核动力工程

Calculation of Source Terms for Water-cooled Fusion Reactor Based on Deviation Effect Nuclide Screening Method

Guo Qingyang, Zhang Jingyu, Zhang Huijie, Wang Qingbin

2022, 43(5): 1-6. doi: 10.13832/j.jnpe.2022.05.0001

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Abstract:
The activated corrosion products are the main radioactive source terms in the normal operation of water-cooled fusion reactor, and are generally solved by analytical methods, but analytical methods cannot improve the calculation speed while meeting the accuracy requirements. In this paper, a nuclide screening method based on quantitative deviation effect analysis is proposed, which defines the two parameters of radioactivity and dose rate as deviation effect indicators. By analyzing the deviation effect indicators, nuclides meeting the acceptance criteria are screened to determine the target nuclides required for calculation. This analysis method can not only meet the accuracy requirements, but also improve the calculation efficiency. This nuclide screening method is applied to the source term analysis of activated corrosion products in the International Thermonuclear Experimental reactor (ITER) limiter-outer cladding water-cooled loop (LIM-OBB), and compared with the high-precision benchmark solution under this issue. The results show that the relative deviations of the specific activity calculation results of important activated corrosion product nuclides such as ⁵⁷Co, ⁵⁸Co, ⁵⁵Fe, and ⁵¹Cr compared with the benchmark solution are all controlled within 1.5%; The calculation efficiency of the nuclide screening method is 279 times higher than that of the benchmark solution.

Numerical Simulations of Flow Field in First-Stage Steam-water Separator of Steam Generator

Hu Shiqu, Gu Kai, Sun Xinyu, Yuan Jingqi, Dong He, Hui Jiuwu

2022, 43(5): 7-11. doi: 10.13832/j.jnpe.2022.05.0007

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Abstract:
Taking the first-stage steam-water separator of the steam generator of PWR nuclear power plant as the research object, the flow field characteristics and steam- water separation performance of wet steam after entering the steam-water separator are simulated by using the computational fluid dynamics (CFD) calculation software ANSYS Fluent. In the simulation process, a calculation model combining the Euler multiphase flow model and the k-ε Realizable turbulence model is used. The simulation results of the working medium flowing through the steam-water separator show that when the vapor phase and the liquid phase flow to their respective outlets through the steam-water separator, there is obvious stratification. Comparing the simulation results under different tangential outlets and droplet diameters, it is shown that the larger the outlet area, the better the separation effect of the separator on droplets; In the diameter range of 0.01-0.10 mm, the larger the droplet diameter, the better the separation effect.The simulation results of the separation efficiency of the steam-water separator under different load conditions show that the separation efficiency slightly decreases with the increase of the unit load.

Transient Characteristics Analysis of Single Parameter Disturbance in Molten Salt Reactor

Chen Guoyu, Li Minghai, Zou Yang, Xu Hongjie

2022, 43(5): 12-19. doi: 10.13832/j.jnpe.2022.05.0012

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Abstract:
As an innovative reactor, the thermal-hydraulic characteristics of molten salt reactor are very different from other reactors. Disturbance transient analysis helps to fundamentally understand its safety characteristics and operating conditions. In order to study the transient characteristics of molten salt reactor operation, this study takes liquid fuel molten salt reactor (MSR) as the research object, and uses the modified RELAP5/MOD4.0 program to carry out disturbance transient analysis under steady-state operation conditions. Disturbance variables include mass flow rate of primary circuit, mass flow rate of secondary circuit, mass flow rate of air radiator and inlet air temperature of air radiator. Main operating parameters, such as power, core inlet and outlet temperature, secondary circuit inlet and outlet temperature, and characteristic time, are analyzed. The results show that the final state of the MSR under various disturbance transients tends to be stable without severe transient changes, which is an intuitive characterization of its inherent stability characteristics. According to the change of power and temperature under disturbance, the control method of power and different circuit temperature is proposed.

Analysis and Research of Coupled Brayton Cycle System for Small Fluorine Salt Cooled High Temperature Reactor

Liu Xiuting, Huang Yanping, Wang Yangle, Liu Guangxu, Zhuo Wenbin, Li Xinyu

2022, 43(5): 20-26. doi: 10.13832/j.jnpe.2022.05.0020

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Abstract:
In order to meet the energy conversion requirements of small fluoride cooled high temperature reactor (FHR), an efficient, compact and water-free cooling power conversion system is developed. In this paper, the thermoelectric conversion efficiency, exergy efficiency and exergy loss distribution of supercritical carbon dioxide (SCO₂), air, argon (Ar), nitrogen (N₂) and xenon (Xe) in different Brayton cycle configurations are compared. It is found that SCO₂ Brayton cycle has the highest thermoelectric conversion efficiency and exergy efficiency compared with other working medium cycles, and its structure is more compact, easy to miniaturization and modularization, and has more advantages in coupling with small fluorine salt cooled high temperature reactor; the configuration of SCO₂ Brayton cycle is optimized, and the optimal cycle configuration mode matching the small fluoride cooled high-temperature reactor is obtained, which constitutes an inherently safe modular small fluoride cooled high-temperature reactor thermoelectric conversion system, providing a new research idea for energy utilization in the west.

