Stress corrosion cracking of Zircaloys. Final report

Technical Report · Sat Mar 01 04:00:00 EST 1980

DOI:https://doi.org/10.2172/5523320· OSTI ID:5523320

Cubicciotti, D.; Jones, R. L.; Syrett, B. C.

The overall aim has been to develop an improved understanding of the stress corrosion cracking (SCC) mechanism considered to be responsible for pellet-cladding interaction (PCI) failures of nuclear fuel rods. The objective of the present phase of the project was to investigate the potential for improving the resistance of Zircaloy to iodine-induced SCC by modifying the manufacturing techniques used in the commercial production of fuel cladding. Several aspects of iodine SCC behavior of potential relevance to cladding performance were experimentally investigated. It was found that the SCC susceptibility of Zircaloy tubing is sensitive to crystallographic texture, surface condition, and residual stress distribution and that current specifications for Zircaloy tubing provide no assurance of an optimum resistance to SCC. Additional evidence was found that iodine-induced cracks initiate at local chemical inhomogeneities in the Zircaloy surface, but laser melting to produce a homogenized surface layer did not improve the SCC resistance. Several results were obtained that should be considered in models of PCI failure. The ratio of axial to hoop stress and the temperature were both shown to affect the SCC resistance whereas the difference in composition between Zircaloy-2 and Zircaloy-4 had no detectable effect. Damage accumulation during iodine SCC was found to be nonlinear: generally, a given life fraction at low stress was more damaging than the same life fraction at higher stress. Studies of the thermochemistry of the zirconium-iodine system (performed under US Department of Energy sponsorship) revealed many errors in the literature and provided important new insights into the mechanism of iodine SCC of Zircaloys.

• Research Organization: SRI International, Menlo Park, CA (USA)
• OSTI ID: 5523320
• Report Number(s): EPRI-NP-1329
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220300 -- Nuclear Reactor Technology-- Fuel Elements
36 MATERIALS SCIENCE
360105* -- Metals & Alloys-- Corrosion & Erosion
ALLOYS
CHEMICAL REACTIONS
CHROMIUM ADDITIONS
CHROMIUM ALLOYS
CLADDING
COMPARATIVE EVALUATIONS
CORROSION
CORROSIVE EFFECTS
CRACKS
DEPOSITION
ELEMENTS
FAILURES
FUEL ELEMENTS
FUEL PELLETS
FUEL RODS
HALOGENS
IODINE
IRON ADDITIONS
IRON ALLOYS
NONMETALS
PELLETS
REACTOR COMPONENTS
STRESS CORROSION
STRESSES
SURFACE COATING
TIN ALLOYS
ZIRCALOY
ZIRCALOY 2
ZIRCALOY 4
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS