Effect of Steady-State Oxidation on Tensile Failure of Zircaloy Cladding

• 발행기관: 한국방사성폐기물학회
• 저자명: Taeho Kim Kyoung Joon Choi Seung Chang Yoo Yunju Lee Ji Hyun Kim
• 간행물 정보: 『Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT)』Vol.20 No.2, 161~170쪽, 전체 10쪽
• 주제분류: 공학 > 공학일반
• 파일형태: PDF
• 발행일자: 2022.06.30

국문 초록

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm−1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm−1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

키워드

Zirconium alloy Oxidation Raman spectroscopy Tensile test Rupture test