Assessment of LUT-2006 Subcooled CHF Prediction at PWR Conditions with Freon CHF Data

Guo Junliang, Kong Huanjun, Gui Miao, Peng Yujiao, Shan Jianqiang

2022, 43(5): 27-33. doi: 10.13832/j.jnpe.2022.05.0027

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Abstract:
An experimental study on the critical heat flux (CHF) is carried out in a circular tube with an inner diameter of 8 mm using R-134a as the modeling working medium. The variation trend of CHF parameters of R-134a is discussed, and Katto’s fluid modeling method is evaluated. The results show that CHF is only affected by local parameters, and the influence of length-diameter ratio can be ignored. The CHF parameter trend of R-134a is similar to that of typical water. Katto's modeling method has high accuracy at low critical air content and even negative critical air content. The CHF experimental data of R-134a are converted into equivalent water data by modeling method and compared with CHF Lookup Table (LUT)-2006. The evaluation results show that LUT-2006 has high prediction accuracy even though there is almost no subcooled CHF data under PWR conditions.

Experimental Study on Saturated Pool Boiling Bubble Behavior of ATF Chromium Coated Zirconium Alloy Cladding at Atmospheric Pressure

Wen Qinglong, Zeng Xiehu, Du Qiang, Chen Zhiqiang, Zhang Ruiqian, Du Peinan

2022, 43(5): 34-42. doi: 10.13832/j.jnpe.2022.05.0034

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Abstract:
Chromium (Cr) - coated zirconium alloy cladding is one of the most promising new cladding materials for accident resistant fuel (ATF). The study of bubble dynamics on the surface of this new material is helpful to evaluate whether it has better heat transfer performance. experiments are carried out on Cr-coated zirconium alloy cladding prepared by different process methods in a pool boiling experimental facility for Cr-coated zirconium alloy cladding under normal pressure. The effects of surface states such as roughness on bubble generation, growth and departure behaviors are investigated.The results show that the bubble contact angle is related to the surface roughness of Cr coating. The larger the roughness, the smaller the bubble contact angle; The bubble departure diameter of the four Cr-coated zirconium alloy cladding samples prepared under different coating processes ranges from 1.256~1.446 mm, and the bubble departure frequency ranges from 29.99~50.97 Hz; the bubble departure diameter is negatively correlated with the roughness, and the departure frequency is positively correlated with the roughness; The deviation between the bubble departure diameter prediction model and the experimental data is ±6%, and the deviation between the departure frequency prediction model and the experimental data is ±3%.

Experimental Study on Flow Patterns and Void Fraction of Non-swirling and Swirling Gas-liquid Two-phase Flow Based on WMS

Liu Shuai, Chen Cong, Liu Li, Gu Hanyang, Zhang Jiarong

2022, 43(5): 43-50. doi: 10.13832/j.jnpe.2022.05.0043

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Abstract:
In order to explore the flow pattern characteristics and the spatial-temporal distribution characteristics of void fraction before and after the gas-liquid two-phase flow pattern changes from non-swirling state to swirling state, based on the high-speed camera and the self-developed wire mesh sensor (WMS) measurement technology, the spatial-temporal evolution characteristics of the phase state of the air-water two-phase flow under the action of a swirling device in a horizontal tube with an inner diameter of 30 mm are studied visually. Under the centrifugal force induced by the swirler, there are obvious bubble coalescence behavior and droplet deposition phenomena in the flow field. Among them, the bubble flow will be transformed into swirling gas column flow, the plug flow into swirling intermittent flow, the slug flow into swirling annular flow, and the annular flow into swirling ribbon flow; compared with slug flow and annular flow, the fluctuation amplitude of the section average void fraction of bubbly flow and plug flow at the swirler outlet is significantly weakened, but the centrifugal force field does not significantly change the section average void fraction of each flow pattern before and after the transition from non-swirling state to swirling state.

Minimum Required Burnup Analysis of Liquid-fueled Molten Chlorine Salt Fast Reactor

Peng Yu, Zhu Guifeng, Niu Miaomiao

2022, 43(5): 51-55. doi: 10.13832/j.jnpe.2022.05.0051

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In order to explore the average discharge burnup depth of liquid-fueled molten chlorine salt fast reactor operating in the breed-and-burn mode, based on the neutron balance analysis method, five common chlorine salts are selected, and the scheme of online removal of fission gas and insoluble fission products is proposed to maintain the breed-and-burn operation mode. The effects of heavy metal density of chloride salt and online treatment scheme on minimum required burnup and acceptable core neutron loss term for maintaining the breed-and-burn mode in the infinite cell model are mainly studied and analyzed. The analysis shows that the minimum required burnup of 68NaCl-32UCl₃ and 20UCl₃-80UCl₄ is 30.47% FIMA (FIMA refers to the ratio of the number of fissioned atoms to the total number of initially loaded metal atoms) and 10.28% FIMA respectively; After removing fission gas and insoluble fission products, the acceptable neutron loss term for 60NaCl-40UCl₃ is increased from 3.49% to 10.68%. The results show that the density of heavy metals in chloride salts has a significant impact on the minimum required burnup, and the removal of fission gases and insoluble fission products can greatly improve the neutron economy of the fuel salt system, and at the same time, improve the acceptable core neutron loss term for the breed-and-burn operation mode.

Research on Single-phase Inter-tube Pulsation Characteristics in Inverted U-tube of Steam Generator Based on Single-phase Compressible Model

Li Zhenzhong, Ma Zaiyong, Zhang Duo, Bu Shanshan, Sun Wan, Zhang Luteng, Zhu Longxiang

2022, 43(5): 56-62. doi: 10.13832/j.jnpe.2022.05.0056

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Abstract:
The inter-tube pulsation under single phase condition can make the backflow of inverted U-tube steam generator occur in advance, thus threatening its safety. In order to explore the inter-tube pulsation characteristics under single phase, based on the single-phase compressible model, the effects of tube length, primary side inlet temperature and pressure, secondary side temperature and heat transfer coefficient on the single-phase inter-tube pulsating critical velocity are studied numerically. The results show that when the tube length is short, the critical velocity increases significantly with the increase of tube length, and the longer the inverted U-tube is, the greater the critical velocity is. The increase of temperature and pressure at the inlet of the primary side will increase the critical velocity, making the single-phase inter-tube pulsation more likely to occur. The increase of the heat transfer coefficient on the secondary side will decrease the critical velocity corresponding to the single-phase inter-tube pulsation. However, the influence of the secondary side temperature on the critical velocity is non-single-valued. With the increase of the secondary side temperature, the critical velocity first increases and then decreases.

Numerical Study on Dryout Critical Boiling in Vertical Circular Tube under Rolling Conditions

Qi Wei, Bu Shanshan, Li Zhenzhong, Ma Zaiyong, Zhang Luteng, Chen Deqi

2022, 43(5): 63-69. doi: 10.13832/j.jnpe.2022.05.0063

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Abstract:
The three-dimensional numerical calculation of the dryout critical heat flux (Dryout CHF) in a circular tube under rolling conditions is carried out. The phase distribution characteristics, the location of critical heat flux (CHF) and the maximum wall temperature in the circular tube under rolling condition are studied, and the characteristics of heat transfer coefficient along the tube wall are analyzed. The results show that under the rolling condition, the phase distribution in the circular tube changes periodically, and the position of CHF also changes periodically; At the same time, it is found that the rolling motion leads to a higher maximum wall temperature, so the rolling condition will make the boiling critical phenomenon more serious. With the change of flow pattern and boiling heat transfer mechanism, the wall heat transfer coefficient will also change significantly along the flow direction. This study can provide a reference for the numerical prediction of dryout CHF under rolling condition.

Uncertainty Analysis and Application of Source Term under Bypass Release Accident of HPR1000

Chang Yuan, Shi Xueyao, Wang Henan, Wang Hui

2022, 43(5): 70-75. doi: 10.13832/j.jnpe.2022.05.0070

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Abstract:
The radioactive source term analysis of serious accidents in nuclear power plants is the focus of attention in the field of nuclear safety, and the source term analysis has great uncertainty. In this paper, based on the best estimate plus uncertainty (BEPU) analysis method, HPR1000 serious accident analysis model is established by using the integrated serious accident analysis program. For the first time, from the point of view of the whole accident process, a process suitable for source term uncertainty analysis of HPR1000 serious accident is developed, and this method is used to analyze the source term uncertainty of containment bypass release category. The research content of this paper enriches the source term analysis of the serious accident of HPR1000, and also lays a foundation for the development of the three-level probabilistic safety analysis (PSA) technology of HPR1000.

Numerical Simulation of Fouling Deposition in Narrow Rectangular Channel

Tan Jiaqi, Liu Dalin, Liu Xiaojing

2022, 43(5): 76-81. doi: 10.13832/j.jnpe.2022.05.0076

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Abstract:
In order to study the fouling caused by the crystallization deposition of CaSO₄ solution in the narrow rectangular channel and its effect on heat transfer, based on the reasonable fouling deposition, fouling erosion and fouling thermal resistance models, the crystallization deposition simulation calculation of fluids with certain heat flux, inlet velocity , inlet temperature and fluid concentration is carried out by using FLUENT software and user-defined function (UDF).The research results show the fouling generation of this working medium and its influence on heat transfer, and the influence of three factors, namely heat flux, inlet velocity and fluid concentration, on fouling deposition is obtained: the fouling resistance increases with the increase of heat flux, decreases with the increase of inlet velocity, and increases with the increase of fluid concentration. This study can be used to simulate the fouling deposition process in the narrow rectangular channel of plate-shaped fuel elements due to crystallization.

Experimental Study on Two-phase Natural Circulation Characteristics under Low-voltage and Low-power Conditions

Liu Xiaoya, Zhang Yongfa, Jiang Lizhi, Jiao Meng, Zhao Xinwen, Wang Xinming, Wang Xiaolong

2022, 43(5): 82-88. doi: 10.13832/j.jnpe.2022.05.0082

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Abstract:
Compared with the land-based nuclear power plant, the passive safety system of marine nuclear power plant has lower operating pressure and frequent changes in operating power. Under the condition of two-phase natural circulation, the flow in the passive safety system is more complex and changeable. In order to further study the circulation characteristics of two-phase natural circulation under the condition of low pressure and low power, the principle test bench of two-phase natural circulation is built based on the proportional analysis method, and the influence of power and initial liquid level height on the characteristics of natural circulation under the condition of low pressure is studied. The results show that under low pressure, the pressure and flow of the system after stable operation are affected by the initial liquid level and power. When the power is 50 kW, the higher the initial liquid level is, the greater the pressure after the system is stabilized, but the flow difference is small; When the initial liquid level is constant, the power is in the range of 40% full power to 100% full power. With the increase of power, the pressure after the system is stabilized also gradually increases. This provides a direction for the follow-up research of two-phase natural circulation of the test bench, and also provides a reference for the research of passive safety system of marine nuclear power plant.

Numerical Analysis of Flow Distribution Characteristics at Core Inlet of Small Reactor under Rolling Condition

Liu Yirui, Dong Xiuchen, Zhang Xin, Yuan Jiangtao

2022, 43(5): 89-94. doi: 10.13832/j.jnpe.2022.05.0089

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Abstract:
In order to study the flow distribution characteristics of coolant at the inlet of reactor core during forced circulation of small reactor under rolling condition, a small reactor model is established by using CFD software STAR-CCM+ with numerical calculation method, model verification is completed, and the flow distribution characteristics at the inlet of reactor core under rolling condition are studied. The results show that the greater the distance from the core inlet to the rolling axis and the greater the rolling amplitude, the greater the coolant flow fluctuation at the core inlet. The long-period rolling has little effect on the flow, but the coolant flow will jump with the decrease of the cycle. The uneven distribution of coolant at the core inlet increases with the increase of rolling angle, but it is not sensitive to the change of rolling cycle.

