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2022 Vol. 43, No. 4

Special Contribution
Research on the Development Trend of Micro Nuclear Reactor Technology
Du Shuhong, Li Yonghua, Sun Tao, Wang Jun, Liu Xiaowen, Su Gang, Zhao Depeng
2022, 43(4): 1-4. doi: 10.13832/j.jnpe.2022.04.0001
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Micro nuclear reactors adopt Generation-IV non-light water reactors, heat pipe reactors and Generation-III light water reactors with high inherent safety, providing long-term and highly reliable power supply for innovative scenario such as remote islands, mining areas, border guard posts and bases, emergency and disaster relief, space exploration and deep-sea exploration. They have broad application prospects, being one of the important technical supports to realize the national strategy. This study summarizes the definition and main R & D reactor types of micro nuclear reactors, and describes the innovative technological characteristics of micro nuclear reactors, such as high inherent safety, easy modularization and expansion, transportability, easy deployment, independent operation and so on, analyzes the development trend of key technologies such as new fuel, integration of main loop, new thermoelectric conversion device, passive safety system, intelligent operation and maintenance and coupling of nuclear energy and other energy sources in China, providing support for the formulation of the technical route for the development of micro nuclear reactors in China.
Reactor Core Physics and Thermohydraulics
Numerical Simulation of Critical Flow in a Narrow Rectangular Channel
Yin Ling, Sun Yunda, Gong Shengjie
2022, 43(4): 5-10. doi: 10.13832/j.jnpe.2022.04.0005
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The main mechanism of coolant leakage through micro-cracks is critical flow, and accurate prediction of critical flow is the key to realize the application of Leak-Before-Break (LBB). Based on the cavitation model, the critical flow in a narrow rectangular channel is numerically simulated, and the influence of the constant term (C) representing the bubble radius and nucleation site density on the critical flow is discussed. The results show that when C=1.25, the simulation results are in good agreement with the experimental values; Within the range of experimental conditions, the error between the simulation results of critical flow and the experimental values is within ± 15%; The modified cavitation model can be used to simulate and calculate the critical flow under upstream temperature and pressure conditions.
Comparative Analysis of Steam Condensation Consisting of Air Outside of Single Tubes and Tube Bundles with Different Tube Diameters and Inclination Angles
Zou Zhiqiang, Wu Lingjun, Li Fangli, Bian Haozhi, Cao Boyang, Ding Ming
2022, 43(4): 11-17. doi: 10.13832/j.jnpe.2022.04.0011
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In order to evaluate the difference of condensation heat transfer law between single tube and steam containing air outside the tube bundle under different heat transfer tube structural parameters, based on a single tube with an outer diameter of 12~19 mm, an inclination angle of 0°~90° and a 3 × 3 tube bundle, an experimental study was carried out in the range of pressure 0.2~1.6 MPa and air mass share 12%~87%. The results show that the influence of tube diameter and inclination angle on single tube and tube bundle shows different rules in different pressure ranges. When the pressure is less than 0.8 MPa, the condensation heat transfer of the tube bundle is affected by the tube diameter and inclination angle, which is the same as that of a single tube. The condensation heat transfer coefficients of both increase with the decrease of tube diameter and inclination angle. At 0.8~1.6 MPa, the condensation heat transfer of tube bundle is significantly different from that of single tube due to the influence of tube diameter and inclination angle. The law of consistency and difference is analyzed according to the mechanism of the effect of non-condensable gas on steam condensation heat transfer.
Analysis and Research on the Influence of Asymmetric Pumping Chamber on the Performance of CAP1400 Nuclear Coolant Pump
Yan Yongqi, Lu Yeming, Liu Haoran, Wang Xiaofang, Zhang Zhigang, Sha Long
2022, 43(4): 18-24. doi: 10.13832/j.jnpe.2022.04.0018
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As the boundary of the nuclear coolant pump, the pumping chamber not only bears the pressure, but also is the only bridge between the guide vane flowing in the circumferential direction and the unidirectional flow pipeline. In order to explore the influence of pumping chamber on the performance of the whole machine, a design method for asymmetric pumping chamber of nuclear coolant pump is proposed based on CAP1400’s 1:2.5 scale model, and four kinds of asymmetric models with different geometric dimensions are designed. With the computational fluid dynamics (CFD) numerical method, the internal flow field, external characteristics and transient load information of the full three-dimensional model of the nuclear coolant pump with orifice ring clearance are obtained. Through comparative analysis, the following conclusions are obtained: four asymmetric pumping chamber models reduce the radial load at the upper cover plate by more than 60%, and reduce the main frequency amplitude of the impeller and the total radial load by more than 13%; While ensuring the obvious improvement of radial load, it can also effectively improve the pump efficiency and head. The former is more obvious, with a lifting range of 0.57% to 0.83%.
Model Development and Transient Analysis of Thermal Stratification Phenomenon in Pool-Type Sodium-Cooled Fast Reactors
Du Peng, Shan Jianqiang, Deng Jian, Liu Yu, Ding Shuhua, Chen Wei, Yuan Peng, Wu Zenghui
2022, 43(4): 25-30. doi: 10.13832/j.jnpe.2022.04.0025
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In this paper, a three-dimensional (3D) system analysis model is established according to the characteristics of the pool-type sodium-cooled fast reactor, and combined with the evolution mechanism of thermal stratification, the key processing method of accurately simulating thermal stratification is proposed, including energy source term processing, 3D momentum equation convection term processing and 3D space inlet effect processing. On this basis, the developed 3D system analysis model is verified by KALIMER and MONJU thermal stratification experiments. The results show that the model effectively solves the problem of 3D thermal hydraulic analysis of pool-type sodium-cooled fast reactor, realizes the rapid and accurate simulation of the transient change of temperature field in the sodium pool and the evolution process of thermal stratification, and can also determine the maximum position of thermal stress on the surface of pool type structure in the process of thermal stratification, which can provide a reference for the safety design of pool-type fast reactor.
