核动力工程

Long-term Operation Practice of Qinshan Nuclear Power Plant

Kong Deping

2022, 43(S1): 1-6. doi: 10.13832/j.jnpe.2022.S1.0001

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Abstract:
Qinshan Nuclear Power Plant is the first nuclear power unit in Chinese Mainland facing the operation license extension (OLE), and the first nuclear power unit in Chinese Mainland approved to extend its operation for 20 years. This paper introduces the operation license technical system and corresponding process of Qinshan Nuclear Power Plant. The successful experience of OLE in Qinshan Nuclear Power Plant can be used for reference and demonstration for the long-term operation of subsequent nuclear power plants.

Research and Application of Technical Route for Operation License Extension in Qinshan Nuclear Power Plant

Tao Jun, Shi Wenxiang, Zhang Jiangtao, Jiang He

2022, 43(S1): 7-10. doi: 10.13832/j.jnpe.2022.S1.0007

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By investigating and studying the main life extension technology routes of nuclear power plants in the world, the technical requirements suitable for life extension of nuclear power plants in China are formulated, and the main contents of safety assessment are determined. Using this technical route, the scope screening, object screening, aging assessment, supplement of final safety analysis report, environmental impact assessment and engineering transformation of safety assessment have been carried out in Qinshan NPP. The practical application shows that this technical route is feasible and meets China’s nuclear safety regulatory requirements. The Operation License Extension (OLE) Project of Qinshan NPP finally passed the review of the regulatory authority and obtained the extended operation license.

Fatigue Analysis and Countermeasure of Pressurizer Surge Line

Zhao Chuanli, Shi Shaobo, Luan Xingfeng, Chen Xueyao, Xu Feng

2022, 43(S1): 11-15. doi: 10.13832/j.jnpe.2022.S1.0011

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In order to establish the analysis method and process of thermal stratification and the impact of operating environment on the fatigue of pressurizer surge line (hereinafter referred to as “surge line”), the fatigue state of surge line is analyzed and predicted through numerical simulation, thermal stratification load analysis and environmental impact coefficient analysis and calculation. The research results show that due to the influence of thermal stratification and operating environment, the cumulative fatigue utilization factor (CUF) of the surge line at the end of its life is close to 1.0, so it is necessary to install an online fatigue monitoring system to monitor the fatigue of the surge line in real time to avoid sudden breakage.

Aging Evaluation of Baffle Bolts for Reactor Internals of Qinshan Nuclear Power Plant

Huang Chao, Xu Feng, Li Shiwei

2022, 43(S1): 16-21. doi: 10.13832/j.jnpe.2022.S1.0016

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In order to evaluate the actual aging state of baffle bolts for reactor internals in Qinshan Nuclear Power Plant, on the basis of absorbing and summarizing the international research results on the aging mechanism of reactor internals, a criterion for judging the aging mechanism of reactor internals is established, and is used to evaluate and identify the aging mechanism of baffle bolts mainly affected by wear, stress relaxation, irradiation swelling, irradiation-promoted stress corrosion cracking and so on. Besides, according to the possible defect types caused by the aging mechanism, the underwater ultrasonic inspection technology is developed to supplement the conventional visual inspection method, so as to formulate the inspection scheme to evaluate the aging state. The evaluation results show that the baffle bolts for reactor internals in Qinshan Nuclear Power Plant are in good aging condition, and no deformation, crack and other aging failure phenomena have occurred. Practice has proved that the evaluation method is effective and can be used for the aging evaluation of PWR nuclear power plant components.

Analysis and Re-evaluation Technology of Qualified Life of Nuclear Safety Grade Cable

Tao Ge, Gao Xuan, Ma Huiming, Zhang Yizhou, Zhao Chuanli, Tao Jun, Kong Deping

2022, 43(S1): 22-26. doi: 10.13832/j.jnpe.2022.S1.0022

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Based on the analysis and study of the qualification method and qualified life of safety grade electrical equipment (1E equipment), the analysis and re-evaluation technology is applied to the life assessment of the cable (EQ cable) with environmental qualification requirements in nuclear power plant, which can reasonably extend the qualified life of EQ cable on the premise of meeting the safety requirements, and ensure that the cable can still achieve its expected function according to the requirements of the specification during the extended period of qualified life. This method can be used not only to evaluate the qualification status of cables during normal operation of power plants, but also to evaluate the ability of cables to achieve their expected functions during the Operation License Extension (OLE) of nuclear power plants. This method has been applied in domestic nuclear power plants.

Study on Creep Model of Pre-stressed Concrete Based on Component Scale

Zhang Jiangtao, Du Yang, Cai Dahua, Tao Jun, Zhao Chuanli, Shi Wenxiang

2022, 43(S1): 27-34. doi: 10.13832/j.jnpe.2022.S1.0027

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To assess long-term pre-stress loss of nuclear power plant containment, the creep prediction model of pre-stressed concrete is established by combining experimental investigation with theoretical analysis by taking pre-stressed concrete beam as the research object. Based on the existing creep test of pre-stressed concrete beams, the shrinkage test of pre-stressed concrete beams under the same environment is carried out with the same concrete materials to measure the actual shrinkage deformation of pre-stressed concrete beams. Considering the coupling effect of concrete shrinkage, creep and pre-stressed tendon relaxation, the age adjusted effective modulus method is introduced to establish the calculation method of concrete creep coefficient derived from the test data. Finally, the creep model of pre-stressed concrete is established and its long-term creep deformation is predicted, which provides a theoretical support for the long-term pre-stress loss assessment of nuclear power plant containment structure.

Practice and Thinking of Environmental Impact Assessment of Qinshan Nuclear Power Plant Operation License Extension Application

Li Zhihua, Kang Yunding, Jiang He, Tao Jun, Qiu Zhijing, Huang Xiaodong, Zhu Kun, Cao Guochang, Shi Wenxiang

2022, 43(S1): 35-39. doi: 10.13832/j.jnpe.2022.S1.0035

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In view of the work related to the Operation License Extension (OLE) for the extended operation of Qinshan Nuclear Power Plant after the expiry of its operation license, the American license renewal (LR) and the long-term operation (LTO) system of the International Atomic Energy Agency (IAEA) are studied and analyzed. Based on the Technical Policy for Extension of Validity Period of Nuclear Power Plant Operation License (Trial) issued by the National Nuclear Safety Administration, the appropriateness, acceptability and compliance of environmental impact assessment of nuclear power plant extended operation are analyzed and studied, and the working methods and processes of OLE environmental impact assessment of nuclear power plants in China are put forward; Combined with the engineering practice experience of Qinshan NPP, the evaluation basis, evaluation criteria, data collection scope and source term optimization issues in the evaluation process are studied, and suggestions for the extended operation of nuclear power plants in China are put forward.

Application of Transient Supervision and Processing Method in Nuclear Power Plant Operation License Extension

Jiang He, Shi Wenxiang, Cao Guochang, Li Zhihua, Xu Feng, Tao Jun, Shi Shaobo

2022, 43(S1): 40-43. doi: 10.13832/j.jnpe.2022.S1.0040

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Before the extended operation of the nuclear power plant, due to the lack of specific understanding of the impact of transient supervision and management on the extended life of the nuclear power plant, the relevant transient supervision and control are only within the design life and do not involve the extended operation. This leads to the failure of the nuclear power plant to manage the consumption of bottleneck transients in the early stage of operation, thus reducing the actual life length of the nuclear power plant; Or the relevant transient data collection is not detailed enough to support more detailed fatigue analysis. In the extended operation evaluation, only more envelope processing can be adopted, which is difficult to achieve a longer evaluation life. In view of the above problems, by learning from the transient related experience in the extended operation research of Qinshan Nuclear Power Plant, this paper studies the transient supervision and data processing of the extended operation of nuclear power plant from the two aspects of daily operation supervision and special continuous evaluation, and forms a transient management technical method suitable for the extended operation of nuclear power plant, which can effectively guide the extended operation of subsequent nuclear power units.

Main Control Room Renewal Technology Research & Planning of Qinshan Nuclear Power Plant

Pu Xiaobin, Ma Jun, Zhou Jianhua, Chen Ziming, Lu Peifang, Xie Rui

2022, 43(S1): 44-50. doi: 10.13832/j.jnpe.2022.S1.0044

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Based on the requirements of Operation License Extension (OLE) and ensuring that the design and equipment of the main control room and emergency control room (referred to as the main control room and the emergency control room) comply with the relevant national nuclear safety regulations and standards, based on the investigation and analysis of the status quo of panels and consoles and man-machine interfaces of the main control room and the emergency control room in Qinshan Nuclear Power Plant, the structures of the panels and consoles, the improvement of equipment and the redesign of the man-machine interface are studied. The solution of the overall renewal of the main control room and the emergency control room is put forward and the feasibility demonstration is carried out. Methods for updating the overall design scheme, validating the process and implementing the controls are planned. The engineering practice and operation performance show that this research method is effective and can provide guidance for OLE research of nuclear power units.

Irradiation Embrittlement Time-limited Aging Analysis of Reactor Pressure Vessel for Qinshan Nuclear Power Plant

Luan Xingfeng, Zhao Chuanli, Xu Feng, Tao Hongxin, Zhang Jiangtao, Gao Xuan, Tao Ge

2022, 43(S1): 51-54. doi: 10.13832/j.jnpe.2022.S1.0051

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Reactor pressure vessel is one of the most important equipment in nuclear power plant, and its irradiation embrittlement state determines the actual operation life of nuclear power plant. A method of irradiation embrittlement time-limited aging analysis (TLAA) for reactor pressure vessel is developed by referring to foreign reactor pressure vessel safety assessment methods. This method evaluates the safety margin of pressure vessel materials under normal and accident conditions from three aspects: upper platform energy, reactor operating pressure-temperature curve and pressure bearing thermal shock. Using this method, the irradiation embrittlement TLAA safety evaluation of the reactor pressure vessel is carried out in the Qinshan Nuclear Power Plant Operation License Extension (OLE) Project. Its evaluation methods and evaluation conclusions have been recognized by the National Nuclear Safety Administration, laying the foundation for the 20-year life extension of Qinshan NPP.

Construction and Analysis of Irradiation Damage Prediction Model for Autonomous Reactor Pressure Vessel

Guo Yanhui, Sun Zaozhan, Sun Haitao, Xu Chaoliang, Liu Xiangbing, Tao Jun

2022, 43(S1): 55-59. doi: 10.13832/j.jnpe.2022.S1.0055

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The problem of irradiation embrittlement of reactor pressure vessel (RPV) is the key to restrict the safe service of RPV, and the building of irradiation damage prediction model is an effective method to predict irradiation embrittlement damage. In this paper, the building mechanisms and methods of typical irradiation damage models such as parameterized prediction model, structured prediction model and artificial neural network prediction model are studied, and the advantages and disadvantages of different prediction models are compared. The results show that the building technology of prediction model based on irradiation mechanism can better reflect the physical mechanism of irradiation embrittlement. Based on this, the construction technology route of RPV independent prediction model is proposed.

Study on Aging Management Review of Electric and I&C Equipment in Evaluation of Nuclear Power Plant Operation License Extension

Kong Jing, Zhang Qi, Chen Zixi, Gao Xuan

2022, 43(S1): 60-64. doi: 10.13832/j.jnpe.2022.S1.0060

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By studying the Technical Policy for Extension of Validity Period of Nuclear Power Plant Operation License (Trial) , referring to the experience of foreign nuclear power plants in life extension activities, and combining with the actual situation of domestic nuclear power plant operation license extension (OLE) evaluation activities, this paper introduces the implementation processes and methods of aging management screening, aging effect identification and evaluation, and management activity review of electric and I&C equipment in OLE of nuclear power plants; The application strategies of “item group”, “hypothetical fault” and “regional space method” are given in the form of examples. The application of the strategy optimizes the screening process and improves the efficiency of aging management review (AMR) activities, which can be used as a reference for nuclear power plant OLE.

Experience Analysis on Safety Review on Operation License Extension of Qinshan Nuclear Power Plant

Sun Haitao, Lyu Yunhe, Gao Chen, Ma Ruoqun, Mao Yuxian, Fang Yonggang, Chai Guohan, Yang Di, Sun Zaozhan

2022, 43(S1): 65-69. doi: 10.13832/j.jnpe.2022.S1.0065

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In order to carry out the nuclear safety review on Operation License Extension application of Qinshan Nuclear Power Plant, document review and site survey are adopted. According to the Technical Policy for Extension of Validity Period of Nuclear Power Plant Operation License (Trial) , and by referring to technical documents such as regulations and standards of American license renewal (LR), in-depth research has been carried out on the screening of aging management review (AMR), AMR results, time-limited aging analysis (TLAA), supplement to safety analysis report, aging management program (AMP), etc., and corresponding review technical opinions and review experience have been formed, which provides important support for the administrative reply of OLE application of QINSHAN Nuclear Power Plant and important reference for OLE application and safety review of subsequent nuclear power plants.

Research and Application of Modification Scheme for Supports and Hangers of Special Pipes for OLE Project of Qinshan Nuclear Power Plant

Feng Zhe, Fang Chengping, Wu Tong, Shao Zhen, Yu Quanzhou, Cheng Xiaowen, Zhang Yunhua

2022, 43(S1): 70-74. doi: 10.13832/j.jnpe.2022.S1.0070

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“Modification of nuclear level 2 and 3 pipe supports and hangers” is one of the modifications of Qinshan Nuclear Power Pant operation license extension (OLE) project. In the project, there are two kinds of special supports and hangers which are common but difficult to modify: one is the rigid bracing piece support not perpendicular to the pipe axis, the other is the guide support using the “guide support base plate”. In order to meet the modification design requirements and relevant laws and regulations, two new standard support and hanger components are designed, standardized and serialized, which can meet the modification requirements with the existing standard components. The two modification design schemes solve the problem of project modification, and have the advantages of convenient modification, strong universality and easy promotion.

Study on Reactor-Mechanical-Electricity for Capacity Expansion and Retrofitting of 320 MWe Steam Turbine Generator Unit in Qinshan Nuclear Power Plant

Li Rupeng, Ye Cheng, Qi Lian, Du Liguo, Huang Jiayun

2022, 43(S1): 75-79. doi: 10.13832/j.jnpe.2022.S1.0075

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On the basis of preliminary analysis on the conditions, traditional methods and characteristics of the reactor-mechanical-electricity matching of 320 MWe unit in Qinshan Nuclear Power Plant, with the application of the current advanced and efficient thermal hydraulic calculation, pipe resistance calculation and heat balance calculation software, taking the required power of the unit capacity expansion and retrofitting as the goal, on the premise of ensuring that the nuclear island reactor power does not exceed the enhanced working condition power, for different pressures and flows of main steam at the outlet of steam generator (SG), according to a certain step length, and considering different feed water temperature, reasonable main steam pipe resistance, optimized cold end parameters, combined with the current advanced design and manufacturing technology of steam turbine generator unit, the thermal hydraulic calculation and thermal balance iterative calculation have been carried out to form the optimized inlet main steam parameters of steam turbine and generating power of unit. While meeting the power target, the efficiency of the unit has also been improved. After the unit capacity expansion and retrofitting, the parameters and performance of the reactor-mechanical-electricity equipment can be better matched, and the operation of the unit is more safe, stable and economical.

Study on Earthquake Interaction between Safety-Related and Non-Safety- Related Items in Equipment Cooling Water System Room of Qinshan Nuclear Power Plant

Tan Yong, Zhang Liang, Peng Jun, Zhang Wangchao, Zhou Wentao, Qu Ting, Xu Gang, Wang Yue

2022, 43(S1): 80-85. doi: 10.13832/j.jnpe.2022.S1.0080

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In order to solve the problems of mixed arrangement of safety-related and non-safety-related items in the equipment cooling water system room of Qinshan Nuclear Power Plant and possible interaction under seismic conditions, by using the method of test and analysis, the safety-related valves and pipes arranged in the equipment cooling water system room are studied and analyzed under the influence of other non-safety-related ducts and pipes under seismic conditions. In the test, the size and height of the items in the equipment cooling water system room of Qinshan Nuclear Power Plant are reproduced 1:1, and the tests are carried out through various interactive methods such as flat fall, single pendulum, and dumping. Taking the pressure retaining performance of pipes and valves as the failure criterion, the deformation and impact acceleration are measured and analyzed. In order to verify the reliability of the test, the numerical simulation method is used to analyze and compare the failure of the duct-pipe interaction conditions, and the results are consistent with the test method. The final research results show that the non-safety-related items such as the duct arranged at 3.6 m above the top of the equipment cooling water system room and the DN50 pipe arranged at 8 m above the top of the equipment cooling water system room in Qinshan Nuclear Power Plant will not damage the function and structural integrity of the safety-related equipment cooling water system header and valve below.

Determination of Scope and Object of Safety Assessment for Operation License Extension of Nuclear Power Plant

Jiang He, Li Zhihua, Shi Wenxiang, Cao Guochang, Zhang Feng

2022, 43(S1): 86-89. doi: 10.13832/j.jnpe.2022.S1.0086

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In order to ensure the integrity and effectiveness of nuclear power plant extended operation safety assessment, through the analysis of the characteristics of extended operation, the key issues that need to be paid attention to in safety assessment are analyzed, and an effective method for determining the scope and object screening is established. Through this method, Qinshan Nuclear Power Plant can identify the scope and object of extended operation safety assessment, so that the follow-up assessment can be carried out effectively. Practice has proved that the results determined according to this can meet the screening requirements of nuclear safety supervision for extended operation safety assessment objects. It shows that the determination principle of the scope and object of extended operation safety assessment of Qinshan Nuclear Power Plant is reasonable, and its method and process can effectively guide the extended operation practice of subsequent nuclear power plants.

Aging Review of Hydraulic Structures in Nuclear Power Plant

Zhang Jiangtao, Wang Zhen, Chen Sen, Cai Dahua, Shi Wenxiang

2022, 43(S1): 90-93. doi: 10.13832/j.jnpe.2022.S1.0090

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Hydraulic structures of the nuclear power plant are important structures that provide water-flooding barrier and cooling water for nuclear safety related systems, equipment and components, and have been included in the scope of assessment and review during the Operation License Extension (OLE) of Qinshan Nuclear Power Plant. Aiming at the hydraulic structures of nuclear power plants, this OLE project determines the management scope and aging effect of hydraulic structures of nuclear power plants, carries out aging management review (AMR) and underwater inspection activities, and grasps the actual service status of hydraulic structures, which proves that hydraulic structures can continue to perform their expected functions during the extended operation of nuclear power plants.

Research on High-Temperature Resistant Strain Sensor Based on Fiber-Optic Fabry-Perot Structure

Liu Linlin, Xu Yugen, Zhang Xiaoling, Yang Heng, Zhu Wanxia, Li Ronghui, Zhao Xin, Sun Lei, Li Pengzhou

2022, 43(S1): 94-98. doi: 10.13832/j.jnpe.2022.S1.0094

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In order to develop a micro strain sensor suitable for mechanical measurement in the high temperature environment of the reactor first loop , based on the principle of optics, demodulation principle and packaging technology of the fiber-optic Fabry-Perot (FP) structure, the thermal-mechanical coupling model of the sensor is proposed, a high temperature resistant strain sensor based on the fiber-optic FP structure and its demodulating system are designed, and its dynamic test is carried out in the normal temperature air, high temperature and medium pressure water environment. The research and test results show that the high temperature strain resistant fiber optic strain sensor and its demodulating system can operate stably in high temperature, medium pressure and water environment, and the measuring range of the strain sensor is 0~4000 με, the accuracy is 0.125%FS, and the speed of the demodulating system is 5 kHz. The high temperature resistant strain fiber optic strain sensor and its demodulating system designed in this study can be used to measure the strain in the high temperature environment of the reactor first loop, thus providing an effective monitoring method for the monitoring in the high temperature environment of the reactor reactor first loop.

Application Research of Similarity Theory in Impact Test

Yang Heng, Shen Shuangquan, Guo Cong, Sun Lei, Zhu Wanxia, Li Pengzhou

2022, 43(S1): 99-102. doi: 10.13832/j.jnpe.2022.S1.0099

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In order to verify the feasibility of the similarity theory in the impact test of large-scale equipment, the key response parameters of the proportional similarity model under the specified impact load are obtained by means of the simulation calculation and analysis of the proportional similarity model and the impact test. The results show that when the impact test is carried out with parameters such as the main pulse peak acceleration and pulse width formulated according to the similarity theory, the strain test results of the same position of each model conform to the similarity theory law. This study verifies the correctness and feasibility of the similarity theory in the impact test, and can be used in the subsequent structural impact test of large equipment.

Study on Fundamental Mechanical Problems in Fluidelastic Instability of Steam Generator Heat Transfer Tube Bundles

Yang Shihao, Lai Jiang, Tan Tiancai, Sun Lei

2022, 43(S1): 103-110. doi: 10.13832/j.jnpe.2022.S1.0103

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In order to explore the mechanism of the fluidelastic instability of the tube bundle and the influence mechanism of the support mode and the internal flow load on the fluidelastic instability of the tube bundle, a theoretical model of the flow induced vibration of the heat transfer tube under complex fluid excitation is established by comprehensively considering the effects of the steady flow elasticity, internal flow excitation and unsteady flow force on the flow induced vibration of the heat transfer tube. Based on the eigenvalue stability theory, the fluidelastic instability mechanism of heat transfer tube under the action of two-phase cross-flow is obtained, and the effects of internal flow excitation and unsteady flow force on the fluidelastic instability mechanism of heat transfer tube are analyzed systematically. The research shows that the support mode will affect the critical velocity of instability, but will not affect the fluidelastic instability mechanism; the internal flow in the tube will couple the modes of the tube bundle, and the high-speed internal flow will change the instability mechanism of the tube bundle; As a kind of forced force, the unsteady flow force may cause the "beating vibration" of the tube bundle before the fluidelastic instability, which shall be avoided in the engineering design.

Numerical Investigation on Flow Induced Vibration of Distributed Circular Tube Bundle

Zhang Yu, He Chao, Sun Lei

2022, 43(S1): 111-115. doi: 10.13832/j.jnpe.2022.S1.0111

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A numerical simulation study based on Computational Fluid Dynamics (CFD) / Computational Solid Dynamics (CSD) coupling method was carried out to investigate the vibration behavior of distributed circular tube bundle under cross flow excitation. In the study, the unsteady lift-drag acting on the tube bundle is obtained by solving the unsteady Reynolds-Averaged NS (URANS) equation, the tube bundle vibration equation is solved by fourth-order Runge-Kutta scheme discretization, a mesh updating strategy based on spring smoothing is adopted to ensure the orthogonality of flow field grid in the process of tube bundle vibration, and the reliability of the numerical method is verified by the experimental and calculation results of the flow around a single tube. Through the above methods, the motion trajectory, flow force and vibration time-frequency characteristics of the central tube are analyzed in detail. The results show that under the fluid excitation, the vertical and transverse vibration frequency of the central tube is consistent with the fluid excitation frequency, showing a typical forced vibration behavior.

Research on the Support of Large Bearing Low-frequency Hard Vibration Isolation Equipment Based on Tilted Support Structure

Liu Tianyan, Han Chao, Shao Xiaolin, Li Pengzhou, Zhang Kun, Sun Yue, Guo Cong

2022, 43(S1): 116-120. doi: 10.13832/j.jnpe.2022.S1.0116

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According to the requirements of large load-bearing and low-frequency vibration isolation of nuclear power equipment and the special working environment in the nuclear island, a metal hard vibration isolation support structure with large load-bearing capacity is designed by using the geometric nonlinearity of typical tilted support structure. The deformation of the structure under rated load is obtained by numerical calculation, and the strength of the structure is verified. Based on the dynamic theory, a dynamic model is established, the vibration isolation effect is analyzed through the vibration level difference, the structural parameters are optimized, and the structural prototype is processed, and the vibration isolation effect of the structure is verified through the vibration reduction test of the real machine. The results show that the structure has good vibration isolation effect on the premise of ensuring the bearing capacity, which can provide a reference for the design of related vibration isolation structures.

Research on Nonlinearity Weakening Method of Quasi-zero-stiffness Vibration Isolator

Han Chao, Liu Guixiang, Shao Xiaolin, Liu Tianyan, Xu Deshui, Zhang Kun, Liu Xueguang

2022, 43(S1): 121-126. doi: 10.13832/j.jnpe.2022.S1.0121

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In order to weaken the nonlinear characteristics of the quasi-zero-stiffness vibration isolator, a nonlinearity weakening method of quasi-zero-stiffness vibration isolator using softening negative stiffness to neutralize hardening negative stiffness is proposed. By adjusting the design parameters, the nonlinear terms of the two negative stiffness cancel each other out, and the system only retains the linear stiffness characteristics. The feasibility of this method is verified by the cases of permanent magnet negative stiffness and three-spring negative stiffness. The results show that the nonlinear stiffness of the quasi-zero-stiffness vibration isolation system designed by this method is greatly weakened and the vibration isolation performance of the system is enhanced.

Mechanical Analysis of Pressure Vessels and Main Pipes under Steam Explosion Loads

Tang Peng, Yao Di, Yu Li, Luo Juan, Zhou Ding

2022, 43(S1): 127-131. doi: 10.13832/j.jnpe.2022.S1.0127

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The mechanical response of HPR1000 reactor pressure vessel (RPV) and its main pipe under the hypothetical steam explosion load is studied. Through the establishment of the finite element model and the numerical analysis according to the transient structure analysis method, the deformation, stress and strain results of RPV and the main pipeline are obtained. The results show that the failure loads of RPV at 600℃, 800℃ and 1000℃ are 1/20, 1/50 and 1/100 design loads respectively; The maximum equivalent stress/strain is located near the nozzle; The stress in most areas of the main pipe does not exceed the yield stress of the pipe. This study can provide technical support for structural integrity analysis of RPV under extreme load.

Study on Flow Induced Vibration Characteristics of Laminated Plate Structure under High Temperature

Jiang Tianze, Li Pengzhou, Ma Jianzhong, Gao Lixia, Zhang Xiaoling

2022, 43(S1): 132-136. doi: 10.13832/j.jnpe.2022.S1.0132

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In order to study the effect of high temperature on the flow induced vibration characteristics of laminated plate structure, a dynamic equation is established for the simplified laminated plate structure considering the action of viscous incompressible fluid. By solving the characteristic matrix, the sensitivity effects of three temperature-affected parameters (elastic modulus of laminated material, cooling water density and saturated water viscosity) on the natural frequency and critical velocity of instability of laminated plate are analyzed. Furthermore, the effect of temperature on the flow induced vibration characteristics of laminated plate structure is studied when the three parameters change at the same time. The results show that the natural frequencies of laminated plate in static and dynamic water decrease with the increase of temperature; the decrease of the elastic modulus of the laminate material leads to the decrease of the rigidity of the laminated plate, which has the greatest effect on the natural frequency of the laminated plate. At the same time, the critical velocity of instability decreases at first and then increases gradually, but the change range is not significant; the additional mass reduction caused by the reduction of cooling water density has the greatest effect on the critical velocity of instability.

Analysis and Research on Influence of Plate-Shell Structure Thickness on Impedance Characteristics

Li Xingzhao, Lu Jun, Li Pengzhou

2022, 43(S1): 137-141. doi: 10.13832/j.jnpe.2022.S1.0137

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The vibration performance of mechanical equipment is closely related to the impedance characteristics of the installation infrastructure, and the thickness of the structural plate is one of the important factors affecting the impedance characteristics of the structure. In this paper, for the typical plate and shell installation infrastructure of marine mechanical equipment, by studying the influence law of the thickness of the structural plate on the impedance characteristics of the structure, it is found that only the thickness of the bearing surface and bottom surface of the structure has a significant impact on the impedance characteristics of the structure, but the thickness of the supporting surface does not. Therefore, the basic method for the simulation design of the impedance characteristics of the structure is obtained. According to this method, the impedance characteristic simulator of the structure is designed, and the impedance characteristic test is carried out. The test results show that the impedance characteristic of the structural simulator is close to that of the actual structure, which verifies the effectiveness of the design method.

Experimental Investigation on Fatigue Crack Propagation Behavior of Nuclear Pipe at High Temperature

Luo Juan, Qi Min, Tang Peng, Tang Long, Yao Di

2022, 43(S1): 142-145. doi: 10.13832/j.jnpe.2022.S1.0142

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In order to study the fatigue crack growth performance of nuclear piping materials at elevated temperature above 500℃, the fatigue crack propagation rate experiments on the materials of base metal, welded seam and heat affected zone of tube were carried out under high temperature conditions. The probabilistic fatigue crack growth curves with different survival rates were obtained according to the method of probability analysis. The results show that there are obvious differences in fatigue crack growth performance of materials in different positions of pipelines at high temperature, and the resistance of fatigue crack propagation for welded seam and heat-affected zone is much better than that of base metal. The results can be used for safety assessment and fracture mechanics analysis of nuclear reactor pipeline structure.

Research on Application of Digital Image Correlation Method in Mechanical Property Test

Yu Li, Luo Jiacheng, Yao Di, Luo Juan

2022, 43(S1): 146-151. doi: 10.13832/j.jnpe.2022.S1.0146

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In order to realize the application of the digital image correlation method in the structural mechanics of the reactor, the non-contact full-field strain measurement technology is used to study the surface speckle treatment of large-scale thin and flexible structures, and to carry out mechanical property tests; through the non-contact full-field strain measurement system-VIC-3D test system, the full-field strain data of the test piece are obtained, and the data are verified with the measurement results of the strain gauge. The results show that the digital image correlation method (DIC) has good strain testing accuracy and wide testing range.

Analysis and Experimental Study on Control Rod Dropping under Seismic Conditions

Zhang Dan, Huang Wenhui, Su Jiexi, Wei Yongtao, Du Jianyong, Sun Lei, Li Pengzhou

2022, 43(S1): 152-156. doi: 10.13832/j.jnpe.2022.S1.0152

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Abstract:
In order to study control rod dropping history under seismic conditions, based on the inertia resistance calculation method, viscous resistance calculation method and collision contact algorithm, the rod dropping history under seismic loads is calculated under the background of a reactor control rod drive line of the third generation nuclear power advanced R &D project. The seismic test of the 1:1 prototype of the reactor drive line is carried out on the multi-point excitation tester. The multi-frequency wave method is used to apply seismic loads at several related points between the drive line and the reactor, including operating basis earthquake (OBE) test and safe shutdown earthquake (SSE) test. The results show that the calculation curves of rod dropping time, rod displacement, velocity and acceleration are in good agreement with the test curves. Therefore, the research methods established in this study can provide necessary analysis methods for the safety analysis of nuclear power plants.

Structural Vibration Analysis of Large Turbogenerator

Bao Yu, He Chao, Zhu Jianbin, Xu Weizu

2022, 43(S1): 157-162. doi: 10.13832/j.jnpe.2022.S1.0157

Abstract(62) HTML (39) PDF(11)

Abstract:
Aiming at the problem that the structural vibration of a large turbogenerator (hereinafter referred to as “generator”) exceeds the standard, the three-dimensional modeling simulation and shutdown modal test of the generator are carried out. The results show that the reason of structural vibration may be the structural resonance caused by the natural frequency of the generator close to the rotor working frequency (50 Hz). Based on the theory and practice of structural resonance, three solutions are proposed to solve the resonance, including increasing the support stiffness, adding the damping vibration absorber and bottom load distribution frequency modulation. Finally, by adding the damping vibration absorber on the generator hydrogen cooler pipe, the bushing vibration at both ends of the generator steam and excitation is greatly reduced, which can ensure that the generator can safely operate for a long time, and solve the long-standing important defect of high generator vibration.

2022 Vol. 43, No. S1