核动力工程

Development and Verification of Two-step Spectrum Unfolding Code

Hu Xiao, Huang Yi, Wang Jie

2024, 45(6): 1-8. doi: 10.13832/j.jnpe.2024.06.0001

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Abstract:
To address the challenge of unknown preset spectra, this paper introduces a two-step spectrum unfolding method that combines the generalized regression neural network (GRNN) and the iterative algorithm. We have independently developed the spectrum unfolding codes for GRNN and iteration and conducted separate and comprehensive validations of the codes. Initially, we utilized activation method data from the Chinese Experimental Fast Reactor (CEFR) to validate the codes. The results indicated that at neutron energies greater than 0.1 MeV, the GRNN results deviated by a maximum of 10.36% from the theoretical spectra. The iterative method’s results deviated by a maximum of 9.15% compared to those obtained using the least squares method. The calculated single nuclear reaction rates showed a maximum relative deviation of 11.71% from the experimental values, indicating good agreement. Furthermore, the GRNN method demonstrated higher accuracy compared to the iterative method without accurate pre-set spectra. Finally, comprehensive validation was performed using Russian boron carbide irradiation data, revealing a maximum deviation of 11.42% in the fast neutron region between the two-step method and the iterative method with pre-set spectra. Therefore, employing a "two-step spectrum unfolding method" to address the challenge of unknown pre-set spectra is feasible, with errors remaining within acceptable limits. The innovative spectrum unfolding method introduced in this paper offers fresh perspectives for the spectrum unfolding of new reactors and offers significant reference value for experiments with unknown pre-set spectra.

Physical Analysis of the Heat Pipe Cooled Micro Nuclear Reactor Based on Thorium-Plutonium Mixed Fuel

Wang Feng, Sun Yuannan, Liu Bin

2024, 45(6): 9-14. doi: 10.13832/j.jnpe.2024.06.0009

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In order to improve the non-proliferation performance of heat pipe cooled micro nuclear reactors and realize the sustainable development of nuclear energy, abundant thorium-based fuel was adopted in this study with the Design A of Idaho National Laboratory (INL) as a reference, and the physical properties of the reactor core in terms of neutron energy spectrum, reactivity coefficients, power distribution and burnup were investigated by using the OpenMC code based on the Monte Carlo method. The results show that compared with UO₂ fuel, thorium-plutonium fuel heat pipe cooled nuclear reactor reduces the fuel loading, extends the operation time, and improves fuel conversion ratio. The overall power distribution of the core is non-uniform, but the axial power deviation is small. Moreover, the reactivity feedback coefficient is negative, ensuring inherent safety of the reactor core. The effective delayed neutron fraction is small. This study will provide reference for the application of thorium-plutonium fuel in heat pipe cooled micro nuclear reactors.

Research on Rapid Analysis Method of Ex-core Detector Response Based on BP Neural-network Algorithm

Li Yingjie, Xia Zhaodong, Zhang Geng, Sun Mingze, Ning Tong, Pan Cuijie, Ma Xiaodi, Sun Xu

2024, 45(6): 15-21. doi: 10.13832/j.jnpe.2024.06.0015

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The response of the ex-core detector describes the corresponding relationship between neutron fluence rate and current signal, which plays a crucial role in the safe operation of the reactor. In response to the problem that both deterministic and Monte Carlo methods cannot balance computational efficiency and accuracy in calculating the response of ex-core detectors, a back propagation (BP) neural network algorithm is used to quickly calculate the response of ex-core detectors. Based on the core design system CMS, the physical modeling of the existing million-kilowatt pressurized water reactor core in China was carried out. The fuel assembly arrangement and burnup changes in the core are used as the inputs of the BP neural network, and the corresponding fuel assembly arrangement and detector responses outside the reactor under different burnup are used as the outputs of the BP neural network. A three-layer BP neural network model is constructed and optimized. After calculation verification, the optimized model can quickly calculate the response of the ex-core detector, and the predicted value has a smaller error compared to the core design system CMS calculation value. It has good engineering application prospects, providing a new idea for calculating the response of the ex-core detector.

Research and Verification of Subcritical Rod Worth Measurement Method for PWRs Based on Bamboo-C Code

Li Zaipeng, Bai Jiahe, Wan Chenghui, Fang He, Pan Zefei, Wu Hongchun

2024, 45(6): 22-29. doi: 10.13832/j.jnpe.2024.06.0022

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The measurement of control rod worth is very important for the validation of nuclear design. Conventionally, the control rod worth measurement is carried out through the boron-dilution method, the rod-swap method, or the dynamic rod worth measurement method. These conventional methods lead to the fact that the zero power physics test occupies too long time. Therefore, this paper proposes the three-dimensional neutron source multiplication method. This method can achieve the measurement of subcriticality as well as control rod worth based on the measured source-range detector signals during the rod withdrawal process and the simulation results by Bamboo-C code. The whole flowchart relies on the current procedure of rod withdrawal process. This method has excellent engineering application ability and can significantly improve the economy of nuclear power plants. The measurement data during the rod withdrawal process in the M310-type reactor of Tianwan NPP is adopted for the verification of the proposed method. Most measured results of control rod worth can meet the engineering acceptance criteria (±10%), which proves that the method proposed in this paper has research value in engineering application.

Acceleration Method and Verification of Theoretical Calculation of Irradiation Supervision for Pressurized Water Reactor

Ye Yaoxin, Zhao Jun, Bao Pengfei, Yu Chao, Jiang Pingting

2024, 45(6): 30-38. doi: 10.13832/j.jnpe.2024.06.0030

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To address the inefficiency inherent in traditional pressurized water reactor irradiation surveillance theoretical calculation methods, which require multiple physical modeling and particle transport calculations, this study introduces an accelerated approach of neutron fluence calculation for pressure vessel irradiation surveillance capsules. The method is based on the Forward-Weighted Consistent Adjoint-Driven Importance Sampling (FW-CADIS) technique, coupling the Monte Carlo (MC) method with the Discrete Ordinates (SN) method. A comprehensive investigation into the influencing factors of computational accuracy and speed was conducted for a CPR1000 pressurized water reactor, validating the applicability of the proposed method under various core parameters and providing recommended values for SN transport simulation parameters. Verification and validation of the proposed method were performed on a CPR1000 pressurized water reactor. The results show that the Figure of Merit (FOM) of neutron fluence rate calculations is improved by approximately 95 to 181 times compared with the direct MC method, with a relative deviation between the calculated neutron fluence and measured values not exceeding 8%. Consequently, the irradiation surveillance theoretical calculation method presented in this study effectively enhances computational efficiency while meeting the precision requirements for engineering applications.

Research on Nuclear Data Target Accuracy Assessment Based on Subspace Method

Qiao Yaxin, Wu Xiaofei, Hou Long

2024, 45(6): 39-46. doi: 10.13832/j.jnpe.2024.06.0039

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According to the target uncertainty limit of reactor physics response calculation, target accuracy assessment solves a problem which identifies the demands for nuclear data uncertainties, which is of great significance for guiding the research direction of nuclear data and improving the economy and safety of reactors. Target accuracy assessment is a nonlinear constrained optimization problem in mathematics. Due to the ultra-large amount of nuclear data, solving the optimization problem in full-space is impossible. Subspace method is an efficient dimensionality reduction method. Through matrix transformation defined by subspace, a high-dimensional problem can be transformed into a low-dimensional problem, while the high-dimensional information is mostly retained, and the stability of numerical calculation is enhanced. Research on ZPPR-9 nuclear data target accuracy assessment based on subspace method shows that, with the 0.3% target uncertainty limit of effective multiplication factor, computational dimension can be reduced from 1170 to 71. The numerical method presented in this paper can be used in future target accuracy assessment calculations.