Study on Neutronic/Thermal-Mechanical Coupling Calculation Method for Fast-neutron Pulse Reactor with Metallic Nuclear Fuel
Guo Shuwei, Chen ZhenPing, Jiang Xinbiao, Li Da, Zhang Xinyi, Wang Lipeng, Hu Tianliang, Xie Jinsen, Yu Tao
2022, 43(4): 31-37. doi: 10.13832/j.jnpe.2022.04.0031
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In order to ensure the operational safety of the fast-neutron pulse reactor and prevent the supercritical pulse from causing physical damage to the material, it is necessary to simulate and analyze the pulse operating conditions of the fast-neutron pulse reactor. In this study, for the fast-neutron pulse reactor with metallic nuclear fuel, the neutronic/thermal-mechanical coupling calculation and analysis of Godiva-I pulse reactor are carried out based on the point reactor dynamics method, Monte Carlo method and finite element mechanics method. The calculation results show that the reactivity temperature coefficient and fission rate are in good agreement with the experimental values, and the reactivity, temperature rise, surface displacement and surface stress are consistent with the actual situation. Therefore, the "neutronic/thermal-mechanical" coupling calculation method established in this paper can be applied to the analysis and calculation of the fast-neutron pulse reactor with metallic nuclear fuel, and has certain reliability.
Nuclear Fuel and Reactor Structural Materials
Study on Decarburization and Oxidation Corrosion Behavior of T-22 Alloy in Impure Helium of High-temperature Gas-cooled Reactor
Li Haoxiang, Zheng Wei, Yin Huaqiang, Du Bin, Wang Qiuhao, He Xuedong, Ma Tao, Yang Xingtuan
2022, 43(4): 38-45. doi: 10.13832/j.jnpe.2022.04.0038
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Helium in the main coolant of high-temperature gas-cooled reactor (HTGR) contains trace impurities, which reacts with the alloy materials of the equipment at high temperature and cause corrosion of the materials. The corrosion test of T-22 alloy, an alternative material for high-temperature gas-cooled reactor steam generator, is carried out in four kinds of impure helium at 950 ℃, and the corrosion time is 50 h. Then the corroded T-22 alloy is characterized by weighing, scanning electron microscope, X-ray energy spectrum, electron probe microanalyzer and carbon sulfur analyzer. The results show that T-22 alloy does not form a continuous dense oxide layer under six corrosion conditions, internal oxidation occurs in the alloy and complete decarburization occurs, and the amount of decarburization is up to 92.86%. The mass change of T-22 alloy after corrosion is very small, and the alloy has been fully decarburized after corrosion for 50 h.
Comparative Validation of Three Dimensional Fuel Rod Fine Simulation Software FUPAC3D and FUPAC
Wang Yanpei, Liu Zhenhai, Qi Feipeng, Tang Changbing, Zhang Kun, Zhou Yi, Wang Peng, Yu Lin
2022, 43(4): 46-52. doi: 10.13832/j.jnpe.2022.04.0046
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To verify the function and accuracy of the three-dimensional fuel rod fine simulation software FUPAC3D based on the three-dimensional finite element analysis platform in analyzing and evaluating the radiation-thermal -mechanical coupling behavior of PWR fuel rods, in this paper, the thermal model, fuel rod mechanical model, fission gas release model and corrosion model adopted by the three-dimensional FUPAC3D software are given. Taking the typical fuel rod parameters and operating conditions of HPR1000 as input parameters, the three-dimensional FUPAC3D software and the 1.5-dimensional FUPAC software that has been applied in engineering are used for modeling and analysis. The calculation results of two kinds of software in the temperature of pellet and cladding, the stress and strain of cladding, and the gap width between pellet and cladding are compared. The results show that FUPAC3D software and FUPAC software have considerable accuracy, and FUPAC3D software has the fine ability to simulate the radiation-thermal-mechanical coupling behavior of PWR fuel rods.
Research on Machine Vision Measurement Method of Rod Spacing Data for PWR Fuel after Irradiation
Li Weicai, Liu Pengliang, Wang Congzheng, Guo Yan
2022, 43(4): 53-59. doi: 10.13832/j.jnpe.2022.04.0053
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To address the difficulty of obtaining and processing fuel rod spacing data after irradiation, based on the geometric characteristics of fuel rods and their arrangement in PWR fuel assemblies, this paper proposes an efficient and reliable method for measuring fuel rod spacing data based on machine vision. The method first uses Retinex algorithm to enhance the preprocessing of the collected images of underwater fuel rods; then, for the front and rear imaging interference problem of the fuel rod array, edge enhancement and line-by-line grayscale feature processing methods are used to effectively separate the fuel rods to be tested from the background fuel rods, and further improve the image clarity; finally, the single-row gray value of the fuel rod image is fitted with quadratic curve to obtain the sub-pixel edge point coordinates of each fuel rod. The field experimental verification results of spent fuel assemblies show that this method can realize the rod spacing measurement of 16 fuel rods at one time, and the measurement accuracy is up to ±0.32 mm, which can provide efficient and reliable data support for fuel performance analysis.
