核动力工程

Nuclear Power AI Applications: Status, Challenges and Opportunities

Zhang Heng, Lyu Xue, Liu Dong, Wang Guoyin, Hang Qin, Sha Rui, Guo Bin

2023, 44(1): 1-8. doi: 10.13832/j.jnpe.2023.01.0001

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Abstract:
In recent years, artificial intelligence (AI) technology has been widely used in the field of nuclear power to promote nuclear power plants to achieve the goal of improving production efficiency, reducing operating costs and improving operating safety through self diagnosis, self optimization and self adaptation. This paper introduces the AI technology often used in the nuclear power field, summarizes its research status in four typical application scenarios of the nuclear industry, namely, intelligent mine, intelligent design, intelligent manufacturing and intelligent operation and maintenance. Finally, it analyzes the challenges and development trends of the application of AI technology in the nuclear power field from three aspects: data samples, network security, and the explanatory nature of deep learning.

Numerical Study on Typical Cell of Fuel Assembly by Turbulence Excitation

Wen Shuang, Zeng Xiehu, Wen Qinglong, Ruan Shenhui, Zhang Ruiqian, Wei Tianguo, Yang Hongyan

2023, 44(1): 9-16. doi: 10.13832/j.jnpe.2023.01.0009

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Abstract:
In order to master the vibration response characteristics of fuel rods in the full length range for fretting wear life analysis of fuel rods, this study is to use the computational fluid dynamics (CFD) method to numerically simulate the turbulence excitation of typical lattice of fuel assembly, and to carry out the transient dynamic analysis of a single rod under different clamping forces through the transient fluctuating pressure distribution on the surface of fuel rod. The results show that: the average turbulent kinetic energy of the upstream section of the grid is about 0.1 m²/s², and the turbulent kinetic energy peaks at 0.65 m²/s² near the grid outlet. Therefore, the existence of grid significantly enhances the turbulence intensity of the flow field, which is the main reason for the turbulence excitation of the fuel rod; the position of the spacer grid with the maximum root mean square amplitude is determined by transient dynamic analysis, and the correlation between the root mean square amplitude and vibration velocity of the grid with the clamping force is established. This study will lay a foundation for the following theoretical calculation and experimental verification of fretting wear.

Calculation Method of Pseudo-critical Line Based on Different State Equations and Division of Pseudo-critical Zone

Zhao Xuebin, Huang Yanping, Liu Shenghui, Zang Jinguang

2023, 44(1): 17-23. doi: 10.13832/j.jnpe.2023.01.0017

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Abstract:
Supercritical fluids undergo drastic changes in thermophysical properties within the pseudo-critical zone. In order to establish the division mode of Pseudo-critical zone, this paper takes supercritical carbon dioxide as the research object. First, the corresponding Pseudo-critical lines are obtained based on three state equations with different accuracy. Then, based on the theory of continuous phase transition, the Pseudo-critical zone is determined by Ehrenfest equation. Finally, according to the division results of the Pseudo-critical zone, matching with the zone where the heat transfer degradation occurs, the internal mechanism affecting the heat transfer degradation behavior in different regions is analyzed.

Experimental Study on Onset of Boiling in Rectangular Narrow Channel with Large Aspect Ratio

Cao Chengque, Kuang Bo, Zhao Yu, Deng Jian, Ding Shuhua, Wu Dan

2023, 44(1): 24-31. doi: 10.13832/j.jnpe.2023.01.0024

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Abstract:
The prediction of onset of nucleate boiling (ONB) in rectangular narrow channel is very important for reactor safety design. For the vertical narrow rectangular channel with the channel size of $50\;\mathrm{ }\mathrm{m}\mathrm{m}\times 3\;\mathrm{ }\mathrm{m}\mathrm{m}\times 1000\;\mathrm{m}\mathrm{m}$, using deionized water as the medium, the position of ONB is confirmed by monitoring the change of wall temperature. The effects of heat flux, mass flow rate, pressure and inlet subcooling on ONB location and wall superheat are studied. Eight existing ONB prediction models are collected and evaluated, and the conclusions are obtained by analyzing the experimental data: The ONB prediction model based on pool boiling and its improved model can not be well applied to rectangular narrow channels, especially for the impact of mass flow rate. Some models developed for prediction of ONB in rectangular channels can reflect the development trend of wall superheat of ONB points with different parameters to a certain extent. However, because the range of experimental parameters is not wide enough, the scope of application and prediction accuracy are still limited. Combined with the main factors affecting the occurrence of ONB in narrow rectangular channels, the analytical solution form suitable for calculating the wall superheat of ONB points in narrow rectangular channels under the conditions of wide spectrum parameters is derived, and the fitting is carried out using experimental data. The deviation between the predicted results of more than 95% of the new relationship and the experimental results is less than ±20%. At the same time, the prediction of ONB data from other relevant published literatures by the new relationship is still in a good error range.

Study on Criticality Safety in the Case of Core Water Inlet Accident During Small Mobile Lead-bismuth Reactor Transportation

Guo Jiaxin, Chen Xiaoliang, Chen Xiaoxian, Xu Jianping, Cheng Yuting

2023, 44(1): 32-36. doi: 10.13832/j.jnpe.2023.01.0032

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Abstract:
For the needs of application scenarios such as islands and remote areas, the safety and feasibility of the whole reactor transportation has become one of the necessary design objectives of small mobile lead-bismuth reactor. Based on the characteristics of small mobile lead-bismuth reactor, the reactivity control method of spectral shift absorber (SSA) is adopted to study the reactivity control scheme to ensure the criticality safety of the whole reactor transportation. MCNP software is used to calculate the effective multiplication factor (k_eff) of fuel rod pellets coated with Gd₂O₃ coating of different thickness during transportation and in case of core water inlet accident, where the coating thickness of 50 μm meets the criticality safety requirements; The burnup characteristics, power distribution and heat transfer of the reactor core after adding SSA are analyzed. The verification shows that it does not affect the normal operation of the reactor core, and the feasibility of this reactivity control scheme is determined.

Experimental Study on Microstructure and Wetting Properties of N36 Zirconium Alloy Oxide Layer

Zhong Lei, Chen Deqi, Yu Hongxing, Liu Hanzhou, Chen Mingjing, Deng Jian, Ding Shuhua, Wu Dan

2023, 44(1): 37-44. doi: 10.13832/j.jnpe.2023.01.0037

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Abstract:
The microstructure and surface wetting properties of domestic N36 zirconium alloy cladding under atmospheric pressure at 600℃, 700℃ and 800℃ are studied. The N36 sample is oxidized, and the oxide layer thickness and surface contact angle are measured. The surface microstructure of the sample is observed by scanning electron microscope (SEM). The type and content distribution of elements are obtained by local scanning of the sample surface by energy dispersive spectrometer (EDS). The effects of oxidation temperature and oxidation time on the surface wetting properties of N36 zirconium alloy are analyzed. The results show that the surface wetting properties of the oxidized sample is enhanced, and the size, depth and internal structure of the surface crack of the oxide layer will affect the surface wetting properties. With the increase of oxidation temperature, the crack size tends to increase. At the same oxidation temperature, the size and number of surface cracks increase with the increase of oxidation time. The study of this paper is helpful to further understand the microscopic characteristics of surface oxidation of N36 zirconium alloy cladding materials.

Analysis of Subcriticality Monitoring Results for First Loading of a PWR

Zheng Junwei, Liu Hang, Cheng Xiongwei, Liu Jikun, Yang Wenqing, Zhang Hengkai

2023, 44(1): 45-53. doi: 10.13832/j.jnpe.2023.01.0045

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Abstract:
An activated secondary neutron source (ASNS) is used to complete the first loading of a PWR. During the first loading, the response test results of the temporary neutron detector (TND) in the reactor are much higher than the simulation results and the counting rate drops significantly. In order to find out the causes of the above problems, the supervision working group analyzed the characteristics of secondary neutron sources and the radiation field established by ASNS, analyzed and verified the influence of subcritical multiplication neutrons of nuclear fuel on TND counting rate, and analyzed the subcritical monitoring data of reactor first loading using ASNS. The results show that the radiation field around TND is a mixed radiation field formed by γ-rays and neutrons. After installing the nuclear fuel assembly between the neutron source assembly and TND, the influence of subcritical breeder neutrons of nuclear fuel on TND counting rate is to increase it; The ASNS decay produced a large amount of γ rays, the peak accumulation of γ pulses from TND in the main amplifier leaded to pulse amplitude distortion. The reason why the TND response test results are much higher than the simulation results is that the pulse amplitude discriminator cannot effectively distinguish the distorted γ pulse and neutron pulse; The reason why the TND count rate drops significantly is that the uranium in the nuclear fuel shields most of the γ rays from ASNS to TND. The operation status of the source range channel and TND meets the requirements of the subcritical monitoring equipment for the first loading program.

Study on Resistance Characteristics of Capillary Flow in Screen Wick Based on CFD Method

Yu Qingyuan, Zhao Pengcheng, Ma Yugao, Zhang Yingnan

2023, 44(1): 54-59. doi: 10.13832/j.jnpe.2023.01.0054

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Abstract:
Screen wick heat pipe is a kind of passive heat transfer equipment based on the principle of two-phase flow phase change cycle. The capillary force and flow resistance in the cycle are closely related to the structure of the screen wick. The study of the resistance characteristics of the screen wick is of great significance to the selection and optimization of the screen wick structure and the improvement of the heat pipe performance. Based on computational fluid dynamics (CFD), a resistance model of capillary flow in screen is established to study the resistance characteristics of capillary flow in multi-layer wire screen wick. The model is used to simulate the capillary lifting experiment, and the relative error between the model and the experimental results is less than 5%. Based on the model, the effects of stacking structure and mesh number (50 mesh, 200 mesh, 400 mesh) on the flow resistance characteristics of screen wick are further analyzed. The results show that the denser the mesh is, the greater the flow resistance is, the viscous resistance coefficient is approximately proportional to the mesh number, and the equivalent inertia resistance increases with the increase of the mesh number. In the low velocity region where Reynolds number is less than 1, viscous resistance plays a dominant role, while in the velocity region where Reynolds number is greater than 1, inertia resistance cannot be ignored; The geometric structure of the screen wick not only affects the flow resistance, but also affects the capillary force. The calculation shows that the capillary pressure and flow resistance of the screen increase with the increase of mesh number, and the capillary performance factor slows down with the increase of mesh number. Considering the process limitation of plain woven screen, 400 mesh screen is ideal.

