Cr涂层锆合金包壳模拟LOCA试验研究

2011年日本福岛核事故暴露传统锆合金燃料包壳在失水事故(LOCA)工况下的安全性问题。为了探究新型Cr涂层锆合金包壳在LOCA工况下的性能，本研究针对物理气相沉积(PVD)工艺涂覆的12～15μm厚度Cr涂层Zr-1Nb合金包壳管，开展模拟LOCA工况下的高温蒸汽氧化-淬火试验，氧化温度为1200℃和1300℃，单面氧化时间为10 min和20 min，淬火温度约800℃，之后对淬火后试样进行环压测试。结果发现，在研究条件下，Cr涂层未出现剥落，涂层完整；Cr涂层锆合金包壳外表面形成较为致密Cr₂O₃层，抑制O原子扩散至锆合金基体，阻止锆合金基体被氧化为ZrO₂层和α-Zr(O)层，环压测试发现淬火后包壳保持良好塑性。研究表明，在本测试工况下Cr涂层锆合金包壳相比传统锆合金包壳具有更强的抗LOCA事故能力。

Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions

The Fukushima nuclear accident in Japan in 2011 exposed the inherent safety problems of traditional zirconium alloy fuel cladding under LOCA conditions. To investigate the performance of a new Cr-coated zirconium alloy cladding under LOCA conditions, high temperature steam oxidation and quenching experiments under simulated LOCA conditions are carried out for 12~15 μm thick Cr-coated Zr-1Nb alloy cladding tube coated by physical vapor deposition (PVD) process, the oxidation temperature and oxidation time were 1200℃, 1300℃ and 10 min, 20 min, respectively, the quenching was performed around 800℃, then ring compression test was performed for the quenched tube. The results indicated that no spalling was found for Cr coatings under experiment conditions, intense Cr₂O₃ layer which formed on the outer surface of Cr-coated tube retarded the diffusion of O into zirconium substrate, protecting the zirconium alloy from oxidized into ZrO₂ and α-Zr(O) layers, Cr-coated zirconium-alloy cladding remained ductile after quenching. It can be concluded that Cr-coated Zirconium alloy behaves better than traditional Zirconium alloy under the experimental conditions.