Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-23
Abstract： In-Service Inpection Rules for Mechnical Componets of PWR(Presurized Water Reactor) Nuclear Islands (RSE-M) is the rules for mechanical equipments during nuclear in-service. With the manufacturing process, welding level, Non-destructive testing method and accumulation of experimental data, RSE-M have gone through multiple versions. Comparing with RSE-M code Ver.1997, some changes for vessel crack growth calculation in addendum version and Ver.2018 are introduced. In the actual example, the crack growth is given by RSE-M code Ver.1997 and Ver. 2010 at the end of life, respectively. The results showed that the method in Ver.2010 is more accurate than in the Ver.1997, which can get more real life prediction and save operation cost of power station.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-23
Abstract： [Background] As a typical representative of passive safety technology, the automatic depressurization system ( ADS ) accelerates the reactor primary loop depressurization after the accident, connects the high-pressure, medium-pressure and low-pressure safety injection systems, and maintains the core cooling. [Purpose] In order to study the opening and closing characteristics of the pressure relief valve of the automatic depressurization system and its influence on the reactor systems. [Methods] Based on the system analysis program, China Advanced Pressurized Water Reactor is taken as the research object, and the typical ADS trigger accident is taken as the initial event. The different opening speeds of the first three valves of ADS and the closing conditions of the fourth pressure relief valve of ADS are simulated, and the response of each system under different working conditions is analyzed. [Results] The results show that, the opening speed of the ADS1-3 valve cannot significantly affect the pressure relief characteristics of the primary loop; the ADS-1 pressure relief valve uses a quick opening method which helps the sprinkler reach the stable critical jet state faster; the ADS-2/3 pressure relief valve uses a slow opening method which is helpful to avoid "sharp" peak points in the pipeline flow curve and reduce the impact of the spraying process on the pipeline and sprinklers while satisfying the economic principle; ADS-4 is crucial for small break LOCA accident and also necessary for the injection of pressure tank at the later stage of the accident. [Conclusions] Through the simulation analysis of the opening and closing characteristics of the ADS pressure relief valve, it provides a reference for the design of the automatic depressurization system, and also provides theoretical and data support for the safety analysis of advanced nuclear power plants.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-21
Abstract： [Background] The helium-xenon mixture closed Brayton cycle system has significant advantages of high cycle efficiency, small specific power volume, and large output specific power, and has broad application prospects in the field of special nuclear power. The coupling of a helium xenon mixed working fluid closed Brayton cycle system with a nuclear reactor to form a megawatt level special nuclear reactor power supply can effectively adapt to high-power energy supply scenarios such as deep space exploration, star catalog nuclear power supply, and deep-sea submersibles. At present, there are few tools available for steady-state simulation research of helium xenon closed Brayton cycles, making it difficult to grasp the steady-state characteristics of the system. [Purpose] In order to grasp the characteristics of equipment and systems before actual engineering design and operation, and reduce research costs, it is necessary to develop relevant helium xenon closed Brayton cycle steady-state simulation tools. [Methods] By establishing component models of core equipment such as reheaters, coolers, turbines, and compressors in the steady-state thermodynamic system of helium xenon closed Brayton cycles, a simulation tool for steady-state analysis of helium xenon Brayton cycles was obtained. And by comparing the design values of the Prometheus project in the United States with the calculated values of simulation software built under the same conditions, the accuracy of the simulation software is verified. The construction model was validated and the effects of parameters such as the highest temperature, lowest temperature, highest pressure, and total thermal conductivity of the regenerator on the cycle were analyzed. The influence of parameters and component performance at various points of the space helium xenon closed Brayton cycle system with an output power of 200kW on system efficiency and specific power was analyzed. [Results] The calculation results of the helium xenon thermodynamic cycle calculation model constructed in this article are in good agreement with the "Prometheus" design values, with a maximum node parameter error of 0.212% and a maximum system parameter error of 3.419%, all of which are within an acceptable range, proving the accuracy and reliability of the helium xenon closed Brayton cycle model. The results also show that there is an optimal pressure ratio for both system efficiency and system specific power, and it is reasonable to use the pressure ratio at the maximum system efficiency in the design; The higher the maximum temperature of the cycle, the lower the minimum temperature, and the greater the system efficiency and specific power. The impact of the minimum temperature of the cycle on the cycle is more significant; The impact of pressure on circulation is not significant. The higher the pressure, the slightly smaller the system efficiency and specific power of the system; The higher the total thermal conductivity of the regenerator, the greater the system efficiency. The specific power of the system remains unchanged, and the higher the pressure ratio, the less significant the impact of the total thermal conductivity of the regenerator on the cycle efficiency. [Conclusions] This study provides a reference and basis for the design and optimization of helium xenon closed Brayton cycles.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-06-20
Abstract： In recent years,the rate of energy cleaning in China has been continuously improved. The output of large-scale new energy represented by wind power and photovoltaic has obvious volatility and uncertainty. While the demand、growth and total amount of power grid load fail to reach the expected level, which poses new tests and challenges to the peak load regulate ability of the power grid. In order to ensure the safety of the power plant and the power grid, the participation of nuclear power units in peak load regulation is becoming increasingly frequent,this poses new challengs to the nuclear safety of nuckear power units, and may even affect the long-term development of chinese nuclear power industry. This paper analyzes the possible effects of nuckear power units participating in power grid peak load regulation on the safety and system equipment of the power plant, focusing on the effects of peak load regulation on the safety、reliability、economy and the treatment of three wastes of the power plant, and puts forward reasonable suggestions and improvement directions, providing a reference for the participation of peak load regulation of nuclear power units in China.
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 2024-06-12
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.
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