Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components最新文献

{"title":"Microstructure, Texture and Mechanical Properties in Different Deformation Regions of the Compressed Zr-4 Alloy With Lamellar Structure","authors":"Jie Ren, Geping Li, Fusen Yuan, Ben Wang, Fuzhou Han, Muhammad Ali, Yingdong Zhang, Wenbin Guo, Bo Gao","doi":"10.1115/icone29-92283","DOIUrl":"https://doi.org/10.1115/icone29-92283","url":null,"abstract":"\u0000 In the present work, the inhomogeneous plastic deformation of the compressed Zr-4 alloy with lamellar structure has been studied systematically. In the compression test of Zr-4 alloy with reduction of 70% at 650°C, the severe deformed area (SDA) and difficult deformed area (DDA) formed. The differences in DDA and SDA including microstructure, texture, and mechanical properties were investigated, using electron backscattering diffraction (EBSD) and Vickers hardness tester. The results showed that a large number of equiaxed grains appeared in SDA due to the occurrence of dynamic recrystallization (DRX), while coarse lamellar grains were still the main microstructure in DDA, showing that the DDA was the intermediate between the initial alloy and SDA. Furthermore, the discontinuous recrystallization mechanism was proposed. According to electron backscattered diffraction (EBSD) analysis, the predominant texture formed in SDA and DDA made normal direction of its {0001} planes of α phase parallel to the compression direction, and another stronger texture in SDA made the normal direction of part {0001} planes ∼15° away from the compression direction. The observed texture difference in SDA and DDA mainly resulted from the DRXed grains. To evaluate the influence of SDA on the mechanical properties of the Zr-4 alloy, microhardness tests were employed. The alloy in SDA had been proved to be strengthened compared to DDA, which was due to both the grain refinement and the texture diversity.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"42 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127086871","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}

{"title":"Analytical and Experimental Research on Reactor Coolant Pump Mechanical Seals With Hydrodynamic Deep-Grooves","authors":"Liming Tang, Yongchun Pan, Xin Tan, Le Jin, Y. Wang","doi":"10.1115/icone29-91237","DOIUrl":"https://doi.org/10.1115/icone29-91237","url":null,"abstract":"\u0000 Hydrodynamic mechanical seal with deep grooves is used on reactor coolant pumps (RCP) of KSB SE. It has several good features, such as simple structure, excellent wear resistance, low pressure leakage rate, etc. In this paper, an analysis method across different calculation platforms is used to analyze the RCP mechanical seals. The finite element analysis based on ANSYS is to simulate the thermal deformation of mechanical seal rings, while an algorithm based on Reynolds equation with JFO cavitation boundary is used to evaluate the seal film, which is less than 1 μm thick. A specific test rig was built up to carry out the operational test for RCP mechanical seals. In this test rig, full pressure, normal temperature and rotary speed can be achieved, meanwhile some special conditions can be simulated, e.g. thermal impact, pressure drop, and loss power. The internal pressure, flow rate, temperature, and rotary speed can be detected in-situ. The low-pressure leakage rate is the flow through the gap between seal faces, which is the most important operational parameter to represent the tribological features of seal faces. In this test, the low-pressure leakage rate can be detected by a weight-based measurement system. After 200 hours test, all detecting parameters kept constant. The detected low-pressure leakage rate was around 80 mL/h, which has good consistency with the analysis results.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"42 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127486336","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}

{"title":"Research on the Application of Cobalt-Based Alloy Laser Additive Manufacturing in Reactor Internals","authors":"Qingtian Wang, Rui Cao, Xianjun Zhang","doi":"10.1115/icone29-90890","DOIUrl":"https://doi.org/10.1115/icone29-90890","url":null,"abstract":"\u0000 In view of the quality problems of manual TIG coating of cobalt-based alloy, the Laser Additive Manufacturing process was studied, especially the optimization of process parameters. Then a series of experiments were carried out on the cobalt-based alloy Additive Manufacturing layer and compared with the traditional manual TIG coating layer. The results show that the microstructure, the hardness, the wear resistance and corrosion resistance are better than.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"49 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132900868","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}

