Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-20
Abstract： [Background]: The single-phase reversed flow in inverted U-tubes of steam generator (SG) leads to increasing flow resistance and decreasing heat transfer area, so it is meaningful to study this phenomenon. [Purpose]: The water level of the secondary side in SG can influence the single-phase reversed flow, it is necessary to clarify its influence mechanism from a more general viewpoint. [Methods]: The dimensionless conservation equations were derived first, and the extreme point was obtained based on the equations. Then the effect of the water level of the secondary side under conditions of different lengths, dimensionless resistance number, and dimensionless heat transfer number was analyzed. [Results]: The decrease in the water level leads to the critical point of the single-phase reversed flow gradually approaching the origin, the influence law of the water level is the same under different pipe length conditions. As the water level decreases, the influence of the dimensionless resistance number and dimensionless heat transfer number on the critical point gradually reduces. [Conclusions]: This study theoretically proves that the effect of secondary water level on single-phase reversed flow is not conducive to the occurrence of backflow, and explains the reasons from a mechanistic perspective, which can assist in accident analysis of related nuclear power plants.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-17
Abstract： [Background]: The criticality assembly is a device with sufficient fissile materials to maintain the self-sustaining chain fission reactions in a controlled way at low power without cooling system, which is widely used in a variety of reactor physics experiments and measurement methods. The neutron source system is an important system equipment of the criticality assembly. When starting up, the neutron source is transported from the shielding tank to the designated position of the core, and a certain number of neutrons are injected into the core continuously, so that the core can maintain a certain number of fission neutrons in the subcriticality state and play the role of "ignition". At the beginning of start-up, the neutron fluence rate in the core is low and cannot reach the source range of detectors. The method of introducing a neutron source into the core is usually adopted to raise the neutron fluence rate to a high enough level before the reactor reaches the criticality state, so that detectors can better monitor the core state and eliminate monitoring blind spots. [Purpose]: The design and development of this neutron source system were based on years of practical experience operating criticality assemblies and utilizing neutron sources. The system serves two primary functions: providing storage and shielding for the neutron source, as well as ensuring its safe transportation between the shield tank and designated location. [Methods]: The components of this neutron source system include shielding tank, connecting structure, neutron source drive with position monitoring, transportation pipeline, pressure detection device, etc., It can reliably enable the neutron source to move smoothly back and forth between the designated positions in the shielded vessel and the core. Real-time fault detection can be achieved through pressure sensors monitoring wire rope tension changes. Additionally, encoder feedback allows real-time positioning determination while terminal switches signal motor stoppage upon reaching specified positions. Prior to official use, tests using both dummy neutron source resembling actual size/mass characteristics and neutron source were conducted，which including motion tests (up/down), wire rope breakage assessment, accessibility checks along with shielding performance measurements inside tanks and pressure value testing. [Results] After multiple rounds of test which included over one thousand back-and-forth movements during more than one hundred experiments on criticality assembly - it has been verified that this system displays real-time positioning accuracy within ±2mm tolerance limits while maintaining normal functionality for upper/lower terminal switches post-installation. Furthermore，the total dose rate requirements for both gamma rays/neutrons ≤10μSv/h have been met by these systems after installation. This comprehensive validation proves once again that our designed scheme offers simple operation procedures alongside high reliability/repeatability levels; rapid troubleshooting capabilities further enhance its practical value. [Conclusions]: This paper gives full consideration to the design of the neutron source system of the criticality assembly, and focuses on optimizing the installation time of the neutron source and solving the problems of the source blockage. The design scheme of the system can be applied to some similar scenarios in industry, such as the transportation of radioactive items (radioactive sources, small spent fuel assemblies), and effectively solve the problem that radioactive items can not be found in time when they fall/get stuck in the pipeline.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-17
Abstract： [Background]: The volume of the nuclear power plant containment is huge, making it difficult to conduct equal-scale or large-scale thermal-hydraulic tests. Currently the test data mainly come from small-scale tests. [Purpose]:To address the applicability of small-scale containment test data in validation process of the containment performance analysis code, the analysis method for applicability of experimental data is proposed and developed on the basis of similarity analysis of the pressure response process in the containment. [Methods]: The applicability study of the test data, which are produced by some scaled containment facilities such as the HDR, Battelle and CVTR, is carried out in combination with the test parameters. The applicability of each test case are obtained respectively when they are applied to validate the containment code in case of the Large Break Loss of Coolant Accident (LBLOCA) and Main Steam Line Break Accident (MSLB) of HPR1000 nuclear power plant. [Results]:The results show that the similarity criteria for pressure response process and key phenomena within the containment vessel under accident conditions can be used to analyze the applicability of different containment tests to the target power plant. [Conclusions]:The proper combination of test cases including HDR ISP-16&23, Battelle CASP-1&2, and CVTR T3 can represent the pressure transient process, results of coupling phenomena such as mass and energy release at the break, condensation near the containment shell and internals, within the HPR1000 containment in case of LOCA or MSLB. The distortion is either within the acceptable range or conservative for design limits of containment pressure, so that the small-scale containment test data are suitable for the verification and validation of the HPR1000 containment thermal hydraulic response analysis code.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-16
Abstract： [Background]:The elastic properties of dispersed fuel serve as crucial parameters in the safety analysis of reactors and the performance assessment of fuel components.[Purpose]:This study considers dispersed nuclear fuel elements as a special type of particulate composite material and employs micromechanics methods to calculate the equivalent elastic properties of the fuel element.[Methods]:Using the universal finite element software ABAQUS and user-defined subroutines, assuming the periodic distribution of fuel particles in the core, a finite element calculation model is established. A representative volume element was selected as the research object, and a thermal mechanical fission gas migration coupling analysis method was established to calculate the equivalent elastic performance of the core. [Results]: The equivalent elastic properties of the fuel element were determined. The effects of particle volume content, particle size, and burnup on the equivalent elastic properties of dispersed nuclear fuel were analyzed and compared. [Conclusions]: The results indicate that the main factors influencing the equivalent elastic properties of the fuel element are particle volume and burnup.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-15
Abstract： The Pressure Water Reactor (PWR) pressurizer surge line connects the pressurizer and the primary circuit of the reactor, where the thermal stratification is easy to occur due to temperature difference of coolant in main coolant pipelines (MCL) and pressurizer (PZR). The phenomenon of thermal stratification could result in unexpected displacements and support loadings of the pipe system. The induced thermal stress may affect the structural integrity of pressurizer surge line, and reduce its fatigue life, posing a serious threat to safe operation of nuclear power plants. Thus, it is necessary to investigate the stress induced by thermal stratification. In this paper, the piecewise linear function and sigmoid function were applied to fit the fluid temperature profile inside the tube, and an analytical method was presented to calculate the temperature and thermal stress field, finally the analytical expressions were derived. The temperature and thermal stress fields obtained by the analytical expressions were compared to those by finite element method (FEM), which showed that the temperature field of the pipeline cross-section calculated by the analytical expression coincides with the FEM result; the thermal stresses obtained by the piecewise linear temperature profile are more conservative than those obtained by temperature profile fitted by sigmoid function; the thermal radial stress and hoop stress by the analytical expressions deviate from the results by FEM, but the axial stress is basically consistent with FEM result; the axial stress on the inner wall of pipeline caused by thermal stratification is most important among the thermal radial stress, hoop stress and axial stress and its peak value in generally located at the upper and lower position of mixing layer. Therefore, in the fatigue analysis, the thermal axial stress induced by thermal stratification and the upper and lower position of mixing layer should be focused. The derived analytical formula is accurate and universal, which can provide technical support for pipeline design considering thermal stratification.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-12
Abstract： Computational Fluid Dynamics (CFD) technology can be used for nuclear reactor core to understand and predict the fine thermal-hydraulic status, to obtain the optimizing design and operation, and improve the safety. However, CFD analysis of reactor core faces challenges such as difficulty in modelling huge amount of meshes, large amount of calculations, time consuming and resource requirements, etc. Moreover, the universality of CFD technology for reactor types is poor so that it requires the whole analysis process again when the reactor type is changed. Based on the characteristics of reactor structure and coolant flow feature, this paper develops a CFD supporting technology that is "specific" to the reactor core and "common" to different reactor types, which can decompose the CFD computing burden and effectively reduce the fine mesh modelling and calculation analysis. It has been successfully applied to the CFD analysis of the reactor cores with full number and whole height of fuel assemblies, such as the reactor core with wire-wound rod bundle assemblies, spacer grid rod bundle assemblies and plate element assemblies.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-07
Abstract： Computational Fluid Dynamics (CFD) technology can be used for nuclear reactor core to understand and predict the fine thermal-hydraulic status, to obtain the optimizing design and operation, and improve the safety. However, CFD analysis of reactor core faces challenges such as difficulty in modelling huge amount of meshes, large amount of calculations, time consuming and resource requirements, etc. Moreover, the universality of CFD technology for reactor types is poor so that it requires the whole analysis process again when the reactor type is changed. Based on the characteristics of reactor structure and coolant flow feature, this paper develops a CFD supporting technology that is "specific" to the reactor core and "common" to different reactor types, which can decompose the CFD computing burden and effectively reduce the fine mesh modelling and calculation analysis. It has been successfully applied to the CFD analysis of the reactor cores with full number and whole height of fuel assemblies, such as the reactor core with wire-wound rod bundle assemblies, spacer grid rod bundle assemblies and plate element assemblies.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-07
Abstract：氢化物是锆合金包壳管在核电厂正常运行过程中与一回路冷却剂发生锆水反应而产生的常见缺陷。本文利用分子动力学方法，采用COMB3势函数，构建含氢化物的锆基模型进行单轴拉伸模拟，探究了氢化物密度对锆力学性能的影响。研究结果表明，当氢化物密度在0~1078 µg/g时，随着氢化物密度的增加，屈服强度、应变和杨氏模量降低。在弹性阶段，氢化物密度的增加使应力集中区域增大，有利于位错形核；在塑性变形阶段，随着氢化物密度的增大，初始位错更倾向于在氢化物周围扩展。当氢化物密度在1078 ~ 2311 µg/g时，随氢化物密度的增加，屈服强度、应变和杨氏模量升高，这是由于氢化物密度较高时产生了大量位错并造成位错塞积。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-05
Abstract： [Background]:The elastic properties of dispersed fuel serve as crucial parameters in the safety analysis of reactors and the performance assessment of fuel components.[Purpose]:This study considers dispersed nuclear fuel elements as a special type of particulate composite material and employs micromechanics methods to calculate the equivalent elastic properties of the fuel element.[Methods]:Using the universal finite element software ABAQUS and user-defined subroutines, assuming the periodic distribution of fuel particles in the core, a finite element calculation model is established. A representative volume element was selected as the research object, and a thermal mechanical fission gas migration coupling analysis method was established to calculate the equivalent elastic performance of the core. [Results]: The equivalent elastic properties of the fuel element were determined. The effects of particle volume content, particle size, and burnup on the equivalent elastic properties of dispersed nuclear fuel were analyzed and compared. [Conclusions]: The results indicate that the main factors influencing the equivalent elastic properties of the fuel element are particle volume and burnup.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-04
Abstract： [Background]:Neural networks, with their powerful fitting capabilities, can learn the relationships between input and output variables based on large amounts of data, often serving as proxy models for physical programs in the field of engineering calculations, including nuclear engineering calculations. Neutron transport calculations, as one of the core links in neutronics simulations, often suffer from lengthy computational times. However, this issue can also be addressed by utilizing neural network models. Nevertheless, neural network models have a series of hyperparameters that need to be set, but manually adjusting these hyperparameters is laborious, repetitive, and reliant only on experience. Moreover, these hyperparameters are not reusable when solving different problems. [Purpose]: By seeking a surrogate model for VITAS, the research can provide some reference for the application of artificial intelligence in core physics calculation theory.[Methods]:This paper proposes the use of the Bayesian optimization algorithm to adjust neural network hyperparameters, combined with learning rate decay and loss function optimization methods. [Results]: By fitting the key core parameters obtained from VITAS's calculation of the TAKEDA benchmark problem, the results show that the average error of the effective multiplication factor is within 150×10-5, and the average error rate of the regional integral flux on the TAKEDA1 dataset is 1.72%, with a maximum error rate of 7.56%. [Conclusions]: This approach can automatically search for the optimal combination of hyperparameters for different datasets to achieve the best performance, demonstrating high flexibility, efficiency, and strong generalization.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-03
