M. Zhang, Jing-Gang Li, Xiaohan Liu, Yong Lu, Yanan Zhu
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).
{"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}
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.
{"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}
The composite of organophosphorus groups loaded on MCM-41(MCM-Zr-TBP) was prepared by multi-steps impregnation method to develop a novel adsorbent for radioactive lanthanides extraction from the secondary contaminated water. The synthesized hybrid material was characterized by SEM and TG. Dy(III) was taken as the representative of trivalent lanthanides. The adsorption performance of Dy(III) on MCM-Zr-TBP composite was systematically studied as the functions of solution pH, initial concentration, interaction time and aqueous temperature. The results showed that Dy(III) adsorption on MCM-Zr-TBP composite was highly dependent on aqueous pH and initial metal ion concentration. Compared with the Freundlich and pseudo-first order models, the Langmuir and pseudo-second order models presented better fitting for the adsorption data. These results indicated that MCM-Zr-TBP was found to be an effective and competent adsorbent, which could be used for the effective removal of lanthanides from the wastewater.
{"title":"Study on Synthesis of the Organophosphorus Functionalized MCM-41 And its Adsorption Property for Dy(III)","authors":"Cong Mao, Hongji Sang, Jiawei Zheng, Yan Wu","doi":"10.1115/icone29-93196","DOIUrl":"https://doi.org/10.1115/icone29-93196","url":null,"abstract":"\u0000 The composite of organophosphorus groups loaded on MCM-41(MCM-Zr-TBP) was prepared by multi-steps impregnation method to develop a novel adsorbent for radioactive lanthanides extraction from the secondary contaminated water. The synthesized hybrid material was characterized by SEM and TG. Dy(III) was taken as the representative of trivalent lanthanides. The adsorption performance of Dy(III) on MCM-Zr-TBP composite was systematically studied as the functions of solution pH, initial concentration, interaction time and aqueous temperature. The results showed that Dy(III) adsorption on MCM-Zr-TBP composite was highly dependent on aqueous pH and initial metal ion concentration. Compared with the Freundlich and pseudo-first order models, the Langmuir and pseudo-second order models presented better fitting for the adsorption data. These results indicated that MCM-Zr-TBP was found to be an effective and competent adsorbent, which could be used for the effective removal of lanthanides from the wastewater.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"66 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78644021","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}
Li Wen, Gao Jinghui, Li Gang, Zhong Shengdong, Xu Jin, Zhao Bingquan
This paper describes the loading pattern, safety evaluation and operating experiences of PWR plant off-site Spent Fuel Dry Storage System (SFDSS). According to the limits from the SFDSS criticality and heat-transfer safety analysis, ALARA principle and AFA series fuel characteristics, technical staff complete the loading pattern and safety evaluation by analyzing influence of assemblies’ initial enrichment, burn up and cooling time. This paper also shows the temperature and dose data during AFA series fuel SFDSS loading, transportation and storage, which can indicate the actual condition of SFDSS criticality, heat-transfer and radiation. In general, AFA series fuel SFDSS is in good condition and meet the expected design and safety requirements.
{"title":"PWR Spent Fuel Dry Storage Loading Pattern and Safety Evaluation","authors":"Li Wen, Gao Jinghui, Li Gang, Zhong Shengdong, Xu Jin, Zhao Bingquan","doi":"10.1115/icone29-93351","DOIUrl":"https://doi.org/10.1115/icone29-93351","url":null,"abstract":"\u0000 This paper describes the loading pattern, safety evaluation and operating experiences of PWR plant off-site Spent Fuel Dry Storage System (SFDSS). According to the limits from the SFDSS criticality and heat-transfer safety analysis, ALARA principle and AFA series fuel characteristics, technical staff complete the loading pattern and safety evaluation by analyzing influence of assemblies’ initial enrichment, burn up and cooling time. This paper also shows the temperature and dose data during AFA series fuel SFDSS loading, transportation and storage, which can indicate the actual condition of SFDSS criticality, heat-transfer and radiation. In general, AFA series fuel SFDSS is in good condition and meet the expected design and safety requirements.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"4 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87870973","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}
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.
{"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}
The application of the dual-cooled annular fuel element can significantly improve the safety and economy of current pressurized water reactors. Due to the geometric differences, the fuel relocation caused by the fuel cracking in annular rods was much more complex and influential on thermal-hydraulic performance than that in solid rods. In this research, the Discrete Element Method (DEM) was applied to perform the simulation of fuel fragment relocation in the annular fuel rod under both long-term normal operation and LOCA conditions, in order to get deep insights into the fuel relocation mechanisms and develop the mathematical relocation model. Under normal operations, it was found that the radial fuel relocation was bidirectional of both inward and outward, resulting in the radial size reduction of both internal and external gas gaps. The maximum reduction of total gaps was about 58% of the as-fabricated value, and the maximum allowable recovery fraction of fuel relocation was about 55%. Under LOCA conditions, the ballooning of the cladding was full considered in the fuel relocation. As a consequence, the substantial axial fuel relocation was found, which resulting a size reduction of active fuel length and a local fuel accumulation at the position of ballooning.
{"title":"Research on Cracked Fuel Relocation of Dual-Cooled Annular Fuel Element Under Normal and LOCA Conditions","authors":"Yangbin Deng, Yuan Yin","doi":"10.1115/icone29-90755","DOIUrl":"https://doi.org/10.1115/icone29-90755","url":null,"abstract":"\u0000 The application of the dual-cooled annular fuel element can significantly improve the safety and economy of current pressurized water reactors. Due to the geometric differences, the fuel relocation caused by the fuel cracking in annular rods was much more complex and influential on thermal-hydraulic performance than that in solid rods. In this research, the Discrete Element Method (DEM) was applied to perform the simulation of fuel fragment relocation in the annular fuel rod under both long-term normal operation and LOCA conditions, in order to get deep insights into the fuel relocation mechanisms and develop the mathematical relocation model. Under normal operations, it was found that the radial fuel relocation was bidirectional of both inward and outward, resulting in the radial size reduction of both internal and external gas gaps. The maximum reduction of total gaps was about 58% of the as-fabricated value, and the maximum allowable recovery fraction of fuel relocation was about 55%. Under LOCA conditions, the ballooning of the cladding was full considered in the fuel relocation. As a consequence, the substantial axial fuel relocation was found, which resulting a size reduction of active fuel length and a local fuel accumulation at the position of ballooning.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"143 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73978261","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}
This paper takes a certain nuclear power plant (NPP) which consists of two 1000 MWe PWR units as the research object, and summarizes various fuel management strategies experienced since it started commercial operation in 2002 and 2003. These fuel management strategies include: 1) high-leakage annual refueling, refueling fuel enrichment is 3.2% and cycle length reaches about 270 EFPD; 2) fuel type mixed refueling (AFA2G and AFA3G mixed), refueling fuel enrichment increases from 3.2% to 3.7% and cycle length reaches about 320 EFPD; 3) low-leakage 1/4 refueling, refueling fuel enrichment increases from 3.7% to 4.2% and cycle length reaches about 316 EFPD that fuel economy has been greatly improved; 4) low-leakage 18-months refueling, refueling fuel enrichment increases from 4.2% to 4.45% and the refueling cycle has been extended from annual refueling to an average of 16 months; 5) two-enrichment refueling, refueling fuel enrichment is extended from single 4.45% to two enrichment 4.45% and 4.00%, which improves the flexibility of fuel management to well meet the load shedding needs. This paper reviews each fuel management strategy’s general refueling pattern, main design parameters and actual operating parameters of the core, fuel economy and the feedback of related problems, finally further suggests a few optimizations of future fuel management strategies.
{"title":"Summary of the History of Improvement and Optimization of Fuel Management Strategies for Two 1000 MWe PWR Units","authors":"Zhou Xiaoling, Xu Zhixian, Zhao Bingquan, Li Wen, Zhong Shengdong, Gao Jinghui","doi":"10.1115/icone29-92954","DOIUrl":"https://doi.org/10.1115/icone29-92954","url":null,"abstract":"\u0000 This paper takes a certain nuclear power plant (NPP) which consists of two 1000 MWe PWR units as the research object, and summarizes various fuel management strategies experienced since it started commercial operation in 2002 and 2003. These fuel management strategies include:\u0000 1) high-leakage annual refueling, refueling fuel enrichment is 3.2% and cycle length reaches about 270 EFPD;\u0000 2) fuel type mixed refueling (AFA2G and AFA3G mixed), refueling fuel enrichment increases from 3.2% to 3.7% and cycle length reaches about 320 EFPD;\u0000 3) low-leakage 1/4 refueling, refueling fuel enrichment increases from 3.7% to 4.2% and cycle length reaches about 316 EFPD that fuel economy has been greatly improved;\u0000 4) low-leakage 18-months refueling, refueling fuel enrichment increases from 4.2% to 4.45% and the refueling cycle has been extended from annual refueling to an average of 16 months;\u0000 5) two-enrichment refueling, refueling fuel enrichment is extended from single 4.45% to two enrichment 4.45% and 4.00%, which improves the flexibility of fuel management to well meet the load shedding needs.\u0000 This paper reviews each fuel management strategy’s general refueling pattern, main design parameters and actual operating parameters of the core, fuel economy and the feedback of related problems, finally further suggests a few optimizations of future fuel management strategies.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"129 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73762617","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}
Yinghong Li, Z. Lv, C. Li, Yuyu Lin, Shaofang Lin, Zhiyi Cai, Yong-jun Deng, Lihong Nie
Manufacturing defects of fuel rod end plug during welding affect fuel reliability. Failure of ring welding or sealing welding of fuel rod end plug during operation will cause fission gas to enter primary coolant of reactor. In order to determine the sensitivity of the influence of relevant characteristic parameters on the performance of fuel rods during end plug welding, based on the Quality Function Deployment method (QFD) and taking tungsten-inert-gas arc welding (TIG) as an example, the characteristic parameters of fuel rod end plug welding are analyzed. Combined with the experimental results, a correlation matrix between welding parameters and welding performance is constructed by LIKERT scale method with focus on the main welding performance criteria such as welding joint size, color, depth of fusion, metallography (bloating, porosity, inclusion, grain boundary separation, etc.), oxidation resistance, etc. And after rating its correlation degree and importance degree, the relevant House of Quality of end plug welding is constructed, and the key impact parameters are identified. Based on the above analysis, some feasible optimization and improvement proposals are put forward for the quality supervision and welding process of fuel element manufacturing factory.