Analysis and Research of Release Behavior on Fission Product of Dispersion Fuel
Tian Chao, Jing Futing, Xia Mingming, Huang Qianming, Liu Jiajia, Xiao Feng, Lyu Huanwen
2022, 43(4): 60-64. doi: 10.13832/j.jnpe.2022.04.0060
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In order to obtain the release characteristics of fission products from dispersion fuel to coolant in the primary loop, the release behavior of fission products from dispersion fuel has been studied, a fission product source term calculation code for dispersion fuel is developed, and the influence of fission product source term is analyzed. The results indicate that the nuclide spectra of fission products after uranium contamination and blistering damage are different; and the release of fission products is proportional to the square of the blister equivalent diameter; and for the dispersion fuel, the proportion of release through recoil in blistering damage is fairly low; and there is a difference in the order of magnitude of fission product release between the dispersion fuel and the ceramic fuel under the same fracture condition. The code developed in this paper can be used to analyze the fission product source term of dispersion fuel, and lay the foundation for the engineering design of follow-up projects.
Structure and Mechanics
Study on Cyclic Relaxation Characteristics of C-ring for Reactor Pressure Vessel
Dong Yuanyuan, Zhang Yabin, Du Hua, Wang Xuxin, Zhao Wei
2022, 43(4): 65-69. doi: 10.13832/j.jnpe.2022.04.0065
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The C-ring for reactor pressure vessel is composed of three layers. In the actual service process, the interaction mechanism between the layers is complex. After multiple compression-rebound cycles, the sealing structure has cyclic relaxation, resulting in the degradation of sealing performance. The experimental study on the above cyclic relaxation is carried out, the degradation law of the sealing performance of C-ring with the number of cycles is obtained, and its cyclic relaxation characteristics are studied. Through theoretical simulation calculation and analysis, the total springback, effective springback, working point line load and other characteristic quantities of the cyclic relaxation law are obtained; By comparing the experimental and theoretical simulation results, the influence of the manufacturing process on the cyclic relaxation characteristics is revealed. The research in this paper can be used to guide the service performance evaluation and manufacturing process control and optimization of the C-ring.
Simplified Fatigue Life Analysis of Pipeline under Super Multi-Support and Multi-Dimension Excitations Based on Pseudo Excitation Method
Long Teng, Yang Shouping, Gong Chun
2022, 43(4): 70-77. doi: 10.13832/j.jnpe.2022.04.0070
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In view of the problem that the calculation scale of the traditional random vibration module is not applicable under the super multi-point random vibration boundary conditions, and the traditional fatigue life analysis method is restricted by the heavy modeling workload, it is unable to complete the fatigue life analysis quickly, based on the virtual excitation method, a simplified fatigue life analysis method for pipeline system with super multi-point and multi-dimension excitation is proposed in this paper. The applicability of the virtual excitation method is verified by comparing the structural dynamics response calculated by the virtual excitation method and the traditional random vibration module, and the fatigue life of straight pipe structure and tee connection position is analyzed by using pipeline simplified fatigue life analysis method and traditional fatigue life analysis method. The results show that the accuracy of random vibration response calculated by the virtual excitation method is consistent with that of the traditional random vibration module, which shows that the method proposed in this paper can break through the limitation of the traditional random vibration module on the number of vibration excitation points and frequency points; The method proposed in this paper does not need to establish a detailed finite element model, and the stress and life analysis results of the straight pipe structure are basically consistent with the fine model, and the stress and life analysis results of the tee connection position are more conservative than the fine model. The research in this paper can provide theoretical guidance for the rapid fatigue life analysis of complex vibrating pipeline systems.
Study on Dynamical Stress Characteristics of DVI Nozzle Subjected to Pressurized Thermal Shocks
Zhao Yanyi, Wang Zewu, Li Junbao
2022, 43(4): 78-85. doi: 10.13832/j.jnpe.2022.04.0078
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The dynamic stress characteristics of direct vessel injection (DVI) nozzle under thermal shock are of great significance for the structural integrity evaluation of reactor pressure vessel (RPV). The thermal-fluid-solid coupling numerical calculation model of the RPV pressurized shell containing the DVI nozzle is firstly established and verified. Then the thermal hydraulic characteristics of RPV cylinder and DVI nozzle of ACC and CMT are studied; Finally, the distribution characteristics of temperature, equivalent stress and equivalent plastic strain in the high stress zone of RPV cylinder and DVI nozzle under thermal shock are analyzed. The results show that there is a large temperature gradient and equivalent stress in the connection area between RPV cylinder and DVI nozzle during ACC safety injection, and local plastic deformation occurs. In case of pressurized thermal shock event, the temperature difference in DVI nozzle area shall be controlled to ensure the structural integrity of reactor pressure vessel. The thermal-fluid-solid coupling numerical calculation model and method developed in this paper can be used for the safety evaluation of DVI nozzle and RPV cylinder in nuclear island, and can also be used to analyze the dynamic stress characteristics of similar pressure-bearing structure under thermal shock.
Study on Modified θ-projection Creep Model of Reactor Pressure Vessel Steel SA533B
Yu Peng, Yang Xiaoming, Ma Rubing, Yuan Yidan
2022, 43(4): 86-90. doi: 10.13832/j.jnpe.2022.04.0086
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Under severe reactor accidents, there is a risk of creep failure of pressure vessel due to insufficient cooling. Based on the modified θ-projection creep model, a creep model for reactor pressure vessel steel SA533B is proposed. The model can completely describe the three-stage creep process, and the simulation results are in good agreement with the experimental result creep curve. At the same time, the creep behavior under any load can be predicted by interpolation method. The model can be further used in the relevant case analysis of pressure vessel failure.