Abstract： [Background]: The volume of the nuclear power plant containment is huge, making it difficult to conduct equal-scale or large-scale thermal-hydraulic tests. Currently the test data mainly come from small-scale tests. [Purpose]:To address the applicability of small-scale containment test data in validation process of the containment performance analysis code, the analysis method for applicability of experimental data is proposed and developed on the basis of similarity analysis of the pressure response process in the containment. [Methods]: The applicability study of the test data, which are produced by some scaled containment facilities such as the HDR, Battelle and CVTR, is carried out in combination with the test parameters. The applicability of each test case are obtained respectively when they are applied to validate the containment code in case of the Large Break Loss of Coolant Accident (LBLOCA) and Main Steam Line Break Accident (MSLB) of HPR1000 nuclear power plant. [Results]:The results show that the similarity criteria for pressure response process and key phenomena within the containment vessel under accident conditions can be used to analyze the applicability of different containment tests to the target power plant. [Conclusions]:The proper combination of test cases including HDR ISP-16&23, Battelle CASP-1&2, and CVTR T3 can represent the pressure transient process, results of coupling phenomena such as mass and energy release at the break, condensation near the containment shell and internals, within the HPR1000 containment in case of LOCA or MSLB. The distortion is either within the acceptable range or conservative for design limits of containment pressure, so that the small-scale containment test data are suitable for the verification and validation of the HPR1000 containment thermal hydraulic response analysis code.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-02
Abstract：氢化物是锆合金包壳管在核电厂正常运行过程中与一回路冷却剂发生锆水反应而产生的常见缺陷。本文利用分子动力学方法，采用COMB3势函数，构建含氢化物的锆基模型进行单轴拉伸模拟，探究了氢化物密度对锆力学性能的影响。研究结果表明，当氢化物密度在0~1078 µg/g时，随着氢化物密度的增加，屈服强度、应变和杨氏模量降低。在弹性阶段，氢化物密度的增加使应力集中区域增大，有利于位错形核；在塑性变形阶段，随着氢化物密度的增大，初始位错更倾向于在氢化物周围扩展。当氢化物密度在1078 ~ 2311 µg/g时，随氢化物密度的增加，屈服强度、应变和杨氏模量升高，这是由于氢化物密度较高时产生了大量位错并造成位错塞积。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-30
Abstract： [Background]: The single-phase reversed flow in inverted U-tubes of steam generator (SG) leads to increasing flow resistance and decreasing heat transfer area, so it is meaningful to study this phenomenon. [Purpose]: The water level of the secondary side in SG can influence the single-phase reversed flow, it is necessary to clarify its influence mechanism from a more general viewpoint. [Methods]: The dimensionless conservation equations were derived first, and the extreme point was obtained based on the equations. Then the effect of the water level of the secondary side under conditions of different lengths, dimensionless resistance number, and dimensionless heat transfer number was analyzed. [Results]: The decrease in the water level leads to the critical point of the single-phase reversed flow gradually approaching the origin, the influence law of the water level is the same under different pipe length conditions. As the water level decreases, the influence of the dimensionless resistance number and dimensionless heat transfer number on the critical point gradually reduces. [Conclusions]: This study theoretically proves that the effect of secondary water level on single-phase reversed flow is not conducive to the occurrence of backflow, and explains the reasons from a mechanistic perspective, which can assist in accident analysis of related nuclear power plants.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-29
Abstract： [Background]: CF series fuel assemblies are the key reactor-core components of the advanced third-generation nuclear power, which are independently developed by China National Nuclear Corporation(CNNC). [Purpose]: The purpose is to analyze the friction force and rod drop performance of CF2 fuel assembly combined control rod drive line moving parts in water and air under different eccentricity. [Methods]: a 1:1 simulated fuel assembly was used in the test with an independently-developed rotatable top cap. The integration of multiple eccentric was initially implemented for scientific and accurate regulation. [Results]: The method to study the performance of the driving mechanism was optimized. The friction force and rod drop performance data of the driving mechanism in water and air at different heights and under different eccentric conditions were obtained. The total rod drop time and the time for rod reaching the buffer increased with the increase of eccentricity while the buffer time was basically constant. The fuel assembly and control rod functioned properly under the maximum eccentricity. The friction did not exceed the allowable limit. And no jamming of control rod occurred under large eccentric condition. [Conclusions]: The experimental results provide an important experimental basis for the design optimization , safe evaluation and software development of CF fuel assembly. The method can be extended to the subsequent CF3 and other fuel assembly scientific research projects.
