核动力工程

Research on Fast Prediction Method of Neutron Flux Based on Hybrid Driven Reduced Order Model

Zhao Ziyan, Xiang Zhaocai, Zhao Pengcheng

2024, 45(4): 1-8. doi: 10.13832/j.jnpe.2024.04.0001

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Abstract:
The accurate prediction of neutron flux and reactor power is very important for the safe operation of the reactor immediately after the disturbance of reactor parameters. The traditional method combining POD and Galerkin projection has the problem of low accuracy due to cumulative error. In this study, the implicit difference method is used to obtain the exact solution of one-dimensional neutron spatiotemporal diffusion. As the reference data, two LSTM neural network terms are introduced to eliminate the cumulative error and truncation error of POD, and to build a hybrid drive model driven by physics and data. The results show that the root-mean-square error of neutron flux, total power and each order modal coefficient is reduced by 1-2 orders of magnitude after adding the neural network correction term, and the calculation time is significantly reduced under the same order of prediction when the neural network extension term is added. The improved model based on 2nd and 3rd order scaling to 6th order is 13% and 7.6% faster than the original 6th order model, respectively. The hybrid drive model can improve the rapid prediction accuracy of POD, and the results have certain reference value.

Research on Industrial Validation of VVER-1000 Based on PWR Core Physics Analysis Code Bamboo-C

Yang Haozhe, He Xudong, Wang Kunpeng, Wan Chenghui, Wu Hongchun

2024, 45(4): 9-16. doi: 10.13832/j.jnpe.2024.04.0009

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Abstract:
This study aims to achieve precise physical analysis of VVER-1000 nuclear reactors. Based on Bamboo-C, an advanced PWR core physical analysis software developed by Xi'an Jiaotong University, a thorough methodological study is carried out. The research encompasses: in the aspect of assembly calculation, methods based on Constructive Solid Geometry (CSG) for hexagonal transport calculations and detailed modeling techniques for heavy reflector layers were studied; in core calculation, a method combining conformal mapping and nonlinear iterative strategies for hexagonal fuel assembly neutron diffusion was investigated. Using Bamboo-C, modeling calculations were performed for three consecutive fuel cycles of start-up physical experiments and power operations of a specific VVER-1000 unit, followed by a comparative analysis with actual measured data. The results indicate: ①In the start-up physical tests, the average error of critical boron concentration is −5.0ppm (1ppm=10^–6); the average errors of the moderator temperature coefficient and the isothermal temperature coefficient are 0.3 pcm/K and 0.9 pcm/K(1pcm=10^–5), respectively; the average error of the boron worth is −5.0%; and the average error in control rod worth is −7.8%. ②During power operations, the average errors of critical boron concentration for three cycles were −2.3ppm, −18.9ppm, and −7.8ppm, respectively; the average errors in the core power distribution for the three cycles were −0.010 (for assembly relative power greater than 1) and 0.012 (for assembly power less than 1). Therefore, Bamboo-C software meets the industrial threshold requirements for calculation errors of key physical quantities in the VVER-1000 reactor core, demonstrating its capability for engineering application.

Verification of PCM Nuclear Design Code for Whole Core Calculations

Lu Gaoqi, Ding Ming, Lan Bing, Pan Xinyi, Li Chun, Wang Chao, Ma Yunfan

2024, 45(4): 17-23. doi: 10.13832/j.jnpe.2024.04.0017

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Abstract:
PCM software package, independently developed by China Nuclear Power Technology Research Institute Co. Ltd., is a nuclear design code consisting of assembly section calculation code PINE and 3D core design code COCO. To validate the whole core calculation capability and accuracy of the PCM nuclear design code, this study conducted verification using internationally recognized benchmark problems such as BIBLIS, IAEA, and LRA, as well as a self-made whole-core problem for the COCO code and the “assembly-core” two-step method within the PCM nuclear design code. The verification results indicated that in the calculations of the light-water reactor core benchmark problems, the average error of the effective multiplication factor (k_eff) for all cases was 6.4pcm, with a maximum error of only 28.2pcm. In all the cases, the deviation of the normalized power distribution of the core fuel assemblies did not exceed 1%. The verification results show that the whole core calculation function of the PCM nuclear design code exhibits high computational accuracy, and the overall calculation accuracy can meet the engineering requirements.

Numerical Study on DNB-Type Critical Heat Flux in Circular Tube under Rolling Condition

Fang Zheng, Du Song, Bu Shanshan, Li Zhenzhong, Chen Deqi

2024, 45(4): 24-31. doi: 10.13832/j.jnpe.2024.04.0024

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Abstract:
A three-dimensional numerical calculation was carried out on the departure from nucleate boiling (DNB) type critical heat flux (CHF) in a vertical circular tube under different rolling conditions. The Euler two-phase flow model and the non-equilibrium wall boiling model were used. By comparing the simulated CHF values of static tubes with experimental values, a sensitivity analysis of different wall boiling sub-models was completed. The CHF of a vertical tube with sinusoidal simple harmonic rolling motion is predicted for 15 combinations of amplitude and period. The results show that all rolling conditions lead to the early occurrence of DNB. In the most "violent" rolling situation, the value of CHF is the smallest. The temperature and heat transfer coefficient in the tube will change periodically with the rolling motion. Within a period, larger amplitude and smaller period will cause the heating wall to have a smaller heat transfer coefficient at a certain moment, resulting in an increase in the maximum temperature of the wall. This study can provide a reference for the numerical prediction of DNB-type CHF under rolling conditions.

