Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water
Abstract In pressurized water reactor (PWR), fretting wear is one of the main causes of fuel assembly failure. Moreover, the operation condition of cladding is complex and harsh. A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cla...
| Main Authors: | , , , , , , |
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| Format: | Article |
| Language: | English |
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SpringerOpen
2023-09-01
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| Series: | Chinese Journal of Mechanical Engineering |
| Subjects: | |
| Online Access: | https://doi.org/10.1186/s10033-023-00931-4 |
| _version_ | 1797453653673508864 |
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| author | Jun Wang Haojie Li Zhengyang Li Yujie Lei Quanyao Ren Yongjun Jiao Zhenbing Cai |
| author_facet | Jun Wang Haojie Li Zhengyang Li Yujie Lei Quanyao Ren Yongjun Jiao Zhenbing Cai |
| author_sort | Jun Wang |
| collection | DOAJ |
| description | Abstract In pressurized water reactor (PWR), fretting wear is one of the main causes of fuel assembly failure. Moreover, the operation condition of cladding is complex and harsh. A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment. It can simulate the fretting wear experiment of PWR under different temperatures (maximum temperature is 350 ℃), displacement amplitude, vibration frequency, and normal force. The fretting wear behavior of Zr-4 alloy under different temperature environments was tested. In addition, the evolution of wear scar morphology, profile, and wear volume was studied using an optical microscope (OM), scanning electron microscopy (SEM), and a 3D white light interferometer. Results show that higher water temperature evidently decreased the cladding wear volume, the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth. |
| first_indexed | 2024-03-09T15:25:00Z |
| format | Article |
| id | doaj.art-0f1fe650dc2b475992e1cb456af3ad6a |
| institution | Directory Open Access Journal |
| issn | 2192-8258 |
| language | English |
| last_indexed | 2024-03-09T15:25:00Z |
| publishDate | 2023-09-01 |
| publisher | SpringerOpen |
| record_format | Article |
| series | Chinese Journal of Mechanical Engineering |
| spelling | doaj.art-0f1fe650dc2b475992e1cb456af3ad6a2023-11-26T12:32:42ZengSpringerOpenChinese Journal of Mechanical Engineering2192-82582023-09-0136111310.1186/s10033-023-00931-4Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized WaterJun Wang0Haojie Li1Zhengyang Li2Yujie Lei3Quanyao Ren4Yongjun Jiao5Zhenbing Cai6Key Lab of Advanced Technologies of Materials, Tribology Research Institute, Southwest Jiaotong UniversityKey Lab of Advanced Technologies of Materials, Tribology Research Institute, Southwest Jiaotong UniversityScience and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of ChinaKey Lab of Advanced Technologies of Materials, Tribology Research Institute, Southwest Jiaotong UniversityScience and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of ChinaScience and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of ChinaKey Lab of Advanced Technologies of Materials, Tribology Research Institute, Southwest Jiaotong UniversityAbstract In pressurized water reactor (PWR), fretting wear is one of the main causes of fuel assembly failure. Moreover, the operation condition of cladding is complex and harsh. A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment. It can simulate the fretting wear experiment of PWR under different temperatures (maximum temperature is 350 ℃), displacement amplitude, vibration frequency, and normal force. The fretting wear behavior of Zr-4 alloy under different temperature environments was tested. In addition, the evolution of wear scar morphology, profile, and wear volume was studied using an optical microscope (OM), scanning electron microscopy (SEM), and a 3D white light interferometer. Results show that higher water temperature evidently decreased the cladding wear volume, the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.https://doi.org/10.1186/s10033-023-00931-4Fretting wearCladdingHigh temperature and high pressureZirconium alloy |
| spellingShingle | Jun Wang Haojie Li Zhengyang Li Yujie Lei Quanyao Ren Yongjun Jiao Zhenbing Cai Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water Chinese Journal of Mechanical Engineering Fretting wear Cladding High temperature and high pressure Zirconium alloy |
| title | Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water |
| title_full | Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water |
| title_fullStr | Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water |
| title_full_unstemmed | Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water |
| title_short | Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water |
| title_sort | fretting wear characteristics of nuclear fuel cladding in high temperature pressurized water |
| topic | Fretting wear Cladding High temperature and high pressure Zirconium alloy |
| url | https://doi.org/10.1186/s10033-023-00931-4 |
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