Effects of an alloying element on a c-component loop formation and precipitate resolution in Zr alloys during ion irradiation. Issue 10 (3rd October 2018)

• Record Type: Journal Article
• Authors: Watanabe, Hideo; Takahashi, Katsuhito; Yasunaga, Kazufumi; Wang, Yun; Aono, Yasuhisa; Maruno, Yusaku; Hashizume, Kenichi
• Is Part Of: Journal of nuclear science and technology. Volume 55:Issue 10(2018)
• Page Start: 1212
• Page End: 1224
• Publication Date: 2018-10-03
• Subjects: Hydrides -- zirconium alloys -- light water reactor -- cladding -- scanning transmission electron microscopy -- X-ray energy-dispersive spectroscopy; Nuclear engineering -- Periodicals; Nuclear physics -- Periodicals; Nuclear energy -- Periodicals; 621.4805
• DOI: 10.1080/00223131.2018.1486244
• Languages: English
• ISSNs: 0022-3131

Abstract

ABSTRACT: It is important to clarify the mechanisms of the dislocation loop formation, dissolution of precipitates to understand the degradation behavior of the fuel cladding tubes in light water reactors (LWR) under neutron irradiation. In this study, 3.2 MeV Ni ion irradiation was carried out at 400°C on Zircaloy-2 and two types of model alloys with and without Fe (Zr-1.5Sn-0.3Fe and Zr-1.5Sn). To understand the effects of hydrogen, 60 and 300 ppm pre-injected Zircaloy-2 samples were also irradiated. The microstructure was observed with a conventional transmission electron microscopy. Additionally, the dissolution of precipitates and the enrichment of the alloying element due to irradiation were analyzed using a spherical aberration (Cs)-corrected high-resolution analytical electron microscope. After ion irradiation at 400°C, the dissolution of Fe-enriched precipitates and the c-component dislocation loops were observed in the region of peak ion damage. Observations by STEM-EDS showed that Fe atoms were enriched in the c-component dislocation loops. On the contrary, the c-component dislocation loops were detected in Fe-containing alloys (Zircaloy-2 and Zr-1.5Sn-0.3Fe alloy) but were not in the Zr-1.5Sn alloy. These results indicate that the dissolution of Fe-enriched precipitates and the enhanced formation of c-component dislocation loops are essential for the degradation of LWR fuel cladding under irradiation.