Experimental and Numerical Simulation Study on Pressure Suppression Characteristics of Small PWR

Qiu Zhifang, Guo Rongda, Cao Xuewu, Yu Hongxing, Sun Hongping, Luo Yuejian

2023, 44(1): 60-66. doi: 10.13832/j.jnpe.2023.01.0060

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Abstract:
In order to study the pressure suppression effect of suppression system of small PWR, a small containment pressure suppression characteristic test device is established. Constant flow mixed gas emission test and variable flow mixed gas emission test are carried out to study the influence of gas-water volume ratio and non-condensable gas on the pressure suppression effect. The experimental result results show that when the gas-water volume ratio is in the range of 2~4.55, the pressure suppression effect gradually increases with the increase of the gas-water volume ratio; The content of non-condensable gas in the mixed gas has a significant influence on the pressure suppression effect. The numerical simulation of the experiment is carried out, and the simulation results can reflect the phenomenon law of the pressure suppression test, but the pressure suppression condensation model still needs to be further optimized to improve the simulation accuracy.

Automatic Processing and Analysis of Submerged Jet Image in Suppression Pool

Peng Chuzhen, Chen Junyu, Wang Shengfei

2023, 44(1): 67-72. doi: 10.13832/j.jnpe.2023.01.0067

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Abstract:
The surge caused by submerged jets in the suppression pool has a major impact on the reliability of the suppression system. And the bubble behavior is directly related to the mechanism of the surge. In order to better identify the bubble behavior in the surge phenomenon, an automatic image processing method of steam submerged jet based on improved watershed image segmentation algorithm is proposed. The differential model is established based on the pixel matrix, the two-dimensional information of the bubble is extracted and the three-dimensional information is approximately estimated, and then the variation of the jet volume is obtained. The bubble condensation period is extracted by using the gray variance information of the image. The algorithm improves the identification accuracy of jet bubbles. Compared with manual calibration, the error of image area recognition is controlled within 10%; It is proved that the relative error of image projection area calculated by the simplified ellipse model is small, and the extraction results of bubble parameters and condensation period are consistent with the experimental characterization and theoretical law.

Ab Initio Molecular Dynamics Calculation of Diffusion Coefficients of Molten Materials

Xu Bo, Zhao Long, Deng Junkai, Li Yang, Guo Kerong, Gong Houjun

2023, 44(1): 73-78. doi: 10.13832/j.jnpe.2023.01.0073

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Abstract:
In order to understand the evolution of the microstructure of the molten materials in the lower chamber of the reactor pressure vessel (RPV) of the nuclear reactor at the later stage of a severe accident, it is necessary to investigate the physical properties of the molten material. The atomic diffusion behavior of molten materials in high temperature liquid is simulated by ab initio molecular dynamics based on the first principles, with the actual materials in the melting process in the melting pool, including fuel pellet UO₂, U-Zr-O materials after melting cladding tubes and U-Fe-O materials after melting stainless steel components as the research objects. The results show that the atomic diffusion coefficients of U, Zr, Fe, O in the high temperature liquid phase are negatively related to the atomic mass, and are less affected by the components at the same temperature, maintaining a relatively stable proportional relationship. The difference of different atomic diffusion coefficients will theoretically lead to the formation of layered structure in the melting pool. Therefore, the diffusion coefficients of various atoms in the above three materials at high temperature can be compared, and the direct quantitative relationship can be determined, which lays a foundation for the further study of the microstructure evolution of molten materials at large scale.

Temperature Field Calculation of Spherical Fuel Element Based on Online Coupling

Gu Chen, He Yanan, Deng Chaoqun, Wu Yingwei, Zhang Jing, Tian Wenxi, Su Guanghui, Qiu Suizheng

2023, 44(1): 79-88. doi: 10.13832/j.jnpe.2023.01.0079

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Abstract:
Due to the complex structure of TRI-Structural Isotropic (TRISO) dispersion fuel elements and its material properties that change under irradiation, it is difficult to determine the equivalent thermal conductivity (ETC) of fuel element under different burnup. In this study, the TRISO particle performance analysis program is developed based on COMSOL software, and compared with the predicted value of BISON program. Then, based on the joint simulation of COMSOL and MATLAB, the calculation method of equivalent thermal conductivity of spherical fuel element is established, and the online coupling calculation between spherical fuel element and TRISO particle model is realized. On this basis, the radial equivalent thermal conductivity distribution and temperature field distribution of fuel element under different boundary temperature and burnup conditions are obtained. The calculation results show that when the fast neutron flux reaches 3×10²⁵ m^–2, the equivalent thermal conductivity of TRISO decreases by about 20%, and the equivalent thermal conductivity of fuel decreases by about 15 W/(m·K). In order to demonstrate the effectiveness of this method, the equivalent thermal conductivity of fuel is calculated by differential-effective medium theory model (D-EMT). The predicted value of spherical fuel center temperature is about 25 K lower than that of this method. The research method in this paper can more truly reflect the temperature field change of spherical fuel element in the reactor.

Study on Uniform Corrosion Behavior of 600 Alloy and 304 Stainless Steel in Supercritical Carbon Dioxide Environment

Liu Zhu, Long Jiachen, Guo Xianglong, Su Haozhan, Wang Peng, Duan Zhengang, Ma Zhaodandan, Zhang Lefu

2023, 44(1): 89-96. doi: 10.13832/j.jnpe.2023.01.0089

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Abstract:
In order to select the structural materials that can be used in the supercritical carbon dioxide nuclear reactor, the uniform corrosion behavior of two kinds of alloys (600 alloy and 304 stainless steel) used in the traditional nuclear reactor in the supercritical carbon dioxide environment at 650℃ and 20 MPa is studied through experiments. The corrosion kinetics of the material is evaluated by weight gain method, and the morphology, structure and chemical composition of the oxide film are analyzed by scanning electron microscope, energy dispersive spectrometer and X-ray diffractometer. The results show that the corrosion weight gain of the two materials obeys the parabolic growth law, and the corrosion resistance of 600 alloy is better than that of 304 stainless steel. After 500 h corrosion, the oxide thickness on the surface of 600 alloy is about 5 μm, the main component is NiCr₂O₄, which is compact in structure and protective. No obvious carburization is found in its oxide film and matrix; After 500 h corrosion, the oxidation film on 304 stainless steel surface can reach about 45 μm. It is a double-layer structure, the outer layer is Fe₃O₄, and the inner layer is NiFeCrO₄. The structure is loose, and significant carburization occurs. This study reveals the corrosion mechanism of the above materials in supercritical carbon dioxide, and provides valuable data support for the selection of structural materials for supercritical carbon dioxide nuclear reactors.

Study on Bearing Wear Life Analysis Method of Nuclear Reactor Control Rod Rotating Device

Zhou Xu, Peng Hang, Du Hua, Deng Qiang, Zhang Zhiqiang, Liu Yanting

2023, 44(1): 97-103. doi: 10.13832/j.jnpe.2023.01.0097

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Abstract:
In order to establish the bearing wear life analysis method of nuclear reactor control rod rotating device and optimize the existing experimental bearing life analysis means of rotating device, by constructing the bearing mechanical analysis model under operating conditions, the contact mechanical characteristics of raceway-roller are obtained. Through the ring control theory, combined with the operating characteristics of the rotating device, the motion characteristics between the bearing roller and the inner and outer raceways are obtained. The wear analysis model of the bearing under the operating condition of the rotating device is constructed based on the Archard wear model, and the wear characteristics between the bearing raceway and the roller are obtained by using the wear iterative method. The research results show that the bearing wear life obtained by the bearing wear life analysis method established in this paper is in good agreement with the test results, and can be used for the bearing wear life analysis of the control rod rotating device in nuclear reactors.

Study on the Operation Characteristics of the Primary Main Circulating Pump of Sodium-cooled Fast Reactor under the Switching Condition of Off-site Main and Auxiliary Power Supplies

Guo Xiaoxian, Gu Jipin, Zhu Hao, Liu Yang, Li Taifeng, Zhang Hu

2023, 44(1): 104-108. doi: 10.13832/j.jnpe.2023.01.0104

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Abstract:
In order to avoid reactor protection shutdown or motor auxiliary winding startup due to the operation of primary main circulating pump (hereinafter referred to as main pump) during the off-site main and auxiliary power supplies switching, the idling characteristics and speed control of main pump are analyzed, the fast speed tracking and frequency search speed tracking modes of the main pump under fast and slow switching conditions are studied respectively, and the prediction model of the minimum speed of the main pump under different modes is given. It is analyzed that the minimum speed of main pump is 708.3 r/min under fast switching condition (power off for 1.5 s) and 341.2 r/min under slow switching condition (power off for 10 s). On the main pump water bench, 1.5 s and 10 s power off tests are used to simulate the fast switching and slow switching conditions of the off-site main and auxiliary power supplies. The test results show that the minimum speed of main pump is 689 r/min under fast switching condition; The minimum speed of main pump under slow switching condition is 346.7 r/min. The predicted minimum speed is in good agreement with the test value, and the deviation is less than 3%.The test verifies the operation characteristics of the main pump under the switching condition of the main and auxiliary power supplies, which can realize that the fast switching does not lead to the reactor protection shutdown and the slow switching does not lead to the startup of the auxiliary winding, which is of guiding significance for the safe operation of the reactor.

Study on Failure Mechanism of Pressurizer Surge Line and Manhole Structure under LOCA

Yu Hang, Zhao Xinwen, Fu Shengwei, Zhu Kang

2023, 44(1): 109-117. doi: 10.13832/j.jnpe.2023.01.0109

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Pressurizer is an important equipment for pressure control and protection in nuclear reactors, and the huge shock generated by loss of coolant accident (LOCA) may cause structural failure of its critical parts. The three-dimensional transient numerical simulation of pressurizer surge line’s flow heat transfer and structural stress, the temperature distribution and sealing performance of manhole structure under the small break LOCA are carried out by the multi-field coupling method, and the failure mechanism is analyzed. The results show that the hot fluid rapidly flowing into the surge line forms a huge instantaneous load, causing a short period of strong vibration of the pipe, the middle part of the pipe deformation is the largest, which may destroy the pipe support structure; The equivalent stress of each part increases rapidly, and concentrated stress occurs at the connecting pipe part of the main pipe. Large stress fluctuation will affect its service life; The manhole structure has a large uneven temperature distribution, and the sealing performance of the gasket under the sealing structure changes the most. The contact pressure of the inner and outer sealing surfaces drops below the design sealing specific pressure before and after 100 seconds, which means leakage occurs. According to the analysis results, this paper puts forward some suggestions for improving the structure of surge line and manhole, which can provide technical reference for accident mitigation after small break LOCA in marine nuclear power plant.