Research on Thermal Response Characteristics of Space Nuclear Power in High-Temperature and High-Velocity Impact Experiment under Accidental Reentry

Zhou Xu, Hu Yupeng, Wang Yijun, Wan Kun, Deng Zhifang, Zhu Changchun, Hu Shaoquan

2024, 45(6): 47-54. doi: 10.13832/j.jnpe.2024.06.0047

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High-temperature and high-velocity impact simulation test is a significant experiment to evaluate the safety of space nuclear power reactor in the accident impact on ground after accidental reentry. In this paper, a finite volume model coupling conduction, convection and radiation is established for the heat loading and high-velocity flight phase of the test, and the thermal response characteristics of the core simulator of the space nuclear reactor in the test are numerically studied, and the effects of loading temperature change rate and diameter-height ratio are analyzed. The results show that during the heat loading phase, the highest temperature and the lowest temperature of the core simulator are located at the junction of the side surface and the bottom surface and at the center of the simulator respectively. The time to reach thermal equilibrium is not only affected by the change rate of loading temperature, but also depends on the diameter-height ratio of the simulator. In the high-velocity flight phase, the highest and lowest temperatures of the core simulator are opposite to those in the heat loading phase, and the lowest temperature decreases with the increase of the diameter-height ratio and flight time. The research results can support the development and experimental design of high-temperature and high-velocity impact simulation test system.

Modified SST k-ω-γ Model and Prediction of Laminar to Turbulent Flow Transition in Helical Tube

Cai Yuntong, Zhao Houjian, Li Xiaowei, Su Yang, Lu Yiming, Guo Zhangpeng, Liu Fang

2024, 45(6): 55-62. doi: 10.13832/j.jnpe.2024.06.0055

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Helical tube heat exchangers are widely used in various fields due to compactness and good heat transfer performance. Different from the flow in a straight pipe, the fluid will be subjected to centrifugal force when flowing in a helical tube. Due to the existence of centrifugal force, the critical Reynolds number for the transition from laminar to turbulent flow in the helical tube increases with the increasing of helical curvature ratios. In the current investigation, the transition process from laminar to turbulent flow in the helical tube is studied by numerical method. By analyzing the relationship between the resistance coefficient and Reynolds number, the accuracy of simulating the transition process by the shear stress transport (SST) k-ω-γ-Re_θ model and SST k-ω-γ model is compared, and the effects of different inlet turbulence intensities (5% and 10%) on the calculation results are analyzed. The SST k-ω-γ-Re_θ model is more sensitive to inlet turbulent intensities, while the SST k-ω-γ model is less affected by inlet turbulent intensities. The critical Reynolds number obtained by SST k-ω-γ model is larger than the critical Reynolds number calculated by the empirical correlation in the literature. In this paper, by adjusting the empirical coefficient of γ transport equation in SST k-ω-γ model, it is found that C_TU1 has a significant influence on the onset of the transition process. With the same helical curvature ratio, the critical Reynolds number increases with the increasing of C_TU1. In the current investigation, based on the existing empirical formula, the C_TU1 for different helical curvature ratios (δ=0.02, 0.04, 0.06, and 0.11) are determined, and the correlation between helical curvature and C_TU1 is obtained by fitting. The accuracy of the modified SST k-ω-γ model in simulating the laminar and turbulent friction factors in helical tubes is verified, and the differences of velocity, turbulent kinetic energy and turbulent viscosity in the calculation results of SST k-ω model and modified SST k-ω-γ model are compared.

Prediction and Analysis of Heat Transfer Characteristics of Supercritical Fluids Based on Interpretable Machine Learning

Li Haozhe, Song Meiqi, Liu Xiaojing

2024, 45(6): 63-74. doi: 10.13832/j.jnpe.2024.06.0063

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The physical properties of supercritical fluids change drastically near the pseudo-critical temperature, making it challenging to accurately predict heat transfer characteristics. In this study, the method of interpretable machine learning was used to predict and analyze the heat transfer characteristics of supercritical fluids. The particle swarm optimization algorithm (PSO) was used to search for the optimal hyperparameters of the back propagation neural network (BPNN) model, the supercritical fluid heat transfer prediction model was established, and its accuracy was compared with the traditional empirical correlation. The global and local interpretation of the BPNN model was carried out by using the SHAP interpretable algorithm, and the supercritical correlation phenomenon and mechanism were found according to the change of feature importance under different conditions. The results show that the MAPE of the established neural network model on the test set is 1.4%, and the coefficient of determination R² is 0.9992, which has higher prediction accuracy compared with the empirical correlation formula. For vertical upward flow, buoyancy effect obviously has higher feature importance in heat transfer deterioration condition, which is the main factor of heat transfer deterioration behavior. Therefore, the research method based on interpretable machine learning established in this study has certain reference significance for further study of the heat transfer characteristics of supercritical fluids.

Study on the Deformation and Breakup Process of Jet Falling Film in Crossflow

Chen Qiuxiang, Hu Hongfei, Wang Haijun, Yang Kuang, Xu Bo

2024, 45(6): 75-83. doi: 10.13832/j.jnpe.2024.06.0075

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To investigate the deformation and breakup process of jet falling film in crossflow, a two-phase numerical simulation based on CFD software is conducted in this study. It is found that the inertia of the falling film flow and the lateral development of the initial Kelvin-Helmholtz (K-H) instability wave lead to the lateral fracture of the falling film in the initial stage of flow, resulting in the formation of several dry spots, while the dry spot occurs earlier on the windward side of the falling film than on the leeward side. The results show that there are two main breakup modes of falling film in crossflow: liquid film breakup and surface breakup. Liquid film breakup refers to the fracture and breakup along the flow direction of the falling film dominated by the Rayleigh-Taylor (R-T) unstable wave, while surface breakup refers to the peeling of liquid filaments and droplets on the windward side of the falling film dominated by the K-H unstable wave. The liquid to gas momentum ratio plays a significant role on the deformation and breakup process of jet falling film in crossflow. When the liquid to gas momentum ratio is greater than 13.16, surface breakup is the main breakup form of falling film. As the liquid to gas momentum ratio decreases, both surface breakup and liquid film breakup of the falling film increase simultaneously, leading to a significant increase of the falling film breakup. The continuous flow length and spanwise width of the jet falling film increase with the increase of the liquid to gas momentum ratio, while the offset distance of the falling film decreases with the increase of the liquid to gas momentum ratio.

Experimental Study on the Influence of Thermal Parameter Deviation on Steady-state Characteristic of Once-through Steam Generator

Du Daiquan, Zhang Ting, Zhuo Wenbin

2024, 45(6): 84-90. doi: 10.13832/j.jnpe.2024.06.0084

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In order to study the influence of design deviation of thermal parameters (average temperature of primary side, feed water temperature, load and steam pressure) on the steady-state characteristics of once-through steam generator (OTSG) under rated conditions, the thermal-hydraulic experiment of typical cells for prototype design of OTSG is carried out. The experimental conditions are as follows: the average temperature of primary side is 297-304℃, the load is 100%FP-104%FP (full power), the steam pressure is 4.0-5.8 MPa, and the feed water temperature is 138-152℃. The experimental results show that the deviation of thermal parameters mainly affects the outlet steam temperature, heat transfer power and secondary side pressure drop of OTSG. The design deviation of the average temperature of primary side has the greatest influence on the steam temperature. The outlet steam temperature decreases with the increase of steam pressure under the condition that the design deviation of each thermal parameter occurs at the same time.

Research on Thermal Diffusion Performance of Fuel Assembly Spacer Grid Based on Numerical Methodology

Chen Xi, Wang Xiaoyu, Cui Cong, Deng Jian, Liu Yu, Liu Luguo, Liang Yu, Peng Huanhuan

2024, 45(6): 91-97. doi: 10.13832/j.jnpe.2024.06.0091

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As one of the most important parameters in subchannel analysis codes used in core thermal-hydraulic design, Thermal Diffusion Coefficient (TDC) is generally acquired by single phase thermal-hydraulic tests, which are costly in time and economy. Based on the mechanism and model of the turbulent mixing, this paper expounds the simulation method of TDC in subchannel code, corrects the problem that TDC was obtained only by temperature field calculation in the past, and puts forward the thermal diffusion factor that characterize the temperature exchange between hot and cold channels. As a result, the whole TDC evaluation method based on CFD technology is formed, and the simulation results are compared to the test results. The results of comparative analysis show that the relative deviation between the results of numerical method and the experimental results is less than 10%, which are in good agreement. Considering certain conservative punishment, it can basically replace the related experiments and greatly improve the efficiency of design and development. Besides, the influences on the TDC of relative parameters, such as thermal hydraulic conditions, grid structure, number of grids, grid spacing, are analyzed. The results demonstrate that the TDC characteristics of fuel assemblies are largely related to the grid structure, and are less affected by thermal parameters.

Study on the Effects of Accident Tolerant Fuels on the Safety of CPR1000 Nuclear Power Plants

Liu Pingping, Liu Mengying, Xu Haode

2024, 45(6): 98-105. doi: 10.13832/j.jnpe.2024.06.0098

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In this paper, CPR1000 is taken as the reference unit. According to the Level 1 CPR1000 Probabilistic Safety Analysis (PSA) results, the Large Break LOCA, Intermediate Break LOCA, Small Break LOCA, Station Blackout (SBO), Total Loss of Feedwater (TLOFW) and Anticipated Transient without Trip (ATWT) with Loss of Main Feedwater (LOMF) are selected as the representative design extend condition (DEC) accident scenarios. Using LOCUST and SPRUCE, the thermal and hydraulic codes developed by China Nuclear Power Technology Research Institute Co., Ltd. based on the performance of accident tolerant fuel (ATF), deterministic calculations are carried out for the five types of ATF under development, namely, ATF-1, ATF-2, ATF-3, ATF-4, and ATF-5. Compared with the traditional UO₂-Zr material, the accident process, core damage time, system success criteria and personnel response time of different ATFs under the above typical accidents are analyzed. It is found that the lower peak cladding temperature and higher cladding limit temperature of ATF in the accident make CPR1000 unit have greater safety margin, which provides support for ATF material selection. Based on the results of deterministic analysis, the Level 1 PSA model is established for different ATF, and the influence of different ATF materials on the safety of CPR1000 unit is given from the perspective of probability theory. The results show that there is no substantial benefit from the direct application of existing ATF to existing reactors. Based on the deterministic and probabilistic analyses, the development direction of reactor based on ATF is given in this paper.

Ultrasonic Measurement Technology and Application on 14-Foot Fuel Assembly Deformation

Zeng Yuan, Liang Jun, Xie Weirong, Wang Gang

2024, 45(6): 106-111. doi: 10.13832/j.jnpe.2024.06.0106

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In order to understand the deformation of the fuel assembly after irradiation and provide an auxiliary means for optimizing the refueling design and adjusting the loading sequence, this measurement technology can quickly measure the deformation of the fuel assembly during the reactor unloading and draw the deformation distribution of the whole core fuel assembly by using ultrasonic ranging and post-image processing technology. The measurement results show that the deformation of the whole core fuel assembly of the same unit tends to be consistent at the end of the cycle, and the peripheral assemblies are deformed to the outside of the reactor to varying degrees at the end of the cycle. This technology can provide a basis for the optimization of refueling design and avoid the risk of rod drop test and the wear of peripheral fuel assembly grid.

Analysis and Research on Axial Stiffness Model and Influencing Factors of PWR Fuel Assembly

Jin Yuan, Gu Chenglong, Tian Wei, Zhang Yuxiang, Li Weicai

2024, 45(6): 112-120. doi: 10.13832/j.jnpe.2024.06.0112

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Currently, the axial analysis model for domestic commercial PWR fuel assembly accident condition mainly relies on foreign technology transfer, and the theoretical mechanism research depth is obviously insufficient. Based on the basic structural characteristics of fuel assembly and the basic mechanical theory, a small basic structure element model is established in this paper and its axial stiffness is analyzed, and then the basic model is reasonably extended to establish the entire fuel assembly axial analysis model. With the obtained analysis model, the sensitivity analysis of axial stiffness is conducted from three aspects of grid number, clamping system and guide thimble thickness. The analysis results show that the number of grid is negatively correlated with the axial stiffness of the assembly, the clamping force of the grid affects the sliding force threshold of the fuel rod, and the axial stiffness increases with the guide thimble thickness. The research results of this paper can provide new ideas for axial model research of new type fuel assembly under accident conditions.

