Thermal creep of irradiated zircaloy cladding.
As part of an effort to investigate spent-fuel behavior during dry-cask storage, thermal creep tests are being performed with defueled Zircaloy-4 cladding segments from two pressurized water reactors - Surry at {approx} 36 GWd/MTU burnup and H. B. Robinson at {approx} 67 GWd/MTU burnup, with corresponding fast (E > 1 MeV) fluence levels of 7 x 10{sup 25} and 14 x 10{sup 25} n/m{sup 2}. The Surry rods are particularly relevant because they were stored in an inert-atmosphere (He) cask for 15 years. The Robinson rods were received after reactor discharge and pool storage. Commensurate with their high burnup, the Robinson cladding has significant waterside corrosion and hydrogen uptake. Test results to-date indicate good creep ductility for both claddings in the 360 400 C and 160-250 MPa (hoop-stress) regime. Partial recovery of radiation hardening may have occurred during the long tests at 400 C, which led to improved creep ductility. Creep-rate sensitivity is significant for stress and even more so for temperature. The higher hydrogen content in the Robinson material appears to have no detrimental effect on creep behavior at the test temperature. One Robinson sample, which ruptured in the weld region at 205 C during cooling from 400 C under stress (190 MPa), precipitated all visible hydrides in the radial direction.
- Research Organization:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USNRC; RW; EPRI
- DOE Contract Number:
- DE-AC02-06CH11357
- OSTI ID:
- 956906
- Report Number(s):
- ANL/ET/CP-110704; TRN: US1002141
- Journal Information:
- ASTM STP, Journal Issue: 2005; Conference: 14th International Symposium on Zirconium in the Nuclear Industry; Jun. 13, 2004 - Jun. 17, 2004; Stockholm, Sweden
- Country of Publication:
- United States
- Language:
- ENGLISH
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