{"title":"Finite Element Analysis and Prediction of the Cross-Section Morphology of Grid Strips","authors":"Jianzhong Mao, Zhi Cheng, Jun Su, Ben Wang, Pengfei Zhang, Zezhi Mei, Jiajian Yuan, Bo Gao","doi":"10.1115/icone29-92344","DOIUrl":"https://doi.org/10.1115/icone29-92344","url":null,"abstract":"\u0000 To investigate the relationships between the quality of the section morphology of the zirconium alloy strips and the blanking parameters as well as the design of the blanking mould, 64 processing parameter combinations are generated by using Latin Hypercube Sampling and the finite element models with different boundary conditions have been constructed to acquire the corresponding section qualities. The calculation results of the finite element model are verified by the experiment data. Furthermore, the random forest algorithm is employed to build the prediction model between the processing parameters and the lengths of the section bright strip. The sensitive analysis of the processing parameters influenced on the section quality have been conducted through the prediction model. The results indicate that the random forest prediction model have a satisfied accuracy, which could effectively forecast the length of the bright band in the section, and then provides guidance on mold design as well as process parameter selections. The blanking clearance included in the design factor has a decisive influence on the length of the bright strip; The friction coefficient has the weakest effect on the results. The effect of die fillet radius and blanking speed on the length of bright band is not significant.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"47 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122695376","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}

{"title":"Die Design and Process Study of Zr-4 Alloy Profile Extrusion","authors":"Peng Qi, Bo Gao, Zhihao Zhang, Jianming Yin","doi":"10.1115/icone29-92961","DOIUrl":"https://doi.org/10.1115/icone29-92961","url":null,"abstract":"\u0000 According to the application requirements of Zr-4 profile with outer square and inner circle for small nuclear reactors, the extrusion die structure and extrusion process parameters are optimized through numerical simulation and calculation. Based on the study of alloy friction boundary conditions and heat exchange conditions, the influences of die inlet angle and working belt structure size on metal flow behavior were clarified. Combined with the analysis of related factors such as flow stress and temperature rise variation law in extrusion process, determine the structure size of the die and the process parameter range of the extrusion, use the on-site 1250T horizontal extruder to carry out the pilot scale experiment of profile extrusion, verify the results of the preliminary analysis and calculation, determine the optimal process technology plan, and trial-produce The required outer square and inner circle zirconium alloy profiles that every sizes and performances meet the design requirements. The cross-sectional size of this profile is 36mm × 36mm, the inner hole diameter is 20mm, and the length is 1500∼2500(mm). It is the first zirconium alloy profile produced by the extrusion process in China. This process can ensure the grains are broken under large deformation, and obtain profiles with good performance, uniform wall thickness and longer length. At the same time, it can also reduce material loss, improve production efficiency.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129658488","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}

{"title":"DFMEA Application in Rotor Design of Million-Kilowatt Nuclear Power Stream Turbine","authors":"Xue Binghan","doi":"10.1115/icone29-90280","DOIUrl":"https://doi.org/10.1115/icone29-90280","url":null,"abstract":"\u0000 DFMEA is a systematic and scientific method for new product design and development, which is applicable throughout the entire process of new product design and development. In this paper, DFMEA method is adopted in the systematic analyzing of the million-kilowatt nuclear power steam turbine rotor new product design and development process. On the basis of the introduction of product characteristics and current development requirement, DFMEA is introduced to the design and development process, applying the block diagram and structure-function matrix to locate the possible failure factors. Subsequently, the evaluation criteria of Severity (S), Occurrence (O), and Detection (D) are introduced and defined. According to the criteria, the Risk Priority Numbers (RPN) and Action Priority (AP) of key risks are sorted, therefore indicating the terms critical to product quality improvement.\u0000 Through the application of DFMEA, an effective risk identification and analysis of the million-kilowatt nuclear power steam turbine rotor is carried out, achieving the results that operational failures are advanced to the design stage to implement countermeasures, therefore realizing the improvement of product and system operation reliability, optimization of project cycle and manufacturing cost reduction. DFMEA promotes the optimization of overall new product development process. Meanwhile it provides ideas for the subsequent design optimization methods of similar kinds.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"11 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127606983","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}

{"title":"Application of Vacuum Hot Helium Leak Examination Technology for ITER Blanket Shield Blocks","authors":"Weiping Li, Guo-shuai Jiang, Wei Luo, Renjun Fang","doi":"10.1115/icone29-91816","DOIUrl":"https://doi.org/10.1115/icone29-91816","url":null,"abstract":"\u0000 The Blanket Shield Blocks is an important part of the International Thermonuclear Experimental Reactor (ITER). In order to ensure its sealing performance, each shield blocks needs to be taken a vacuum hot Helium leak testing. A method for high-sensitivity examination using a vacuum thermal helium leak test system is proposed in this paper. The shield blocks can remove its own impurity gas and increase the activity of Helium atoms by increasing the environment temperature. The examination sensitivity is improved by increasing the pressure difference between the inside and outside of the shield block. The sensitivity of this method can realize the helium leak test with the system sensitivity of 5 × 10−11 Pa·m3/s under the conditions of examination environment temperature of 250°C, ambient pressure of 1 × 10−5Pa, and products internal helium pressure of 4MPa.\u0000 The results indicate that the technology has a high degree of automation and data traceability, which can avoid the influence of the product’s own outgassing on the examination, and has a high sensitivity to the examination of subtle air leakage.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"35 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134082965","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}

