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2023 Vol. 44, No. 2

Special Contribution
Thoughts on the Application of Artificial Intelligence in Nuclear Energy Field
Tan Sichao, Li Tong, Liu Yongchao, Liang Biao, Wang Bo, Shen Jihong
2023, 44(2): 1-8. doi: 10.13832/j.jnpe.2023.02.0001
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Under the new wave of global artificial intelligence, the nuclear energy industry has gradually started the process of integrating with the development of artificial intelligence. This paper discusses some problems arising from the combined application of artificial intelligence and nuclear energy. First of all, it clarifies the application advantages of artificial intelligence in the field of nuclear energy. Artificial intelligence technology can enhance the economical efficiency and functionality of nuclear energy by reducing the operating costs, improving the power generation efficiency and optimizing the control strategies. Secondly, it holds the key to the integration of artificial intelligence and nuclear energy, that is, applying key supporting techniques such as big data, cloud computing, and the Internet of Things, and realizing the best fitting of artificial intelligence technology to nuclear engineering problems according to the application scenarios and boundaries in the nuclear energy field. Then, it determines the personnel-led issues in the process of nuclear energy intelligentialization, where the nuclear industry personnel will lead the realization of the effective fitting and integration of artificial intelligence and nuclear engineering problems, thereby promoting the development of nuclear energy intelligence. Finally, it realizes people's recognition and acceptance of nuclear energy intelligence and discusses how to build an intelligent and trusted security system for nuclear energy from the perspectives of data, algorithms, standardization, security, and public acceptance so that nuclear industry personnel and the public accept nuclear energy intelligence. Through the elaboration of several issues in the process of nuclear energy intelligentialization, it is expected to arouse the common thinking of nuclear industry personnel and the public, promote the cross-disciplinary deep integration of artificial intelligence and nuclear energy science and technology and then realize the in-depth empowerment of artificial intelligence to the nuclear energy industry.
Reactor Core Physics and Thermohydraulics
Pebble-Bed High-Temperature Gas-Cooled Reactor Burnup Uncertainty Analysis Based on Fine Burnup History and Fine Burnup Chains
Cui Menglei, Guo Jiong, Wang Yizhen, Liu Baokun, Kong Boran, Zhu Kaijie, Li Fu
2023, 44(2): 9-14. doi: 10.13832/j.jnpe.2023.02.0009
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The fuel element burnup history becomes extremely complex as a result of the multi-pass of the fuel pebble through the reactor core in fuel management of the pebble-bed high-temperature gas-cooled reactor (PB-HTGR). The core physical design program VSOP of PB-HTGR can provide a fine burnup history of fuel elements, but it involves only a small number of burn-up chains and nuclide species. The burnup calculation program NUIT developed independently by Tsinghua University can realize fine burnup calculation and contains complete burnup chain and nuclide information, but it cannot track the fine burnup history. In this paper, we will develop the fine burnup calculation function of PB-HTGR based on NUIT and the fine burnup history of PB-HTGR provided by VSOP, build a burnup analysis process with fine burnup history simulation and fine burnup chain nuclides and realize a burnup uncertainty analysis function. On this basis, the contribution of fission yield uncertainty to the burnup calculation contribution of PB-HTGR is studied, and the results are compared with the VSOP calculation results. The calculation and analysis results show that the fine burnup calculation results based on NUIT and the burnup calculation results based on VSOP are mutually verified, and more nuclide information can be obtained. Such calculation results can be used as a basis for the decay heat uncertainty study of PB-HTGR.
Application Research on VITAS—a General-purpose Neutron Transport Code
Zhang Tengfei, Yin Han, Sun Qizheng, Xiao Wei
2023, 44(2): 15-23. doi: 10.13832/j.jnpe.2023.02.0015
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In order to improve the applicability of deterministic whole-core neutron transport code, a general-purpose neutron transport code VITAS is developed. The verification results of TAKEDA3 benchmark problem (rectangular assembly), the TAKEDA4 benchmark problem (hexagonal assembly), the Dodds benchmark problem (R-Z geometry) and the C5G7-TD5 benchmark problem (PWR high fidelity calculation) show that higher-order spatial and angular basis functions can make the results converge asymptotically and steadily to the reference solution, reaching the calculation accuracy level equivalent to that of multi-group Monte Carlo method. Compared with the reference solution, the deviations are less than 60pcm (1pcm = 10−5) in effective multiplication coefficient (keff), −3pcm in control rod worth and 1.03% in neutron flux distribution root mean square (RMS) for TAKEDA3 benchmark problem; less than 20pcm in keff, 32pcm in control rod worth and 0.70% in neutron flux distribution RMS for TAKEDA4 benchmark problem; less than 1% (maximum) in power for Dodds benchmark problem; and less than 0.9% in power for C5G7-TD5 benchmark problem. The research in this paper shows that VITAS has the potential to become a general-purpose calculation tool for accurately solving the neutron transport problems.
Adjoint Neutron Flux Calculation Technique Based on Improved Variational Nodal Method
Liang Boning, Wu Hongchun, Li Yunzhao
2023, 44(2): 24-29. doi: 10.13832/j.jnpe.2023.02.0024
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The adjoint neutron flux is of great significance for nuclear safety and detector calculation in pressurized water reactor (PWR). However, existing nodal methods would cause a big error due to heterogeneous nodes, including heterogeneous cross sections and discontinuity factors, which will appear frequently with the control rod moving. In this paper, an improved variational nodal method (VNM) is proposed to reduce the error. It determines the continuous conditions for adjoint nodal methods that are different from forward equation. Unlike traditional VNM, which establishes a functional method globally, this paper establishes a functional method for each node. It constructs a multiplier term with a heterogeneous discontinuity factor to explicitly deal with the adjoint neutron flux with surface discontinuity. Apart from the expansions of adjoint neutron flux, cross section and surface partial neutron current densities, the surface discontinuity factor (DF) is also expanded into pieces-wise orthogonal polynomials to construct the nodal response matrixes. The numerical results of the BEAVRS benchmark problem with heterogeneous nodes existing demonstrate that compared with the traditional VNM, the improved VNM can reduce the error by two orders of magnitude for the adjoint neutron flux in fuel area and the adjoint effective multiplication factor, which can help realize high accuracy calculation for the inner product of forward and adjoint neutron flux.
Development and Validated Application of Calculation Function of High Fidelity Refueling Cycle for Pressurized Water Reactor
Wang Xining, Liu Zhouyu, Zhou Xinyu, Wen Xingjian, Cao Lu, Zhang Sifan, Xu Xiaobei, Yi Siyu, Li Shuaizheng, Li Fan, Su Xin
2023, 44(2): 30-36. doi: 10.13832/j.jnpe.2023.02.0030
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The refueling cycle calculation function for pressurized water reactor (PWR) is developed on the basis of the self-developed numerical nuclear reactor physics calculation code NECP-X. Startup physics experiments are conducted for the first, second and third cycles of an M310 reactor, and fine modeling calculation is carried out for the first two cycles. By comparing the calculated values with the measured values, it shows that the errors of calculation results of critical boron concentration, control rod worth and temperature coefficient in the startup physics experiments for the first, second and third cycles are relatively small, which meet the acceptance criteria. The results of comparison of the critical boron concentration and core power distribution with the measured values at different burnup levels show that the maximum boron concentration deviation at the stable burnup point is −39ppm (1ppm = 10−6), and the maximum assembly power error is less than 4.5%. With the increase of burnup level, the core power distribution flattens out and the error decreases gradually. The calculation results show that NECP-X already has the calculation function for the startup physics experiments and multi-fuel cycle of commercial PWRs.
Numerical Analysis of Influence of Positioning and Wrapping Wire Structure on Thermohydraulic Characteristics of Rod Bundle Channel
Liu Sichao, Liu Yu, Tian Ruifeng, Yang Xiaolei, Chen Xi, Li Xiaochang
2023, 44(2): 37-42. doi: 10.13832/j.jnpe.2023.02.0037
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The positioning and wrapping wire design is widely applied in the core design of metal cooled fast reactor and gas cooled fast reactor. In this paper, the effects of pitch, number and shape of positioning and wrapping wires on the flow and heat transfer of supercritical carbon dioxide in rod bundle channel are simulated and analyzed based on three-dimensional fine mesh model of wrapping wire positioning rod bundle channel. The simulation results show that positioning and wrapping wire pitch has a greater influence on the temperature field and flow field than the positioning and wrapping wire number and shape. When the positioning and wrapping wire pitch is less than 200 mm, pressure drop at the inlet and outlet increases significantly, surface heat transfer coefficient increases, and temperature non-uniformity decreases significantly and the temperature unevenness decreases greatly; with the increase in the number of the positioning and wrapping wires, pressure drop at the inlet and outlet increases linearly, but the surface heat transfer coefficient changes little; circular positioning and wrapping wire can achieve a similar effect to that of rectangular positioning and wrapping wire with a smaller cross-sectional area, and trapezoidal positioning and wrapping wire has less influence on flow field than rectangular positioning and wrapping wire.
CHF Mechanism Model in Narrow Rectangular Channel Based on Energy Balance on Heating Wall
Yan Meiyue, Deng Jian, Pan Liangming, Ma Zaiyong, Li Xiang, Wan Lingfeng, He Qingche
2023, 44(2): 43-47. doi: 10.13832/j.jnpe.2023.02.0043
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Narrow rectangular channel is widely used in various fields because of its compact structure and large heat transfer area. The safety and economy of reactor can be improved by improving the prediction method of critical heat flux (CHF) in the narrow rectangular channel and establishing a CHF mechanism model. In this paper, a visual experimental study is conducted on the CHF flowing vertically upward in a narrow rectangular channel. On this basis, a CHF mechanism model based on the heating wall energy balance is developed. A set of constitutive relations are provided to close the developed model, and the experimental data are used to compare and evaluate the new model. The narrow rectangular channel and it has good accuracies of less than ±20% as relative to the experimental values.
Development of Reduced-Order Thermal Stratification Model for Upper Plenum of Lead-Bismuth Fast Reactor Based on CFD
Yang Tao, Zhao Pengcheng, Zhao Yanan, Yu Tao
2023, 44(2): 48-53. doi: 10.13832/j.jnpe.2023.02.0048
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After the emergency shutdown of the lead-bismuth fast reactor, the thermal stratification in the upper plenum has an important impact on the integrity of the reactor structure and the residual heat removal capacity of the natural circulation, which requires special attention. In order to overcome the defects of traditional thermal stratification analysis method, a high-precision full-order snapshot is obtained based on computational fluid dynamics (CFD) code, and a reduced-order thermal stratification model is built by combining proper orthogonal decomposition (POD) with Galerkin projection. After conducting a comparative analysis of thermal stratification with the full-order model of CFD, the results show that the reduced-order thermal stratification model developed can effectively simulate the temperature distribution in the upper plenum and carry out a quick research on the thermal stratification interface characteristics in case of the lead-bismuth fast reactor accident. The research in this paper provides an important analytical tool for studying the thermal stratification mechanism and effectively curbing the thermal stratification.