Research on Characteristics of Secondary Side Passive Residual Heat Removal System of Lead-bismuth Reactor under SBO

Qian Yalan, Lin Qian, Yang Zijiang, Chen Kang, Zhan Wenhui, Tang Chuntao, Yang Bo

2024, 45(4): 32-37. doi: 10.13832/j.jnpe.2024.04.0032

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Abstract:
The secondary side passive residual heat removal (PRHRS) of Russian SVBR-100 lead-bismuth reactor was chosen as the research object, and the RELAP5/MOD4.0 code was used to model and evaluate the heat removal capacity and parameter sensitivity of PRHRS under station blackout (SBO) accident. The results show that the key parameter of the peak cladding temperature is 816.35 K during the whole SBO accident, which is within the safety limit. PRHRS can timely remove the residual heat. By increasing the heat exchange area of the built-in condensing heat exchanger in PRHRS water tank, the residual heat removal capacity of PRHRS can be enhanced. The safety analysis model and evaluation method for secondary side PRHRS established in this study can provide technical reference for the design and application of PRHRS of lead-cooled reactor in China.

Study on Uncertainty of Two-Phase Flow Parameter Detection Based on Monte Carlo Method

Liu Li, Zhu Longxiang, Zhang Luteng, Ma Zaiyong, Sun Wan, Pan Liangming, Deng Jian

2024, 45(4): 38-44. doi: 10.13832/j.jnpe.2024.04.0038

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Abstract:
Bubble velocity and bubble number are key parameters for calculating the phase characteristics such as interfacial area concentration, so it is necessary to study the uncertainty of bubble velocity and number measured by the conductivity probe. The Monte Carlo method is adopted to generate a large number of random bubble motion samples, and the statistical law of1~6 mm bubbles captured by the probe is obtained. By introducing the relative velocity fluctuation component H, the influence of bubble transverse velocity on the effective bubble number and bubble velocity is studied. The results show that the presence of bubble transverse velocity alleviates the inability to measure smaller bubbles due to the probe transverse spacing. However, the number of effective bubbles significantly decreases and the number of escaping bubbles increases with the increase of bubble transverse velocity. Meanwhile, the velocity error only comes from the probe transverse spacing when H=0, and the velocity error decreases with the increase of the bubble diameter. When H≠0, for bubbles with a diameter greater than 3 mm, the transverse movement of the bubbles to the left or right makes the probe pass through the bubbles far from the central axis, which leads to the increase of the actual moving distance of the bubbles and the increase of the velocity error. This study can provide a reference for determining and correcting the uncertainty of two-phase flow parameters such as interfacial area concentration.

Study on CHF Relational Expression Development Based on High-precision Subchannel Code

Wu Change, Zhang Yuxiang, Chen Changyi, Jiang Li, Shan Jianqiang, Fu Xiangang

2024, 45(4): 45-52. doi: 10.13832/j.jnpe.2024.04.0045

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Abstract:
Three groups of CHF test data of non-uniform heating typical grid and guide tube grid were adopted, and the local parameters were obtained by using high precision subchannel code ATHAS. The development of CHF relation suitable for the analysis of fuel assembly deviation from nucleate boiling ratio (DNBR) was completed, and the cold wall effect factor and DNBR limit of guide tube were obtained. Compared with the results of the correlation developed by FLICA, the DNBR limit calculated by the relationship developed by ATHAS is lower, and the prediction rate of axial position of burn-out (BO) point and CHF is higher.

Numerical Simulation on Flow Heat Transfer Characteristics of Helium-Xenon Mixture in Tight Lattice Rod Bundle Channel

Zhang Jiaxin, Bao Hui, Cong Tenglong, Gu Hanyang

2024, 45(4): 53-60. doi: 10.13832/j.jnpe.2024.04.0053

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In response to the design and analysis requirements for the core of the helium-xenon cooled high temperature gas cooled reactor, this study has established a comprehensive three-dimensional heat transfer model for helium-xenon mixture. This model encompasses property model, turbulent model, and turbulent Pr model. Utilizing this model as a foundation, numerical analysis of the flow and heat transfer characteristics of helium-xenon mixture in the fuel rod bundle channel has been conducted. This research investigates the influence of geometric parameters and operational parameters on relevant characteristics. The results reveal that the presence of cladding will bring significant circumferential non-uniformity to the flow and heat transfer in the tight lattice rod bundle channel, necessitating consideration of cladding in both subchannel and three-dimensional numerical simulation. In addition to cladding, heat transfer in the tight lattice rod bundle channel is primarily influenced by the rod diameter ratio. Under identical simulation conditions, a larger rod diameter ratio leads to enhanced convective heat transfer of the mixed gas.

Experimental Study on Boiling Two-Phase Flow Instability in a Single Helically Coiled Tube

Zheng Pengde, Tang Qifen, Li Zhenzhong, Wang Ningyuan, Chen Deqi

2024, 45(4): 61-68. doi: 10.13832/j.jnpe.2024.04.0061

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Abstract:
The boiling phase change happened in the heating channel will induce flow instability. Thus, studying the two-phase flow instability in helically coiled tubes is of great significance in the design and operation of helical-coil one-through steam generators. In this paper, the boiling two-phase flow experiment in a single helically coiled tube is carried out on the thermal experimental platform, and the flow instability phenomenon when boiling two-phase flow occurs in the helically coiled tube is studied. In the experiment, the boiling two-phase flow instability is classified by analyzing the variation and spectral characteristics of parameters such as flow rate and pressure drop in the helically coiled tube at different times during the rise of heating power. The results show that when the experimental parameters are in the range of 0.1~3 MPa for pressure, 300~1200 kg/h for flow rate, 20~100℃ for inlet subcooling and 0~200 kW for heating power in the experimental section, the helically coiled tube heating channel exhibits flow drift instability as the power increases. Once the flow drifts to another flow value, low-amplitude high-frequency density wave oscillations occur under low steam quality conditions, while high-amplitude low-frequency density wave oscillations occur under high steam quality conditions. In addition, the increase of inlet subcooling, inlet flow rate and system pressure will improve the stability of the system.