Key words CF2 fuel assembly, Rod Drop Performance, Friction
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-28
Abstract： The search and localization of unknown radioactive sources is an important research topic in the field of nuclear security inspection and nuclear emergency response. In order to improve the source finding efficiency and adapt to the multi-source environment detection, a particle filtering source finding method integrating the angle of arrival is proposed. Firstly, a hardware platform combining autonomous localization and angle-of-arrival sensing is constructed to introduce position and angle information to the detector; secondly, the angle-of-arrival information is taken into account on the basis of particle filtering, which can dynamically shrink the source searching area and improve the searching efficiency; lastly, the angle-of-arrival-guided robot attitude adjustment is adopted in the path planning of the autonomous source searching, which can enhance the flexibility of the robot in searching for the source. Simulation experiments prove that this method can work correctly and effectively, and tests using radioactive sources further verify the practicality of this method for multi-source search.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-26
Abstract： [Background]: CF series fuel assemblies are the key reactor-core components of the advanced third-generation nuclear power, which are independently developed by China National Nuclear Corporation(CNNC). [Purpose]: The purpose is to analyze the friction force and rod drop performance of CF2 fuel assembly combined control rod drive line moving parts in water and air under different eccentricity. [Methods]: a 1:1 simulated fuel assembly was used in the test with an independently-developed rotatable top cap. The integration of multiple eccentric was initially implemented for scientific and accurate regulation. [Results]: The friction force and rod drop performance data in water and air at different heights and under different eccentric conditions were obtained. The total rod drop time and the time for rod reaching the buffer increased with the increase of eccentricity while the buffer time was basically constant. The fuel assembly and control rod functioned properly under the maximum eccentricity. The friction did not exceed the allowable limit. And no jamming of control rod occurred under large eccentric condition. [Conclusions]: The experimental results provide an important experimental basis for the design optimization , safe evaluation and software development of CF fuel assembly. The method can be extended to the subsequent CF3 and other fuel assembly scientific research projects.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-25
Abstract： Due to only extracting a single feature of the vibration signal of the nuclear power unit, the detection effect of the vibration fault detection method for nuclear power units is poor. Therefore, a vibration fault detection method for nuclear power units based on DBN and multi-sensor data decomposition was designed. Obtain vibration signal data of nuclear power units, smooth and fuse the obtained multi-sensor data, extract multiple features of the vibration signal of nuclear power units under the action of DBN, calculate the sensitivity index and fuzzy entropy of different features, analyze the characteristics of the vibration signal, construct a corresponding vibration fault detection model, and solve the vibration fault signal of nuclear power units. The experimental results show that in the practical application of this method, the AUC-ROC curve area is closer to 1, and the detection effect is better.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-24
Abstract： [Objective] In order to solve the problem that the existing R&D level of mobile reactors cannot meet the relevant requirements of the GB11806-2019 Regulations for the Safe Transport of Radioactive Materials, and cannot obtain transportation permits under the domestic regulatory system, [Method] In this study, according to the relevant requirements of 10 CFR 71 and §12 of the United States, the types of accidents that may be encountered in the transportation of mobile reactors were sorted out, and the most serious accident impact with the tanker truck was used as the design benchmark accident. The accident risk of a mobile reactor with an assumed power of 20 MWt under the design baseline accident is calculated. [Result] It was calculated that the probability of the design benchmark accident of the mobile reactor was 9.7×10-6/year under the condition of annual transportation, and the irradiation dose to the staff under the design benchmark accident after the reactor was cooled for one year was 810mSv. After 5 years of cooling, the radiation dose to workers under the design baseline accident was 590 mSv. [Conclusion] The accident consequences of mobile reactors under the design benchmark accident far exceed the