Cr涂层锆合金包壳模拟LOCA试验研究

王占伟; 严俊; 彭振驯; 任啟森; 廖业宏; 李思功; 赵亚欢

doi:10.13832/j.jnpe.2023.02.0122

Cr涂层锆合金包壳模拟LOCA试验研究

doi: 10.13832/j.jnpe.2023.02.0122

中广核研究院有限公司核燃料与材料研究所，广东深圳，518026

详细信息

作者简介:
王占伟（1987—），男，硕士研究生，现主要从事核燃料性能分析及燃料材料试验研究，E-mail: zhanwei_wang@163.com

中图分类号: TL334
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出版历程
- 收稿日期: 2022-05-17
- 修回日期: 2022-12-04
- 刊出日期: 2023-04-15

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

Department of Nuclear Fuel & Material, China Nuclear Power Technology Research Institute Co., Ltd., Shenzhen, Guangdong, 518026, China

摘要

摘要: 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12~15 μm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。
- Cr涂层锆合金包壳 /
- 失水事故（LOCA） /
- 高温蒸汽氧化 /
- 淬火 /
- 塑-脆性转变
Abstract: 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.
- Cr-coated zirconium alloy cladding /
- Loss of coolant accident (LOCA) /
- High temperature steam oxidation /
- Quenching /
- Ductile-to-brittle transition

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图 1 试验样品

Figure 1. Test Sample

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图 2 不同工况下高温蒸汽氧化-淬火后宏观试样

Figure 2. Macrograph of Test Sample after High Temperature Steam Oxidation and Quenching under Different Conditions

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图 3 未氧化和氧化后Cr涂层锆合金包壳表面XRD图谱

Figure 3. XRD Spectrum of Unoxidized and Oxidized Cr-coated Zirconium-alloy Cladding Surfaces

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图 4 氧化样品表面形貌

Figure 4. Surface Morphology of Oxidized Sample

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图 5 氧化样品横截面形貌

Figure 5. Cross-section Morphology of Oxidized Sample

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图 6 氧化样品Cr元素EDS面分析

Figure 6. EDS Surface Analysis of Cr Element of Oxidized Sample

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图 7 氧化样品横截面EDS线分析

Figure 7. EDS Line Analysis of Oxidized Sample Cross-section

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图 8 淬火后样品载荷-位移曲线

Figure 8. Load-displacement Curves of Post-quench Sample

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图 9 环向压缩样品断口形貌

Figure 9. Fracture Morphology of Ring Compression Sample

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表 1 氧化样品表面EDS点分析

Table 1. EDS Point Analysis of Oxidized Sample Surface

取样位置	元素含量^①/%
取样位置	Cr	O
P1	38.05	61.95
P2	38.99	61.01
P3	39.11	60.89
注：①原子百分比

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表 2 氧化样品横截面EDS点分析

Table 2. EDS Point Analysis of Oxidized Sample Cross-section　　　　　　

取样位置	元素含量^①/%
取样位置	Cr	O	Zr	Fe	Nb
1	44.41	55.59
2	97.04	1.53	1.42
3	69.37		28.75	0.94	0.94
4	1.67	7.88	88.22	0.19	2.04
5	10.88	7.96	78.66	0.53	1.98
6	1.13	9.36	87.51	0.13	1.87
7	7.59	8.38	82.13	0.40	1.50
8	2.40	7.94	87.76	0.12	1.78
9	6.07	8.24	83.34	0.43	1.91
注：①原子百分比

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