Study on Three-dimensional Thermal-hydraulic Characteristics of a Space Reactor based on Open Lattice Structure
Wang Zhipeng, Zhao Jing, Shi Lei
2023, 44(2): 54-61. doi: 10.13832/j.jnpe.2023.02.0054
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High temperature gas cooled reactor combined with magnetohydrodynamic (MHD) power generation is an efficient space power system. It can meet the requirements in space tasks for high power and high efficiency and thus has broad application prospects. In this paper, a core scheme composed of 217 fuel rods in a triangular arrangement is proposed in accordance with the design conditions to be met for MHD power generation and with the reference to the open lattice scheme in Prometheus Project. The three-dimensional modeling of the space reactor is carried out after determining the flow model through the experimental data. The thermal-hydraulic characteristics are studied on the basis of taking into consideration the gap structure, the fuel rod power distribution and the in-reactor radiation. Finally, sensitivity analysis on thermal parameters is carried out mainly for the ambient temperature and the external surface emissivity. The calculation results show that the thermal design of the core meets the requirements of material temperature and pressure drop limit. The transverse flow of coolant in the fuel area is not obvious and there is no complex vortex structure. The flow phenomenon is relatively simple. The steady-state thermal calculation results are not sensitive to the change of ambient temperature, but the change of emissivity has a relatively large impact.
Study on the Application of Interfacial Area Transport Equation in One-dimensional Two-fluid Model
Shen Mengsi, Lin Meng
2023, 44(2): 62-68. doi: 10.13832/j.jnpe.2023.02.0062
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In order to resolve the drawbacks of flow regime map used in the one-dimensional two-fluid model based nuclear power plant system analysis code and improve the accuracy of the system analysis code, this paper explores the application of the interfacial area transport equation (IATE) in the one-dimensional two-fluid model to predict the two-phase flow. The one-dimensional two-fluid model solver coupled with IATE (Solver-IATE) is developed and verified with FORTRAN. The numerical simulation of upward bubbly flow in the small adiabatic circular tube is conducted based on Solver-IATE, and the results are compared with the simulational results from the flow regime map. The study shows that the phase interfacial area concentration results calculated using IATE are closer to the experimental value than that using the flow regime map. Thus, the application of IATE in the one-dimensional two-fluid model can improve the accuracy of the calculation of phase interfacial area concentration, thereby improving the accuracy of one-dimensional two-fluid model based nuclear power plant system analysis code in calculating interaction terms between two phases and more accurately predicting the transient response characteristics of reactor.
Model Study on Bubble Slide and Early-Stage Condensation Growth in Rectangular Narrow Channel
Zhang Lin, Liu Hanzhou, Liu Xiaojing, Chen Yong, Chen Deqi
2023, 44(2): 69-76. doi: 10.13832/j.jnpe.2023.02.0069
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Bubble growth in rectangular narrow channel can directly change the phase interfacial area concentration and thus affect the flow channel's heat and mass transfer performance. In order to obtain a model for different types of bubble growth in a narrow flow channel, wall boiling flow heat transfer experiments are carried out based on a through-body visible test section. The bubble growth models are studied for both bubble slide and early-stage condensation conditions in subcooled boiling. The experimental results show that there are two forms of bubble growth, namely, bubble slide growth and early-stage condensation growth. Based on the heat transfer energy equation, the bubble growth models under the two conditions are established, and it is verified through the experimental data that the model error is within 20%. Therefore, this study can provide a more refined model of bubble growth for two-phase numerical simulation of boiling, thus improving the accuracy of bubble behavior prediction.
Study on the Influence of Bionic Guide Vane on the Performance of CAP1400 Main Pump
Liu Haoran, Lu Yeming, Wang Xiaofang, Li Jialing, Zhang Zhigang
2023, 44(2): 77-83. doi: 10.13832/j.jnpe.2023.02.0077
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In order to explore the influence of bionic guide vane on the overall performance of the main pump, a new bionic structural design of guide vane is proposed in this paper by taking the scale model (1:2.5) of CAP1400 main pump as the study object, and an optimized model (the optimal solution of the bionic guide vane) is obtained by optimizing the design platform. By using the numerical method, the hydraulic performance and safety performance of the full three-dimensional model of the main pump are obtained. By comparing and analyzing the performance difference between the original model and optimized model, the following conclusions are drawn: under the design condition, the optimized model can improve the head and efficiency of the main pump by 1.7% and 1.9% respectively; the optimized model can reduce the internal flow field noise and improve the stress distribution on the guide vane surface; the optimized model has little effect on the cavitation performance of the main pump. This paper can provide a reference for the subsequent hydraulic design and acoustic prediction of main pump.
Experimental Study on Steam Critical Flow Leakage from a Small Break in Pipeline of Pressurized Water Reactor
Zhu Mengxin, Yin Songtao, Wang Haijun, Wang Ningning
2023, 44(2): 84-90. doi: 10.13832/j.jnpe.2023.02.0084
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In order to explore the characteristics of steam critical flow leakage from a small break in pipeline during the loss of coolant accident of pressurized water reactor (PWR) nuclear power plant, small-break leakage experiments of pipelines are carried out in this paper to explore the characteristics of saturated/superheated steam critical flow leakage. Based on pressure pipeline fatigue through crack (microchannel), the steam critical flow leakage experiment is carried out within the fluid pressure range of 3~12 MPa and the fluid temperature range of 240℃~320℃. The experimental results show that the critical mass flow rate of steam is positively correlated with the initial fluid pressure and negatively correlated with the initial fluid superheat degree. Compared with the critical flow leakage of supercooled water, the critical mass flow rate of steam is less affected by inlet pressure loss, friction effect and acceleration effect. The one-dimensional isentropic model is used to predict the critical mass flow rate of steam. The mean relative deviation between the predicted value and the experimental value is 14.17%, which indicates that the one-dimensional isentropic model can accurately predict the critical mass flow rate of steam.
Experimental Study on Pool Boiling Heat Transfer of Cr-coated Zirconium Cladding
Zeng Xiehu, Chen Zhiqiang, Wen Qinglong, Du Qiang, Zhang Ruiqian, Du Peinan
2023, 44(2): 91-97. doi: 10.13832/j.jnpe.2023.02.0091
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Chromium (Cr) coated zirconium alloy cladding is considered as one of the most promising cladding materials for accident tolerant fuel (ATF). The degree of influence of the surface state of this material on the heat transfer performance will greatly affect the process optimization direction of coated zirconium cladding. In this paper, experiments are conducted on the Cr-coated zirconium alloy claddings prepared under atmospheric pressure in the pool boiling experimental device for Cr-coated zirconium alloy cladding. The influence law and mechanism of surface states such as roughness on heat transfer are studied. The results show that the improvement of surface roughness can reduce the conditions for the formation of vaporization core and significantly enhance heat transfer under the same wall superheat degree. Within the range of parameters studied in this paper, with the increase of heat transfer surface roughness, CHF shows an upward trend accordingly. Increasing the surface roughness can effectively improve the CHF value. On this basis, the prediction relational expression of the influence of roughness on heat transfer coefficient is also developed in this paper.
Optimization of Turbulent Prandtl Numbers and RANS Models for Liquid Lead-bismuth Eutectic
Deng Shiyu, Lu Tao, Deng Jian, Zhang Xilin, Zhu Dahuan
2023, 44(2): 98-103. doi: 10.13832/j.jnpe.2023.02.0098
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In the engineering field, the RANS turbulence models are often used for thermal and hydraulic numerical simulation. However, the liquid lead-bismuth eutectic (LBE) has unique thermophysical properties, and the applicability of conventional turbulent Prandtl number models and RANS turbulence models to its flow and heat transfer simulation needs to be studied. In order to more accurately describe the flow and heat transfer process of LBE in wire-wrapped fuel assembly, the turbulent Prandtl number models and RANS turbulence models are optimized in this paper based on the large eddy simulation. First, four different turbulent Prandtl number models are used to carry out the large eddy simulation of the flow and heat transfer process of LBE in wire-wrapped fuel assembly, and the experimental data and simulation results are compared and analyzed to optimize these models. Then, based on the optimized turbulent Prandtl number model, the applicability and accuracy of the RANS turbulence models to the numerical simulation of LBE are evaluated. The results show that Cheng's turbulent Prandtl number model and SST k-ω model have the highest accuracy and applicability to the simulation of flow and heat transfer of LBE.
Nuclear Fuel and Reactor Structural Materials
Study on the Effect of C-ion Irradiation on Hardness and Young's Modulus of Nuclear-grade Graphite
Guo Lina, Bian Wei, Peng Shunmi
2023, 44(2): 104-108. doi: 10.13832/j.jnpe.2023.02.0104
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In order to ascertain the effect of ion irradiation dose and temperature on the hardness, Young's modulus and microstructure of nuclear-grade graphite, 0.02 dpa, 0.2 dpa and 2 dpa C4+ are used in this paper to irradiate nuclear-grade graphite at room temperature and 180℃, respectively. The properties and microstructure of nuclear-grade graphite under different ion irradiation conditions are studied by using nano indentor and transmission electron microscopy. The results show that the hardness and Young's modulus increase with the increase of irradiation dose at room temperature, when the irradiation dose is 2 dpa, the peak values of hardness and Young's modulus increas sharply from 0.51 GPa and 15.52 GPa without irradiation to 2.51 GPa and 37.73 GPa respectively. When the irradiation dose is 0~0.2 dpa at 180℃, the hardness and Young's modulus increase with the increase of irradiation dose, and they are higher than those at room temperature with the same irradiation dose, when the irradiation dose reaches 2 dpa, the peak values of hardness and Young's modulus decrease rapidly from 1.72 GPa and 31.53 GPa of 0.2 dpa to 1.32 GPa and 25.91 GPa. The increase of the hardness and Young's modulus is due to the micro-crack closure and the increase of matrix defects in graphite caused by irradiation. The sharp decrease of the hardness and Young's modulus at 180℃ is due to the amorphization of graphite structure caused by irradiation.
Study on Seismic Test of PWR Fuel Assembly
Guo Yan, Zhang Guoliang, Zhang Yanhong, Li Weicai, Hu Xiao, Gu Chenglong
2023, 44(2): 109-115. doi: 10.13832/j.jnpe.2023.02.0109
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The seismic behavior of fuel assembly, as Class I seismic item, is directly related to the operation safety of nuclear power plants. It is usually necessary to verify the reasonableness of the seismic analysis method for reactor fuel assembly through seismic test. By simulating the actual reactor core fuel assembly installation method and designing the pressurized water reactor (PWR) fuel assembly seismic test specimens and devices, this paper carries out an experimental study for different fuel assembly quantity arrangement schemes on the high-performance seismic simulation vibration table. The results show that the first-order frequency of the fuel assembly in the water medium is 2.96 Hz, and the maximum impact force occurs in the position close to the middle of the fuel assembly. The lattice impact force and the relative displacement of the fuel assembly and the accelerations of the simulated core plate and baffle under the action of earthquake are obtained. The test results can be used to establish the seismic analysis model of fuel assembly and verify the analysis software under the designed reference accident condition.
Influence of Neutron Irradiation on Mechanical Properties of Cr-coated Zirconium Alloy
Wu Yazhen, Xi Hang, Li Guoyun, Liu Xiaosong, Zhang Haisheng, Sun Kai, Ning Zhien, Fang Zhongqiang, Liu Shasha
2023, 44(2): 116-121. doi: 10.13832/j.jnpe.2023.02.0116
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In order to study the influence of neutron irradiation on the mechanical properties of Cr-coated zirconium alloy, a neutron irradiation test is carried out in this paper for the Cr-coated zirconium alloy prepared by multi-arc ion plating technology. The mechanical properties were conducted by in situ tensile tests and the coating adhesion was analyzed by micron morphology characterization. Results show that Cr coating zircaloy has a similar irradiation effect with commercial zircaloy cladding. Compared with uncoated zircaloy, Cr coating shows that the yield strength and tensile strength increased but percentage elongation after fracture decreased.The Cr coating shows cracks till a large tensile strain without peeling off before rupture. The Cr coating exhibits good mechanical properties and excellent adhesion with Zircaloy Tubes.
Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions
Wang Zhanwei, Yan Jun, Peng Zhenxun, Ren Qisen, Liao Yehong, Li Sigong, Zhao Yahuan
2023, 44(2): 122-128. doi: 10.13832/j.jnpe.2023.02.0122
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The Fukushima nuclear accident in Japan in 2011 exposed the inherent safety problems of traditional zirconium alloy fuel cladding under LOCA conditions. To investigate the performance of a new Cr-coated zirconium alloy cladding under LOCA conditions, high temperature steam oxidation and quenching experiments under simulated LOCA conditions are carried out for 12~15 μm thick Cr-coated Zr-1Nb alloy cladding tube coated by physical vapor deposition (PVD) process, the oxidation temperature and oxidation time were 1200℃, 1300℃ and 10 min, 20 min, respectively, the quenching was performed around 800℃, then ring compression test was performed for the quenched tube. The results indicated that no spalling was found for Cr coatings under experiment conditions, intense Cr2O3 layer which formed on the outer surface of Cr-coated tube retarded the diffusion of O into zirconium substrate, protecting the zirconium alloy from oxidized into ZrO2 and α-Zr(O) layers, Cr-coated zirconium-alloy cladding remained ductile after quenching. It can be concluded that Cr-coated Zirconium alloy behaves better than traditional Zirconium alloy under the experimental conditions.
Simulation Research on Additional Mass of PWR Fuel Assembly
Guo Yan, Zhang Guoliang, Liu Huan, Li Weicai
2023, 44(2): 129-135. doi: 10.13832/j.jnpe.2023.02.0129
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In order to accurately explore the influence of fluid-structure interaction behavior between reactor coolant and fuel assembly on the vibration characteristics of fuel assembly, this paper takes the pressurized water reactor (PWR) fuel assembly as the research object, applies the computational fluid dynamics (CFD) software FLUENT platform and the dynamic mesh technology thereof and calculates the additional mass of fuel assembly under separate motion conditions of fuel assembly and core baffle by establishing the fuel assembly simulation rod bundle, core baffle and coolant model. The results show that the mean value of the fuel assembly's additional mass coefficient is 2.4712 under the fuel assembly motion condition and –3.4713 under the baffle motion condition, both with a deviation of less than 5% from the literature value. After superimposing of additional masses, the deviation between the calculated value of the fuel assembly vibration frequency and the underwater vibration test result is less than 5%, which verifies the rationality of the analysis method. The simulation calculation method established in this study can be used to calculate the additional mass of PWR fuel assembly.
Analytical Study on Accident Tolerant Fuel Used in the High Performance Pressurized Water Reactor
Yin Chunyu, Gao Shixin, Qian Libo, Qin Xue, Wu Lei, Zhang Yu, Cui Huaiming, Xiao Zhong, Su Guanghui
2023, 44(2): 136-144. doi: 10.13832/j.jnpe.2023.02.0136
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In order to determine the accident tolerant fuel (ATF) element design schemes for future high performance pressurized water reactor (PWR), this study comprehensively analyzes several potential ATF design schemes from the perspectives of safety, economical efficiency and fuel performance by using the methods applicable to the analysis of fuel performance, nuclear design and reactor thermal safety. The results show that the scheme of using SiC composite cladding + high uranium density fuel is good. High uranium density fuel includes UN, U3Si2 and UN-U3Si2 composite fuels, each of them has distinct characteristics, Among them, UN-U3Si2 composite fuel can theoretically develop its strengths and avoid its weaknesses and become one of the characteristics of high uranium density fuel. However, from the perspective of neutron economy, it is necessary to enrich 15N in UN, and the current enrichment technology will greatly increase the manufacturing cost of this type of fuel. Therefore, the design scheme of SiC composite cladding + U3Si2 fuel should be selected for the design of ATF fuel element of high-performance PWR.
Structural Mechanics and Safety Control
Structure-Performance-Cost Integration Multi-Objective Optimization Design for HTR Fuel Storage Canister
Hao Yuchen, Li Yue, Wang Jinhua, Gong Menghang, Wu Bin, Wang Haitao, Ma Tao, Liu Bing
2023, 44(2): 145-151. doi: 10.13832/j.jnpe.2023.02.0145
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Fuel storage canister is a key equipment in the fuel supply system for high temperature reactor (HTR). In order to explore the optimal design scheme, a structure-performance-cost integration multi-objective optimization design method for fuel storage canister is proposed as follows: select the structural plate thickness of fuel storage canister as the design variable, use the Latin hypercube sampling (LHS) method to generate uniform sampling points, obtain the drop response through numerical calculation, and employ the hybrid radial basis function neural network (RBFNN)-feedforward neural network (FFNN) to construct a surrogate model. With the optimization design objectives of minimizing the maximum plastic deformation, the cost and the mass, constrain the radial displacement expansion under the action of pebble bed, and solve the optimization problem by using the strength Pareto evolutionary algorithm (SPEA-Ⅱ). The results show that the safety of the fuel storage canister is significantly improved, and the maximum plastic deformation can be reduced by 20.17%; it has good economical efficiency and lightweight effect, the cost of a single canister can be reduced by 2,128 yuan, and the mass can be reduced by 12.54%. The integration optimization method proposed in this paper can provide reference for the fuel storage canister design.