Research on Stiffness Characteristics of Hold-down Spring Based on Johnson-Cook Constitutive Model
Yang Ningrui, Wu Xingwen, Liang Shulin, Qin Mian, Zhu Fawen, Wang Haoyu, Liu Menglong
2022, 43(4): 91-98. doi: 10.13832/j.jnpe.2022.04.0091
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The stiffness design of hold-down spring in fuel assembly plays an important role in its safe service. By introducing the Johnson-Cook nonlinear constitutive relation of INCONEL 718 Alloy, the Johnson-Cook nonlinear constitutive model of INCONEL 718 Alloy under different neutron irradiation doses is fitted; the finite element model of the hold-down spring system is established, and the influence of different factors on the stiffness characteristics of the hold-down spring is studied. The results show that the increase of temperature, loading times and amplitude lead to different degrees of softening of the hold-down spring under cyclic loading, while the increase of the loading rate makes the stiffness of the hold-down spring harden. The study in this paper can provide a reference for the stiffness design of the hold-down spring in the fuel assembly.
Research on Anti-Impact Scaling Test Method for Supporting Structure of Large Equipment
Li Pengzhou, Li Yilei, Sun Lei, Qiao Hongwei, Chen Xuede, Zhang Kun, Li Xihua
2022, 43(4): 99-105. doi: 10.13832/j.jnpe.2022.04.0099
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In order to provide experimental basis for anti-impact design of the supporting structure of large nuclear power equipment and ensure that its impact resistance meets the requirements of relevant specifications, it is necessary to select the scaling model of the large equipment and its supporting structure as the research object of impact resistance technology and carry out impact test on standard medium-sized impact machine. According to the π theorem, the impact resistance test of the scaling model of the supporting structure of the large equipment is theoretically analyzed, and three optional test conditions are obtained; By changing the acceleration value of the scaling model, the maximum stress strength at the root of the supporting structure of the model is consistent with that of the prototype, and the difficult problem of test conditions in the derivation of π theorem is solved. The analysis method established in this study can provide a reference for the scaling test of the impact machine of the supporting structure of large equipment.
Structural Integrity Analysis of Fuel Element Cladding Based on Fluid-Solid-Heat Coupling
Yu Hang, Zhao Xinwen, Fu Shengwei, Zhu Kang
2022, 43(4): 106-112. doi: 10.13832/j.jnpe.2022.04.0106
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When a transient condition occurs in the reactor system, the transient and great change of the coolant will impact the structural integrity of the fuel element cladding and endanger the safety of the reactor. In this paper, taking the 3×3 fuel assembly of a PWR as the object, 3D refined simulation of the flow and heat transfer characteristics of fuel assembly and temperature, deformation and stress of fuel element cladding under cold water accident is carried out with the fluid-solid-thermal coupling method. The results show that the spacer grid can enhance the convective heat transfer intensity on the fuel rod surface; When the cladding is deformed, it bends to the side in contact with the rigid convex and bulges to the side in contact with the spring; the temperature and the maximum equivalent stress of the contact part between the cladding and the spacer grid increase with the accident time, and the maximum equivalent stress exceeds the yield strength of the cladding material, which will cause strength failure and affect its structural integrity. The research in this paper can provide reference for the integrity evaluation of reactor fuel element cladding under transient condition.
Safety and Control
Calculation and Analysis of Rod Dropping Process of Improved Control Rod Assembly for Supercritical Pressurized Water Reactor
Zheng Lele, Yue Ti, Zhu Fawen, Qin Mian, Li Xiang, Liu Menglong, Huang Shan
2022, 43(4): 113-117. doi: 10.13832/j.jnpe.2022.04.0113
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In order to verify whether the improved control rod assembly of supercritical pressurized water reactor (SCWR) can achieve the expected hydraulic buffering function, the rod dropping process is analyzed by using Fluent based on 6 degree of freedom model, which is an option of the layering dynamic mesh method. The control rod assembly dropping time and rod dropping final velocity are analyzed. The results show that compared with the design before the improvement, the dropping time of the improved control rod assembly is increased, but it can still meet the safety requirements; Compared with that before the improvement, the rod dropping final velocity is significantly reduced, and the impact force of rod dropping is thus reduced, so as to ensure the structural integrity of control rod assembly and fuel assembly. The design of the improved control rod assembly can achieve the expected hydraulic buffering function and can be used in the core design of supercritical pressurized water reactor.
Fault Diagnosis Method of Nuclear Power Plant Based on Adaboost Algorithm
Li Xiangyu, Cheng Kun, Tan Sichao, Huang Tao, Yuan Dongdong
2022, 43(4): 118-125. doi: 10.13832/j.jnpe.2022.04.0118
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At present, most of the nuclear power plant fault diagnosis algorithms based on ensemble learning pay attention to improving the identification accuracy of various machine learning algorithms, while ignoring the integration method of the underlying base learner, which makes it difficult to improve the accuracy of the ensemble learning algorithm in identifying accident types, and there is a problem of whether the identification results are credible. In this paper, based on Adaboost algorithm, a machine learning algorithm model is designed to enable the control system of a nuclear power plant to identify fault types independently. By reasonably allocating weight coefficients for various fault identification algorithms of ensemble learning, the algorithm model improves the identification accuracy and reliability of the whole ensemble learning algorithm for nuclear power plant accident types. At the same time, the test results show that the average identification accuracy of Adaboost algorithm for seven typical nuclear power plant operation or accident conditions can reach more than 95%; And when the accident occurs 150 seconds, the identification accuracy can reach 100%. Therefore, the integration method of Adaboost algorithm to the base learner can be used to optimize the algorithm structure of ensemble learning and improve the identification accuracy of the algorithm for the types of nuclear power plant accidents.