{"title":"Study on Forging Technology of Integrated Topcover Forging for AP1000 Reactor Pressure Vessel","authors":"Zhaoyin Li, Kai Dong, Zhiyong Tao","doi":"10.1115/icone29-92966","DOIUrl":"https://doi.org/10.1115/icone29-92966","url":null,"abstract":"\u0000 AP1000 topcover forging, as a key component in the Reactor Pressure Vessel (RPV), is located at the top of the RPV and is an important reactor primary component within the pressure boundary of the primary circuit, and its integrity is of great significance to the safe operation of nuclear power equipment. The traditional topcover part is welded by three parts: the upper head, the topcover flange and the stacking tube socket. When the components are in service, the in-service inspection workload of the equipment is large, and the safety risk is high under complex working conditions. Therefore, the study of a large forging integrated top cover with a stacking tube socket is carried out to realize the upper head, the topcover flange and the stacking tube as integrated topcover. The integral forging of the integrated topcover is an urgent problem to be solved.\u0000 This paper mainly studies the development of forging process for topcover forging of AP1000 RPV with SA508 III material. Based on the shape characteristics and forging rheological characteristics of the topcover forgings, the topcover tube seat was identified as the key point of forming, so the concave round cake preform was designed, and the final forging was formed by spinning the preform. The preform forming simulation and the final spinning forming simulation were carried out by DEFORM software respectively. The simulation results show that the metal filling is good, the forging allowance is sufficient, but the fracture factor of the bottom head is high. In order to more comprehensively verify the feasibility of the forming scheme, Trial production of 1:20 small scale simulation parts was carried out, and the forming results showed that the small scale parts of the integrated topcover were well filled after spinning. Through the manufacture of theoretical and practical simulation parts, the forging process scheme of spinning preforms through two sets of dies is verified, which effectively solves the problem that the size of the top cover forgings is difficult to guarantee. It was manufactured according to the designed forging process using the 300-ton series of steel ingots, and the final product size and non-destructive testing met the requirements, and the qualified product was delivered.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133640559","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}

{"title":"Radiographic Examination of Electric Heater Tube Weld for HPR1000 Pressurizer","authors":"Jintao Yu, Dandan Lu, Jianlei Zhang","doi":"10.1115/icone29-92958","DOIUrl":"https://doi.org/10.1115/icone29-92958","url":null,"abstract":"\u0000 HPR1000 (Hua-long Pressurized Reactor) nuclear power technology is a third generation nuclear power technology independently developed by China and has completely independent intellectual property rights. The pressurizer is one of the most important equipments in nuclear power plants. Compared to the regulator of the CPR1000 pressurizer, the HL-1000 pressurizer designed by CNPDC has as many as 108 electric heaters. Each electric heater contains a weld of electric heating sleeve to connection. RT should be completed at least 216 exposures, if each weld is exposure with a minimum of two 90 degree shots, in this paper, the authors have improved RT examination technique of these welds and verified it on the actual products. The results show that the new technique can reduce the number of exposures by more than one third and speed up the manufacture of the whole pressurizer regulator.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"9 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127759653","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}

{"title":"Experimental Study on the Feasibility of Dry Grinding of MOX Core Blocks","authors":"Jianzhong Mao, Xin Liu, Ben Wang, Pengfei Zhang, Bo Gao","doi":"10.1115/icone29-92415","DOIUrl":"https://doi.org/10.1115/icone29-92415","url":null,"abstract":"\u0000 The MOX core block is the main component of the MOX fuel assembly. In this paper, a resin-bonded diamond cup wheel was used for grinding experiments on three ceramic materials, ZrO2, CeO2+AL2O3 and CeO2+MgO, with similar properties to the MOX core block. A grinding temperature prediction model was established using the circular contact moving heat source model, and the grinding temperature of the workpiece surface was predicted and the control strategy was studied in conjunction with the grinding experimental results to evaluate the dry grinding adaptability of the MOX core block based on the workpiece surface grinding temperature. The results show that when the grinding wheel linear speed Vs = 60 m/s, the workpiece feed linear speed Vwt = 627mm/s, and the grinding depth is in the range of 0.01mm∼0.04mm single side grinding depth, the grinding temperature of the workpiece surface is below 200°C, and the oxidation temperature of the workpiece is not reached, therefore, the dry grinding method is considered suitable for the machining of MOX core blocks.","PeriodicalId":317622,"journal":{"name":"Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components","volume":"51 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"122700267","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}