Research on Prediction and Sensitivity Analysis of Minimum Film Boiling Temperature of Quenching Boiling Based on Machine Learning

Zhang Junquan, Deng Jian, Luo Yan, Lu Tao

2024, 45(4): 69-76. doi: 10.13832/j.jnpe.2024.04.0069

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Abstract:
Quenching boiling is widely used in the cooling process of fuel rods after the loss of coolant accident in nuclear reactor. The determination of the minimum film boiling temperature (T_min) is very important for the safe operation of nuclear reactors. Based on the experimental data in the literature, this paper selects three typical machine learning models: Random Forest (RF), Artificial Neural Network (ANN) and eXtreme Gradient Boosting (XGBoost) to predict T_min during quenching boiling and conduct a sensitivity analysis of influencing factors. The results show that the machine learning method can effectively improve the accuracy of T_min prediction compared to the traditional empirical correlation. Among the models, the RF model exhibits the best predictive performance with a coefficient of determination R² of 0.9770. By combining the RF model with the Sobol’ global sensitivity method, the study identifies the coolant subcooling as the most influential parameter on T_min, followed by initial wall temperature, while length-diameter ratio, pressure and thermophysical properties have a smaller impact. The findings of this research will provide theoretical guidance for improving the safety of nuclear reactors.

Study on Jet Mixing Characteristics of Lead-Bismuth Eutectic Cooled Reactor Assembly Head Based on CFD Method

Zhang Ji, Wang Yingjie, Wang Mingjun, Tian Wenxi, Qiu Suizheng, Su Guanghui

2024, 45(4): 77-86. doi: 10.13832/j.jnpe.2024.04.0077

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Abstract:
In the upper chamber of the Lead-Bismuth Eutectic (LBE) cooled reactor, fluid temperature fluctuations during the LBE mixing process from different power assemblies may lead to thermal fatigue of the solid structure, threatening the safety of LBE reactor operation. Based on the Computational Fluid Dynamics (CFD) method, this paper established a computational model suitable for the jet simulation of the LBE reactor assembly head and validated it through experimental data. Then, simulations of jet flow conditions with various inlet parameters were carried out. The research results show that the increase in the inlet temperature difference will lead to a significant increase in the temperature distribution inhomogeneity on the axial section downstream of the assembly head, and the influence range continues to the position of the measuring column. Within the calculation range, when the temperature difference decreases by 20 K, the root mean square temperature of each downstream section decreases by approximately 23.5%. The increase in inlet velocity causes the secondary flow to increase, but the intensity of the secondary flow will decrease. Within the calculation range, the degree of mixing first decreases and then increases as the inlet velocity increases. This paper can provide reference for research on flow field analysis downstream of the assembly head, structural optimization design of the assembly head, and LBE reactor core flow distribution design.

Numerical Study on Characteristics of Subcooled Flow Boiling with the Coupling Effect of 3×3 Petal-Shaped Fuel Rods and Coolant

Du Lipeng, Song Shangdian, Cai Weihua, Jiang Zeping, Cheng Qi, Zhang Wenchao, Jin Guangyuan

2024, 45(4): 87-95. doi: 10.13832/j.jnpe.2024.04.0087

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Abstract:
In order to promote the engineering application of petal-shaped fuel rods in water-cooled reactors, it is necessary to understand the subcooled flow boiling characteristics of coolant in the sub-channels of petal-shaped fuel rod bundles. The Euler model and wall boiling model were applied to numerically simulate the subcooled flow boiling under the coupling effect of 3×3 petal-shaped fuel rods and coolant. Using the simulation results, the distribution of parameters such as void fraction, wall temperature and transverse flow velocity in different sub-channels, as well as the effects of uniform heating mode and axial cosine heating mode on flow and heat transfer were explored. The research results indicate that subcooled boiling occurs first on corner fuel rods, and with the increase of heating power, the position unevenness of onset of nucleate boilding (ONB) of subcooled boiling of the corner, edge and center fuel rods decreases. Under the same heating conditions, the wall superheat at ONB on the corner fuel rod is the largest, followed by the edge fuel rod and the center fuel rod is the smallest. The surface heat flux at the inner concave arc of the fuel rod is greater than that at the outer convex arc. Under the condition of constant total heating, cosine heating reduces the non-uniformity of wall temperature compared with uniform heating.

Semi-Analytical Solution of Temperature Rise Caused by Irradiation Effect for Perforated Plates with Square Penetration Patterns in Reactor Vessel Internals

Lin Bingchi, Xu Xiao, Kong Xiaofei, Liu Pan, Jin Ting, Nie Zhaoyu, Lu Zhicheng, Yao Bowei

2024, 45(4): 96-102. doi: 10.13832/j.jnpe.2024.04.0096

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In Reactor Vessel Internals (RVI), the temperature rise caused by the irradiation effect is significant for the thick perforated plates. Due to the complexity of the structure, its temperature rise is generally calculated by finite element method. In this paper, the semi-analytical method is used to calculate the maximum temperature rise of the perforated plate with square penetration patterns under irradiation effect. The calculation formula of the temperature rise of the structure is obtained, and the calculation results are compared with the finite element results. The results show that the deviation between the maximum temperature rise calculated by the formula and the finite element results is less than 2% when the radial range of the peak value of γ heat release rate is greater than 4 times the spacing between holes, and the temperature distribution in the thickness direction where the maximum temperature rise is calculated by the formula is in good agreement with the finite element results.

Analysis on Flow Distribution Characteristics of Steam Generator under Natural Circulation Condition

Luan Xingjian, Wang Wen, Song Jiahao, Han Fei, Jiang Erhui, Cheng Kun, Yang Fan

2024, 45(4): 103-110. doi: 10.13832/j.jnpe.2024.04.0103

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Natural circulation is a particular operating condition of nuclear power system, and reversed flow occurs in the inverted U-shaped tube bundle of the steam generator, which affects the heat transfer between the primary and secondary sides and the operation stability. In this research, In this study, the calculation code of primary flow distribution in steam generator of nuclear power unit is developed, and the accuracy of the calculation code is verified by the Finnish PACTEL pressurized water reactor experiment. The effects of loop mass flow, tube height and primary inlet temperature on the flow distribution of inverted U-tube bundle steam generator are discussed. The results show that the lower the tube height and the lower the inlet temperature of the primary side of the inverted U-shaped tube bundle, the smaller the critical pressure drop and critical flow velocity will be. Compared with the change of the height of the inverted U-shaped tube, the change of the inlet temperature of steam generator primary side affects the flow distribution of the inverted U-shaped tube bundle more significantly. The reversed flow could be suppressed by increasing the circulation mass rate, and it will disappear when the mass rate increases to a threshold.

Verification and Uncertainty Evaluation of LOCUST Reflood Model

Xu Rongshuan, Xia Hang, Xu Caihong, He Dongyu, Wang Ting, Li Jinggang

2024, 45(4): 111-117. doi: 10.13832/j.jnpe.2024.04.0111

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Abstract:
The reflood stage is an important stage after a large break loss of coolant accident (LBLOCA) in pressurized water reactor. In order to evaluate the simulation ability of the system code LOCUST, the verification and uncertainty research of LOCUST reflood model are carried out. Based on the experimental results of RBHT test bench, the LOCUST reflood model is verified. At the same time, the uncertainty of the reflood model is analyzed by response surface method, and the response function of the maximum temperature of the heating rod surface at three heights of RBHT test section is obtained by response surface method with the wall-liquid film boiling heat transfer coefficient, wall-vapor film boiling heat transfer coefficient and interface friction coefficient as input parameters. The calculated results are in good agreement with the experimental results, and the maximum temperature deviation is within 40 K. Based on the calculation results of response surface method, the maximum deviation of the maximum temperature of the heating rod surface at three heights is about 20 K with 95% probability and 95% confidence. Meanwhile, the highest temperature values calculated for the three heights are basically consistent with the experimental values when the dimensionless factors of the three input parameters are 1.951, 1.233, and 0.1, respectively.

Experimental Study of the Influences of CRUD layer on Bubble Departure Diameter and Bubble Departure Frequency on Fuel Cladding Surface

Cai Jiejin, Hu Zhiping, Deng Rining

2024, 45(4): 118-126. doi: 10.13832/j.jnpe.2024.04.0118

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Chalk River Unidentified Deposits (CRUD) is naturally formed on the fuel cladding during the routine operation of pressurized water reactor (PWR), and its influence mechanism on the boiling heat transfer behavior of cladding is still unclear. In order to investigate the influence of CRUD on the bubble departure diameter (BDD) and bubble departure frequency (BDF), based on the flow boiling visualization experimental platform under atmospheric pressure, the cladding material Zr-4 alloy is used as the substrate, and SiO₂ deposition layers with different thicknesses are deposited layer by layer to simulate the CRUD. The bubble dynamics analysis of BDD and BDF is carried out through the flow boiling experiment, with their relationship with the wall superheat considered, and compared with the existing prediction model. It is found that compared with the surface without SiO₂ deposit, the BDD and BDF on the surface with SiO₂ deposit are larger, and the increase of wall superheat will cause the BDD to become larger and accelerate the bubble departure. Under the same conditions, the influence of fluid subcooling and Reynolds number on the BDF is greater than that on the BDD. The improved prediction equations of BDD for all conditions in this experiment are put forward. The error between the predicted value of the improved prediction equations and the experimental value is less than 30%.

Study on Adaptability of Heat Transfer Model and Oxidation Relationships Based on COSINE Sub-channel Code

Cheng Yixuan, Meng Zhaocan, Zhang Hao, Zhang Yilin, Zhao Meng, Yang Yanhua

2024, 45(4): 127-133. doi: 10.13832/j.jnpe.2024.04.0127

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In view of the urgent need of heat transfer model and oxidation relationships in pressurized water reactor nuclear subchannel software to improve the core safety and the accuracy of simulation and prediction of domestic software, we used numerical simulation technology to analyze the heat transfer model and oxidation relationships in COSINE subchannel software, and used experimental data to study the influence of different theoretical relationships on boiling heat transfer performance and oxidation amount. The results indicate that the software has the ability to simulate the heat transfer before and after the criticality in the rod bundle, and the simulation results are in good agreement with the experimental values. Before the superheat degree is less than 4 K, the MAX model is suitable for calculating nucleate boiling. When the superheat degree is greater than 4 K, the PLUS model has good applicability. Dougall-Rohsenow model is suitable for calculating film boiling. Baker-Juster model slightly overestimated the oxidation amount before the temperature was lower than 1374 K; When the temperature is higher than 1374 K, the oxidation amount is underestimated.

Preliminary Analysis of the Core Factors for Improving Efficiency of Thermionic Energy Conversion

Ni Wentao, Luo Qi, Zhong Wuye, Lyu Zheng

2024, 45(4): 134-141. doi: 10.13832/j.jnpe.2024.04.0134

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Efficient thermionic energy conversion technology is the key technology for improving the thermoelectric conversion efficiency of thermionic fuel elements and promoting the space thermionic reactor power supply towards higher power and longer lifetime. To explore the key factors that improve the efficiency of thermionic energy conversion, this article starts from the basic principle of thermionic energy conversion and summarizes the methods to improve the thermoelectric conversion efficiency of thermionic fuel elements from three aspects: improvement of the emitter, improvement of the collector, and reduction of arc voltage drop. Analysis shows that the clear direction for significantly improving thermoelectric conversion efficiency is the improvement of the collector, and the key is the development of a new generation of low absorption cesium work function collector materials.

A Reduced-Order Model of Mode Characteristics and Flow-Induced Vibration Response of Fuel Rod Based on POD Method

Min Guangyun, Jiang Naibin

2024, 45(4): 142-149. doi: 10.13832/j.jnpe.2024.04.0142

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Abstract:
A Reduced-Order Model (ROM) for swiftly predicting both the modes and flow-induced vibration response of fuel rods is introduced in this study. Firstly, modes for fuel rods with varying stiffness are obtained using ANSYS-Batch, and these modes are then compiled into a Snapshots matrix. Subsequently, leveraging MATLAB and semi-empirical formulas, the flow-induced vibration responses of fuel rods are batch-calculated, and the resulting data is assembled into another Snapshots matrix. The Proper Orthogonal Decomposition (POD) method is then applied to project the Snapshots matrix into a lower-dimensional space, with the POD modes having the highest energy contribution being selected based on the magnitude of eigenvalues. Finally, the Snapshots matrix is reconstructed back into physical space using the selected POD modes, enabling the rapid calculation of both mode and flow-induced vibration responses. Our study reveals that for reconstructing the first-order mode and flow-induced vibration responses, a smaller stiffness necessitates a greater number of POD modes. Furthermore, in the reconstruction of different-order modes with the same stiffness, higher-order modes require a greater number of POD modes. The findings of this study hold significance for the swift calculation of fuel rod mode characteristics and flow-induced vibration responses.