irradiation dose limits for workers in SSG-26 Advisory Material for the IAEA Regulations For the Safe Transport of Radioactive Material and the GB18871 Safety Standards for Protection against Ionizing Radiation and Radiation Sources. Moreover, the probability of occurrence of design benchmark accidents does not meet the screening requirements of 10-7 years of over-design benchmark accidents in China. Therefore, it is necessary to take management measures, including armed escort and route planning, during the transportation of mobile reactors, to reduce the probability of design baseline accidents, and take corresponding protective measures to mitigate the consequences of design baseline accidents to meet the regulatory requirements of the nuclear safety department.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-23
Abstract： In-situ leaching is the main process of natural uranium production in China. The environmental protection problems in the process of construction, operation and decommissioning have been widely concerned. The production of wastes was introduced in this paper, and ten issues of concern were raised, included bleed ratio, drilling mud disposal, monitoring well setting, groundwater baseline determination, abnormal data determination of monitoring well, wastewater treatment, evaporation pool operation, target value determination of groundwater restoration, groundwater remediation and waste minimizaton. the present situation of technology and management was discussed, and the main task and research direction in the field of environmental protection of in-situ leaching of uranium mining were pointed out, in order to provide some suggestions for the sustainable development of in-situ leaching of uranium mining。
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-05-23
Abstract： With the development of digital control technology, the traditional instrument and control system based on analog quantity in nuclear power plant is gradually replaced by full digital technology, and it is possible to use more complex and efficient advanced control technology. Making full use of the advantages of system information in the process of digitization of the whole plant to improve the automation level of nuclear power plant has gradually become the focus of research on pressurized water reactor control system. The control systems of Pressurizer Water Reactor (PWR) nuclear power plant are based on traditional Proportional Integral Derivative (PID) controller. Although there are studies on improving the control performance of PWR NPP control systems by advanced control algorithms, such as neural network control, fuzzy control and model predictive control, most of them only focus on the control system itself without considering the interconnection and coupling among multiple control systems. The operation task of PWR nuclear power plant needs to be coordinated by multiple control systems at the same time, and the effect of improving the overall performance by simply improving the performances of the controllers are limited. [Purpose]: To comprehensively consider the coupling effect amonge control systems, coordinate multiple control systems from the top level to optimize the overall control performances and achieve better task execution results, a setpoint decision optimization system is proposed. [Methods]: The intelligent decision system for PWR control system was optimized based on particle swarm optimization (PSO) method. The decision objective function and operation constraint conditions of the intelligent decision system were proposed. Considering the actual operation of PWR, the system optimized the setpoint offline and the intelligent decision operation was performed online according to the operation condition to provide the directions and amplitudes of the control targets for the underlying control systems. The typical operation process of the PWR NPP was taken as an example to carry out the simulation of the deigned intelligent decision-making system, and the simulation results were analyzed. [Results]: Compared with the control scheme using traditional setpoints, the ITSE (Integral of Time multiplied by the Square Error) of average coolant temperature, pressurizer level, pressurizer pressure and steam generator level decreased by 58.9%, 67.7%, 99.9% and 83.3%, respectively. The peak value decreased by 62.4%, 3.0%, 100% and 66.3% respectively. The simulation results show that the system proposed in this paper can effectively reduce the ITSE and peak value of the system. The overall control performances and safety margin of the control systems of PWR NPP are improved. In practical engineering practice, it can be combined with digital twin technology to use the characteristics of the twin that can synchronously reflect the real state of the system for more accurate online setpoint optimization, so as to achieve better control performance.
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