The recovery of irradiation damage for Zircaloy-2 and Zircaloy-4 following low dose neutron irradiation at nominally 358 °C
- Bechtel Marine Propulsion Corporation, West Mifflin, PA (United States). Bettis Lab.
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
The recovery of irradiation damage in wrought Zircaloy-2 and Zircaloy-4 was determined in this paper following a series of post-irradiation anneals at temperatures ranging from 343 °C to 510 °C and for time periods ranging from 1-h to 500 h. The materials had been irradiated at nominally 358 °C in the High Flux Isotope Reactor (HFIR) at neutron fluences of nominally 3 × 1025 n/m2 (E > 1 MeV). Irradiation at nominally 358 °C resulted in a coarser distribution of < a > loops that result in a 25–45% lower irradiation hardening than reported in the literature for irradiations at 260–326 °C. The irradiation hardening and recovery were determined using tensile testing at room-temperature. Post-irradiation annealing at 343–427 °C was shown to result in an increase in irradiation hardening to values even higher than for the as-irradiated material in the first 1–10 h of annealing. This Radiation Anneal Hardening (RAH) was followed by a relatively slow recovery of the irradiation damage. Much faster recovery with no RAH was observed for post-irradiation annealing at temperatures of 454–510 °C. Irradiation at 358 °C was shown to result in different recovery kinetics than observed in the literature for irradiation at 260–326 °C. While the general trend described above is true for the four materials tested (alpha-annealed and beta-treated Zircaloy-2 and Zircaloy-4), notable and yet unexplained differences in RAH and in recovery are observed between the materials that might be a result of differing solute effects. Examinations of microstructure using Transmission Electron Microscopy were used to investigate the RAH and recovery mechanisms. Finally, agreement between the measured and calculated irradiation hardening using a generalized Orowan hardening model to account for the observed loop structure was not as close for the post irradiation annealed condition as for the as-irradiated condition, which can likely be attributed to unaccounted for changes in the configuration of the < a > loops to dislocation lines, segregation of solutes to dislocation loops, and the potential for the formation of fine clusters of point defects or solutes during annealing.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Organization:
- USDOE Office of Science (SC)
- Contributing Organization:
- Bechtel Marine Propulsion Corporation, West Mifflin, PA (United States)
- DOE Contract Number:
- AC05-00OR22725
- OSTI ID:
- 1360084
- Journal Information:
- Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Vol. 461; ISSN 0022-3115
- Publisher:
- Elsevier
- Country of Publication:
- United States
- Language:
- English
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