Optimization Research of SGTR Procedure for Nuclear Power Plant
Liu Lixin, Wang Zhe
2022, 43(4): 126-130. doi: 10.13832/j.jnpe.2022.04.0126
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Nuclear power plant mitigates SGTR accident by emergency operating procedure (EOP). SGTR analysis results show that after the operator opens the pressurizer relief valve to decrease reactor coolant system pressure in the process of accident mitigation, the safety injection flow increases quickly causing the pressurizer level increasing substantially, there may be a potential danger. The purpose of this paper is to better mitigate the SGTR accident, so that the pressurizer level will not rise too high during the accident mitigation process to keep the nuclear power plant safe. By optimizing the EOP mitigation steps, cutting off a row of safety injection in advance, and analyzing and calculating the accident mitigation process of the optimized EOP, the final results show that the maximum level of the pressurizer decreases, which reduces the risk of excessive level of the pressurizer, which provides a basis for the follow-up improvement of the nuclear power plant procedure.
Study on Operation of Small Lead-Based Fast Reactor Based on Constant Ratio of Flow to Power
Hu Yang, Liu Yinuo, Zeng Wenjie, Liang Lehua, Li Chuhao
2022, 43(4): 131-135. doi: 10.13832/j.jnpe.2022.04.0131
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In order to study the operation strategy of small lead-based fast reactor, the core transfer function model is established, and the operation schemes of constant ratio of core flow to power and stable core nuclear power are designed by using proportional-integral-derivative (PID) controller and control rod drive mechanism. The control systems with different operation strategies are established respectively, and the simulation of primary loop flow step and core reactivity disturbance is carried out. The results show that under the condition of introducing primary loop flow step down, the stable nuclear power operation scheme leads to the high core outlet temperature due to the constant core power; in the case of constant ratio of flow to power, the core power decreases with the primary flow, so that the core outlet temperature is stable in a safe range. When there is step reactivity disturbance, both schemes can quickly control the overshoot and overshoot of core power, and the core outlet temperature is basically constant.
Realization of Modbus/TCP Protocol Library Functions Across Embedded Platforms
Cheng Yangjie, Qin Fan, Xu Yonghong, He Xiaopeng, Dai Kailei, Li Lu, Zheng Xiao
2022, 43(4): 136-142. doi: 10.13832/j.jnpe.2022.04.0136
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In order to apply Modbus/TCP protocol in different localized CPUs, it is necessary to design and realize cross-platform protocol library functions. According to the specification of Modbus application protocol, the cross-platform library functions of the ten function codes that Modbus/TCP protocol needs to support are developed without relying on operating system calls. The library functions can be used by the communication application layer for the design and development of Modbus/TCP protocol client and server according to its own needs. After secondary development on localized CPU and embedded platform, by calling the library function in this study, the developer can effectively collect the data of the pressurizer pressure and water level of the reactor coolant system and the power regulation of the rod control system through the Modbus/TCP protocol.
Study on the Habitable Dose Model of Internal Leakage in Main Control Room of Nuclear Power Plant
Wang Chao, Shi Yanming, Zhang Yanle
2022, 43(4): 143-146. doi: 10.13832/j.jnpe.2022.04.0143
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The radioactive impact of unfiltered air leakage (internal leakage) in the main control room of a nuclear power plant is an important part of habitability evaluation. At present, the dose model for this part is too simplified and does not accord with the actual engineering design. Combined with the actual design characteristics of a nuclear power plant, this study studies the migration mechanism of internal leakage source terms, deduces the differential equation of radioactivity, establishes the habitable internal leakage dose model of the main control room, selects typical design basis loss of coolant accident (LOCA) and large break loss of coolant accident (LB-LOCA) with reactor melting to carry out the application of the model, and compares it with the commonly used simplified model at present. The results show that the dose result of the simplified model under LB-LOCA condition is smaller than that of this model, and the simplified model can not cover all accident scenarios. After analysis, the internal leakage dose model established in this study is more in line with the actual scenario, suitable for the impact evaluation of internal leakage in the habitable area of the main control room, and can be used for the verification of internal leakage test results and engineering project design.
Development and Application of Structural Method for Uncertainty Evaluation of Constitutive Models
Xiong Qingwen, Gou Junli, Du Peng, Deng Jian, Liu Yu, Chen Wei, Dang Gaojian
2022, 43(4): 147-153. doi: 10.13832/j.jnpe.2022.04.0147
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The best estimate plus uncertainty (BEPU) analysis is recommended by IAEA for safety analysis of nuclear power plants, and has become the mainstream method for license application of nuclear power plants. Typical BEPU method relies on the best estimation program to propagate the uncertainties of input parameters to the output, while the uncertainties of the program constitutive model are often not properly considered. In this study, a structural method is proposed to evaluate the uncertainties of program constitutive model. Based on this method, constitutive models are classified according to characteristics, and different evaluation methods are adopted for different model types. The model evaluation methods used in this study include the non-parametric curve estimation method in the forward method and the Bayesian calibration method and coverage calibration method in the reverse method, as well as alternative model construction methods. The structural method is used to quantify the uncertainties of important models in LOCA, and the quantified model uncertainties are transmitted to the peak cladding temperature through sampling calculation. The results show that both the sampling calculation values and the experimental values are smaller than the conservative calculation value, and the propagation calculation results after considering the model uncertainties can well cover the experimental values, and the safety margins can be effectively increased after considering the model uncertainties.