Study on Nucleation Site Density of Surfaces Modified by Femtosecond Laser

Tang Wuyu, Zhou Lei, Zhang Junyi, Yan Xiao

2022, 43(5): 95-99. doi: 10.13832/j.jnpe.2022.05.0095

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Using stainless steel as the experimental piece, femtosecond laser technology is used to fabricate ordered micro-nano structures to prepare modified surfaces. The pool boiling nucleation site density experiment is carried out with deionized water as the working medium, and the nucleation site density experimental data of three different experimental surfaces (conventional surface, modified surface 1, and modified surface 2) under different thermal parameters are obtained. The variation law of nucleation site density with wall superheat degree is quantitatively analyzed, and an improved nucleation site density model is obtained by fitting on the basis of the Liquan model. It is found that the nucleation site density of the three experimental surfaces increases with the increase of wall superheat degree, and the nucleation site density of the modified surface is significantly higher than that of the conventional surface under the same thermal parameters.The improved model optimizes the predicted value of nucleation site density, and the predicted value is in good agreement with the experimental data.

Research on Aerosol Behavior of Combustion Products under Sodium Pool Fire Accident

Sun Hongping, Den Jian, Luo Yuejian, Zhang Ming, Xu Youyou, Wu Xiaoli, Liu Lili, Chen Chong, Qiu Suizheng, Su Guanghui

2022, 43(5): 100-108. doi: 10.13832/j.jnpe.2022.05.0100

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Abstract:
In the safety assessment of sodium-cooled fast reactor (SFR), it is particularly important to analyze the aerosol behavior of combustion products in sodium pool fire accident caused by sodium leakage. In this paper, the analytical program REBAC-SFR for aerosol behavior of combustion products in sodium pool fire accident is developed by coupling the sodium pool fire combustion model with the aerosol dynamics model. Based on this program, SAPFIRE-D1 and ABCOVE sodium pool fire experiments are simulated and compared with the experimental data. The results show that the program developed in this paper has good reliability and correctness, and can provide a theoretical tool for the analysis of the aerosol behavior of combustion products under the sodium pool fire in the sodium process room.

Experimental Study on Effect of Subcooling Degree and Roughness on Quenching Boiling of FeCrAl Plate

Zhang Qiqi, Luo Yan, Lu Tao, Deng Jian, Zhang Xilin, Zhou Zhaochun

2022, 43(5): 109-114. doi: 10.13832/j.jnpe.2022.05.0109

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In order to investigate the effect of subcooling degree and surface roughness on the quenching boiling process of FeCrAl plate, a visual experimental study is carried out on the quenching boiling process of FeCrAl plate under different subcooling degree and surface roughness. The temperature inside the plate is measured by thermocouple, and the surface temperature and heat flux of the plate are solved analytically by using the inverse problem of heat conduction; Through the comparative analysis of the experimental phenomena, the effect of subcooling degree and surface roughness on the plate quenching boiling process is investigated, and the relationship between subcooling degree and the minimum film boiling temperature is established. The results show that during the quenching boiling process, the Kelvin-Helmholtz unstable wave is formed on the surface of the plate, and the sudden cooling front caused by the rupture of the gas film is in the shape of “parabola”. With the increase of subcooling degree, the minimum film boiling temperature and critical heat flux increase, the cooling rate of plate surface increases, and the duration of quenching boiling process shortens; Larger surface roughness can promote the progress of quench boiling on flat surfaces but the effect is small.

Diagnostic Method for Damage Equivalent of Reactor Fuel Elements

Lin Xiaoling, Liu Cuihong

2022, 43(5): 115-118. doi: 10.13832/j.jnpe.2022.05.0115

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Damage equivalent is an important index to measure the severity of reactor fuel element damage. However, the damage equivalent is not operable in the decision-making application because it cannot be directly measured. It is necessary to establish the monitoring index corresponding to the damage equivalent. Based on practical experience, the typical nuclides that can be used for damage diagnosis are analyzed and determined, and the transfer relationship between the activity concentration of fission product nuclides in the reactor primary circuit coolant and the damage equivalent of fuel elements is established; the experimental method of sampling and analyzing the primary circuit coolant is given, and the issues that should be paid attention to in the experimental process are pointed out; a method for diagnosing the damage equivalent by monitoring the nuclide activity concentration of typical fission products in the primary circuit coolant is established, and the main factors affecting the uncertainty in the diagnosis are analyzed. This study provides a technical method for diagnosis of reactor fuel element damage equivalent.

Design and Research of Data Management System for the Whole Life Cycle of Small Modular Reactor Structural Materials

Song Danrong, Xu Bin, Liu Jia, Qin Dong, Zhang Xianjun, Wang Zhuo, Zhu Hong

2022, 43(5): 119-125. doi: 10.13832/j.jnpe.2022.05.0119

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The structural materials of small modular reactors ( SMR) are characterized by a wide variety of types, sources and formats. Based on the background of modern information technology and big data, combined with the particularity of small reactor structure material data, a special data management system covering the whole life cycle of small reactor structure materials is designed and constructed from the point of view of material data management, which realizes the transformation from fragmented data acquisition to massive data integration, processing and fusion. The whole system not only realizes the design of user-defined database, but also realizes the management and application of the whole life cycle data of small reactor structural materials, meeting the user's needs for data query, data retrieval, visual analysis, etc., which is conducive to promoting the standardized and intelligent development of small reactor structural materials data management. At the same time, the data management system breaks through the technical bottleneck of multi-scale material data management, enhances the security and reliability of material data, and provides important support for the digital R&D and design of small reactors.

