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2022 Vol. 43, No. 1

Special Contribution
Evelopment Characteristics and Inspiration of Marine Nuclear Power
Lu Chuan, Wang Zhonghui, Yu Junchong
2022, 43(1): 1-6. doi: 10.13832/j.jnpe.2022.01.0001
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The marine nuclear power technology of the United States and Russia has been leading the world for a long time, and their development experience and technical context have high reference value. Through the analysis and research on the main development process and technology of marine nuclear power in the United States and Russia, this paper innovatively summarizes the common development laws of marine nuclear power in the United States and Russia, such as basic type of reactor system, general test platform and differential configuration from the aspects of technical route and trend, a series of common and differential characteristics followed by marine nuclear power technology in the United States and Russia are excavated and refined, which can provide some reference and enlightenment for the development of marine nuclear power.
Reactor Core Physics and Thermohydraulics
Research and Verification of High-Fidelity Physics Calculation Method for Hexagonal Reactor
Chen Junji, Liu Zhouyu, Cao Lu, Cao Liangzhi
2022, 43(1): 7-14. doi: 10.13832/j.jnpe.2022.01.0007
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Russian commercial pressurized water reactor VVER and most experimental reactors adopt hexagonal compact lattice arrangement. To realize the high-fidelity numerical simulation analysis of the VVER and the hexagonal experimental reactor, in this paper, the high-fidelity calculation method and program development of hexagonal core are carried out based on NECP-X. First, the global-local coupled resonance self-shielding calculation method is extended to the hexagonal core to realize the high-precision resonance calculation of hexagonal core fuel rods; Second, based on 2D/1D coupled transport calculation method, the high-fidelity calculation method of hexagonal core is studied; Finally, in order to improve the calculation efficiency of the full-core calculation, the coarse-mesh finite-difference (CMFD) acceleration method based on the loosely coupled unstructured grid of domain decomposition is studied, which can realize pin-by-pin coarse-mesh finite-difference acceleration based on rectangular, hexagonal and other polygonal cells. In order to verify the accuracy and efficiency of the hexagonal core high-fidelity calculation method, the hexagonal C5G7 benchmark problem is calculated, and the calculation accuracy of the hexagonal transport calculation method and the acceleration effect of the CMFD method are analyzed; NECP-X is applied to the two-dimensional full core calculation of a pulsed reactor in Xi’an. The comparison with the results of Monte Carlo program shows that the eigenvalues and power distribution calculated by NECP-X have high accuracy. Therefore, the high-fidelity calculation method of hexagonal core established in this paper can be applied to the analysis and calculation of hexagonal core.
Large-Eddy Simulation Numerical Study on Phase Change Heat Transfer Characteristics of Melting Pool
Xi Zhiguo, Zhang Luteng, Hu Yuwen, Gong Houjun, Ma Zaiyong, Sun Wan, Zhou Wenxiong, Pan Liangming
2022, 43(1): 15-21. doi: 10.13832/j.jnpe.2022.01.0015
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The research of flow and heat transfer characteristics in reactor melting pool is of great significance to ensure the in-vessel retention. Based on the open source software OpenFOAM platform, combined with the large-eddy simulation turbulence method and the phase change process of the melting pool, this paper establishes the heat transfer model of the melting pool, carries out numerical calculation for the LIVE working condition of the typical melting pool heat transfer experiment, and obtains the velocity field and temperature field in the melting pool, as well as the thickness and heat flux distribution of the hard shell on the inner wall of the lower head. The results show that the velocity, temperature and heat flux density in the melting pool increase with the increase of height or radial angle, the thickness of the hard shell decreases with the increase of radial angle, and the heat load on the wall of the lower head accumulates at the top. The heat transfer parameter calculation results are in good agreement with the experimental data as a whole, which can effectively reflect the natural convection and phase change process in the melting pool, verify the reliability of the calculation model, and provide a reference for further research on the phase change heat transfer characteristics of the melting pool.
Study on Optimization Design of Fuel Assembly under Low Flow Condition
Zheng Xiao, Luo Hanyu, Du Peng, Qiu Zhifang, Tian Ye
2022, 43(1): 22-27. doi: 10.13832/j.jnpe.2022.01.0022
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In order to explore the optimization design of fuel assembly for modular small reactor (SMR), in this paper, the thermal-hydraulic performance of truncated CF2 fuel assembly under the range of SMR parameters is analyzed and studied, and the influence law of the spacing and layout of mixing grid on the thermal-hydraulic performance of fuel assembly is obtained. The results show that: ① Under low flow condition, too long or too short spacing of the mixing grid will reduce the thermal performance of the fuel assembly, and the spacing of the mixing grid shall be reasonably considered in the design. ② The arrangement of the mixing grid in the upstream region of the heating section of the fuel assembly is not obvious for improving the thermal performance of the SMR fuel assembly, so the layout shall be simplified in this region. ③ Under normal operation and accident conditions of SMR, the state point parameters of the middle and downstream regions of the fuel assembly are relatively bad. Reasonably designing the spacing and arrangement of the mixing grid in the middle and downstream regions can significantly improve the thermal performance of the fuel assembly and improve the thermal safety margin. The results of this paper can provide a reference for the design optimization of SMR fuel assembly.