{"title":"Key Impact Factors Analysis of Fuel Rod End Plug Welding Based on QFD","authors":"Yinghong Li, Z. Lv, C. Li, Yuyu Lin, Shaofang Lin, Zhiyi Cai, Yong-jun Deng, Lihong Nie","doi":"10.1115/icone29-92803","DOIUrl":"https://doi.org/10.1115/icone29-92803","url":null,"abstract":"\u0000 Manufacturing defects of fuel rod end plug during welding affect fuel reliability. Failure of ring welding or sealing welding of fuel rod end plug during operation will cause fission gas to enter primary coolant of reactor. In order to determine the sensitivity of the influence of relevant characteristic parameters on the performance of fuel rods during end plug welding, based on the Quality Function Deployment method (QFD) and taking tungsten-inert-gas arc welding (TIG) as an example, the characteristic parameters of fuel rod end plug welding are analyzed. Combined with the experimental results, a correlation matrix between welding parameters and welding performance is constructed by LIKERT scale method with focus on the main welding performance criteria such as welding joint size, color, depth of fusion, metallography (bloating, porosity, inclusion, grain boundary separation, etc.), oxidation resistance, etc. And after rating its correlation degree and importance degree, the relevant House of Quality of end plug welding is constructed, and the key impact parameters are identified. Based on the above analysis, some feasible optimization and improvement proposals are put forward for the quality supervision and welding process of fuel element manufacturing factory.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"2 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74364125","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}
In the core of pressurized water nuclear reactor, coolant flow-induced vibration of Grid to rod fretting (GTRF) is the dominant factor leading to fuel rod damage. pre-oxidization treatment of zirconium cladding forming a ceramic layer on its surface is the main way to reduce the GTRF wear. In this paper, the growth law of CZ2 alloy cladding pre-oxidization zirconia ceramic layer formed in air was studied. The micro-hardness and elastic modulus of CZ2 alloy cladding and zirconia ceramic layer were measured by in-situ nano-mechanical testing system., while the morphology of these pre-oxidization zirconia ceramic layer were observed by scanning electron microscope. The fretting wear properties of the pre-oxidization zirconia ceramic layer were studied by high temperature and high pressure fretting wear tests. The results show that the pre-oxidization zirconia ceramic layer growth law of CZ2 alloy cladding at 560°C and 600°C is consistent, and the pre-oxidization zirconia ceramic layer are compact and crack-free. The pre-oxidization zirconia ceramic layer can improve the fretting wear resistance of CZ2 alloy cladding at high temperature and high pressure, and the maximum wear depth were reduced by 80%.
{"title":"Study of Pre-Oxidization Law and Fretting Wear Resistance of CZ2 Alloy Cladding","authors":"Guocheng Sun, Shi Lin, Xu Wang, Liu-tao Chen","doi":"10.1115/icone29-93804","DOIUrl":"https://doi.org/10.1115/icone29-93804","url":null,"abstract":"\u0000 In the core of pressurized water nuclear reactor, coolant flow-induced vibration of Grid to rod fretting (GTRF) is the dominant factor leading to fuel rod damage. pre-oxidization treatment of zirconium cladding forming a ceramic layer on its surface is the main way to reduce the GTRF wear. In this paper, the growth law of CZ2 alloy cladding pre-oxidization zirconia ceramic layer formed in air was studied. The micro-hardness and elastic modulus of CZ2 alloy cladding and zirconia ceramic layer were measured by in-situ nano-mechanical testing system., while the morphology of these pre-oxidization zirconia ceramic layer were observed by scanning electron microscope. The fretting wear properties of the pre-oxidization zirconia ceramic layer were studied by high temperature and high pressure fretting wear tests. The results show that the pre-oxidization zirconia ceramic layer growth law of CZ2 alloy cladding at 560°C and 600°C is consistent, and the pre-oxidization zirconia ceramic layer are compact and crack-free. The pre-oxidization zirconia ceramic layer can improve the fretting wear resistance of CZ2 alloy cladding at high temperature and high pressure, and the maximum wear depth were reduced by 80%.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"86 8","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72394323","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}
B. Du, Huaqiang Yin, Tengyu Ma, Haoxiang Li, Weizhen Zheng, Xuedong He
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.
{"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}
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