Effect of Thermal Aging Temperature on Precipitation Behavior of Laves Phase and Impact Performance in High Si Content Ferritic Martensitic Steels

Zhou Jun, Qiu Shaoyu, Qiu Risheng, Zeng Wen, Wang Hao, Shu Ming, Yang Canxiang

2022, 43(5): 126-132. doi: 10.13832/j.jnpe.2022.05.0126

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Ferritic martensitic steel (F/M steel) is one of the main candidate materials for lead-cooled fast reactor core. Increasing Si content can improve its corrosion resistance, but at the same time, it can promote the precipitation of Laves phase, thus affecting the toughness and plasticity of the material. For a F/M steel with a Si content of 0.98%, thermal aging experiments for 5000 h at three temperatures (500, 550 and 600℃) are carried out to study the effect of temperature on the precipitation behavior of Laves phase and impact performance. The results show that heating can promote the nucleation and coarsening of Laves phase, and with the temperature increasing from 550℃ to 600℃, the coarsening rate of Laves phase increases from 3.7 nm/h^1/3 to 9.0 nm/h^1/3. On the other hand, the increase of thermal aging temperature will accelerate the degradation of impact performance. After thermal aging at 550℃ and 600℃ for 500 h, the impact work (A_KV) decreases to 51% and 39% of that before thermal aging, respectively, while the A_KV value remains 75% of that before thermal aging for 2500 h at 500℃. The precipitation of Laves phase has a strong corresponding relationship with the degradation of impact performance, which is the main reason for the degradation of impact performance. egradation of impact performance.

Research on the Thermal-Mechanical Performance of Hollow Hexagonal Fuel Element

Liu Shichao, Li Quan, Huang Yongzhong, Pang Hua, Li Yuanming, Chai Xiaoming, Qiu Xi, Zhao Yanli, Liao Nan, Ran Renjie

2022, 43(5): 133-137. doi: 10.13832/j.jnpe.2022.05.0133

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Hollow hexagonal fuel elements are widely used in high temperature gas-cooled reactor. In order to study the in-reactor performance of the hollow hexagonal fuel element and evaluate its failure probability, the thermo-mechanical behavior analysis of the hollow hexagonal fuel element for high temperature gas-cooled reactor is carried out. The thermal-mechanical behavior of the hollow hexagonal fuel element is calculated by using the multi-physical field coupling method, and the temperature field, deformation, stress distribution and failure probability of the hollow hexagonal fuel element under the condition of low neutron fluence are analyzed. The results show that the maximum operating temperature of the hollow hexagonal fuel element is about 1020 K, the maximum stress of SiC matrix is about 107.32 MPa, and the failure probability is 3.52×10⁻⁴. The low failure probability of SiC matrix ensures the structural integrity of fuel elements. Under the lower neutron fluence, the operating temperature and stress of the hollow hexagonal fuel element are lower, the structural integrity can be ensured, and the fuel element has a good operating state in the reactor.

Calculation of Stress Intensity Factor of External Surface Crack on the Nozzle of Steam Generator in Nuclear Power Plant

Zhang Ruikai, Liu Pan, Tan Jianping, Li Yue, Wang Dasheng, Tu Shandong

2022, 43(5): 138-146. doi: 10.13832/j.jnpe.2022.05.0138

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Due to the particularity of its structure, the nozzle of the nuclear steam generator (SG) is prone to defects during manufacturing and operating process. In order to evaluate the safety of the defect, the stress intensity factor solution available in engineering is required. In this paper, taking the external surface crack of nuclear SG nozzle as the research object, the equivalent stress intensity factors of cracks in different directions and sizes under internal pressure, bending moment and temperature loads are calculated by finite element method and RSE-M code, and the distribution law of equivalent stress intensity factors at the crack front under different loads is analyzed. By comparing the calculation results with the straight pipe stress intensity factor solution of RSE-M code, it is found that the straight pipe stress intensity factor calculation method of RSE-M code can be conservatively applied to the cracks at SG nozzle, and the conservation increases with increasing crack depth. In order to realize the accurate evaluation of the defect safety of SG nozzle, a calculation method of stress intensity factor applicable to the cracks on the external surface of SG nozzle is given based on finite element calculation and RSE-M influence coefficient method, which can provide guidance for the design and maintenance of SG nozzle.

Numerical Simulation of Melting Behavior of 2×2 Rod Bundle Structure Based on MPS Algorithm

He Mengxuan, Fu Shengwei, Yu Hang

2022, 43(5): 147-153. doi: 10.13832/j.jnpe.2022.05.0147

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Taking the 2 × 2 rod bundle structure of typical pressurized water reactor fuel assembly as the research object, the three-dimensional model of rod bundle with and without positioning grid is established. Based on the semi-implicit moving particle algorithm (MPS), the melting behavior of the rod bundle structure under the background of serious accidents is numerically simulated, and the influence of the positioning grid on the flow channel blocking process in the rod bundle melting process is analyzed. The results show that the MPS algorithm can better simulate the melting behavior of the rod bundle structure, and the positioning grid will speed up the melting process of the core and the blocking speed of the cooling channel. The research results in this paper are beneficial to the optimization and improvement of the core melting model under severe accident.

Reliability Evaluation Method of Bolts of Reactor Internals in High-neutron-fluence-rate Region Based on XGBoost

Wang Wenhui, Wan Anping, Deng Chaojun, Gong Zhipeng, Zhang Hongliang, Ye Yanghan, Wang Pengfei, Liu Canxian, Li Yuezhang

2022, 43(5): 154-162. doi: 10.13832/j.jnpe.2022.05.0154

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The bolts of reactor internals are in high temperature, high pressure and high radiation environment for a long time, and the bolts connecting the shroud and the forming plate are subject to irradiation assisted stress corrosion cracking (IASCC). In order to predict the service life of bolts in stress corrosion environment in advance and reduce the spare parts inventory of nuclear power plants, this paper uses the XGBoost(eXtreme Gradient Boosting) model to predict the service life of bolts of reactor internals in high irradiation environment. First, the whole-cycle residual life evolution data in the high-neutron-fluence-rate region of PWR are analyzed and processed, and the correlation model is obtained. Then, XGBoost model based on data driven is proposed to predict the residual life of bolts. This method has strong generalization and high accuracy, and can well evaluate the reliability of bolts in high-neutron-fluence-rate region; Finally, 35000 samples are used as the training set and 15000 samples as the test set, which are compared with the calculated values of the International Atomic Energy Agency (IAEA) empirical formula. The results show that the prediction accuracy of the XGBoost model is as high as 99.93%, which is better than the multiple linear regression method, AdaBoost (using a linear loss function), AdaBoost (using a squared loss function) and AdaBoost (using an exponential loss function) methods.

Study on Influence of Containment Venting on Hydrogen Risk of Spent Fuel Building under Severe Accident

Yang Zhiyi, Shi Xueyao, Zhang Jiajia, Ding Chao, Chong Yimin

2022, 43(5): 163-167. doi: 10.13832/j.jnpe.2022.05.0163

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Taking a domestic Ⅲ generation pressurized water reactor nuclear power plant as an example, two typical severe accident conditions are selected, modeling and calculation with Modular Accident Analysis Program(MAAP) are carried out, and the process of containment venting and the hydrogen risk to the spent fuel building are analyzed. The results show that if the ventilation system of the spent fuel building is not considered, the mixed gas released from the containment will cause a certain hydrogen risk to the spent fuel building due to the condensation of water vapor. If the role of the ventilation system of the spent fuel building is considered, the hydrogen risk in the spent fuel building will be eliminated.

SPSA-based Performance Optimization Method for Steam Generator MPC System

Geng Pengcheng, Shi Changqing, Kong Xiangsong, Liu Hang, Liu Jiabin, Jiang Shaobo

2022, 43(5): 168-175. doi: 10.13832/j.jnpe.2022.05.0168

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Abstract:
The level change process of steam generator (SG) in nuclear power plant has strong nonlinearity and the phenomenon of "false water level" exists. Traditional SG level control systems mostly use fixed-parameter proportional-integral-derivative (PID) controllers, but traditional PID control methods do not have the capabilities of self-optimization, self-adaptation, and self-learning, making it difficult to achieve and maintain optimal control system performance. In order to improve the transient response ability of the unit and the stability, safety and economy of the nuclear power plant, a model predictive control (MPC) strategy based on simultaneous perturbation stochastic approximation (SPSA) algorithm is proposed. In this method, MPC system is used to replace the traditional PID control system, and SPSA is used to optimize and set the parameters of the level control system, so as to optimize the performance of the SG level control system. The simulation results show that this method can effectively improve the setting efficiency of SG level control parameters and the stability of the control system.

Research on Calculation Method of Containment Spray Coverage Rate Based on Monte Carlo Simulation Method

Huang Jieqing, Wu Xiaotian, Wen Liang, Peng Yue

2022, 43(5): 176-180. doi: 10.13832/j.jnpe.2022.05.0176

Abstract(109) HTML (16) PDF(26)

Abstract:
In order to establish a reliable and fast calculation method of containment spray coverage rate, and provide a new auxiliary means for the design and research of the containment spray system, the calculation model of containment spray coverage rate based on Monte Carlo simulation is established in this study by using the method of theoretical analysis. Compared with the calculation results of containment spray coverage rate based on computer aided design (CAD), the applicability of Monte Carlo spray coverage rate calculation method is verified. The results show that the calculation errors of the two methods are within 1%.Therefore, the calculation method of containment spray coverage rate based on Monte Carlo simulation method is reliable and widely applicable. Compared with the spray coverage rate calculation method based on computer aided design (CAD) software, the new method has faster calculation speed and lower human error rate, which is conducive to sensitivity analysis and can greatly improve the design capacity of the containment spray system.

Research for Spent Fuel Pool Self-safe Cooling Technology after Serious Accident

Zhang Ming, Ge Yunzheng, Zhang Shoujie, Liu Weimin

2022, 43(5): 181-187. doi: 10.13832/j.jnpe.2022.05.0181

Abstract(111) HTML (21) PDF(27)

Abstract:
In view of the loss of internal and external power supplies after a serious accident in a nuclear power plant, a scheme for long-term self-safe cooling of the spent fuel pool is proposed by extracting the waste heat of the spent fuel pool for power generation. Based on the thermal process analysis, working medium selection, thermal analysis of key equipment, and system design research based on the waste heat of the spent fuel pool, the feasibility of using the waste heat of the spent fuel pool to achieve long-term self-safe cooling of the spent fuel pool after a serious accident is discussed. The research shows that according to the working environment and the system output power after the serious accident of the nuclear power plant, the waste heat self-generating system of the spent fuel pool can be constructed by the Shangyuan cycle or the Guohai cycle. For both in-service reactor types and new reactor types, the system can ensure the continuous removal of waste heat from the spent fuel pool, meeting the requirement that the temperature of the spent fuel pool is lower than 80°C, thereby realizing self-safe cooling of the spent fuel pool.

