Mechanistic modeling of Zircaloy deformation and fracture in fuel element analysis

Conference · Sat Jun 01 04:00:00 EDT 1985

OSTI ID:6490305

Nichols, F A

A review is given of the comprehensive model developed in the 1960's 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 ..cap alpha..-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 which was presumed to operate independently. The overall scheme was shown to be good agreement with all the in-pile data then available and implemented into the computer analysis of fuel element behavior. 48 refs., 1 fig.

• Research Organization: Argonne National Lab., IL (USA)
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 6490305
• Report Number(s): CONF-850601-3; ON: DE85018415
• Country of Publication: United States
• Language: English

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36 MATERIALS SCIENCE
360103* -- Metals & Alloys-- Mechanical Properties
ALLOYS
BWR TYPE REACTORS
CREEP
DEFORMATION
FAILURES
FRACTURES
FUEL ELEMENTS
HARDNESS
MATERIALS
MATHEMATICAL MODELS
MECHANICAL PROPERTIES
PHYSICAL RADIATION EFFECTS
PWR TYPE REACTORS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTOR MATERIALS
REACTORS
TENSILE PROPERTIES
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCALOY
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS