{"title":"Zircaloy-4 燃料销在模拟失壳事故条件下失效:氧脆性","authors":"","doi":"10.1016/j.pnucene.2024.105485","DOIUrl":null,"url":null,"abstract":"<div><div>An extensive experimental investigation was performed to study the oxygen embrittlement of the Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under simulated Loss-of-Coolant Accident (LOCA) conditions. Zircaloy fuel cladding experiences creep and corrosion simultaneously during service and LOCA conditions. Zircaloy-4 fuel pins were pre-oxidized to attain different oxide layer thicknesses, achieving in-service conditions. These pre-oxidized tubes were then subjected to burst tests in the steam environment to mimic the LOCA scenario. The present study aims to improve the understanding of the effect of oxidation on the cladding microstructure and the mechanical response of the fuel pin in a LOCA scenario by accounting for the cross-influence, during transient heating, of oxidation and deformation on the behavior of the clad in the LOCA domain. The oxide layer morphology in pre- and post-burst samples was studied using FESEM, XRD, and Raman spectroscopy. In some cases, the inner oxide layer grew faster than the outer oxide layer when the fuel pin was heated in steam during the burst test. The evolution during transient heating of radial and circumferential crack growth in the oxide layer and the occurrence of delamination facilitated faster oxygen and hydrogen uptake. The hydrogen uptake in pre and post-burst samples was related to the oxygen uptake. The hydrogen concentration increases with the oxygen concentration in the pre-oxidized samples. Small oxygen and hydrogen concentrations were found in the post-burst as-received samples due to the formation of a protective oxide layer. The high-temperature oxide layer was formed at extremely high heating rates.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":3.3000,"publicationDate":"2024-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"Zircaloy-4 fuel pin failure under simulated loss-of-coolant-accident conditions: Oxygen embrittlement\",\"authors\":\"\",\"doi\":\"10.1016/j.pnucene.2024.105485\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><div>An extensive experimental investigation was performed to study the oxygen embrittlement of the Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under simulated Loss-of-Coolant Accident (LOCA) conditions. Zircaloy fuel cladding experiences creep and corrosion simultaneously during service and LOCA conditions. Zircaloy-4 fuel pins were pre-oxidized to attain different oxide layer thicknesses, achieving in-service conditions. These pre-oxidized tubes were then subjected to burst tests in the steam environment to mimic the LOCA scenario. The present study aims to improve the understanding of the effect of oxidation on the cladding microstructure and the mechanical response of the fuel pin in a LOCA scenario by accounting for the cross-influence, during transient heating, of oxidation and deformation on the behavior of the clad in the LOCA domain. The oxide layer morphology in pre- and post-burst samples was studied using FESEM, XRD, and Raman spectroscopy. In some cases, the inner oxide layer grew faster than the outer oxide layer when the fuel pin was heated in steam during the burst test. The evolution during transient heating of radial and circumferential crack growth in the oxide layer and the occurrence of delamination facilitated faster oxygen and hydrogen uptake. The hydrogen uptake in pre and post-burst samples was related to the oxygen uptake. The hydrogen concentration increases with the oxygen concentration in the pre-oxidized samples. Small oxygen and hydrogen concentrations were found in the post-burst as-received samples due to the formation of a protective oxide layer. The high-temperature oxide layer was formed at extremely high heating rates.</div></div>\",\"PeriodicalId\":20617,\"journal\":{\"name\":\"Progress in Nuclear Energy\",\"volume\":null,\"pages\":null},\"PeriodicalIF\":3.3000,\"publicationDate\":\"2024-10-13\",\"publicationTypes\":\"Journal Article\",\"fieldsOfStudy\":null,\"isOpenAccess\":false,\"openAccessPdf\":\"\",\"citationCount\":\"0\",\"resultStr\":null,\"platform\":\"Semanticscholar\",\"paperid\":null,\"PeriodicalName\":\"Progress in Nuclear Energy\",\"FirstCategoryId\":\"5\",\"ListUrlMain\":\"https://www.sciencedirect.com/science/article/pii/S0149197024004359\",\"RegionNum\":3,\"RegionCategory\":\"工程技术\",\"ArticlePicture\":[],\"TitleCN\":null,\"AbstractTextCN\":null,\"PMCID\":null,\"EPubDate\":\"\",\"PubModel\":\"\",\"JCR\":\"Q1\",\"JCRName\":\"NUCLEAR SCIENCE & TECHNOLOGY\",\"Score\":null,\"Total\":0}","platform":"Semanticscholar","paperid":null,"PeriodicalName":"Progress in Nuclear Energy","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0149197024004359","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
Zircaloy-4 fuel pin failure under simulated loss-of-coolant-accident conditions: Oxygen embrittlement
An extensive experimental investigation was performed to study the oxygen embrittlement of the Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under simulated Loss-of-Coolant Accident (LOCA) conditions. Zircaloy fuel cladding experiences creep and corrosion simultaneously during service and LOCA conditions. Zircaloy-4 fuel pins were pre-oxidized to attain different oxide layer thicknesses, achieving in-service conditions. These pre-oxidized tubes were then subjected to burst tests in the steam environment to mimic the LOCA scenario. The present study aims to improve the understanding of the effect of oxidation on the cladding microstructure and the mechanical response of the fuel pin in a LOCA scenario by accounting for the cross-influence, during transient heating, of oxidation and deformation on the behavior of the clad in the LOCA domain. The oxide layer morphology in pre- and post-burst samples was studied using FESEM, XRD, and Raman spectroscopy. In some cases, the inner oxide layer grew faster than the outer oxide layer when the fuel pin was heated in steam during the burst test. The evolution during transient heating of radial and circumferential crack growth in the oxide layer and the occurrence of delamination facilitated faster oxygen and hydrogen uptake. The hydrogen uptake in pre and post-burst samples was related to the oxygen uptake. The hydrogen concentration increases with the oxygen concentration in the pre-oxidized samples. Small oxygen and hydrogen concentrations were found in the post-burst as-received samples due to the formation of a protective oxide layer. The high-temperature oxide layer was formed at extremely high heating rates.
期刊介绍:
Progress in Nuclear Energy is an international review journal covering all aspects of nuclear science and engineering. In keeping with the maturity of nuclear power, articles on safety, siting and environmental problems are encouraged, as are those associated with economics and fuel management. However, basic physics and engineering will remain an important aspect of the editorial policy. Articles published are either of a review nature or present new material in more depth. They are aimed at researchers and technically-oriented managers working in the nuclear energy field.
Please note the following:
1) PNE seeks high quality research papers which are medium to long in length. Short research papers should be submitted to the journal Annals in Nuclear Energy.
2) PNE reserves the right to reject papers which are based solely on routine application of computer codes used to produce reactor designs or explain existing reactor phenomena. Such papers, although worthy, are best left as laboratory reports whereas Progress in Nuclear Energy seeks papers of originality, which are archival in nature, in the fields of mathematical and experimental nuclear technology, including fission, fusion (blanket physics, radiation damage), safety, materials aspects, economics, etc.
3) Review papers, which may occasionally be invited, are particularly sought by the journal in these fields.