Study on Mechanism and Model Influence of Loop Seal Clearing in PWR SBLOCA Accident
Zhu Donglai, Yang Jun, Zhou Xiafeng, Deng Chengcheng, Ding Shuhua, Li Zhongchun, Huang Tao
2022, 43(1): 28-34. doi: 10.13832/j.jnpe.2022.01.0028
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Loop seal clearing (LSC) is one of the typical accident characteristics of cold leg small break loss of coolant accident (SBLOCA) in PWR. In order to determine the influence of the physical model of LSC phenomenon and explore the setting of the physical model for accurately reproducing the LSC phenomenon, this study combs and analyzes the main physical models affecting LSC from the perspective of the physical mechanism of LSC phenomenon. The physical mechanism of LSC phenomenon and the influence of physical model are simulated and verified by SBLOCA series experiments of LOBI bench. The results show that after reasonable setting of the physical model affecting the LSC phenomenon, the RELAP5 program model can better reproduce the LSC phenomenon in the LOBI bench experiment, which verifies the rationality of the physical model influence and model setting of the LSC phenomenon.
Visual Experimental Study on Boiling Crisis Induced by Flow Oscillation in a Single Rod Channel
Liu Haidong, Chen Deqi, Qin Jiang, Liu Hanzhou, Yan Peigang, Liu Wei
2022, 43(1): 35-41. doi: 10.13832/j.jnpe.2022.01.0035
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In order to deeply analyze the influence characteristics of boiling crisis induced by boiling two-phase flow oscillation, this paper takes deionized water as the working medium, a single rod channel with a cross section of 19 mm×19 mm and an outer diameter of 9.5 mm at the center as the research object. Through visual experimental research on boiling two-phase flow characteristics under different thermal parameters, combined with the behavior of bubbles and vapor-liquid interface characteristics, the influence characteristics of boiling crisis induced by flow oscillations are analyzed. The results show that flow oscillation is easy to occur at low pressure, low mass flow rate and low inlet subcooling, which leads to the early occurrence of boiling crisis, and the critical heat flux is significantly lower than that under stable conditions; With the increase of wall heat flux, the two-phase flow patterns in the channel appear bubble flow, slug flow, combined slug flow, stirred flow, violent stirred flow and unstable annular flow; When the flow oscillates violently, there is reflux in the channel; When the boiling crisis occurs, the pressure drop fluctuation in the channel is the largest, and the corresponding flow pattern is unstable annular flow. Therefore, the flow oscillation in the single rod channel may lead to the early occurrence of boiling crisis.
Experimental Study on Critical Heat Flux of Vertical Square Channel with Single Rod
Liu Wei, Guo Junliang, Zhang Dan, Gui Miao, Hu Ying, Liu Yang
2022, 43(1): 42-47. doi: 10.13832/j.jnpe.2022.01.0042
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The critical heat flux (CHF) of vertical square channel with single rod is experimentally studied by using R134a as the working fluid. A square channel with a flow channel cross section of 19 mm×19 mm and a single heating rod with an outer diameter of 9.5 mm are used to simulate the typical cell channel in PWR. The experimental conditions cover the typical operating conditions of PWR by fluid modeling method. The experimental results show that the CHF parameter trend of R134a in the square channel is the same as that of water in the circular tube, and R134a can replace water as a modeling fluid; After corrected with cold wall factor, the circular tube Bowring relation and Katto & Ohno relation can be used to predict CHF in square channel with cold wall; Katto’s fluid modeling method is suitable for square channel with cold wall.
Model for Calculating Idling of Nuclear Power Coolant Pump Considering Coolant Kinetic Energy in Pipeline
Jiao Zhe, Cai Long, Zhang Liping, Hu Lei, Liu Xiangsong
2022, 43(1): 48-51. doi: 10.13832/j.jnpe.2022.01.0048
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If the idling inertia of the nuclear main pump is designed too small, nuclear accident will be caused once the whole plant power failure occurs, and if the idling inertia is designed too large, the unit efficiency will be greatly reduced. Therefore, the accuracy of the idling calculation model is very important to ensure the safety of nuclear power plant and improve the unit efficiency. In this paper, the influence of coolant kinetic energy in the pipeline on the idling process of reactor coolant pump is considered. Through the power conservation equation and pump similarity law during start-up and shutdown, the idling transient calculation model considering the influence of pipeline coolant is deduced and established, and the simple calculation formulas of idling inertia and idling time of pump unit are given, which makes the calculation results more accurate and has a wider range of engineering application, and can be applied to the accurate design of idling inertia and the accurate calculation of idling time in nuclear and non-nuclear engineering.
Neutronics Influence Research on Axial Grid of PWR Fuel Assembly
Huang Xing, Wan Chenghui, Li Yunzhao, Wu Hongchun, Gu Weiquan, Cai Guangming, Xu Jialong
2022, 43(1): 52-56. doi: 10.13832/j.jnpe.2022.01.0052
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The heterogeneous modeling method and homogenous modeling method have been respectively proposed for the axial grids based on the advanced fuel-management code for PWR named Bamboo-C. Based on these two different modeling methods for axial grids, modeling analysis for the fuel assemblies of M310 reactor in Fuqing Nuclear Power Plant is carried out. By comparing with the measured data of the reactor core, the effects of two different modeling methods on the critical boron concentration, axial power distribution and axial power offset are tested. The numerical results show that the axial heterogeneous modeling method of PWR fuel assembly can significantly improve the calculation accuracy of the key physical parameters of the reactor core.