Application of Green Function Method Considering Thermal Stratification Effect in Rapid Calculation of Fatigue Monitoring System

Chen Rong, Zhang Guihe, Liang Enming

2024, 45(4): 150-154. doi: 10.13832/j.jnpe.2024.04.0150

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This paper analyzes how to realize fast calculation when thermal stratification effect is considered in fatigue monitoring system. A new Green's function considering thermal stratification effect is proposed. At the same time, the thermal stratification stress of elbow model under assumed transient state is calculated by this method and compared with the finite element method, which shows that this method is reasonable and accurate. The results show that the calculation method based on Green's function and considering thermal stratification is efficient, fast and accurate, and can be applied to the stress servere position at the corner of elbow. At present, the method has been applied to the fatigue monitoring system of nuclear power plant equipment.

Study on Load-following Operation Mode of Small Sodium-Cooled Fast Reactor Nuclear Power System

Yin Kai, Gong Lin, Duan Tianying, Hou Bin, Dai Raoqi, Liu Yong, Hu Jiayong

2024, 45(4): 155-161. doi: 10.13832/j.jnpe.2024.04.0155

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Based on the platform of RELAP/GSE combined with MATLAB/Simulink, a simulation model is established for the scheme of coupling Stirling thermoelectric conversion module in small sodium-cooled fast reactors, so as to study the load following operation capability and operation mode of the nuclear power supply system of a small sodium-cooled fast reactor. In the extreme case of load step and no control system intervention, the load-following capability of each loop system of nuclear power supply is tested in a hierarchical manner, and a control scheme suitable for load-following mode is proposed and verified. The simulation results show that the power supply system has the ability to withstand ±10% load step changes and has strong load-following capability to cooperate with grid peak shaving. At the same time, this research proposes two coordinated control schemes for the load-following operation mode of the power supply system, and provides a coordinated control scheme suitable for the load-following operation mode of the small sodium reactor power supply through simulation analysis and comparison.

Research on General Design Method of Emergency Communication System in Nuclear Power Plant

Yu Yun, Guo Shujie, Lu Wenkui, Wang Gaopeng, Liu Jing

2024, 45(4): 162-165. doi: 10.13832/j.jnpe.2024.04.0162

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In order to further improve the emergency communication capability of the communication system of nuclear power plant and support the emergency response under nuclear accident conditions, this study analyzes the current relevant standards and requirements of the communication system design of nuclear power plant, and puts forward the overall design flow of the emergency communication system. The communication function requirements under accident conditions are studied, and the design principles and functional orientation of emergency communication system for nuclear power plants are determined, which have been applied to the communication system design of various reactor nuclear power plants such as Generation Ⅲ nuclear power plants. The emergency communication ability and safety level of nuclear power plant are further enhanced, and the economy of communication system is improved.

Research on Dynamic Modeling and Control Method of Heat Pipe Reactor

Yin Shaoxuan, Yu Ren, Sheng Dongjie, Mao Wei

2024, 45(4): 166-172. doi: 10.13832/j.jnpe.2024.04.0166

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In order to study the nuclear power control method of heat pipe reactor (HPR), the lightweight dynamic model of MegaPower HPR is constructed, and then the designed controller is verified by simulation. Based on lumped parameter method, a heat transfer model from core to heat pipe and then to heat exchanger is studied, and Simulink model is established. Aiming at the control method design of HPR, the control objectives of load tracking and keeping the heat exchanger outlet temperature constant are determined. Then, two controllers, the parallel proportional-integral-derivative (PID) and the cascade PID, are designed, and their control effects are compared and analyzed. The results show that the steady-state error of the model is less than 0.05%, and the parameter response trend of the model without control is consistent with the theoretical analysis, and the simulation speed is faster. In terms of control, the two controllers can achieve the control objectives, and the adjustment time of nuclear power and heat exchanger outlet temperature is less than 150 s with small fluctuation amplitude. Therefore, the dynamic model of HPR established in this paper can be used to verify the control method, and the two controllers based on PID design have great control effect, and the cascade PID controller has better control performance.

Research on Optimization of Core Power Regulation System of Swimming Pool Reactor Based on PSO-BP Neural Network

Peng Zhiwen, Chen Xiaoliang, Zhu Jiachen, Wang Feng

2024, 45(4): 173-180. doi: 10.13832/j.jnpe.2024.04.0173

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Abstract:
Based on the MATLAB/Simulink, the simulation model of the power regulation system and the primary heat transfer system of the 49-2 swimming pool reactor was constructed, and the external reactive disturbance simulation test was carried out to verify the accuracy of the model. The proportion integration differentiation (PID) controller combined with particle swarm optimization (PSO) and BP neural network was used as the main controller, and the response of the regulating system under core reactivity and core inlet temperature disturbance was simulated, which was compared with that of the original controller of swimming pool reactor and the traditional BP neural network controller. The results show that the PID controller based on PSO-BP neural network can make the core reach a stable state quickly, with shorter regulating time and smaller overshoot, and has better robustness and stability.

Study on Orthogonal Experiments of Jet Breakup and its Modeling Based on MPS Method

Peng Cheng, Meng Xianpin, Deng Jian

2024, 45(4): 181-189. doi: 10.13832/j.jnpe.2024.04.0181

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In order to study the primary influencing factors of jet breakup length during the core melt jet breakup process and their ranking, 25 sets of experiments with 3 factors and 5 levels were designed based on the orthogonal experimental method, and the jet breakup lengths were obtained for each condition by using the Moving Particle Semi-implicit method (MPS). The simulation results were analyzed by polarity and variance analysis, and it was concluded that the primary influencing factors of jet breakup length were jet to coolant density ratio, jet velocity and jet diameter, and all of them have a second level of significance (**). Their ranking in the same level was in the following order: jet velocity > jet to coolant density ratio > jet diameter. Moreover, a new empirical relationship for predicting jet breakup length was fitted using simulated data, and the error margin of the model was kept within ±30% when the ratio of jet to coolant density was 1.1~13.6.

Research on Efficient Verification and State Recognition Method for the Action Reliability of Manual Globe Valve

Zhou Suting, Zhang Lin, Nie Changhua, Fan Wenyutao, Huang Yanping, Liu Jie, Yuan Kai

2024, 45(4): 190-195. doi: 10.13832/j.jnpe.2024.04.0190

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Abstract:
As a typical valve in primary system, manual globe valve is of great importance to maintain system operation and protect system safety. In order to verify the action reliability of the nuclear-grade manual globe valve and determine its operation state accurately and quantitatively, this paper studies and establishes an integrated intelligent operation device for manual globe valve action test, and proposes a method for identifying the state of the manual globe valve based on the combination of wavelet packet decomposition and support vector machine (SVM). Firstly, the torque signal is employed as the characteristic curve and the wavelet packet decomposition technique is utilized to extract the time-frequency domain features. The time domain and time-frequency domain features are integrated to construct the hybrid feature vector. Secondly, the Principal Component Analysis (PCA) is used to perform the dimensionality reduction analysis on the feature vectors to obtain fault feature vectors. Finally, the support vector machine (SVM) method is employed to identify the action state of valve. The results shows that the device constructed in this study solves the problems of long time-consuming and low efficiency in verifying the reliability of manual globe valve actions, as well as the difficulty in quantifying the evaluation of the action process. The proposed method can identify the three action states of the valve accurately and efficiently.

Design and Multi-Objective Optimization Study of Liquid Lead-Supercritical Carbon Dioxide Heat Exchanger

Li Liangxing, Shi Shang, Zhao Haoxiang, Zhao Jiayuan

2024, 45(4): 196-204. doi: 10.13832/j.jnpe.2024.04.0196

Abstract(281) HTML (20) PDF(28)

Abstract:
In order to improve the comprehensive heat transfer performance of the primary heat exchanger in lead-cooled fast reactors, the present study established a thermal-hydraulic model for a spiral-coil primary heat exchanger using liquid lead and supercritical carbon dioxide (S-CO₂) as working fluids. A design code was developed in MATLAB language, and a multi-objective optimization design was conducted on the heat transfer area and comprehensive performance evaluation factor of the primary heat exchanger by employing the Non-dominated Sorting Genetic Algorithm-II (NSGA-II). The results showed that the optimization design method proposed in this paper can effectively reduce the heat transfer area of the heat exchanger and improve its comprehensive performance. In the design of the primary heat exchanger, priority should be given to the outer diameter of tubes, the number of spiral tube layers and the number of spiral tubes in the first layer, so as to reduce the heat exchange area and improve the comprehensive heat exchange performance.

Impact of Decay on the Transport of Radioactive Aerosols in Long Square Tubes

Liu Man, Xia Mingming, Chen Zhi

2024, 45(4): 205-212. doi: 10.13832/j.jnpe.2024.04.0205

Abstract(50) HTML (24) PDF(13)

Abstract:
Decay radiation can cause accumulation of surface charges of radioactive aerosol particles, and then affect their migration process. However, the charge effect of decay is not considered in the current radionuclide transport simulation. In this study, a particle decay charging model was established based on Python, and a coupled scheme of particle decay charging and flow field was proposed and implemented in Fluent. The results of decay charge model of particles containing ¹⁰⁶Ru, ¹³¹I, ¹³²Te and ¹³⁷Cs are analyzed respectively, and the results show that the particle charge will reach the equilibrium value in a short time. The flow of particles containing ¹³²Te in a long square tube is simulated in Fluent. The results show that the electric field force mainly exists near the tube wall, pointing to the direction of particle concentration decline, which indicates that decay charge will promote the diffusion of aerosol and fill the whole space more quickly. This study provides a reference for the coupling scheme and simulation results of decay, electric field, and flow field in subsequent simulations of radioactive nuclide transport.

Capacity Configuration and Operation Optimization of a Low-Temperature Reactor Nuclear Heating System with Heat Storage

Liu Weiqi, Wang Jinshi, Xue Kai, Sun Zhiyong, Liu Xingmin, Li Gen, Yan Junjie

2024, 45(4): 213-220. doi: 10.13832/j.jnpe.2024.04.0213

Abstract(95) HTML (40) PDF(14)

Abstract:
In order to meet the growing demand for low-carbon heating and improve the operational flexibility and economic benefits of the heating system, a nuclear heating system (DHGHS) integrating a "Yanlong" pool-type low-temperature heating reactor (DHR-400), a heat storage pool, and a gas boiler was proposed. A central heating region in Liaoyuan City was taken as the application scenario of DHGHS, the equipment capacity and operation optimization with the goal of minimizing the annual cost were carried out. A comparison between DHGHS and four heating schemes including DHR-400 and heat storage pool, DHR-400 and gas boiler, gas boiler, and ground source heat pump was conducted. The results show that the heat storage pool with a rated volume of 3.15×10⁵ m³ and the gas boiler with a rated capacity of 82.79 MW can achieve the flexible operation of DHGHS throughout the heating period. The total number of power adjustments of DHR-400 in the whole heating period is only 177 times. The optimal annual cost of DHGHS is RMB 1.16×10⁸, which is lower than the other four heating schemes. The optimal heating scale of DHGHS is 1.18×10⁷ m². The work in this paper can provide theoretical guidance for the design and operation optimization of the multi-heat source nuclear heating system.

Research on Vibration Measurement Method of Nuclear Power Plant Pipeline Based on Unmarked Vision Algorithm

He Mengfu, Zhang Yiming, Qin Manqing, Xu Zili, Liao Tongtong

2024, 45(4): 221-227. doi: 10.13832/j.jnpe.2024.04.0221

Abstract(51) HTML (16) PDF(12)

Abstract:
In order to improve the problem that the vibration response of thin-walled pipes and small branch pipes is difficult to be effectively measured by contact measurement method, this paper proposes to calculate the optical flow of adjacent frames at different times based on the camera calibration algorithm and optical flow algorithm under the condition of visual measurement, so as to realize the unmarked visual structure motion measurement in the two-dimensional direction of the pipeline. Experimental verification was conducted on two typical structures, cantilever beam and nuclear power pipeline, and the measurement results of random points were compared with those of laser displacement sensor and acceleration sensor. The results indicate that the results of visual measurement of pipeline vibration are basically consistent with those of laser displacement sensor and acceleration sensor, and the relative error is less than 4.9%. Therefore, the unmarked visual structure motion measurement algorithm proposed in this paper can be used as a non-contact measurement option for pipeline vibration measurement.

Dissolution Behavior of Steam Generator Deposit in EDTA Solution

Song Lijun, Xiao Yan, Sun Yun, Tian Zhaohui, Liu Canshuai, Zou Wei

2024, 45(4): 228-234. doi: 10.13832/j.jnpe.2024.04.0228

Abstract(40) HTML (15) PDF(8)

Abstract:
In order to explore the solubility of chemical cleaning agent EDTA on simulated deposit Fe₃O₄ and steam generator sludge, and to guide the selection of chemical cleaning processes, XRF and ICP-OES were used to analyze the dissolution effects of solution temperature, EDTA concentration, and dissolution time on simulated deposit Fe₃O₄ and SG sludge. The chemically modified electrode was prepared by Fe₃O₄, and the cyclic voltammetry test and AC impedance test of the modified electrode were carried out by using a three-electrode system. The results indicate that the higher the solution temperature, the higher the EDTA concentration, and the stronger the solubility of Fe₃O₄. The longer the dissolution time, the higher the dissolution rate of Fe₃O₄ and SG sludge. Due to the differences between SG sludge and Fe₃O₄, EDTA solution has a weaker ability to dissolve SG sludge than Fe₃O₄. The electrochemical reaction process of Fe₃O₄ modified electrode in EDTA solution is diffusion-controlled process.

Research and Application of Influence of Black Rod and Gray Rod on Control Rod Drop Time in Nuclear Power Plant

Zhang Hengkai, Liu Hang, Liu Jikun, Liu Shuangjin, Zhao Yuntao

2024, 45(4): 235-240. doi: 10.13832/j.jnpe.2024.04.0235

Abstract(76) HTML (44) PDF(22)

Abstract:
In order to achieve more accurate and detailed drop time of control bank in nuclear power plant, based on the force analysis of control rod assembly in Chinese Pressurized Reacter 1000 MW (CPR1000) and Advanced Chinese Pressurized Reactor 1000 MW (ACPR1000⁺) nuclear power units and the test results of 11 units, it is found that the influence of core flow distribution difference on the drop time of control rod assembly is negligible during the commissioning and start-up of the units, but the influence of mass difference due to different materials of black rod and gray rod on the drop time is obvious. The average drop time of gray rod is about 4.6% longer than that of black rod because the mass of gray rod is 8.5 kg less than that of black rod. The test results are consistent with the theoretical expectation. Accordingly, it is suggested that the drop time consistency check of black rod and gray rod should be considered separately, and the acceptance criterion of drop time consistency check with 5 times standard deviation is put forward for the first time, which can achieve more accurate and detailed drop time consistency valuation than the original acceptance criterion.

Study on Operation Reliability of Reactor Trip Circuit Breaker in Nuclear Power Plants

Li Huwei, Zhang Yangcheng, Li Bin

2024, 45(4): 241-244. doi: 10.13832/j.jnpe.2024.04.0241

Abstract(64) HTML (19) PDF(18)

Abstract:
The reactor trip circuit breaker (RTCB) is a very important actuator in the reactor protection system (RPS) of nuclear power plant, and its reliability is directly related to whether the reactor can achieve and maintain a stable safety state. In this paper, the operation status of RTCB in domestic nuclear power plants is fully investigated, the operation reliability of RTCB is deeply analyzed and studied, the failure modes and causes are summarized, and the trend and influence of its operation reliability change are analyzed. The existing problems in management and safety awareness are put forward, and suggestions for follow-up work are given from the aspects of equipment manufacturing, operation and maintenance. This study has referential significance for the related work of RTCB in domestic operating nuclear power plants.

ACP100 Radiation Streaming Shielding Design Based on Discrete Ordinates Visual Modeling Technology

Tang Songqian, Chen Xin, Liu Jiajia, Wen Xingjian, Tian Chao

2024, 45(4): 245-247. doi: 10.13832/j.jnpe.2024.04.0245

Abstract(101) HTML (48) PDF(23)

Abstract:
The design characteristics of ACP100 make the radiation steaming become one of the radiation protection problems that need to be paid attention to, and it is necessary to carry out targeted shielding design. In order to improve the accuracy of discrete ordinates method in small modular reactor modeling calculation, a visual modeling plug-in of discrete ordinate method based on the code NX is developed in this study, which can directly define all attributes in NX graphical interface, and automatically complete meshing to form a complete calculation input. In this study, the radiation streaming calculation of ACP100 is completed based on the visual modeling technology and the self-developed three-dimensional discrete ordinates transport calculation code Hydra, and a special shielding module is set for the radiation streaming problem, which greatly reduce the thermal neutron flux in the main pump room and the radiation dose of the operating platform.

Research on Dynamic Parameter Model of Electrical Performance of Reactor Control Rod Drive System

Li Mengshu, Tang Shihan, Zheng Gao, He Zhengxi, Li Qing, Fu Guozhong, Peng Ziheng, Chen Shuaijun, Zhang Yun

2024, 45(4): 248-254. doi: 10.13832/j.jnpe.2024.04.0248

Abstract(625) HTML (20) PDF(32)

Abstract:
Due to the lack of accurate parameter model of electrical performance in the dynamic change process of the reaction system, the control effect of the existing control rod drive system in nuclear power plant is not good under different environmental conditions. In this paper, the static electromagnetic simulation model of the system is established, and the electrical performance parameters (resistance, inductance) and electromagnetic force discrete data of the system under all conditions are obtained by state classification. Based on the data, the dynamic parameter model and algorithm of current, electromagnetic force, armature displacement and acceleration in the dynamic process of the control rod drive system are established. The algorithm is finally verified by the combination of electromagnetic model, circuit model and numerical simulation, and it can simulate the running state of the actual system and has the ability to predict the electrical performance parameters of the system at the next moment only by relying on control instructions and current. The dynamic parameter model constructed in this paper can lay a foundation for the formulation of efficient intelligent control strategy.

Automatic Control Method of Nuclear Thermal Propulsion System Based on vPower

Ma Xinyi, Han Wenbin, Deng Jian, Huang Shanfang, Qi Zhichao

2024, 45(4): 255-261. doi: 10.13832/j.jnpe.2024.04.0255

Abstract(59) HTML (18) PDF(18)

Abstract:
Nuclear thermal propulsion has the advantages of large thrust, high specific impulse, high energy conversion efficiency and long operating time, with broad prospects in the field of deep space exploration. Automatic reactor control can reduce the misoperation accident caused by human, improve economic performance and reliability, and reduce unnecessary losses. Based on vPower simulation support platform, the automatic control simulation of typical nuclear thermal propulsion system is carried out to study the automatic control method of nuclear thermal propulsion system. The simulation control system is designed and added by determining the control strategy, selecting the proportional differential integral (PID) as the main control method, and adding the reactive feedback module to improve the system simulation model. The simulation results show that with the designed control system, the automatic control of nuclear thermal propulsion system start-up and shutdown can be realized, and the power, specific impulse and thrust can be automatically controlled and adjusted according to the external target requirements.

Numerical Simulation and Experiment Research on Radioisotope Thermoelectric Generator

Huang Xueliang, Li Mancang, Chen Zhang, Zhang Xinhu, Zhou Daijie, Chen Zhiyu, Xie Yunli, Guo Rui, Wang Yu

2024, 45(4): 262-266. doi: 10.13832/j.jnpe.2024.04.0262

Abstract(117) HTML (29) PDF(20)

Abstract:
Radioisotope thermoelectric generator is a device that converts the thermal energy produced by the decay of radioactive isotopes into electric energy. It involves the strong coupling of thermal-electric physical fields and is difficult to simulate accurately. In this paper, based on a ⁹⁰Sr Radioisotope thermoelectric generator prototype, firstly, the digital simulation model of the direct coupling of thermal-electric physical fields of the prototype is established, and the model parameters are optimized by combining the experimental data. Then, the accuracy of the model simulation under the steady and dynamic operation of the prototype is verified by simulation tests. Finally, the thermal-electric field analysis of the whole prototype and the output power research with different load resistance are carried out by using the model. The results show that the heat leakage of the prototype system accounts for 26% and the electric energy loss of the circuit is 10% under given operating conditions. Under the best matching load resistance, the maximum output power of the whole prototype can reach 96 mW, and the thermoelectric conversion efficiency is 2%.

Experimental Investigation on Minimum Film Boiling Temperature during Quenching of FeCrAl

Wang Zefeng, Deng Jian, Qiu Zhifang, Chen Xi, Wang Xiaoyu, Chen Jianda, Xiong Jinbiao

2024, 45(4): 267-273. doi: 10.13832/j.jnpe.2024.04.0267

Abstract(65) HTML (35) PDF(12)

Abstract:
FeCrAl is proposed as one of the candidate materials of accident tolerant fuel (ATF) cladding, which can suppress hydrogen generation under severe accident condition, and improve reactor accident tolerance. In this paper, the boiling heat transfer behavior of FeCrAl and Zr-4 during quenching is studied based on visualization method. The surface heat flux and temperature of FeCrAl are calculated by solving one-dimensional inverse heat conduction problem, and the effects of surface oxidation and solid thermophysical properties on the quenching behavior of FeCrAl are analyzed. The results show that as the liquid subcooling degree increases, the quenching duration of FeCrAl is decreased, and the minimum film boiling temperature increases. With the increase of solid thermophysical property (ρc_p)_w, the quenching duration increases and the minimum film boiling temperature decreases. Because of the excellent high-temperature oxidation resistance of FeCrAl, the effect of surface oxidation on its boiling heat transfer behavior during quenching can be ignored.

Anti-Noise Coding Research for Next-Generation Nuclear Power DCS Communication System

Shan Weiwei, Ren Jie, Peng Weilun, Zeng Hui, Li Sixing, Xiao Anhong, Feng Jintao, Deng Yuhao

2024, 45(4): 274-279. doi: 10.13832/j.jnpe.2024.04.0274

Abstract(1212) HTML (19) PDF(26)

Abstract:
In the process of introducing wireless signals in the rupgrading of wireless circuits in nuclear power plants or the design of the next generation distributed control system (DCS), it is necessary to improve the quality of wireless communication through error-correcting codes. This paper investigates the communication performance of encoded wireless signals in the vicinity of nuclear power plant instrumentation and control devices. Firstly, the problems faced by wireless communication signals in nuclear power plant are expounded. Secondly, the communication channel model around instrument and control devices is established, and the performance of low-density parity-check code (LDPC) coding in 5G enhanced mobile bandwidth (5GeMMB) scenario is analyzed by Monte Carlo simulation. Finally, the simulation analysis results are further verified through equipment development and field experiments. The research results show that the applicability of LDPC coding in DCS wireless communication environment in 5GeMMB scenario is insufficient, and the communication design needs to be further improved to enhance the availability of wireless communication in instrument and control devices of nuclear power plant. This study can provide some design reference for introducing wireless signals into the production system of nuclear power plants.

2024 Vol. 45, No. 4