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}
During the service of the reactor, the zirconium alloy will inevitably interact with hydrogen to form zirconium hydride, which will adversely affect its mechanical properties. In this paper, the first-principles software VASP is used to analyze the phase transition of zirconium hydride, and based on the results, the subprogram of MAAP4 integrated analysis for severe accidents is optimized to further analyze the interaction between zirconium alloy cladding and hydrogen. Verification. The results show that the diffusion equation has large calculation potential and high calculation accuracy for the effect of zirconium hydride with obvious phase transition, and can realize the hydrogen performance analysis of the cladding material in the reactor. At the same time, VASP shows that δ – ZrH1.5 is the main component of cladding material hydrogenation It is also the main reason for brittle transition.
{"title":"Analysis of Hydrogenation Properties of Nuclear Fuel Zirconium Alloy Cladding","authors":"Lin Qin","doi":"10.1115/icone29-89580","DOIUrl":"https://doi.org/10.1115/icone29-89580","url":null,"abstract":"\u0000 During the service of the reactor, the zirconium alloy will inevitably interact with hydrogen to form zirconium hydride, which will adversely affect its mechanical properties. In this paper, the first-principles software VASP is used to analyze the phase transition of zirconium hydride, and based on the results, the subprogram of MAAP4 integrated analysis for severe accidents is optimized to further analyze the interaction between zirconium alloy cladding and hydrogen. Verification. The results show that the diffusion equation has large calculation potential and high calculation accuracy for the effect of zirconium hydride with obvious phase transition, and can realize the hydrogen performance analysis of the cladding material in the reactor. At the same time, VASP shows that δ – ZrH1.5 is the main component of cladding material hydrogenation It is also the main reason for brittle transition.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"18 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88577169","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 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. 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.
{"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}
Dongqing Tian, Li Shi, Libin Sun, Keya Shen, Kun Xu
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.
动态抗拉强度是高温气冷堆(HTGR)石墨构件设计中评价堆芯支撑结构完整性的重要参数。采用Digital Image Correlation and Split Hopkinson Pressure Bar (DIC-SHPB)测试系统，按照ASTM D8289-20的要求，进行圆盘压缩试验，研究不同晶粒尺寸石墨的动态劈裂拉伸强度。用高速摄像机捕捉断裂过程，用DIC计算拉伸应变。结果表明，SHPB试验方法能够进行石墨的圆盘压缩试验。石墨的动态拉伸应力-应变曲线经历了压缩阶段、近弹性阶段、裂纹发展阶段和裂纹非稳定扩展阶段四个阶段。当应变速率为330 s−1 ~ 630 s−1时，石墨的动态抗拉强度随应变速率的增大而增大。细粒石墨的动态增长因子(DIF)在1.05 ~ 1.2之间，粗粒石墨的动态增长因子在1.2 ~ 1.6之间。随着应变速率的增大，石墨的动态抗拉强度有较大的分散。结果表明，随着应变速率的增加，石墨的抗拉强度显著提高，而断裂应变略有下降。
{"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}
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.
{"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}
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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