Journal of Nuclear Fuel Cycle and Waste Technology最新文献_第6页

A Bi-Layered Three-Dimensional Mechanical Modeling of the Cladding and Its Creep Deformation Analysis 熔覆层的双层三维力学建模及蠕变分析

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-88944

M. Zhang, Jing-Gang Li, Xiaohan Liu, Yong Lu, Yanan Zhu

{"title":"A Bi-Layered Three-Dimensional Mechanical Modeling of the Cladding and Its Creep Deformation Analysis","authors":"M. Zhang, Jing-Gang Li, Xiaohan Liu, Yong Lu, Yanan Zhu","doi":"10.1115/icone29-88944","DOIUrl":"https://doi.org/10.1115/icone29-88944","url":null,"abstract":"\u0000 The cladding of the fuel rod is a cylinder-shaped structure made of Zirconium alloy, which will collapse due to structural creep under the extremely service conditions such as high temperature, high pressure, and high irradiation. The collapse of the cladding results in losing its structure function and threatening the safety of the reactor. Based on the commercial finite element software ABAQUS with its user subroutine CREEP, a bi-layered (coating and matrix material) three-dimensional (3D) cylindrical cladding model is established with thermal and irradiation finite creep behavior. The external pressure is assumed to be constant acting on the outer surface. The deformation and the rate of deformation increase with the increasing of the irradiation time in the reactor, which leads to the collapse of the cladding eventually. The initial ovality has a positive effect on the creep deformation. Compared with the single-layered model, the coating of the bi-layered cladding can prevent the Zirconium alloy matrix from excessive creep deformation and thus can protect the cladding. The thicker the coating, the stronger the protective effect from the mechanical point of view. A qualitative case of the cladding creep burst was simulated, and the behavior of the creep burst and creep collapse is similar. The corrosion and oxidation behavior are not considered herein for simplicity. The current bi-layered 3d model can be extended to the structural design, safety analysis, as well as life evaluation of some multi-layered cladding of the accident tolerant fuel (ATF).","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"29 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74434447","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Microstructural Evolution of SiC-Coated C/C Composites Under Argon Ion Irradiation 氩离子辐照下sic包覆C/C复合材料的微观组织演变

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-90325

Xiangmin Xie, Long Yan, Guodong Cheng, Xian Tang

{"title":"Microstructural Evolution of SiC-Coated C/C Composites Under Argon Ion Irradiation","authors":"Xiangmin Xie, Long Yan, Guodong Cheng, Xian Tang","doi":"10.1115/icone29-90325","DOIUrl":"https://doi.org/10.1115/icone29-90325","url":null,"abstract":"\u0000 SiC coatings have been used to improve the oxidation resistance and stability of C/C composites in high-temperature reactors. However, the irradiation-induced surface structural transformations of SiC-coated C/C composites have been rarely studied. Herein, chemical vapor reaction (CVR) SiC-coated C/C composites were irradiated with 300 keV argon ions at room temperature with irradiation doses ranging from 5 × 1015–1 × 1017 ions·cm−2. The damage patterns of the pristine C/C composites and SiC-coated C/C composites were observed using scanning electron microscopy, and the shape and size evolutions of the CVR-SiC particles were investigated as a function of the irradiation dose. The results revealed that the pristine C/C composites were severely damaged after ion irradiation, and a large number of defects and pores formed on the surface. In contrast, the ion-irradiated SiC-coated C/C composites showed an undamaged surface. As the irradiation dose increased from 0 to 1 × 1017 ions·cm−2, the CVR-SiC particles were transformed from irregular to spherical shapes, and the average size of the SiC particles was reduced from 22 to 5 μm. The size reduction and spheroidization of the SiC particles under irradiation were attributed to the amorphous transformation of SiC. This study can provide deeper insight into the irradiation behavior of SiC-coated C/C composites in high-temperature reactors.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"102 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80599450","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study on Synthesis of the Organophosphorus Functionalized MCM-41 And its Adsorption Property for Dy(III) 有机磷功能化MCM-41的合成及其对Dy(III)的吸附性能研究

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-93196

Cong Mao, Hongji Sang, Jiawei Zheng, Yan Wu

引用次数: 0

PWR Spent Fuel Dry Storage Loading Pattern and Safety Evaluation 压水堆乏燃料干贮存方式及安全性评价