Circulation and Equipment
Research on Modeling and Simulation of Nuclear Power System Based on APROS
Tian Peiyu, Li Yi, Liang Tiebo, Wang Changshuo
2022, 43(4): 154-161. doi: 10.13832/j.jnpe.2022.04.0154
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Based on the simulation software APROS, this paper establishes a simulation model of the primary loop and secondary loop coupling system of the two-loop nuclear power system, and simulates the steady-state condition of power operation and the dynamic condition of linear variable load. The results show that the maximum steady-state relative error of the model simulation results is less than 5%, which is in good agreement with the design value; the dynamic simulation trend is basically consistent with the simulation trend of the thermal-hydraulic calculation program RELAP5, which verifies the effectiveness of the model. Therefore, the primary and secondary loops of the nuclear power system match well, and the system model established in this paper can accurately simulate the operation of the nuclear power system.
Development of In-Service Inspection System for Flange Bolts of Reactor Coolant Pump in AP1000 Nuclear Power Plant
Zhou Lusheng, Wang Liren, Liu Yizhou, Chen Junrong
2022, 43(4): 162-167. doi: 10.13832/j.jnpe.2022.04.0162
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The flange bolts in-service inspection of reactor coolant pump in AP1000 Nuclear Power Plant is one of important supervision items. At present, there is no in-service inspection system and application case in China for this component. Combined with the structural characteristics of AP1000 reactor coolant pump flange bolts, the analysis of on-site high-dose environment and complex inspection conditions, this paper designs and develops a set of in-service ultrasonic inspection system for reactor coolant pump flange bolts, which implements ultrasonic inspection from the inner wall of the bolt center hole and is suitable for in-service inspection requirements. The debugging and test results on the simulation body of the reactor coolant pump show that the system can realize the functions of circumferential operation, vertical obstacle avoidance, precise alignment adjustment between the special ultrasonic probe and the bolt hole, etc., and then realize the ultrasonic scanning of the reactor coolant pump flange bolts. The engineering application results show that the system meets the requirements of in-service inspection of flange bolts of reactor coolant pump in AP1000 Nuclear Power Plant, and has high reliability and good applicability.
Research on Multi-parameter Synchronous Monitoring Technology of Nuclear Power Plant Equipment Status
Shen Jiangfei, Wang Shuangfei, Huang Lijun, Ling Shuanghan, Zhang Sheng
2022, 43(4): 168-173. doi: 10.13832/j.jnpe.2022.04.0168
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Abstract:
The stable operation status and long-term operation data accumulation of nuclear power plant equipment have established a good data foundation for realizing data-driven intelligent monitoring of equipment status. In this paper, an intelligent monitoring method of equipment condition based on multi-parameter correlation is proposed, which includes three steps: modeling, training and inference, and a data-driven intelligent monitoring and early warning model of equipment condition is established. First, the system equipment monitoring parameters, parameter monitoring contents and correlation are identified and analyzed, and the correlation model of monitoring parameters is designed and established. Second, the historical data of normal operation of the equipment are collected and selected as training data, and the correlation model is trained based on BP feed forward neural network; Finally, the measured values of the monitoring parameters of the equipment are collected in real time, and the predicted values of each parameter are inferred based on the model, and the deviation between the measured value and the predicted value is monitored, and an early warning message is given when the deviation exceeds a predetermined threshold. This paper takes the heat exchanger and main feed pump of a power plant as an example to model and verify. The results show that the monitoring model proposed in this paper can effectively monitor the small and abnormal changes of equipment parameters synchronously, give early warning against early abnormalities, and maintain a very low false alarm rate.
Operation and Maintenance
Study on Stability of Gaps of Main Equipment Support of Nuclear Power Plant Nuclear Island
Chen Sun, He Yingyong, Zhang Xinghui, Zhou Yupeng, Zhao Xiaohong
2022, 43(4): 174-177. doi: 10.13832/j.jnpe.2022.04.0174
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Abstract:
There is a certain gap between the main equipment and its support in a nuclear power plant nuclear island to allow the free thermal displacement and thermal expansion caused by the change of temperature and pressure of the reactor coolant system. The stability of gap is important for the safe operation of the nuclear power unit. Therefore, the support gap needs to be measured and evaluated in each refueling cycle of the in-service nuclear power unit, which takes a long time and has high radiation risk. In this paper, the influencing factors of support gap composition are analyzed. Combined with the historical data of gap measurement and engineering experience, the concept of gap stability and its acceptance criteria are proposed, and the evaluation process of gap stability is determined. Based on the measured data of the support gap of the voltage stabilizer in a running nuclear power plant, it provides a basis for shortening the period of gap measurement. The method can shorten the cycle of shutdown and refueling, reduce the radiation dose of the overhaul and measurement personnel, and ensure the safe operation of the unit and improve the economy.