PSA-Based Uncertainty Analysis of Pressurized Water Reactor LBLOCA

Deng Jian, Xiong Qingwen, Gou Junli, Liu Yu, Bao Hui, Shen Danhong, Zhou Jiayue

2022, 43(5): 188-194. doi: 10.13832/j.jnpe.2022.05.0188

Abstract(113) HTML (38) PDF(32)

Abstract:
In order to combine deterministic and probabilistic analysis to carry out more realistic safety analysis of nuclear reactor accident conditions, a method combining probabilistic safety analysis (PSA) and best estimate plus uncertainty (BEPU) analysis is proposed, the extreme accident of the double-ended fracture large-break loss of water accident (LBLOCA) in a typical three-loop pressurized water reactor cold pipe section is taken as the object. First, the accident failure analysis of the emergency core cooling system is carried out based on PSA. Then, combined with BEPU analysis, the cladding peak temperature (PCT) distribution and conditional core damage probability (CCDP) of each accident sequence in the event tree are evaluated, and the core damage frequency of the PWR in this accident condition is finally determined. The analysis results show that the emergency core cooling system of PWR can ensure the safety of the reactor under the condition of double-ended fracture of cold pipe section, and a row of low-pressure safety injection systems are sufficient to remove the residual heat of the core and ensure the safety of the reactor.

Simulation Study on Fluid-elastic Instability of Two-Dimensional Tube Bundle Based on UDF

Tan Wei, Zhao Chengzhuo, Huang Xuan, Shen Pingchuan, Zhang Ke, Zhu Guorui

2022, 43(5): 195-202. doi: 10.13832/j.jnpe.2022.05.0195

Abstract(79) HTML (49) PDF(23)

Abstract:
In view of the low calculation accuracy and huge calculation cost of the existing fluid-solid coupling model in the simulation study of fluid-elastic instability, a two-dimensional unidirectional fluid-solid coupling model which can predict the critical velocity of tube bundle is established. This model is based on the commercial ANSYS Fluent software, the flow field is calculated by the SST k-ω turbulence model, and then the fluid force on the tube is extracted by a self-compiled user-defined function (UDF), and the fourth-order Runge-Kutta method is used to solve the structural dynamic equation and realize the unidirectional fluid-solid coupling calculation. Using this model, the fluid-solid coupling calculation is carried out on the triangularly-arranged tube bundle with a pitch-diameter ratio of 1.5, and the critical velocity, the amplitude time-history curve and the amplitude spectrum of the central tube are obtained, which are verified by water tunnel experiments.The results show that this model can accurately predict the critical velocity with low calculation cost, and also obtain the true vibration characteristics of the tube. The amplitude time-history curve and amplitude spectrum of the central tube calculated by simulation are similar to those of the experiment.In addition, the resistance and lift coefficient data obtained by simulation calculation show that the resistance and lift coefficient time-history curves change from disorder to regularity with the increase of the velocity. When the converted velocity reaches 2.44, the main frequency of the resistance and lift coefficient includes the component of the natural frequency of the tube in still water.

Analysis of Liquid Film Characteristics of Deep-groove Reactor Coolant Pump Mechanical Seal

Jin Le, Wang Yan, Cui Huaiming, Zhu Xiangdong, Zhang Chaonan, Mao Yuanfan

2022, 43(5): 203-210. doi: 10.13832/j.jnpe.2022.05.0203

Abstract(167) HTML (43) PDF(25)

Abstract:
By combining the three-dimensional finite element analysis method and the sealing liquid film flow field analysis based on the classical friction theory, six sealing surface schemes of the mechanical seal of a new type of nuclear reactor coolant pump (referred to as the nuclear main pump) are analyzed and studied. The key parameters such as liquid film thickness, contact load, nominal wear rate, and low pressure leakage rate of each scheme are compared. The calculation results show that the design scheme with a groove width of 6 mm is a set of designs with relatively balanced performance, and the performance output characteristics of the sealing surface are similar to that of an imported mature mechanical seal and slightly better than that of the imported model; The linear groove scheme with low-pressure compensation can greatly prolong the service life of the sealing surface, but also bring a higher low-pressure leakage rate.

Design and Experimental Research of Reactor Irradiated Sample Delivery System

Hu Ruirong, Zhu Shifeng, Wang Naxiu, Zhang Lina, Xu Bo, Cao Yun, Wang Xiaoyan

2022, 43(5): 211-216. doi: 10.13832/j.jnpe.2022.05.0211

Abstract(96) HTML (17) PDF(27)

Abstract:
In order to realize the rapid delivery of reactor irradiated samples under the conditions of small space inside the reactor and long distance outside the reactor, an irradiated sample delivery system combining steel wire rope drum and pneumatic delivery is designed. In this paper, the structural design of the key components in the irradiated sample delivery system is described, and the theoretical calculation of the pneumatic part is carried out. An experimental study of the delivery system is carried out to further verify the design scheme. The research results show that the delivery system has certain feasibility in the conditions of small space inside the reactor and long distance outside the reactor.

Requalification Experimental Study on Life Extension of Electrical Instrument Equipment in Nuclear Power Plant

Chen Qing, Guo Xing, Gao Xuan, Qiu Xinyuan, Wang Guangjin, Duan Xuxing, Zhao Chuanli, Jiang Shenghan

2022, 43(5): 217-222. doi: 10.13832/j.jnpe.2022.05.0217

Abstract(127) HTML (24) PDF(23)

Abstract:
On the basis of the research results of environmental qualification of electrical instrument equipment in nuclear power plant, the life extension requalification analysis and experimental research of electrical instrument equipment in nuclear power plant are carried out. Taking the DDG-1 electrical penetration assembly (EPA) of Qinshan Nuclear Power Co., Ltd. as the research object, the principles for requalification experimental study are formulated according to the actual operation. Based on these principles, the test scheme, test item sequence and EPA repair basis and scheme are determined by combining the analysis method, and requalification experimental study is carried out on this basis. The properly repaired DDG-1 EPA passed the test of equipment performance with time, seismic test, thermodynamic test under design basis accident (DBA) conditions and ultimate electrical performance test after DBA in sequence according to the test scheme. Its state is intact after the test, which indicates that the DDG-1 EPA can achieve the expected goal of continuing to extend its life for 20 years after proper repair, and can provide guidance and reference for the requalification experimental study of other electrical instruments in nuclear power plants.