Dynamic Reliability Analysis of CRDM Pressure Shell under Multiple Failure Modes

Chen Peng, Zhu Yizhou, Zhang Kai, Xie Yongcheng

2023, 44(1): 118-123. doi: 10.13832/j.jnpe.2023.01.0118

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Abstract:
To investigate the structural reliability of the control rod drive mechanism(CRDM), multiple failure modes of the CRDM pressure shell is considered, while the structural dynamic reliability model related to the CRDM step number is built by the stress-strength interference theory. The dynamic distribution model of the stress amplitude under the strength failure mode is represented by the order statistic. Based on Miner cumulative damage theory and fatigue equivalent stress distribution model, the relationship between fatigue life, cumulative damage distribution and the number of step impact loads is established. The results show that under the action of step impact load, the early structural reliability of pressure shell is mainly determined by the reliability of strength failure mode. When the step action reaches a certain number of times, the failure rate of fatigue failure mode starts to increase significantly; Compared with the fatigue failure mode, the reliability of the strength failure mode is more sensitive to the change of the mean stress.The results can provide reference for the reliability design and maintenance management of CRDM pressure shell.

An Instantaneous Method for Reactors Risk Calculation Applies to Level Three Probabilistic Safety Assessment

Cheng Jie, Tang Xiuhuan, Chen Shanqi, Wang Jin, Li Yazhou, Li Jie, Wang Fang, Jiang Jieqiong

2023, 44(1): 124-128. doi: 10.13832/j.jnpe.2023.01.0124

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Abstract:
Risk monitor is a very important technical basis for nuclear reactor safety supervision and nuclear emergency decision-making. Conventional risk monitors only support the core meltdown risk calculation based on the level-1 probabilistic safety assessment. Focusing on the real-time risk monitoring of reactor based on the level-3 probabilistic safety assessment, an improved method is proposed in this paper for risk calculation. For the actual configuration of reactor operation, the accidents frequencies are calculated based on real-time risk model and on-line condition monitoring, and supports the real-time calculation of released source term and off-site dose by accident categorization and dose factors. The case study of reactor risk model shows the effectiveness and provides instantaneous calculation for core damage frequency, large release frequency and off-site risk. The proposed method can provide technical support for nuclear reactor safety regulation and nuclear emergency.

Application of Power Function Model of Gas Leakage in Tightness Test of Double-wall Containment Annulus

He Rui, Shen Dongming, Chen Wei, Huang Xiaoming, Zhang Bo

2023, 44(1): 129-133. doi: 10.13832/j.jnpe.2023.01.0129

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The tightness of the annulus of the nuclear power unit with double-wall containment is an important guarantee of nuclear safety, which needs to be tested before the unit is loaded. The power function model of gas leakage has a wider range of pressure application and higher accuracy than the quadratic function model for calculating the leakage of civil buildings. In this paper, based on the power function model of gas leakage, the calculation model of annulus leakage rate and the power function fitting scheme with annulus negative pressure correction are derived, and the calculation model is verified by the measured data of a double-wall containment annulus tightness test. The results show that the calculation model in this paper has higher accuracy than the quadratic function fitting scheme with annulus negative pressure correction and the linear fitting scheme used by a foreign company, and is more consistent with the actual leakage of double-wall containment.

Study on the Smoothness of the Flow Channel of Rapid Pressure Relief Pipeline in Severe Accident of HPR 1000

Lu Xifeng, Wang Xinjun, Ai Honglei, Lyu Yongbo, He Feng, Li Bingjin, Zhang Quan

2023, 44(1): 134-140. doi: 10.13832/j.jnpe.2023.01.0134

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Both pipelines and the equipment experience extreme high temperature and pressure in sever accident of the nuclear power plant. As the only way to relieve pressure in case of severe accident, it is very important to ensure the smoothness of the flow channel of rapid pressures relief pipeline. In this paper, the study on the smoothness of the flow channel of rapid pressure relief pipeline in severe accident of HPR 1000 is carried out. Heat transfer analysis is done for rapid pressure relief pipeline and pressurizer in severe accident. The temperature changes of pipeline and pressurizer are obtained. The deformation process of rapid pressure relief pipeline in severe accident is simulated by elastic analysis method. The relationship between temperature and deformation is obtained. The 3D model is established, and the material non-linearity is introduced. The simulation study on the position of spring support and damper of rapid pressure relief pipeline is carried out, and the status of spring support and damper of pipeline under severe accident is obtained. The influence of high temperature creep on pipeline integrity is analyzed for the case of temperature higher than 450℃. 10 positions with the largest deformation on the rapid pressures relief pipeline are selected to study the residual area of the pipeline section, and the minimum residual area ratio and the minimum flow area of the pipeline under severe accidents are obtained. The research results show that the flow channel of the rapid pressures relief pipeline can still ensure the smoothness under the severe accident of HPR 1000, and the rapid pressures relief pipeline of HPR 1000 can ensure that the reactor core will not melt.

Research on Calculation Method of Leakage Rate of Graphite Gasket Seal

Jiang Lu, Fu Xiaolong, Zhang Liping, Zhang Ying, Yu Mingda, Tian Jun

2023, 44(1): 141-147. doi: 10.13832/j.jnpe.2023.01.0141

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Abstract:
In order to improve the seal analysis and design method of nuclear grade main equipment, a numerical analysis model of the seal is established based on the voltage regulator manhole seal structure, and the contact stress of the graphite gasket seal is analyzed and studied; The theoretical prediction model of graphite gasket seal mass leakage rate is established by combining the parallel circular plate flow model and porous medium seepage model; Based on the theoretical prediction model, the mass leakage rate under design conditions, test conditions and startup transient conditions is calculated, and the main influencing parameters are analyzed and discussed. The results show that the contact stress of the graphite gasket seal is uniformly distributed along the circumferential direction, while the contact stress in the radial middle region of the graphite ring is slightly lower than that on both sides of the graphite ring; In the transient state of rising temperature and pressure, the seal contact stress decreases with time, and the seal mass leakage rate is negatively correlated with the contact stress. Increasing the seal contact stress can reduce the mass leakage rate, but the efficiency decreases gradually. Reducing the roughness can significantly reduce the mass leakage rate.The analysis method in this paper can provide an important reference for seal leakage rate analysis and tightness evaluation of nuclear grade main equipment.