Research on Radioactive Contaminated Soil Sorting and Volume-reducing Device
Luo Feng, Li Zhenchen, Zeng Guoqiang, Li Wenyu, Hu Bo, Zhu Xin, Gu Min, Wan Qianyin
2022, 43(4): 178-184. doi: 10.13832/j.jnpe.2022.04.0178
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Abstract:
The remediation and restoration of radioactive contaminated sites has become an important support to ensure the healthy and sustainable development of the nuclear industry. In this paper, according to the treatment requirements of radioactive contaminated soil in typical areas, the source term analysis and sorting principle test are carried out to determine the radioactive contaminated soil sorting and volume-reducing process scheme and the device design indicators. Then a novel radioactive contaminated soil sorting and volume-reducing device is designed, which could realize the functions of radioactive contaminated soil drying, screening, online detection and separation according to disposal requirements. The performance verification results show that the theoretical detection limit of 137Cs in radioactive contaminated soil is 20.7 Bq/kg, and the treatment capacity can reach 106 kg/h, which meets the design indicator. The device is expected to realize the downgrade of part of the contaminated soil in a typical area from low-level radioactive waste to extremely low-level radioactive waste or extremely low-level radioactive waste to exempt waste in the follow-up project implementation. The research can provide theoretical guidance and experimental basis for the process design and engineering verification of radioactive contaminated soil treatment.
Research on Diagnosis Method of Operational Events of Nuclear Reactor Based on Convolutional Long Short-term Memory Network and Artificial Whale Algorithm
Sun Yuanli, Song Zhihao
2022, 43(4): 185-190. doi: 10.13832/j.jnpe.2022.04.0185
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Abstract:
In case of abnormality in the nuclear power plant, the cause shall be diagnosed in time to avoid serious consequences to the operators and the surrounding environment. In this paper, convolutional neural network (CNN) and long short-term memory (LSTM) network can better extract the local characteristics of data and the characteristics of memory time series information, and study the operational event diagnosis technology of nuclear reactor based on convolutional long short-term memory (CLSTM) network and artificial whale algorithm. The method described in this paper was tested by the simulation experiment of nuclear power plant reactor simulator, and the final test accuracy is 99.91%, which proves the effectiveness of the research method described in this paper. The relevant research results can be used as a diagnosis method of nuclear power plant operational events, which is conducive to improving the intelligence and information level of operational event diagnosis, providing a technical basis for few or no people on duty in nuclear power plants, and improving the public's understanding and trust in nuclear power plants.
Quantitative Study on 5% Damage Rate of Steam Generator Heat Transfer Tube in M310 Unit
He Feng, Zhu Jianping, Lu Yuechuan, Lu Xifeng, Wang Xinjun, Li Xiao
2022, 43(4): 191-195. doi: 10.13832/j.jnpe.2022.04.0191
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Abstract:
For M310 nuclear power unit, Part 3 ( Radiochemistry Code ) of the Technical Specification of Chemistry and Radiochemistry specifies that when “133Xe>92500 MBq/t or 133Xe>37000 MBq/t and 131I/133I>1.5”, execute the command "if the damage rate of heat transfer tube of at least one steam generator (SG) exceeds 5%, reduce the load to NS/SG mode at the rate of 50 mw/min". However, in the actual operation process, since it is impossible to judge whether the damage rate of SG tube exceeds 5%, it is impossible to determine whether to execute the instruction of "reducing the load to NS/SG mode at a rate of 50 MW/min". Therefore, for M310 nuclear power unit, the operation instruction in the Radiochemistry Code is not operable. This paper investigates the relevant radiochemistry operation instructions at home and abroad, fully understands the meaning of "the damage rate of heat transfer tubes exceeds 5%”, and makes quantitative analysis. Through the analysis of the crack stability of the SG heat transfer tube under the leakage rate of 5, 24, 44 L/h, it is found that the unstable fracture of a single heat transfer tube will not occur at these leak rates, and then an operational suggestion is given: using the indicator "If the leak rate of at least one SG from the primary side to the secondary side exceeds 5 L/h" instead of the unquantifiable "If the damage rate of at least one SG heat transfer tube exceeds 5%", the safety and economy of nuclear power plant operation can be ensured.
Research on Halcon-based Image Positioning Technology of Reactor Core Detector Assemblies
An Yanbo, Yu Zhiwei, Wang Bingyan, Chen Shuhua, Zhang Anrui, Xu Shichao
2022, 43(4): 196-200. doi: 10.13832/j.jnpe.2022.04.0196
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Abstract:
The reactor core detector assemblies of HPR1000 reactor need to be removed and replaced after their life cycle expires. Due to the large deviation of the detector assemblies themselves and other factors, the theoretical coordinates during installation cannot be directly used as the positioning coordinates during the removal of the detector assemblies. In this paper, an algorithm for image positioning of core detector assembly is developed based on machine vision software Halcon. The algorithm uses the principle of template matching to search in the images of the detector assemblies captured by the camera to obtain the exact coordinates of the detector assemblies. Experiments show that the algorithm has high positioning accuracy and can meet the requirements of the image positioning algorithm for the removal of the detector assemblies.
Investigation on Screening and Ranking of Life Cycle Management Objects in Nuclear Power Plant
Lyu Fangming, Lei Cheng, Guo Lixia, Shi Yan, Wang Yadong
2022, 43(4): 201-205. doi: 10.13832/j.jnpe.2022.04.0201
Abstract(174) HTML (53) PDF(27)
Abstract:
The objects of life cycle management (LCM) are systems or equipment that have an important impact on the safety, availability and economy of power plants. How to identify these objects is the basis of LCM. Based on the study of the traditional life cycle management object screening and ranking method of Electric Power Research Institute (EPRI), and combined with the present situation and demand of equipment management of Chinese nuclear power enterprises, a screening method which takes into account both equipment criticality and economic factors is put forward. Besides, the hierarchical attributes that can reflect the functional importance and the actual operation and maintenance status of the equipment are established, and the weight of each attribute is calculated by the analytic hierarchy process, and then the priority rankings of the LCM objects are determined. The application results in a nuclear power unit show that this quantitative screening and ranking method is in line with industry experience and the actual operation and maintenance status of equipment, and has important application value and practical guiding significance.