Research on the Influence of 56Fe Evaluation Cross Section of CENDL-3.2 and ENDF/B-Ⅷ.0 on Shielding Calculation
Zhang Bin, Ma Xubo, Hu Kui, Chen Yixue, Wu Haicheng
2022, 43(1): 57-63. doi: 10.13832/j.jnpe.2022.01.0057
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CENDL-3.2 Library updated the 56Fe inelastic scattering cross section. In order to verify the difference of cross-section and the shielding computing ability between the CENDL-3.2 Library and ENDF/B-Ⅷ.0 Library, NJOY2016 program was used to compare the microscopic cross section of 56Fe after resonance reconstruction, such as inelastic scattering, total cross section, and the multi-group cross section library was also made. Three series of shielding benchmarks, ILL-Fe, OKTAVIAN-Fe and IPPE-Fe, are selected in the range of 56Fe inelastic scattering energy, and 56Fe is used as the main nuclide. The results show that the inelastic cross section of CENDL-3.2 Library is lower than that of ENDF/B-Ⅷ.0 Library in the range of 4MeV~12MeV. The multi-group cross-section benchmark verification showed that the calculated results of CENDL-3.2 Library are in good agreement with the experimental values. For OKTAVIAN-Fe benchmark, the results of the two Libraries agree well in the range of 0.1MeV~1MeV. In addition, the same phenomenon is found in all the benchmark tests, that is, in the energy range of 1MeV~10MeV, where the 56Fe inelastic cross section is the main contribution, the results of CENDL-3.2 Library are higher than those of ENDF/B-Ⅷ.0 Library.
Research on Verification Methodology of Applicability of Integral Effect Test Data Based on Dimensionless Criterion Numbers
Zhang Xueyan, Deng Chengcheng, Zhu Donglai, Chen Wei, Ding Shuhua, Yang Jun
2022, 43(1): 64-71. doi: 10.13832/j.jnpe.2022.01.0064
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In the process of evaluating the safety of nuclear reactors and nuclear power plants, it is generally necessary to establish integral effect test or separate effect test bench based on the similar scaling laws to provide data support for safety performance verification and evaluation. As important parameters to measure the degree of similarity of scaling, dimensionless criterion numbers can characterize specific physical phenomena independent of the bench characteristics and the device size. Therefore, they can be used to verify the rationality of scaling design and evaluate the applicability of experimental data. The cross-bench application of dimensionless numbers can not only avoid excessive repetitive experiments, but also assist in the evaluation of a physical phenomenon that can not be accurately reproduced by a single bench. In order to explore the application methods and principles of dimensionless numbers in scaling analysis and applicability evaluation of experimental data, in this paper, aiming at SBLOCA of the traditional PWR, based on the numerical simulation results of RELAP5, the top-down scaling analysis method is used to calculate the dimensionless parameters and compare the data of the integral effect test bench LOFT and LOBI. The analysis results show that the dimensionless numbers related to important phenomena and parameters such as break mass outflow, core decay heat and primary circuit pressure are in good agreement with the bench; the dimensionless number ratios related to loop friction resistance and loop buoyancy have great distortion. The dimensionless analysis method used in this paper is expected to be used for mutual verification of experimental data on the same type of test bench, and to provide a reference for the development and verification of new reactor types.
Research on Calculation of Interfacial Resistance in One-Dimensional Two-Fluid Model
Ye Tingpu, Cheng Cheng, He Hui, Dong Xianhong, Xu Haiyan
2022, 43(1): 72-77. doi: 10.13832/j.jnpe.2022.01.0072
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In one-dimensional two-fluid model, the interfacial resistance is a key parameter for determining the degree of coupling between phases. At present, the calculation methods include drift-flux model method and resistance coefficient method. In this study, by using the subchannel program, the two calculation methods are evaluated based on the circular tube air-water two-phase experimental data. The results show that the prediction ability of the drift-flux model method in the one-dimensional two-fluid model is better than that of the resistance coefficient method. At the same time, the influence of distribution effect on the calculation of interfacial resistance in two-phase flow is evaluated. The results show that the effect is small in the low void fraction region, but obvious in the high void fraction region.
Preliminary Analysis of Unprotected Reactivity Introduction Accident of Liquid Molten Salt Reactor with Graphite Channel based on TREND Program
Yu Wen, He Long, Zou Yang, Xu Bo, Dai Ye, Xu Hongjie
2022, 43(1): 78-83. doi: 10.13832/j.jnpe.2022.01.0078
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Using the self-developed system analysis code TREND and based on the liquid point reactor dynamics model, aiming at the design of 10 MW liquid molten salt reactor with graphite channel, the transient changes of core power, graphite temperature and molten salt temperature at the core outlet of 10 MW liquid molten salt reactor with graphite channel with different reactivity under step introduction and linear introduction are studied and analyzed. The results show that the reactor system can operate safely without protection when the reactivity step introduction is lower than 570pcm (1pcm=10−5). When about 800pcm is introduced due to loss of lift of single control rod, the reactivity introduction rate does not exceed 8pcm/s, and the reactor can effectively control the peak power and reduce the core outlet temperature by making use of its own negative feedback characteristics of temperature and power, so as to ensure the safe operation of the reactor without protection. Therefore, liquid molten salt reactor has good inherent safety.
Research on Void Distribution Measurement of 5×5 Rod Bundle Channel Based on Wire Mesh Sensor
Wang Yinglong, Xie Hao, Xiong Jinbiao, Yang Yiang, Cheng Xu
2022, 43(1): 84-91. doi: 10.13832/j.jnpe.2022.01.0084
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In order to study the two-phase distribution in the rod bundle channel, a wire mesh sensor module suitable for the rod bundle channel was designed and fabricated, and the experiment of measuring the void distribution of air-bubble flow in a 5×5 rod bundle channel was carried out. The distribution law of void fraction in rod bundle channel and the effect of bubble size on void distribution are analyzed. The experimental results show that the small bubbles with reversal of transverse lift direction gather near the wall, and the large bubbles gather in the center of the subchannel; The critical bubble diameter of transverse lift direction reversal at room temperature and pressure is between 4~6 mm, which proves the applicability of transverse lift model in rod bundle channel.
Study on the Effects of Tube Diameter and Inclination Angle on the Condensation Heat Transfer of Air-Containing Steam Outside the Tube Bundle
Liu Shiwen, Li Yi, Cheng Xiang, Bian Haozhi, Cao Boyang, Ding Ming
2022, 43(1): 92-96. doi: 10.13832/j.jnpe.2022.01.0092
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Through condensation tests of air-containing steam outside the tube against the 3×3 tube bundle with different tube diameters and inclination angles, the basic laws of heat transfer tube diameter and inclination angle affecting the condensation heat transfer of air-containing steam outside the tube bundle are studied. The results show that the effects of tube diameter and inclination angle are significantly different in different pressure ranges. At pressures below 0.8 MPa, the condensation heat transfer coefficient increases with the decrease of tube diameter and inclination angle, and the condensation heat transfer coefficient of 12 mm and 0° inclination angle heat transfer tube is 29% higher than that of 19 mm and 90° inclination angle. At pressures above 0.8 MPa, the condensation heat transfer coefficient decreases with the decrease of the tube diameter, up to 18%; with the decrease of the inclination angle, the condensation heat transfer coefficient first decreases and then increases, and the condensation heat transfer coefficient is the smallest at an inclination angle of about 60°.
Nuclear Fuel and Reactor Structural Materials
Effect of Neutron Irradiation on Mechanical Properties of Accident-Tolerant Fuel FeCrAl Alloys
Lei Yang, Zhang Haisheng, Mao Jianjun, Liu Xiaosong, Qiao Yingjie, Wang Peng, Wu Yazhen, Xiao Wenxia
2022, 43(1): 97-101. doi: 10.13832/j.jnpe.2022.01.0097
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FeCrAl alloy has good high-temperature oxidation resistance and mechanical properties and can be used as fuel cladding material. In order to study the irradiation mechanical properties of FeCrAl alloy, FeCrAl alloy mechanical properties test with different element contents and 2×1019 cm−2 and 8×1019 cm−2 neutron fluence irradiation was carried out, the tensile properties of FeCrAl alloys were tested at room temperature and 380℃, and the tensile strength and yield strength of FeCrAl alloys with different Cr and Al contents were obtained. The effects of Al content, Cr/Al content ratio and neutron irradiation on the mechanical properties of FeCrAl alloy were studied. The results show that the strength of FeCrAl alloy generally increases with the increase of Al content; After 2×1019 cm−2 neutron irradiation, the strength of FeCrAl alloy is greatly improved; After 8×1019 cm−2 neutron irradiation, the strength of FeCrAl alloy does not increase significantly. The research results provide important data support for the R&D and selection of accident-tolerant fuel cladding.
Fuel Performance Analysis of Light Water Reactor Based on the Combination of U3Si2 Fuel and Two-Layer SiC Cladding Based on Multi-Physical Field Coupling
Yin Chunyu, Liu Rong, Jiao Yongjun, Qiu Chenjie, Liu Zhenhai, Qiu Bowen, Gao Shixin, Xing Shuo
2022, 43(1): 102-109. doi: 10.13832/j.jnpe.2022.01.0102
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Based on COMSOL platform, a fuel performance analysis program based on multi-physical field full coupling is developed, and the correctness and accuracy of the program are verified by comparing with the radial power distribution model; Then the performances of the combinations of U3Si2 fuel and two-layer SiC cladding, U3Si2 fuel and zirconium alloy cladding under normal reactor operating conditions are further analyzed and compared with the combination of UO2 fuel and zirconium alloy. The calculation results show that the combination of U3Si2 fuel and zirconium alloy cladding has lower fuel center temperature, fission gas release and internal pressure than that of UO2 fuel and zirconium alloy, but the air gap closing time will be earlier; The combination of U3Si2 fuel and two-layer SiC cladding has higher fuel center temperature, greater fission gas release and internal pressure than the combination of U3Si2 fuel and zirconium alloy, and its fuel center temperature increases significantly with the increase of burnup. Compared with zirconium alloy cladding, two-layer SiC cladding can effectively delay the closing of air gap and alleviate the mechanical interaction between fuel and cladding.
Theoretical Calculation of High Temperature Potential-pH Based on Fe-H2O, Cr-H2O and Zr-H2O Systems
Yang Yang, Yang Yu, Chen Yunmin, Cao Qi, Xiong Wei, Lu Yunyun, Wu Xiaoyong
2022, 43(1): 110-115. doi: 10.13832/j.jnpe.2022.01.0110
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From the perspective of thermodynamics, this paper uses the Nernst equation to calculate the potential-pH diagrams (E-pH diagrams) of the Fe-H2O system, Cr-H2O system and Zr-H2O system at temperatures of 423 K and 573 K. This paper theoretically explain the corrosion behavior tendency of the main constituent elements of reactor structural materials of iron, chromium and zirconium in high temperature and high pressure water, which are affected by potential and pH. This provides data reference for the materials electrochemical corrosion tests in the water chemical environment in reactor, preventing the corrosion of the material, and extending the service life of the material.
Study on the Determination Method of the Residual Fluorine Content on the Inner Surface of the Fuel Rod Cladding Tube
An Shenping, Li Shuliang, Liao Zhihai, Huang Xinshu, Ning Wei, Ren Liping
2022, 43(1): 116-121. doi: 10.13832/j.jnpe.2022.01.0116
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The residual fluorine on the inner surface of the zirconium alloy pipe might accelerate the stress corrosion of the microcracks on the surface of the zirconium alloy. In order to accurately determine the residual fluorine content in the zirconium alloy pipe, this paper developed a special internal surface residual fluorine extraction device through experimental research, and high-temperature hydrolysis conditions were studied; Ion chromatography was used to determine the extracted fluoride ions. The established analysis method could achieve rapid and continuous determination, and the analysis range was 0.05~1.0 μg/mL; finally, this method was used for zirconium alloy pipes. The residual fluorine content on the inner surface was measured. The results should that the extraction efficiency of the developed device for the residual fluorine on the inner surface could reach 101%; the recovery rate of the method was 98%~104%, the maximum relative standard deviation was 3.9%; with high precision and accuracy, the measurement results met the production needs.
Research on Cooperative Design of Anti-Hangup and Thermal Performance of Spacer Grid
Chen Jie, Chen Ping, Pang Hua, Lei Tao, Pu Zengping, Deng Shuang, Peng Yuan, Ren Quanyao, Su Min
2022, 43(1): 122-126. doi: 10.13832/j.jnpe.2022.01.0122
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Spacer grid is an important part of the fuel assembly skeleton structure, whose main function is to clamp and locate the fuel rods. The anti-hangup performance and thermal performance also shall be considered. Based on the feedback of the operation experience of the mainstream fuel assembly, this paper uses the three-dimensional modeling software (UG) to simulate the relative motion of the grid, analyzes the causes of the hangup, and makes it clear that the continuous arrangement of the outer strip guide vane can effectively improve the anti-hangup performance, which is verified by the spacer grid hangup test; Through the computational fluid dynamics (CFD) simulation analysis of the side cell where the spacer grid is located, it is found that the continuous arrangement of guide vanes based on the traditional anti-hangup design is not conducive to the thermal performance between adjacent grids; On this basis, a scheme of alternating-height arrangement of guide vanes is designed. Theoretical analysis and experimental verification results show that the scheme realizes the cooperative design of anti-hangup and thermal performance of the fuel assembly spacer grid.
Safety and Control
Analysis of Influencing Factors of Control Rod Downward Motion Characteristics
Li Tian, Wei Bingqian, Li Yan, Yang Zhendong, Li Shishuang
2022, 43(1): 127-132. doi: 10.13832/j.jnpe.2022.01.0127
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Liquid suspension passive shutdown technology is one of the research hotspots in the field of nuclear reaction safety in recent years. It is important to study the downward motion characteristics of control rods for the safe operation of nuclear power plants. Two typical unprotected transient loss of flow accidents (ULOF) were selected as the basis to design various working conditions for model experiments. Combined with the experimental data, the force analysis of the downward motion of the control rod was carried out to obtain the time-history change of its resistance. The influence of outlet aperture and initial circumferential position on the time history and buffering effect of the control rod was analyzed by the control variable method, and the function relationship between resistance coefficient and Reynolds number was obtained. It can provide the basis for optimizing the structure of control rod assembly and the selection of resistance coefficient in the study of the downward motion force of control rod.
LSSVM Based Covert Attack Method Research on Pressure Control System of Pressurizer in Nuclear Power Plant
Wang Dongfeng, Li Qixian, Huang Yu, Xu Jing, Wang Biao
2022, 43(1): 133-140. doi: 10.13832/j.jnpe.2022.01.0133
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Covert attacks pose a serious threat to the safe, stable and efficient operation of the control system of pressurizer in a large nuclear power plant. The key to realize covert attack is to establish a high-precision object estimation model. This paper proposes a covert attack method based on least square support vector machine (LSSVM). The system identification is carried out by LSSVM algorithm, and the high-precision estimation model of the attacked area of the pressurizer is obtained. Then the estimation model is combined with the hidden controller to realize the covert attack of the pressurizer pressure control system in the case of no noise, noise and with nonlinear links. The simulation results show that the attack mode not only causes some damage to the pressure control system of the pressurizer, but also has a high degree of covertness.
Test and Data Analysis of Response Time of Control Rod Drive Mechanism
Jiang Xuejun, Hou Jie, Li Qin, Wei Yongbo, Lai Wei
2022, 43(1): 141-147. doi: 10.13832/j.jnpe.2022.01.0141
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In this paper, a set of a fast test system for electromechanical response time, using PXI data acquisition platform and MATLAB for data analysis, is built to measure the fast response time of the control rod drive mechanism (CRDM) and filter out the noise interference in the test environment. This test system can collect the real-time and high-frequency position signals of the control rod quickly and uses the moving average filtering and segmental curve fitting to filter and fit the position data of the control rod. Besides, the test system uses the FFT spectrum analysis to verify the effect of signal denoising. This test system is used in the electromechanical response time test of the CRDM in the first scram system of the thorium molten salt reactor-liquid fuel 1 (TMSR-LF1), demonstrating the test system in this study is consistent to the requirement of the CRDM electromechanical response time test on feasibility, reliability and applicability.
Research on Fault Detection and Identification Method of Small PWR Based on Principal Component Analysis
Cao Huasong, Sun Peiwei
2022, 43(1): 148-155. doi: 10.13832/j.jnpe.2022.01.0148
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Fault detection and identification are important for the safety and economy of small PWRs. The fault detection and identification method based on signal and expert knowledge and experience is usually applied in nuclear reactors. However, operators are often unable to identify the fault type and trace the fault cause in time and accurately from the massive fault data information. A method of fault detection and identification of small PWR based on principal component analysis is presented in this paper. First, the model of a small PWR is established by RELAP5 code, and the sample data of typical faults is obtained. Second, the dimension of sample data is reduced by using principle component analysis method. T2 and Q statistics are calculated to detect the reactor operation condition by judging whether the thresholds are exceeded. Then, the contribution rate of process variables to statistics is analyzed by using the contribution rate graph method, so as to determine the variables that play a major role in the change of fault characteristics and realize the identification of different faults. Finally, the effectiveness of the method is verified by comparing with the actual physical process analysis results.