Study on the Effect of Boron-coated Thickness of Boron-coated Proportional Counter Tube on the Detection Efficiency

Zhu Chaoyang, Li Litao, Wang Zhentao

2023, 44(1): 148-153. doi: 10.13832/j.jnpe.2023.01.0148

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Abstract:
The boron-coated proportional tube is a key device for monitoring the neutron fluence outside the reactor pressure vessel. The boron-coated thickness of the boron-coated proportional tube affects its intrinsic detection efficiency. Simulating the transport process of the ⁷Li and α generated by ¹⁰B(n,α)⁷Li in the boron layer with different thicknesses, and calculating the overall ion emission rate on boron layer interface and further the calculation method of the intrinsic detection efficiency of the boron-coated proportional counting tube is proposed. The results of simulation and calculation show that: when ¹⁰B(n,α)⁷Li is under unit nuclear reaction rate, the overall ion emission rate on boron layer interface is positively linearly correlated with the thickness of the boron layer when it is less than 1.5 μm, and increases more slowly with the increase of the thickness of the boron coating when it is greater than 1.5 um, finally the overall ion emission rate on boron layer interface reaches the maximum─1.3×10⁻⁴(cm²·s)⁻¹ when the thickness of the boron coating is 3.6 μm. Based on the results of the overall ion emission rate, the relationship curve between the boron coating thickness and the intrinsic detection efficiency of the boron-coated proportional counting tube was further calculated, which can provide reference value for selecting an appropriate thickness of boron coating to define the best detection efficiency in the manufacturing of the boron-coated proportional tube.

Study on Method of Pneumatic Test for Pressure Vessels in NPP Nuclear Island

Zhao Weihua, Shao Chunbing, Feng Huixing, Xie Jianfang, Jiang Kuirong, Li Jiuqiang

2023, 44(1): 154-158. doi: 10.13832/j.jnpe.2023.01.0154

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According to the requirements of the in-service inspection code and program, periodic hydrostatic test should be carried out for the nuclear island pressure vessels of the French-system nuclear power plant. However, due to the reasons of system design, some vessels can not be subject to hydrostatic test with liquid, but can only be subject to pneumatic test. In this paper, the implementation requirements of pneumatic test in domestic and foreign codes are compared and analyzed, and combined with the process of pneumatic test in the stage of nuclear island installation, the test pressure, test medium and acceptance standard of nuclear island pressure vessel are selected. At the same time, combined with the risk analysis and radiation protection requirements of the vessel hydrostatic test, the protective measures for the pneumatic test are formulated. According to the above test parameters and risk protection measures, the pressure vessel pneumatic test is successfully carried out in the nuclear island of a nuclear power plant, providing important reference for the pressure vessel pneumatic test of the nuclear island in the subsequent in-service phase.

Design and Development of Intelligent Operator Support System for Nuclear Power Plants

Xu Renyi, Wang Hang, Peng Minjun, Liu Yongkuo, Yu Yue, Ai Xin

2023, 44(1): 159-166. doi: 10.13832/j.jnpe.2023.01.0159

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In order to reduce the nuclear power plants operator ’s work pressure and psychological burden when dealing with abnormal and accident conditions, and avoid misjudgment or misoperation, this research has designed and developed a set of nuclear power plant intelligent operator support system (NPPIOSS) for the reactor primary circuit system and its key auxiliary system, which integrates data acquisition and storage, online monitoring, fault detection and diagnosis, key parameter trend prediction and other functions. The simulation results show that NPPIOSS system can accurately detect and identify typical faults of nuclear power plant, so as to help operators accurately judge the state of power plant and reduce human error. Therefore, NPPIOSS system can assist operators in subsequent decision-making after the failure of nuclear power plant, so as to improve the operation safety of nuclear power plant.

Study and Verification of Diagnosis Method for Secondary Neutron Source Breakage

Zheng Junwei, Liu Hang, Li Wenhai, Liu Jikun

2023, 44(1): 167-170. doi: 10.13832/j.jnpe.2023.01.0167

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Abstract:
The secondary neutron source used in the PWR nuclear power unit has a risk of breakage. Physical inspection of the status of the secondary neutron source is not possible at reactor power operation condition. According to the activation characteristics of the secondary neutron source, ¹²²Sb and ¹²⁴Sb are used as the characteristic nuclides to diagnose the damage of the secondary neutron source. The feasibility of the method for diagnosing the damage of secondary neutron source by using the on-line monitoring data of γ radioactivity of primary coolant and the specific activity of ¹²²Sb and ¹²⁴Sb in primary coolant is analyzed, the secondary neutron source damage diagnosis process is designed, and the secondary neutron source damage problem of the second generation improved 1000 MW PWR nuclear power unit is diagnosed using the above diagnosis methods. The verification results show that the specific activity change trend of ¹²²Sb and ¹²⁴Sb obtained by the primary coolant sampling analysis after the damage of the secondary neutron source should be consistent with the γ radioactivity change trend of the primary coolant monitored by the nuclear radiation monitoring equipment.Therefore, the secondary neutron source damage diagnosis method proposed in this study is effective.