Research on the Closure Effect of Circumferential Through-Wall Cracks in Stainless Steel Piping under Residual Stress
Liu Zhenshun, Zhang Sheng, Mao Qing, Zheng Xiangyuan
2023, 44(2): 152-158. doi: 10.13832/j.jnpe.2023.02.0152
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The predicted value of the opening displacement of the circumferential through-wall crack (CTWC) in piping under different load levels is a critical parameter for the application of the leak-before-break (LBB) technology. In this paper, both numerical analysis and comparative verification are adopted to study the variation law of critical closure stress of CTWC under typical welding residual stress (WRS) based on actual measured material property curve in engineering for austenitic stainless steel piping with representative geometric dimensions. The analysis results reveal that both the current GE/RPRI method and the NUREG/CR-6837 correction method have underestimated the closure effect of CTWC in piping caused by simplified residual stress field recommended by the Task Group on Codes of American Society of Mechanical Engineers (ASME). In addition, the failure mode of piping under CTWC closure state is explored. On this basis, the influence of crack closure effect on the application of LBB technology is further discussed, which provides technical ideas that can be used as a reference for subsequent engineering practice.
Study on Reliability Evaluation Model for the Reactor Protection System Shutdown Function Considering Self-diagnostics
Wang Mingyang, Zhang Wei, Xu Dongling, Cheng Yuyu, Zheng Mingguang
2023, 44(2): 159-165. doi: 10.13832/j.jnpe.2023.02.0159
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As an important feature of digital instrument control system, online self-diagnostics plays an important role in the reliability analysis of shutdown function of reactor protection system (RPS) in nuclear power plant. By analyzing the influence of self-diagnostics on human factors and periodic testing, the component-level spurious actuation model was established. Taking a typical RPS TX as an example, the dynamic TX sequence-level and system-level module spurious actuation models are established by Markov method. The relationship between TX shutdown function reliability and self-diagnostics is quantitatively calculated by using the system-level module spurious actuation model. Through qualitative analysis and quantitative calculation, the necessity of comprehensive consideration of self-diagnostics in RPS shutdown function reliability analysis is demonstrated, which provides a reference for the subsequent reliability evaluation of RPS shutdown function in China.
Study on Dose Criteria in Safety Classification of Nuclear Power Plant Items
Zhao Danni, He Fan, Pang Zongzhu, Sun Zaozhan, Liu Yu, Yang Zhiyi
2023, 44(2): 166-171. doi: 10.13832/j.jnpe.2023.02.0166
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In the design of nuclear power plant, the safety classification of items is carried out to ensure that the design, manufacturing and construction of the items meet appropriate requirements and that the reliability consistent with their functions can be achieved. This paper briefly describes the method of safety classification of nuclear power plant items according to safety importance and the factors that should be considered. For the consequences of failing to perform a safety function that should be considered in the safety classification of the items, the dose criteria for the public and staff under different working conditions of nuclear power plant are used to divide them into high, medium and low consequences. Through the study, some suggestions on the quantification of high, medium and low radioactive consequences are put forward so as to make this method more operable in the safety classification of nuclear power plant items.
Circuit Equipment and Operation Maintenance
Study on Test Scheme for Friction Properties and Service Life of Secondary Seal of Reactor Coolant Pump Hydrodynamic Shaft Seal
Cong Guohui, Zhang Yixun, Duan Yuangang
2023, 44(2): 172-176. doi: 10.13832/j.jnpe.2023.02.0172
Abstract(113) HTML (20) PDF(26)
Abstract:
The friction properties and service life of the secondary seal are the key factors affecting the service life of the hydrodynamic shaft seal of the reactor coolant pump. In order to study the long-term service life of the secondary seal, a high-frequency reciprocating test device is established to simulate the reciprocating motion of the secondary seal under the high pressure medium condition, and the friction change data of the O-shaped rubber sealing ring and dual metal parts of the secondary seal under three working conditions of increasing frequency, increasing displacement amplitude and increasing medium pressure are obtained. The results show that the secondary seal of hydrodynamic shaft seal of the reactor coolant pump is in a fretting friction state under the normal operating condition frequency of 25 Hz, amplitude of about 30 μm and medium pressure of 5.3 MPa. The friction properties of the secondary seal will remain basically unchanged when the frequency is increased to a level no higher than 300 Hz under the normal operating conditions, but the friction properties will change significantly when the frequency is increased to a level higher than 500 Hz. The friction properties of the secondary seal will remain basically unchanged when the amplitude or the fluid pressure is increased under the normal operating conditions. It can be seen that the effect of increasing the frequency to a level below 300 Hz on the service life of secondary seal can be considered as linear increase, which can effectively reduce the service life verification time.
Study on Hydrodynamic Characteristics of Transient Process of Reactor Coolant Pump Shaft Stuck Accident
Li Yibin, Qu Zehui, Guo Yanlei, Li Donghao, Yang Congxin, Pan Jun, Wang Xiuyong
2023, 44(2): 177-184. doi: 10.13832/j.jnpe.2023.02.0177
Abstract(1209) HTML (43) PDF(40)
Abstract:
In order to explore the hydrodynamic characteristics of transient process of the reactor coolant pump shaft stuck accident, a full three-dimensional simplified model of the reactor primary circuit system was established by dynamically matching the hydraulic characteristics of the reactor coolant pump and the resistance characteristics of the system pipeline. The transient numerical simulation of the reactor coolant pump shaft stuck accident condition is carried out by using the computational fluid dynamics (CFD) method, and the transient variations of external characteristics, internal pressure field, impeller blade load and force of the reactor coolant pump under different shaft stuck conditions are obtained. The study shows that the shorter the shaft stuck time, the more dramatic the transient variation of the reactor coolant pump characteristic parameters, and the more serious the impact of the accident. Taking the moment when the impeller speed just drops to 0 r/min as the node, under three shaft stuck conditions (i.e. shaft stuck time = 0.1 s, 0.3 s and 0.5 s), the flow rate decreases to 82.3%, 61.4% and 49.6%, respectively, of that under the normal operation. The head of the reactor coolant pump reaches the reverse extreme value, i.e. −137.7%, −87.4% and −56.9% , respectively, of the value under the normal operation. The pressure difference between the two sides of the impeller blade reaches the maximum, i.e. 1.34 MPa, 0.73 MPa and 0.47 MPa, respectively, and a relatively concentrated low-pressure area is formed on the side of the blade working surface in the impeller blade and in the middle part of the guide vane flow channel. The reverse extreme value of the axial force on the impeller reaches −159.3%, −96.5% and −65.5%, respectively, of the value under the normal operation. The numerical prediction method provides certain data support for the dynamic safety assessment of the reactor hydrodynamic system.