Column of Science and Technology on Reactor System Design Technology Laboratory
Study on Reactivity Variation and Its Influencing Factors of Lead-cooled Traveling Wave Reactor
Qin Tianjiao, Xia Bangyang, Li Qing, Li Sinan, Zhang Ce, Lu Di
2022, 43(4): 206-212. doi: 10.13832/j.jnpe.2022.04.0206
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Abstract:
Lead-cooled traveling wave reactor has outstanding advantages of good safety, long refueling and shuffling cycles, and high utilization rate of uranium resources. It is one of the key development directions of advanced nuclear energy system. Realizing the small change of reactivity is the key technical problem in the core scheme design of lead-cooled traveling wave reactor. Taking the physical scheme of lead-cooled traveling wave reactor with thermal power of 700MW and metal fuel as the research object, this paper focuses on the influence of the design parameters of core ignition zone and breeder zone on the effective multiplication factor (keff), and analyzes the change trend of core reactivity in the whole life. The numerical results show that ignition zone design parameters significantly affect the initial keff of the core. The larger the amount of fissile nuclides in the ignition zone, the larger the initial keff. The variation range of reactivity can be effectively reduced by adjusting the axial position of the ignition zone in the core and its fuel enrichment; The higher the ratio of convertible nuclides to fissile nuclides loaded in the core, the more 239Pu produced by breeding, and the better the overall breeding performance; The longer the breeder zone, the longer the duration of equilibrium state and the longer the core life. The conclusions of this paper can provide an important theoretical basis for the physical scheme design of lead-cooled traveling wave reactor core and the selection of key parameters.
Research on Wide Energy Spectrum Neutron Detection Technology Based on 6LiF/ZnS(Ag) and Scintillating Fiber
Xiong Bangping, Wu Zhiqiang, Wan Bo, Yang Daibo, Li Kun, Li Gang, Zhang Hu
2022, 43(4): 213-217. doi: 10.13832/j.jnpe.2022.04.0213
Abstract(279) HTML (55) PDF(49)
Abstract:
To solve the shortcomings of traditional neutron detectors in neutron detection in complex environments such as narrow space, strong electromagnetic interference, and long-distance transmission, this study combines 6LiF/ZnS(Ag) mixed materials and scintillating fiber to design a new type of scintillator fiber neutron detector that can be used for wide energy spectrum neutron measurement. Based on the FLUKA software, the neutron detection performance of the new fiber neutron detector was simulated and the optimization design of the scintillator fiber probe was completed. The results show that when the energy of the incident neutron is 0.01~10eV and 0.5~10MeV, the new neutron detector has a high neutron detection efficiency, it can detect the neutron in the wide spectrum of thermal neutron and fast neutron. In addition, by comparing the difference in pulse amplitude, the new neutron detector can realize the discrimination of n-γ signals.
Research on Dual-restricted Nuclide Selection and Burnup Chain Compression Algorithm
Hu Yuying, Liao Hongkuan, Yao Dong, Yu Yingrui, Zhou Bingyan
2022, 43(4): 218-222. doi: 10.13832/j.jnpe.2022.04.0218
Abstract(551) HTML (188) PDF(21)
Abstract:
In the burnup calculation of the reactor core, due to the large difference in the reaction cross sections and lives of nuclides in the evaluation nuclear database, the burnup matrix has large scale and strong rigidity. In order to reduce the scale of burnup matrix and improve the ill conditioned degree of matrix, it is necessary to study the burnup chain compression algorithm suitable for various core design and R & D requirements, and form a compression burnup chain and database. First, the nuclide selection criteria are established, and the nuclide importance is sorted and screened according to the contribution rate of each nuclide to neutron absorptivity and important nuclide nucleon density. The dual-restricted burnup chain compression algorithm based on neutron absorption rate and contribution rate of important nuclide production is studied, and the development of related program modules is completed. Through the calculation and analysis of Kylin-2 program database compression, the feasibility of the burnup chain compression algorithm is verified. The use of compression database can greatly reduce the calculation time and improve the calculation efficiency on the basis of keeping the original calculation accuracy. Through the research of burnup chain compression algorithm and the implementation of compression database, it provides technical support for making compression database from the evaluation database.
Research on Fault Prediction of Reactor Power Measurement Circuit Based on Relevance Vector Machine
Min Yuan, Chen Zhi, Wan Bo, Yang Cheng, Han Wenxing, Yuan Yannan
2022, 43(4): 223-229. doi: 10.13832/j.jnpe.2022.04.0223
Abstract(160) HTML (35) PDF(25)
Abstract:
In order to improve the supportability and maintainability of the nuclear measuring device, taking the reactor power measurement amplifier circuit as the object, this paper predicts the typical faults of the circuit through the multi-kernel relevance vector machine model based on quantum particle swarm optimization algorithm. From the pulse response signal of the power measurement amplifier circuit, the feature information is extracted by the wavelet packet decomposition method, and the Euclidean distance between the feature and the normal state feature of the circuit is calculated as the fault degree indicator of the circuit. Multi-kernel relevance vector machine is selected to establish circuit fault prediction model. The influence of kernel function type and parameter optimization algorithm of relevance vector machine model on the prediction effect of the model is analyzed. The research results show that the multi-kernel relevance vector machine model optimized by quantum particle swarm algorithm has better prediction accuracy for the future running state of the circuit, and can accurately predict the changing law of the fault degree of the circuit.