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-93351

Li Wen, Gao Jinghui, Li Gang, Zhong Shengdong, Xu Jin, Zhao Bingquan

引用次数: 0

Corrosion Behavior of a Novel Alumina Forming Austenitic Steel Exposed to Supercritical Water 一种新型氧化铝形成奥氏体钢在超临界水中的腐蚀行为

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-92471

Dayun Sun, Yang Gao, S. Cong, Xianglong Guo

{"title":"Corrosion Behavior of a Novel Alumina Forming Austenitic Steel Exposed to Supercritical Water","authors":"Dayun Sun, Yang Gao, S. Cong, Xianglong Guo","doi":"10.1115/icone29-92471","DOIUrl":"https://doi.org/10.1115/icone29-92471","url":null,"abstract":"\u0000 The corrosion behavior of novel alumina-forming austenitic steel Fe-26Ni-19Cr-2.5Al-1Nb-0.5Si and 310S steel in aerated supercritical water (SCW) at 550 °C/25 MPa was investigated. The AFA and 310S steels has been exposed in supercritical water for up to 1000h. The results show that both the weight gain curves of AFA and 310S steels follow near-parabolic law. Although the weight gain of 310S steel exposed in supercritical water after 1000h was up to 37 mg/dm2, the weight gain of AFA steel exposed in supercritical water after 1000h was near 18 mg/dm2, which was only half of that of 310S steel. The weight gain curves indicating that the AFA steel has better corrosion resistance than 310S steel. Besides, microstructure characterization of two steels has been conducted by SEM, EDS and XRD. SEM images shows that there are some differences between surface morphology of 310S steel and AFA steel. The microstructure results show that 310S steel has a double oxide layer: a Fe-riched outer layer and a Cr-riched inner layer, while a multilayer structure mainly composed of Fe-riched oxide layer, Cr-riched oxide layer and Al-riched oxide layer was formed on AFA steel, indicating a different corrosion process from 310S steel. The corrosion mechanism of two steels based on the microstructure is discussed in detail.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"87 S1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"91030956","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Analysis of Hydrogenation Properties of Nuclear Fuel Zirconium Alloy Cladding 核燃料锆合金包层加氢性能分析

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-89580

Lin Qin

引用次数: 0

Potential Crud Related Margin Gains With Chromium Coated Zirconium Cladding 潜在的原油相关边际收益与镀铬锆包层

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-94477

W. Byers, Guoqiang Wang

{"title":"Potential Crud Related Margin Gains With Chromium Coated Zirconium Cladding","authors":"W. Byers, Guoqiang Wang","doi":"10.1115/icone29-94477","DOIUrl":"https://doi.org/10.1115/icone29-94477","url":null,"abstract":"\u0000 The build-up of deposits (Crud) on fuel cladding can cause several issues for PWRs. The crud becomes activated and is transported to other locations in the RCS increasing radiation fields, it can collect boron causing power shifts (CIPS, Crud Induced Power Shifts) and if thick enough localized corrosion will be increased (CILC, Crud Induced Localized Corrosion). The application of a chromium coating to zirconium alloy cladding is being pursued to improve the corrosion performance of the cladding both during normal operation and in the event of accident conditions producing high cladding temperatures. The change will also influence the crud affinity of the cladding. The deposition of corrosion products from the ex-core portions of the plant will be altered. Also, corrosion of the cladding itself contributes to the crud burden in a PWR, and this factor will undoubtedly change with the application of a chromium coating. Changes in the surface texture produced by the coating will also influence crud deposition by changing boiling nucleation and thermal hydraulic factors.\u0000 This paper discusses how a change to chromium coated fuel cladding will likely affect crud deposition and will give the results of crud deposition testing of chromium coated cladding in the WALT loop.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"10 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81263354","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 1