Study on Seismic Design Method of Class II Research Reactor Based on TMSR-LF1

Liu Yicheng, Wang Xiaoyan, Wang Xiao, Zhang Xiaochun, Gong Wei, Dai Rencong

2022, 43(5): 223-228. doi: 10.13832/j.jnpe.2022.05.0223

Abstract(117) HTML (18) PDF(24)

Abstract:
Seismic design is an important part of nuclear facilities design to meet the comprehensive requirements of safety and economy. At present, the seismic design of research reactors lacks corresponding specifications and studies, and a relatively complete method system has not yet been found. In this paper, a seismic design method for class II research reactor matching structure and equipment is recommended. Taking the earthquake motion with a 50 a exceedance probability of 2% as the safe shutdown earthquake (SSE), and taking the 2 MW liquid fuel thorium based molten salt experimental reactor (TMSR-LF1) as an example, the design response spectra obtained by using this method and various specification methods are compared and analyzed, and they are applied to the seismic design and calculation of structures and equipment. The results show that the recommended method has better economy than the nuclear power code and better conservatism than the civil code on the premise of meeting the seismic design matching of structures and equipment.

Numerical Simulation Research on Reforming Fluidization of Radioactive Waste in Nuclear Power Plant

Zeng Shenfu, Liu Xiajie, Lin Peng, Zheng Wei, Qiao Baoquan

2022, 43(5): 229-232. doi: 10.13832/j.jnpe.2022.05.0229

Abstract(121) HTML (29) PDF(26)

Abstract:
During the high-temperature reforming process, the radioactive organic nuclear waste discharged from the nuclear power plant will exist in a fluidized state. In order to obtain the detailed parameters of this process and design better operating conditions, a new numerical simulation method for reforming fluidization of radioactive waste in nuclear power plant is proposed. The numerical calculation model is designed, the governing equations of gas-solid hydrodynamics are established, and the turbulence model is established; The geometric model of radioactive organic chemical waste in nuclear power plant is established, the object geometric model is established, the grid structure is divided, and the initial boundary conditions are set. The effects of different height-diameter ratios on the radial velocity, radial solid holdup and radial gas holdup of particles are studied by numerical calculation. The calculation results show that when the height-diameter ratio is 1.0, the fluidization effect of gas-solid particles in the fluidized bed is the best.

Research on the Application of Laser Additive Manufacturing in Reactor Internals

Wang Qingtian, Li Hao, Hu Xuefei, Wang Zhonghui, Zhu Mingdong, Zhao Wei

2022, 43(5): 233-237. doi: 10.13832/j.jnpe.2022.05.0233

Abstract(151) HTML (23) PDF(36)

Abstract:
In view of the quality problems of manual TIG overlaying of cobalt-based alloy, the additive manufacturing process of cobalt-based alloy is studied, especially the optimization of laser additive manufacturing process parameters. Then a series of experiments, including hardness, wear resistance, corrosion resistance etc., are carried out on the cobalt-based alloy additive manufacturing layer and compared with the manual TIG surfacing layer. The results show that the microstructure of the laser additive manufacturing layer is more refined, the hardness is more uniform, the wear resistance and corrosion resistance are better than that of the manual TIG surfacing layer.

CORCA-based Software Implementation of Core Solution and Conjugate Calculation with Fixed Neutron Source

Zhou Nan, Yu Yingrui, Zhao Wenbo, Liao Hongkuan, Lu Di, Chen Feifei, Liu Jiayi

2022, 43(5): 238-244. doi: 10.13832/j.jnpe.2022.05.0238

Abstract(155) HTML (51) PDF(23)

Abstract:
In the deep subcritical state, the traditional source multiplication method has the characteristic of low accuracy in the measurement of nuclear reactor reactivity. In order to improve the measurement accuracy, this paper expands CORCA software, develops CORCA-FIX software with fixed source problem solving function and neutron value solving function with discontinuous factor, and uses the comparison program and real reactor data to calculate and verify CORCA-FIX software. The verification results confirm that CORCA-FIX has high accuracy in solving the deep subcritical state of the core with fixed source. After the output results are applied to the real reactor data, better subcritical degree measurement results are obtained, and the deviation criteria of reactivity measurement in engineering applications are met.

Optimization Design for Reactor Control System of the First Overseas HPR1000 Reactor

Zhang Ying

2022, 43(5): 245-249. doi: 10.13832/j.jnpe.2022.05.0245

Abstract(151) HTML (30) PDF(60)

Abstract:
Reactor control system is an important instrument and control system and plays a significant role in the normal operation of the nuclear power plant (NPP). In order to guarantee the good control performance of the control system during the normal operation of NPP and to decrease the on-site commissioning time, it is necessary to optimize the control parameters through simulation study on the control system in the design phase. This paper analyzes the reactor control system functions of the first overseas third-generation HPR1000 reactor, and describes the controlled variables of each control system. On this basis, the optimization of control system parameters is described. Using the NPP digital simulation tool, the sensitivity analysis of the control system parameters is carried out through system modeling and simulation, and the optimal control system parameters are obtained according to the static and dynamic response characteristics of the system under different parameter values, and such parameters meet the design requirements through performance verification. The obtained reactor control system parameters have been used in the design of the reactor control system of first overseas HPR100 reactor, and used to guide the on-site commissioning and operation of NPP.

2022 Vol. 43, No. 5