EMI Analysis and Location of I&C Equipment in Nuclear Power Plant

Huang Su, He Xiaodong, Han Dadi, Liu Xinyuan, Li Benrong

2023, 44(1): 171-176. doi: 10.13832/j.jnpe.2023.01.0171

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Abstract:
Electromagnetic interference (EMI) has a significant impact on the I&C equipment of nuclear power plants. In some nuclear power plants, EMI leads to false alarm of protection equipment, misoperation or refusal to operate of protection system, resulting in false shutdown and even reactor shutdown accident. Therefore, it is necessary to make targeted solutions for each interference source of EMI. Taking the interference of temperature measuring elements in a nuclear power plant as an example, this paper uses the method of layer by layer evolution to find out the interference source, analyze the interference characteristics, and finally propose a solution. Measures such as improving the grounding of important protection signals, and issuing the regulations on EMI prevention management of operating units have greatly reduced the impact of EMI on operating equipment in nuclear power plants.

Phased Array Ultrasonic Inspection Technology for BOSS Weld Overlaying Repair in Nuclear Power Plant

Tang Jianbang, Wu Yukun, Yu Zhe, Zhu Jiazhen, Kang Zhiping

2023, 44(1): 177-182. doi: 10.13832/j.jnpe.2023.01.0177

Abstract(145) HTML (34) PDF(30)

Abstract:
Nuclear power plant reactor refueling pool and spent fuel pool cooling and treatment system (PTR) and equipment circulating cooling system (RRI) contain a large number of BOSS welds, whose safety and reliability directly affect the safety status of stored nuclear fuel. Defect troubleshooting and on-line repair are the key points and difficulties of in-service inspection and supervision of nuclear power plants. In this paper, a set of special phased array ultrasonic probe and inspection process are designed to meet the special requirements and inspection difficulties of online overlaying repair of BOSS weld and the limitations of radiographic inspection. The test results meet the requirements of overlaying repair, and the nondestructive inspection methods and in-service inspection and supervision strategies for BOSS weld overlaying repair in nuclear power plants are formulated.

Method and Device Development of Output Test for Pyrotechnic Actuator on Squib Valve in AP/CAP Nuclear Plant

Li Wei, Zhou Qiangqiang, Zhang Ji, Yan Guohua, Yu Zhaohui

2023, 44(1): 183-186. doi: 10.13832/j.jnpe.2023.01.0183

Abstract(105) HTML (19) PDF(14)

Abstract:
According to the test program of the squib valve in AP/CAP nuclear power plant, the output test of the pyrotechnic actuator of the squib valve is required, and a special test device is developed to complete the test. The principle and method of output test of the pyrotechnic actuator of the squib valve in AP/CAP nuclear power plant and the development content of test device are introduced. The developed test device ignites the pyrotechnic actuator of the squib valve to produce high temperature gas. By collecting the gas pressure data contained in the device and drawing and analyzing the pressure-time curve, the output of the pyrotechnic actuator is evaluated. Different types and specifications of pyrotechnic actuators are tested respectively. The results show that the test device is applicable to the test of the pyrotechnic actuator of the squib valve with multiple types and specifications, and its performance is reliable and operation is convenient.

Treatment of Simulated Radioactive Cobalt Containing Wastewater by Ultra-low Pressure Reverse Osmosis Process

Zhang Wenxiu, Zhang Guanghui

2023, 44(1): 187-191. doi: 10.13832/j.jnpe.2023.01.0187

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Abstract:
The radioactive wastewater produced by the operation, maintenance and decommissioning of nuclear power plants is seriously harmful to the social environment and life and health safety. A pilot plant (144 L/h) is used to study the removal effect of cobalt (Co) from simulated radioactive wastewater by ultra-low pressure reverse osmosis (ULPRO) process. The effects of operating pressure, recovery rate, influent concentration, coexisting ions and natural organic pollutants on it are determined. The results show that the removal rate of Co is positively correlated with the operating pressure and negatively correlated with the recovery rate, the removal rate of Co is constant at 99.5% with the increase of influent concentration, and 10 mg/L is the critical concentration for its change; As for coexisting ions, only Ca²⁺^{will inhibit the removal of Co, and the promotion effect of other ions is in the order of Na⁺>Mg²⁺, SO₄²⁻>Cl⁻>NO₃⁻; Organic pollution reduces the membrane flux by 9.4%, but the removal rate of Co increases to 99.97%. The experimental results show that the ULPRO process has stable operation effect and high removal rate for the treatment of simulated radioactive wastewater containing Co, which can provide guidance for industrial application.}

Design and Optimization of Electromagnetic Ultrasonic Longitudinal Guided Wave Transducer for Heat Exchange Tube of Stream Generator

Wang Libo, Fang Zhihong, Wang Fangfang, Wang Fei, Han Zhixiong, Zhu Yi, Zhao Yajun

2023, 44(1): 192-197. doi: 10.13832/j.jnpe.2023.01.0192

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Abstract:
The special spiral structure of the heat exchange tube of the steam generator of the high-temperature gas-cooled reactor makes it difficult for the traditional external electromagnetic ultrasonic guided wave transducer to be effectively detected. Aiming at the defect detection of stainless steel heat exchange tube of stream generator, a new built-in electromagnetic ultrasonic longitudinal guided wave transducer is developed, a finite element multi physical field coupling model is established, the static magnetic field distribution of ferromagnetic structure of the transducer is studied, and the longitudinal guided wave excited by the transducer is simulated in time domain. The results show that: the use of the squeezed magnetization transducer structure can ensure that the vertical magnetic field near the coil is much larger than the horizontal magnetic field, and can efficiently excite a single mode of longitudinal guided waves inside the tube. The optimized probe can detect through-hole defects with a diameter of 5mm and circular groove defects with a length × width × depth of 20 mm×1.5 mm×1.2 mm. Therefore, the new electromagnetic ultrasonic longitudinal guided wave transducer can effectively excite the longitudinal guided wave, and is expected to be applied to the in-service defect detection of the heat exchange tube of the steam generator of the high-temperature gas-cooled reactor.