Research on Degradation Measures of Critical Components in Daya Bay Nuclear Power Plant Digital Transformation Project
Xu Ying, Zhang Guojun, Zhao Hao, Wang Zhixian, Zhao Yan
2023, 44(2): 185-190. doi: 10.13832/j.jnpe.2023.02.0185
Abstract(145) HTML (59) PDF(23)
Abstract:
The analog control system of Daya Bay Nuclear Power Plant consists of discrete electrical components integrated by hard wiring, and it is planned to undergo digital transformation during the 30a overhaul. Due to the functional limitation of the analog platform, the single failure of equipment has a great impact on the control loop, and a large number of instrumentation and control devices are identified as critical components. Reducing the number of critical components is a key goal of the Daya Bay Nuclear Power Plant Digital Transformation Project. The Project takes pressurizer water level control loop—the important control loop of nuclear island as the research object and systematically studies and practically applies the degradation measures of critical components. By using the automatic voting function and fault diagnosis function of the reconstructed digital control system (DCS), the voting optimization scheme of measurement channel and the double redundancy design scheme of output channel are put forward. Using the principle of interface optimization, main control unit and auxiliary control unit of the pressurizer water level control loop are concentrated in a functional subgroup, and a large number of cross-subgroup interfaces are cancelled. With the application of the above measures, the number of critical components for instrument control of the pressurizer water level control loop has been reduced from 17 to 0, which has comprehensively improved the reliability of control functions, reduced the management cost of plant equipment and and provided an important reference scheme for achieving the key objectives of the Project.
Study and Prevention of Steam Flow Induced Vibration of Nuclear Power Plant Condenser
Zu Shuai, Chen Jie, Che Yinhui, Wang Guoshan, Zhao Qingsen, Zhang Qiang, Wu Zhenpeng
2023, 44(2): 191-197. doi: 10.13832/j.jnpe.2023.02.0191
Abstract(152) HTML (57) PDF(23)
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In view of a number of titanium tube cracking events caused by steam flow induced vibration of a certain type of nuclear power condenser under half-side condenser operation conditions, this paper uses the computational fluid dynamics (CFD) method based on porous medium model to conduct a full three-dimensional numerical simulation of the steam-side flow field in the throat and tube bundle area of the condenser and calculates and obtains the steam-side velocity field of the condenser under half-side condenser operation conditions and the steam flow induced vibration risk coefficient distribution of titanium tubes. The simulation calculation results show that the steam flow induced vibration risk of the surface titanium tube in the finger gap area above the air-cooling area of the heat exchange module close to the vertical centerline of the condenser is relatively high under half-condenser operation conditions, and it is necessary to consider installing anti-vibration bars or taking preventive tube plugging measures to reduce the steam flow induced vibration risk. According to the numerical simulation calculation results under the leakage conditons of the condenser during half-side condenser operation and the actual operation records of the nuclear power unit, it is suggested that the electric power limits for the safe operation of the nuclear power plant condenser under the half-condenser operation conditions in summer and winter should be 900 MW and 600 MW respectively. Since the span of the condenser is a bit too large, the span of titanium tube should be shortened to less than 610.5 mm so as to avoid steam flow induced vibration.
Impact Analysis of Single Fuel Rod Damage during Fuel Assembly Repair
Chen Xiaoqiang, Yin Shuhua, Wei Xuehu, Lyv Weifeng, Xiong Jun
2023, 44(2): 198-202. doi: 10.13832/j.jnpe.2023.02.0198
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Abstract:
The cumulative effective dose to fuel assembly repair workers, the total radioactive activity of gaseous effluent released to the environment and the impact on gaseous effluent discharge monitoring are calculated and evaluated by taking the radioactive substances released from single fuel rod damage during the fuel assembly repair in nuclear power plant as the analysis object. The analysis results show that in case of single fuel rod damage, the cumulative effective dose received by each worker involved in the fuel assembly repair is 12.2 mSv, which is lower than 20 mSv—the annual average effective dose limit for workers’ occupational exposure stipulated in GB 18871-2002; the total radioactive activities of noble gases and iodine in the gaseous effluent released to the environment are 3.51 × 1011 Bq and 2.17 × 108 Bq respectively, which are far lower than the annual emission control values (6.0 × 1014 Bq and 2.0 × 1010 Bq) specified in GB 6249-2011. The reading of noble gas exhaust measuring instrument is lower than 1.0 × 1011 Bq/h 40 minutes after the fuel rod damage, so the nuclear power plant does not need to be put on emergency standby.