Experimental Study on Dynamic Tensile Strength of Graphite 石墨动态拉伸强度的实验研究

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-90670

Dongqing Tian, Li Shi, Libin Sun, Keya Shen, Kun Xu

{"title":"Experimental Study on Dynamic Tensile Strength of Graphite","authors":"Dongqing Tian, Li Shi, Libin Sun, Keya Shen, Kun Xu","doi":"10.1115/icone29-90670","DOIUrl":"https://doi.org/10.1115/icone29-90670","url":null,"abstract":"\u0000 Dynamic tensile strength is an important parameter in the design of graphite components for High Temperature Gas-Cooled Reactors (HTGR) to evaluate the integrity of core support structures. The Digital Image Correlation and Split Hopkinson Pressure Bar (DIC-SHPB) test system was used to perform the disc compression tests according to ASTM D8289-20 to study the dynamic splitting tensile strengths of graphites with different grain sizes. The fracture process was captured by a high-speed camera and the tensile strain was computed by DIC. The results show that the SHPB test method is capable of performing the disc compression tests of graphite. The dynamic tensile stress-strain curve of graphite underwent four stages: compression stage, near-elastic stage, crack development stage and crack non-stable extension stage. When the strain rate is in the range of 330 s−1 to 630 s−1, the dynamic tensile strength of graphite increases with increasing strain rate. The dynamic increase factor (DIF) of fine-grained graphite ranged from 1.05 to 1.2, while that of coarse-grained graphite was about 1.2 to 1.6. There is a large dispersion in the dynamic tensile strength of graphite when the strain rate is increased. It was found that the tensile strength of graphite improved considerably with increasing strain rate, while the fracture strain decreased slightly.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"3 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90649197","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Numerical Study of the Effect of δ-Ferrite on Crack-Tip Mechanical Field δ-铁氧体对裂纹尖端力学场影响的数值研究

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-92656

Yule Wu, Le-fu Zhang

{"title":"Numerical Study of the Effect of δ-Ferrite on Crack-Tip Mechanical Field","authors":"Yule Wu, Le-fu Zhang","doi":"10.1115/icone29-92656","DOIUrl":"https://doi.org/10.1115/icone29-92656","url":null,"abstract":"\u0000 A dual-phase microstructure with 5–12% δ-ferrite is needed to prevent hot cracking of stainless steel weld metal. However, the δ-ferrite also makes it susceptible to long-term thermal aging embrittlement. Recent studies revealed that the stress corrosion cracking (SCC) susceptibility might either be reduced or increased by the δ-ferrite, depending on the embrittlement degree of the ferrite during the operation. One possible explanation was that the harder δ-ferrite inhibited the crack growth by changing the cracking direction and creating a highly branched crack path. To further reveal the mechanism of δ-ferrite effect on SCC behavior of those materials in high temperature water, a finite element investigation for the distribution of crack-tip stress and plastic strain field with and without δ-ferrite under a constant load was conducted. The result shows that the crack tip stress and plastic strain decrease when the crack encounters the δ-ferrite, indicating a lower cracking susceptibility, and the enhancement of δ-ferrite hardness can further reduce the crack-tip plastic strain. the severe stress and strain states occur along the ferrite edge, this indicates that the cracks are more tend to initiate and develop along the interface of δ-ferrite and austenite matrix.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"43 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81284545","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Preliminary Study on Corrosion Behavior and Mechanical Properties Of Inconel 617 in Impure Helium Environment of VHTR Inconel 617在VHTR非纯氦环境中的腐蚀行为和力学性能的初步研究

IF 0.4

Journal of Nuclear Fuel Cycle and Waste Technology Pub Date : 2022-08-08 DOI: 10.1115/icone29-88943

B. Du, Huaqiang Yin, Tengyu Ma, Haoxiang Li, Weizhen Zheng, Xuedong He

{"title":"Preliminary Study on Corrosion Behavior and Mechanical Properties Of Inconel 617 in Impure Helium Environment of VHTR","authors":"B. Du, Huaqiang Yin, Tengyu Ma, Haoxiang Li, Weizhen Zheng, Xuedong He","doi":"10.1115/icone29-88943","DOIUrl":"https://doi.org/10.1115/icone29-88943","url":null,"abstract":"\u0000 Helium is generally used as coolant in Very-High-Temperature Reactor (VHTR), but a small amount of impurity gas will inevitably be mixed in the primary coolant during construction, operation and maintenance. At high temperature, these impurity gases will corrode with the alloy materials of the intermediate heat exchanger, resulting in the decline of the properties of the superalloy materials. This paper mainly studied the changes of microstructure and mechanical properties of Inconel 617, a candidate material for high temperature reactor intermediate heat exchanger, after aging for 50 hours in impure helium at 950 °C. The microstructure of the alloy was characterized by weighing, scanning electron microscope and energy dispersive X-ray spectroscopy. After corrosion, Cr oxide layer was formed on the alloy surface, and Al internal oxidation appeared below the oxide layer. The mechanical test results showed that the strength and plasticity of Inconel 617 decreased significantly, which was related to the intergranular oxides and carbides. The fracture morphology of the alloy is mainly brittle fracture.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"6 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81047289","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0