Research on Data-driven Human Error Causal Mechanism in Nuclear Power Plants

Chang Meng, Wang Ru, Li Pengcheng, Liu Zhen, Liu Xiaohui

2023, 44(1): 198-203. doi: 10.13832/j.jnpe.2023.01.0198

Abstract(158) HTML (62) PDF(31)

Abstract:
In order to identify human error mechanism, the Organization-oriented human error analysis (OTHEA) technique is used to analyze the 137 domestic human factor event reports that occurred in nuclear power plants (NPPs) from 2010 to 2017. By using the methods of correlation and factor analysis, the correlation between the influencing factors of human error is identified, and the scene that triggers human error is identified. On this basis, the causal mechanism model of human error is established to reveal the mechanism of human error. The results show that the main combination modes affecting human error are knowledge and experience level, information display quality, pressure level, attention and vigilance, and safety attitude. The knowledge and experience level is mainly affected by the level of training and communication; The information display quality is mainly affected by the technical system, human-machine interface, procedures and organizational design; The pressure level is mainly affected by tasks, procedures, man-machine interface, technical system, organizational design, knowledge and experience level and information display quality; Attention and vigilance is mainly affected by the work environment, work organization and management, and the information display quality; The safety attitude is mainly affected by the organizational safety culture, work organization and management and the quality and ability of operators. The above research can provide a theoretical basis for the precise prevention and control of human error in nuclear power plants, and improve the safety level of nuclear power plants.

Research on Space Mapping Control Algorithm of Master-slave Equipment in Teleoperation Mode of Underwater Robot

Yang Junhao, Wang Bingyan, Yu Zhiwei, Pu Yaozhou, Li Hao, Chen Qian, Ma Shanlin

2023, 44(1): 204-209. doi: 10.13832/j.jnpe.2023.01.0204

Abstract(223) HTML (71) PDF(46)

Abstract:
In order to complete the underwater cleaning work safely and efficiently, this paper proposes an underwater robot cleaning device for nuclear power plants, which is based on UR5 underwater serial robot. The control system of the robot is introduced, and the joint space and workspace of the master and slave are analyzed. At the same time, in the teleoperation mode, the master-slave joint space mapping algorithm and workspace mapping algorithm are analyzed, and teleoperation experiments are carried out based on the two mapping control algorithms, and the test results are compared comprehensively. Finally, according to the needs of the underwater cleaning task of the nuclear power plant, the joint space mapping algorithm with fast response is selected for the master-slave control in the teleoperation mode.

Simulation Study on Switching Operation Law of Feed Water Pumps in Multi-pump Parallel Feed Water System

Tian Peiyu, Li Yi, Liang Tiebo, Wang Changshuo, Fang Huawei

2023, 44(1): 210-216. doi: 10.13832/j.jnpe.2023.01.0210

Abstract(147) HTML (21) PDF(29)

Abstract:
As one of the main subsystems of the nuclear power system, the switching operation law of the feed water pumps in the multi-pump parallel feed water system is critical to the system operation economy and system operation characteristics. In this study, the simulation model of multi-pump parallel water supply system is established by using the system simulation support software APROS, and the accuracy of the model is verified according to the rated design value. Based on this, the switching operation law and system dynamic characteristics of feed water pump are studied through the simulation of linear rising and falling load under different switching conditions. The research results show that, for this research object, when the switching point under high load condition is selected as 70% of the rated flow, and the switching point under low load condition is selected as 30% of the rated flow, good system dynamic response can be obtained, and high economy of feed water pump operation can be maintained. In addition, the low load condition is more sensitive to the disturbance caused by the switching of feed water pump. Under low load conditions, if the switching conditions are not selected properly, the system will trigger overpressure discharge in the process of load reduction.

Mechanical Behavior Analysis of Heat Pipe Reactor Core Matrix Structure at High Temperature

Tian Jun, Su Dongchuan, Li Hui, Xiong Furui, Liu Changjun, Bi Penghua, Tan Jianping

2023, 44(1): 217-221. doi: 10.13832/j.jnpe.2023.01.0217

Abstract(163) HTML (23) PDF(41)

Abstract:
In order to study the thermal stress failure behavior of the core matrix of the heat pipe reactor at high temperature, two high temperature test schemes were developed based on the simplified porous matrix structure and the design parameters of Megapower 5 MWt heat pipe reactor: normal condition and abnormal condition, in which the failure of a single heat pipe is considered under abnormal condition. The macroscopic test results show that there is no obvious deformation and failure of the matrix structure. Combined with numerical analysis method, the temperature distribution and stress-strain response of the matrix structure under two conditions are obtained, which further explains that the static strength failure and plastic collapse failure of the matrix structure will not occur under the test conditions. This study lays a foundation for defining the strength design criteria of the core matrix structure of the heat pipe reactor.

2023 Vol. 44, No. 1