Marginal Research on IVR Capability of Alumina Nanofluid Enhanced Spherical Lower Head
Song Jian, Yu Hongxing, Deng Jian, Xiang Qing'an, He Xiaoqiang
2022, 43(1): 156-162. doi: 10.13832/j.jnpe.2022.01.0156
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In order to evaluate the extent to which alumina nanofluids can extend the IVR capability of the spherical lower head relative to pure water working medium, in this paper, the critical heat flux (CHF) mechanism model of alumina nanofluid based on bubble force balance and the wall heat flux partitioning CHF model are used to calculate the CHF of nanofluid on the outer surface of spherical lower head. The decay heat distribution sampling calculation is carried out by using the IVR analysis software CISER, and the random distribution of CHF on the wall of the lower head with the inclination angle is obtained. Compared with the theoretical value of the nanofluid CHF model, the influence of nanofluid on the marginal expansion of IVR capability is studied and judged by taking the CHF ratio less than 1 as the IVR success criterion. The results show that if no optimization measures are taken for the internal and external heat transfer composition of the lower head and only nanofluid is used to replace pure water working medium, the IVR capability margin of PWR nuclear power plant can be expanded to 1300 MW rated power level.
Research on Preliminary Application of MBSE in Nuclear Power Design
Zhu Junzhi, Yang Jue, Wan Lei, Shi Weili, Liu Yongkang
2022, 43(1): 163-168. doi: 10.13832/j.jnpe.2022.01.0163
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In order to deal with the challenges brought by the increasing user requirements to the “requirements-satisfying” backward design, in this paper, the model based system engineering is applied to the architecture design of “safety injection system”. The design process includes requirement analysis, function analysis and design synthesis: the system requirements are obtained through the sequence diagram and requirement diagram of the requirement analysis process, and the tracking relationship between the requirements is established to facilitate the impact analysis of requirement changes. The system function architecture design and early verification and confirmation are realized through the activity diagram, state diagram and sequence diagram of the function analysis. Through the trade-off analysis of the design synthesis process, the key function alternatives are selected, and the system architecture model and dynamic operation process are displayed with the help of block definition diagram to ensure that the designed system meets the expectations of stakeholders. The application results show that model-based systems engineering (MBSE) is suitable for the existing nuclear power design and can effectively improve the problems existing in the traditional design.
Synthesis of Two-Dimensional Tungsten Trioxide (WO3) Nanostructure and Its Application in Photodetectors
Sun Congjian, Huang Youjun, Wang Yinli, Bao Chao, Jiang Tianzhi, Xu Qinglan, Lei Wen, Zhao Li
2022, 43(1): 169-174. doi: 10.13832/j.jnpe.2022.01.0169
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Abstract:
Tungsten oxide (WO3) is ideal for manufacturing transistors and photodetectors. Although some work has been done on growth of WO3 nanostructures, making ideal long enough nanowires is still a challenge. Among many synthesis methods, the author chose to use chemical vapor deposition (CVD) to synthesize WO3 nanowires and study their application to photoelectric sensors. The main factors affecting WO3 nanowires products are precursor material temperature, substrate position, carrier gas flow rate and growth duration time. The greatest length of nanowires grown under appropriate growth condition is approximately 100 μm. WO3 nanowires can be used for fabricating high-performance photodetectors. WO3 photodetector has excellent device performance including high sensitivity, fast response speed and module miniaturization, which indicate that two-dimensional WO3 nanowires have great advantages in the manufacture of photodetectors.
Analysis of Ex-Vessel Steam Explosion under RPV Side Break Cases
Chen Peng, Zhao Xinhai, Zhan Dekui, Xia Shaoxiong
2022, 43(1): 175-182. doi: 10.13832/j.jnpe.2022.01.0175
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The calculation and analysis of steam explosion were carried out for the pressure vessel (RPV) side breach conditions that are more likely to occur in the actual process. According to the Phenomena Identification and Ranking Table (PIRT) issued by Organization for Economic Cooperation and Development (OECD), the sensitivity analysis parameters for ex-vessel steam explosion are selected. By establishing 3D local break and 2D circular break geometric models with MC3D, the sensitivity analysis under RPV side break conditions is carried out for the important parameters affecting the calculation results (break size, pit water level, break position, trigger conditions, liquid column fragmentation and droplet fragmentation model), and the worst calculation conditions are obtained. The sensitivity analysis results show that under the condition of large break loss of coolant accident (LBLOCA), when the reactor pit is at full water level, 2D side circular break occurs in RPV, steam explosion is triggered when contacting the reactor pit side wall, and CONST model and Classical model are adopted, the calculation results of pressure load on the reactor pit side wall are the most conservative and pose the greatest threat to the integrity of the reactor pit and containment.
Risk Analysis of Late Reinjection Pressure in the Primary Circuit under IVR Strategy
Wang Xiaoji, Wu Lingjun, Zhu Dahuan, Deng Jian, Liu Lili, Xu Youyou
2022, 43(1): 183-186. doi: 10.13832/j.jnpe.2022.01.0183
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Based on the design of a million-kilowatt nuclear power plant, this article analyzed the possible late primary circuit reinjection scenarios in severe accidents when the In-Vessel Retention (IVR) measure is taken (that is, when the lower head of the pressure vessel has formed a melting pool), and studied the primary circuit pressure respond to the late reinjection. Through the comparative analysis of the condition with no reinjection, the influence of reinjection time, reinjection flow and the opening number of pressure relief valves in serious accidents on the pressure of the primary circuit were comprehensively evaluated, the influence law of each measure was obtained, and suggestions were put forward for severe accident management strategies.
Circulation and Equipment
Numerical Simulation Research on Flow-Induced Vibration Characteristics of Fluid-Conveying Pipe Network
Liu Shiwen, He Ronghui, Yang Zhao, Wang Jiarui, Chen Shuang, Lai Jianyong, Li Yi
2022, 43(1): 187-191. doi: 10.13832/j.jnpe.2022.01.0187
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To study flow-induced vibration characteristics of pipeline system for pipeline optimization design, two-way fluid-structure interaction numerical simulation for the typical fluid-conveying pipe network with various fluid excitation was conducted based on Ansys Workbench, and the flow-induced vibration characteristics of the pipeline structure were obtained. The effects of excitation type, medium temperature, flow field structure and natural frequency of structure on the fluid-induced vibration characteristics in the pipe were analyzed and discussed. The results show that the structural amplitude of pipeline under excitation of pulsating flow is significantly larger than that under constant flow. When the frequency of fluid excitation is close to the natural frequency of the pipeline structure, the structure and fluid tend to resonate, leading to increased structural vibration. By applying constraint support at the appropriate position of the pipeline, the natural frequency of the structure is far away from the fluid excitation frequency, which can effectively reduce the vibration of the pipeline. In addition, the medium temperature and flow velocity have a great influence on the structural amplitude.
Research on Design of HPR 1000 Steam Generator Upper Support Based on High Seismic Performance
Tang Chenhang, Huang Yan, Shen Pingchuan, He Gening, Yu ping, Su Tong
2022, 43(1): 192-196. doi: 10.13832/j.jnpe.2022.01.0192
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In order to improve the seismic performance of ZH-65 steam generator of HPR 1000 nuclear power plant, a new support scheme of steam generator is put forward, that is, the structure of connecting rod and hydraulic damper is used for the upper support of steam generator, and the overall design scheme and the thermal expansion compatibility of connecting rod are designed and studied. Compared with the original steam generator upper support of the second generation plus nuclear power unit, the upper support designed and studied in this paper is greatly optimized in terms of equipment weight, number of welds and difficulty of installation and commissioning; It can effectively reduce the supporting load, and the maximum reduction range is about 24%; The weld load of steam generator nozzle can be reduced, and the maximum reduction range is about 28%.
Other Columns
Application of Digital Twin Technology in the Design Phase of Floating Nuclear Power Plants
Li Kaiyu, Cai Qi, Cai Xinxin, Chen Yuqing, Peng Liu, Zhang Yifang
2022, 43(1): 197-201. doi: 10.13832/j.jnpe.2022.01.0197
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The research object is the reloading operation process of a floating nuclear power plant. A technical route of digital twin technology is proposed for application in floating nuclear power plants. Design data is integrated and used to develop a digital twin system for nuclear fuel handling cabins. The virtual cabin model is established, and the new fuel transportation process control program is developed. Based on the OPC Unified Architecture (OPC UA) communication protocol, the two-way interaction between the virtual cabin model and the control system is realized. The results show that the digital twin system can provide a reference for the determination of personnel positions and the optimization of the reloading operation control process. Therefore, digital twin technology can serve the design phase of floating nuclear power plants, and can provide a reference for the application of digital twin technology in the full life cycle of floating nuclear power plants.
Development of γ Radiation Field Calculation Platform for Nuclear Equipment Identification Device
Yang Yushu, Liu Jizhen, Wang Xu, Chen Jialang, Qi Mingsen, Zhang Ying
2022, 43(1): 202-207. doi: 10.13832/j.jnpe.2022.01.0202
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In order to solve the problem of radiation field calculation in practical work more widely and conveniently, an effective and reliable calculation model, computer data management and data visualization technology are adopted, and greedy algorithm is used to quickly solve complex equations. A general γ radiation field calculation platform is developed, which realizes the functions of radiation source management, three-dimensional display of radiation source layout, dose rate level of radiation field and automatic generation of layout scheme. Experiments show that the γ radiation field automatically generated by the platform can meet the requirements of γ radiation identification of nuclear equipment.
Development and Verification of Radiation Shielding Optimization Design Platform for Marine Reactor
Li Yuehang, Yu Tao, Chen Zhenping, Gan Bin, Xian Xirui, Niu Haoxuan
2022, 43(1): 208-214. doi: 10.13832/j.jnpe.2022.01.0208
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Marine reactor puts forward higher requirements for nuclear reactor shielding design, and the traditional radiation shielding design methods and design software can no longer meet the requirements. In order to obtain more accurate radiation shielding design, this paper establishes a multi-objective optimization platform for radiation shielding of marine reactor-MOSRT, which integrates the functions of “geometric modeling-material modeling-shielding optimization-result visualization” based on the open source SALOME framework. MOSRT platform can realize 3D CAD solid modeling of shielding structure, multi-objective optimization of radiation shielding based on genetic algorithm and 3D visualization of dose field of shielding calculation results. Based on Savannah and MRX marine reactor model, the radiation shielding optimization verification of MORST platform is carried out. Compared with the initial scheme, the optimization scheme has achieved good optimization results in dose, mass and volume. It is proved that MOSRT platform initially has the function of radiation shielding optimization design, which can provide auxiliary design means for marine reactor engineering and conceptual shielding design.
Research on Radiation Risk of On-Site Workers Based on PSA Accident Sequence
Zhou Jing, Lyu Weifeng, Ran Wenwang, Gong Quan, Xiong Jun
2022, 43(1): 215-220. doi: 10.13832/j.jnpe.2022.01.0215
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From the perspective of total risk control, the dose and radiation risk acceptance criteria of on-site workers under accident conditions are put forward, and the corresponding assessment methods are established. Taking a typical PWR nuclear power plant as an example, the full-scope accident sequence of probabilistic safety analysis (PSA) was used for verification and assessment, and the radiation dose and the risk of radiation lethality of on-site workers after a typical PWR nuclear power plant accident were assessed. The verification results show that the total radiation lethality risk of the on-site workers after the accident is much lower than the total death risk value of the public due to natural disasters, diseases, traffic accidents and different industries; After the accident, the radiation lethality risk accounts for the highest proportion when the workers operate in the fuel building. Therefore, when the workers carry out relevant operations in the fuel building, they can formulate corresponding radiation protection measures in advance to reduce the radiation risk; Other personnel in the working group and accidental exposure personnel account for a relatively high risk of radiation lethality after the accident, airborne radioactivity can be protected by air masks and other methods to reduce its radiation risk. The corresponding analysis results can provide reference for the formulation of the post-accident treatment plan of the nuclear power plant and the radiation protection measures for the on-site workers after the accident.