Study on Molecular Dynamics of the Adsorption and Film Formation of Octadecylamine on Carbon Steel Surface
Li Chao, Huang Junlin, Wang Lu, Zhou Keyi
2023, 44(2): 203-209. doi: 10.13832/j.jnpe.2023.02.0203
Abstract(1161) HTML (18) PDF(22)
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P280GH carbon steel pipe is widely used in the main feedwater system of Hualong One unit, and octadecylamine (ODA) is applied to form a corrosion inhibition film by adsorption on the inner wall of the pipe. ODA can effectively inhibit corrosion and avoid pipe failure and serious scaling of steam generator (SG). However, the mechanism of adsorption and film formation of ODA on carbon steel surface is still unclear, which seriously restricts the performance optimization, optimization and application of ODA. In order to solve this problem, molecular dynamics (MD) simulation method is used to carry out the study. The results show that the nitrogen atoms at the head of the ODA molecule form coordination bonds with the iron atoms on the carbon steel surface, which promotes the adsorption and anchoring of the ODA molecule. The microscopic configuration of the corrosion inhibition film formed is dependent on the concentration of ODA. At low concentration, the corrosion inhibition film has a poorly interwoven single-layer configuration between ODA molecular tail chains, and as the concentration increases, the corrosion inhibition film evolves into a tightly interwoven complex bilayer configuration between ODA molecular tail chains. After the concentration of ODA exceeds a certain threshold, the configuration of the corrosion inhibition film no longer changes significantly, and the ODA molecules that are not adsorbed to form the film eventually accumulate to form colloidal micelles.
Study on General Layout of Main Plant Building of Thorium-Based Molten Salt Experimental Reactor with Liquid Fuel
Bei Chen, Jia Xiaopan, Xue Jing, Wang Zhenzhong
2023, 44(2): 210-215. doi: 10.13832/j.jnpe.2023.02.0210
Abstract(133) HTML (21) PDF(41)
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In order to realize the reasonable and compact general layout of the main plant building for 2 MW thorium based molten salt experimental reactor (TMSR) with liquid fuel, the overall design characteristics of the main plant building are determined in this paper according to the type characteristics, top-level design and system functional requirements of the molten salt reactor, and the relative position characteristics of the key equipment and items of TMSR are discussed. Besides, the general layout scheme of the plant is finally obtained through reasonable planning of the functional zoning and equipment layout of the plant. Through the implementation of this project, it provides a basic platform for the system integration and verification of TMSR and provides technical support and experience for the design and construction of small modular thorium-based molten salt demonstration reactor.
Column of Science and Technology on Reactor System Design Technology Laboratory
Study on Thermal Hydraulic Characteristics of Two-phase Discharge Process under Different Initial Events
Yu Na, Wu Dan, Huang Tao, Wang Zefeng
2023, 44(2): 216-221. doi: 10.13832/j.jnpe.2023.02.0216
Abstract(112) HTML (21) PDF(22)
Abstract:
This paper studies the complex two-phase thermal hydraulic process after the opening of the pressurizer safety valve so as to determine the two-phase discharge characteristics of the pressurizer safety valve under different initial events. In this paper, the autonomous system analysis program ARSAC is used to model and analyze the upstream and downstream of the pressurizer safety valve, and three typical valve discharge processes are selected, including the accidental opening of the pressurizer safety valve, the complete loss of main steam flow that causes the opening of one or more safety valves, and the accidental starting of the safety injection pump that causes the intermittent opening of the pressurizer safety valve under the condition of low temperature overpressure protection. The complex two-phase thermal hydraulic characteristics involved in the process of water seal and steam (or water) discharge after the opening of the pressurizer safety valve are studied. The results show that the ARSAC program can capture the flow pattern changes in the pipeline during two-phase discharge; the water seal will form an obvious flow peak through the downstream pipeline, and the flow peak and time characteristics of the downstream pipeline caused by the discharge process are different under different upstream initial conditions. The study in this paper can provide guidance for load analysis, safety evaluation and design optimization.
Preliminary Conceptual Design of Ultra-high Flux Fast Neutron Test Reactor Core
Cai Yun, Wang Lianjie, Wang Liangzi, Xia Bangyang, Lou Lei, Zhang Bin, Zhang Ce, Hu Yuying
2023, 44(2): 222-226. doi: 10.13832/j.jnpe.2023.02.0222
Abstract(159) HTML (57) PDF(47)
Abstract:
To meet the development need of advanced nuclear systems, an ultra-high flux reactor (UFR) core design concept is proposed in this paper. In this concept, plate-type fuel and square fuel assembly design is adopted, and a wide flow channel is provided to ensure a high volume share of the core coolant. The core is provided with 52 boxes of fuel assemblies, 8 boxes of control rod assemblies and a thick reflective layer. The results show that the cycle length of the core can reach 100 equivalent full power days (EFPD) by the core conceptual design scheme evaluation. The maximum neutron fluence rate of the proposed ultra-high flux reactor can reach 1.0×1016 cm−2·s−1.
Preliminary Conceptual Design of Ultra-high Flux Reactor Core with Annular Elements
Wang Lianjie, Cai Yun, Wang Liangzi, Xia Bangyang, Lou Lei, Zhang Bin, Zhang Ce, Hu Yuying
2023, 44(2): 227-231. doi: 10.13832/j.jnpe.2023.02.0227
Abstract(204) HTML (34) PDF(38)
Abstract:
Based on the annular fuel elements, a conceptual design of ultra-high flux reactor (UFR) is proposed. The fuel assembly design adopts a hexagonal assembly composed of 61 fuel elements. The core is designed with 52 boxes of fuel assemblies, 9 boxes of control rod assemblies and a thick reflecting layer. The key parameters such as the core cycle length, neutron fluence rate, neutron energy spectrum and neutron spatial distribution are given from the evaluation of the core conceptual design scheme. The results show that the maximum neutron fluence rate of the proposed ultra-high flux reactor can reach 1.0 × 1016 cm−2·s−1 under the current general parameters.