Investigation of the Performance of Different Types of Zirconium Microstructures under Extreme Irradiation Conditions E.M. Acosta, O. El-Atwani Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, West Lafayette, IN The safe and continued operation of the US nuclear power plants requires improvement of the radiation resistant properties of materials used in nuclear reactors. Zirconium is a material of particular interest due to its use in fuel cladding. Studies performed on other materials have shown that grain boundaries can play a significant role on the radiation resistant properties of a material. Thus, the focus of our research is to investigate the performance of different zirconium microstructures under irradiation conditions similar to those in commercial nuclear reactors. Analysis of the surface morphology of zirconium both pre- and post-irradiation was conducted with Scanning Electron Microscopy (SEM). Cold-rolled (small-grain microstructure) and annealed (large-grained microstructure) zirconium samples were mechanically polished in order to be irradiated. Room temperature irradiation of zirconium samples was conducted at energies of 100 eV and 1 keV with He + ions at a flux of 1 x10 20 m -2 using a gridded ion source. High temperature (350 ⁰ C and 700 ⁰ C) He + irradiations were performed at an energy of 100 eV using a gridless end- hall ion source at the same flux. Transmission Electron Microscopy (TEM) was conducted to determine the grain size of the zirconium samples. Preliminary results show greater surface damage on the rolled zirconium samples than on the annealed samples for all irradiation cases. The difference in damage was most evident in high temperature irradiations. These findings suggest that large-grained zirconium may be more suitable for fuel cladding applications. Further testing will be performed using higher fluxes, temperatures and energies. ABSTRACT MOTIVATION The safe and continued operation of nuclear reactors requires radiation resistant materials. Zirconium is one of the most important materials due to its use in fuel cladding. Studies performed by El-Atwani et. al. and Murty et. al. on tungsten and carbon steels have shown that grain boundaries can play a significant role on the radiation resistant properties of a material. This research focuses on the performance of different zirconium microstructures under irradiation conditions similar to those in nuclear reactors. METHODS Cold-rolled (small-grained microstructure) and annealed (large-grained microstructure) zirconium samples were tested under extreme irradiation conditions. Room Temperature (RT) irradiations of both types of zirconium samples were conducted at energies of 100 eV and 1 keV using He + ions at a flux of 1x10 20 s -1 m -2 with a gridded ion source. The total fluence was 5 x10 22 m -2 High temperature (350°C and 750°C) He + irradiations were conducted at an energy of 100 eV using a gridless end-hall ion source at a flux of 1x10 20 s -1 m -2 to a fluence of 5 x10 22 m -2 Rolled and annealed zirconium samples Mechanical polishing of zirconium samples Irradiation of zirconium samples Zirconium surface morpholog y analysis performed with SEM. Cross-section imaging performed with Focused Ion Beam (FIB) and SEM. Grain size determined with TEM. Figure 1: Scanning Electron Microscope (Purdue University n.d.) Figure 2: Transmission Electron Microscope (University of Delaware n.d.) Figure 3: Focused Ion Beam (Wikipedia 2014) TEM: GRAIN SIZE TEM micrographs showed that the annealed zirconium has large grains (5-10 µm) with sharp grain boundaries. The rolled samples were found to have ultrafine grains, though the size of the grains is difficult to determine due to non- sharp grain boundaries. Figure 4: TEM micrographs of the rolled and the annealed samples. SEM: UN-IRRADIATED, RT SEM images of the zirconium samples were taken to compare with the irradiated samples. Figure 5: SEM micrographs of the un-irradiated rolled and annealed samples SEM: 100eV, RT Surface damage on both samples is characterized by large holes and small blisters, but the damage is greater on the rolled sample. Defects are formed primarily due to helium bubble formation. However, only small bubbles form at RT. Figure 6: SEM micrographs of both samples irradiated with 100 eV He at RT. SEM: 100eV, 350 ⁰ C Figure 7: SEM micrographs of both samples irradiated at 350 ⁰ C and 100 eV Temperature relevant to Boiling Water Reactors (BWR). The surface of the rolled sample is covered by blisters. Faceted voids can also be seen which were most likely created by helium bubbles bursting on the surface. The annealed sample shows less damage, but several surface voids are still observed. SEM: 100eV, 700 ⁰ C Figure 8: SEM micrographs of both samples irradiated with 100 eV He at 700 ⁰ C. Temperature relevant to Pressurized Water Reactors (PWR). The damage in the rolled sample becomes much more significant at this temperature. The density of blisters on the surface is increased. Micro-cracks are formed due to bubble formation at the grain boundaries along with irradiation enhanced diffusion. The annealed sample shows some blisters, voids and cracks on the surface, but the damage remains less than that in the rolled sample. SEM: 1keV, RT Figure 9: SEM micrographs of both samples irradiated with 1 keV He at RT. At this temperature and energy, the damage on both samples is comparable. FIB/SEM: CROSS-SECTION DAMAGE COMPARISON Figure 10: Cross-sectional SEM images of the samples irradiated with 1 keV He at RT and 100 eV He at 700 ⁰ C At 1 keV and RT, the damaged layer is comparable on both samples. At 100 eV and 700 °C, the thickness of the damaged layer is greater in the rolled sample. CONCLUSIONS These preliminary results suggest that large-grained zirconium is more suitable for nuclear reactor applications. This deviates from the expectation that increasing the grain boundary density will improve radiation resistant properties. ONGOING WORK Both types of zirconium samples will be tested under higher fluxes, temperatures and energies. Thermal Desorption Spectroscopy (TDS) will be conducted on the zirconium sample. TDS functions by heating the sample and releasing helium that is trapped in the zirconium sample. The helium that is released is detected with a Residual Gas Analyzer. The resulting data is a intensity vs. temperature profile that shows where the trapped helium is coming from depending on the peak location. Figure 11: TDS chamber front view Figure 12: TDS chamber side view ACKNOWLEDGEMENTS I would like to thank Purdue University and the Summer Undergraduate Research Fellowship for providing me with the opportunity to conduct research. I would also like to thank Dr. El- Atwani for providing me with the SEM and TEM micrographs of zirconium. Finally, I would like to thank Grant Hosinski for the collaborative effort to assemble the TDS chamber. REFERENCES Schwope, A. D., & Chubb, W. (1952). Zirconium alloys for nuclear reactor applications. Columbus, Ohio: United States Atomic Energy Commission. El-Atwani, O., Hinks, J., Greaves, G., Gonderman, S., Qiu, T., Efe, M., et al. In-situ TEM observation of the response of ultrafine- and nanocrystalline-grained tungsten to extreme irradiation environments. Scientific Reports, 4. Uberuaga, B. P., & Bai, X. The Influence of Grain Boundaries on Radiation-Induced Point Defect Production in Materials: A Review of Atomistic Studies. JOM, 65, Murty, K., Alsabbagh, A., & Valiev, R. Z. Influence of grain size on radiation effects in a low carbon steel. Journal of Nuclear Materials, 443, Scanning Electron Microscope. (n.d.). Purdue University -. Retrieved July 24, 2014, from Transmission Electron Microscope. (n.d.). University of Delaware. Retrieved July 24, 2014, from Focused ion beam. (2014, May 21). Wikipedia. Retrieved July 24, 2014, from