Study on Adsorption Kinetics of Gaseous Iodine by 8% Silver Loaded Mordenite
Xiong Wei, Zhang Jinsong, Cao Qi, Chen Yunming, Yang Yu, Lu Yunyun, Yang Yang, Tang Jia, Wang Haijun, Liu Chenlong
2022, 43(1): 221-225. doi: 10.13832/j.jnpe.2022.01.0221
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To remove radioactive iodine from spent fuel reprocessing, in this study, 8% silver loaded mordenite was used as the adsorbent, and the dynamic adsorption method was used to study the effects of gas linear velocity, adsorption column height, humidity and NO2 volume fraction on the adsorption effect of gaseous iodine. The experiment results show that the gas linear velocity has little effect on the saturated adsorption capacity, and the mass transfer resistance of the adsorption column increases at lower gas linear velocity; The height of the adsorption column has little effect on the mass transfer resistance. With the increase of the height of the adsorption column, the saturated adsorption capacity increases to a certain extent; Humidity does not affect the adsorption effect; NO2 can promote the adsorption effect; Yoon-Nelson model can well fit the adsorption process of iodine on 8% silver loaded mordenite adsorption column.
Column of Science and Technology on Reactor System Design Technology Laboratory
Optimization Design of Passive Residual Heat Removal System for MSR Based on Air Cooling
Zhang Zhuohua, Fu Yao, Sun Wei, Ran Xu, Li Feng, Xian Lin, Su Dongchuan, He Xiaoqiang
2022, 43(1): 226-231. doi: 10.13832/j.jnpe.2022.01.0226
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Shanghai Institute of Applied Physics proposed a technical solution based on the Thorium-based Molten Salt Solid-state Test Reactor (TMSR-SF1) based on TRISO-coated spherical particulate fuel and liquid fluoride salt. One of the most important works is the design of passive residual heat removal system (PRHRS). Because of the incompatibility between molten salt and water and its high operating temperature, it is necessary to use air as the final heat sink to design PRHRS. In order to achieve the design objectives of system simplification, volume minimization and consideration of heat removal and insulation, starting with the model of heat transfer process from MSR core active zone to external air heat sink, this paper establishes the PRHRS optimization design model, obtains the optimization design scheme, and based on the improved RELAP5/MOD4.0 code (special improved code for TMSR-SF1) carries out the demonstration and evaluation of PRHRS capacity. Calculation and analysis show that the design of PRHRS capacity is reasonable, which can ensure the heat removal safety after the reactor SBO.
Study on the Creep Damage Constitutive Models of 16MND5 Steel for Domestic Reactor Pressure Vessel
Su Dongchuan, Zhang Ying, Du Juan, Sun Yingxue, Fu Xiaolong, Li Hui, Shao Xuejiao, Guo Sujuan
2022, 43(1): 232-237. doi: 10.13832/j.jnpe.2022.01.0232
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In order to obtain the creep behavior of reactor pressure vessel (RPV) materials at high temperature and ensure the integrity of RPV under severe accident conditions, the high temperature creep properties of 16MND5 steel for domestic RPV were tested, and the creep property of the material at 600~900℃ was obtained. Based on the basic creep constitutive model of strain strengthening and the creep damage model based on ductile depletion theory, a creep damage constitutive model suitable for 16MND5 material is established, and the creep damage model parameters are given. Results show that the finite element simulation data of the creep damage constitutive model proposed in this paper are in good agreement with the experimental data, which verifies the correctness of the creep damage model. This method can be used for the creep damage analysis of RPV under serious accidents, and provide support for the integrity analysis of RPV.
Validation and Analysis of Test Results of Critical Mass Measurement of Hexagonal Casing Type Fuel Reactor
Wei Yanqin, Huang Shien, Wang Lianjie, Lou Lei, Cao Jiebao, Cai Yun
2022, 43(1): 238-241. doi: 10.13832/j.jnpe.2022.01.0238
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In order to verify the calculation reliability of nuclear design code for fuel assembly, beryllium assembly and aluminum assembly, the verification calculation and deviation analysis of test data of critical mass measurement of hexagonal casing type fuel reactor are carried out. By analyzing the reactivity differences of aluminum assemblies in different locations, a new calculation model of aluminum assemblies near active zone is proposed, which reduces the calculation deviation of aluminum assemblies near active zone from 2.2% to 0.1%. It lays a good foundation for the engineering verification of the core nuclear design code.
Experimental Verification of Special Nuclear Design Code for Hexagonal Casing Type Fuel Reactor
Wang Lianjie, Wei Yanqin, Huang Shien, Lou Lei, Ma Yongqiang, Cao Jiebao
2022, 43(1): 242-245. doi: 10.13832/j.jnpe.2022.01.0242
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Based on the experimental data of seven Zero Power physical test schemes of Hexagonal Casing Type Fuel Reactor (HCTFR), the calculation accuracy of the nuclear design codes CELL and CPLEV2 was engineering verified. According to the verification calculation results, the calculation deviations of the critical rod position effective multiplication factor (keff) of the seven critical test schemes are all within ±0.8%, which is in good agreement with the experimental results. The value of control rods and the calculation deviation of shutdown depth are also in the acceptable range, indicating that CELL+CPLEV2 has high calculation accuracy and reliability and can be used in HCTFR core design.