.
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