Simulation Study on Concentration Polarization and Electrode Kinetics during Electrorefining of Uranium

Liang Bo, Zhang Meng, Sun Lanxin, Wang Jingyang, Lin Rushan, Han Wei, Jiao Caishan

2024, 45(6): 121-131. doi: 10.13832/j.jnpe.2024.06.0121

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A concentration-dependent Butler-Volmer electrode kinetics equation was established by correlating concentration with the overpotential through the Nernst equation. The mass transfer equation and potential distribution equation were optimized based on the supporting electrolyte theory, and the uranium electrorefining model was improved. The cyclic voltammetry curve, constant potential deposition process and constant current deposition process were simulated by using the new model, and the concentration polarization phenomenon and electrode dynamic behavior under different electrolytic conditions were quantitatively analyzed. The simulated cyclic voltammetry curve is in good agreement with the experimental results, verifying the accuracy of the model. Through modeling, the distributions of U(III) concentration, potential, and current density in the molten salt and on the surface of electrode were obtained. The key parameters such as diffusion layer thickness, limiting diffusion current and deposition layer thickness were predicted, and the driving force changes caused by concentration polarization during constant current deposition and constant potential deposition were compared. The numerical model established in this study can be used as a powerful tool to optimize process parameters and design process equipment, and has important physical significance for deepening understanding of uranium electrorefining mechanism.

Research on Vibration-induced Noise Simulation of Main Control Room of High-temperature Gas-cooled Reactor

Zhu Tenghao, Wang Hongtao, Wang Haitao

2024, 45(6): 132-138. doi: 10.13832/j.jnpe.2024.06.0132

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Noise in the main control room is one of the major concerns for operational safety of the nuclear power plant. In this paper, the impact of the main steam pipeline vibration on the noise level in the main control room of the high temperature gas-cooled reactor (HTGR) is investigated by using a structural finite element model and an acoustic boundary element model. A finite element model for frequency response analysis of the nuclear island of one HTGR conceptual design, and a boundary element model for acoustic analysis of the main control room in the frequency domain, are established respectively, to predict the noise levels in the main control room dominated by vibration transfer from the main steam pipelines. The influence patterns of various main steam pipes are explored, and the wall vibration that contributes most to the noise level of the main control room is identified based on acoustic contribution analysis. A method for optimizing the main control room noise through physical partitioning is proposed. The results show that, horizontal vibrations of the main steam pipeline generate higher noise levels in the control room compared to vertical vibrations. The maximum noise caused by the vibration of the main steam pipeline exceeds 60 dB. The walls near the main steam isolation valve room and the ceiling contribute the most to the indoor noise in the control room. Through physical partitioning, the noise level of the control room can be reduced significantly.

Research on Sensitivity Analysis Model of Grounding Measurement Capacitance Rod Position Sensor

Li Yanlin, Qin Benke, Bo Hanliang

2024, 45(6): 139-146. doi: 10.13832/j.jnpe.2024.06.0139

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Control rod position sensor is one of the six core components of the control rod hydraulic drive system, which provides the only true rod position indication for nuclear reactors. The grounding measurement capacitance rod position sensor has the advantages of high precision and strong anti-interference ability. Control rod positions can be measured step-by-step by this kind of sensors. In order to clarify the capacitive sensitive mechanism of this kind of capacitance rod position sensors, the sensitivity analysis model of the sensor is established by conformal mapping. Model modification and model validation are conducted by the numerical simulation and results of the static calibration experiments. The results show that the static measurement characteristics of the grounding measurement capacitance rod position sensor can be accurately analyzed by the sensitivity analysis model. The relative error between the results obtained by the sensitive analysis model and experiments is 3.4%. The sensitivity analysis model can be used for structural analysis and optimal design of the sensor.

Study on Limitation Size of Fretting Wear of Inconel 690 Steam Generator Tube

Li Tian, Xue Donglin, Chen Yanhui, Shao Chunbing, Huang Song, Hui Hu, Jiao Peng

2024, 45(6): 147-152. doi: 10.13832/j.jnpe.2024.06.0147

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This paper focuses on the burst pressure of Inconel 690 tubes with fretting wear defects. High-temperature burst test and eddy current testing on the defect size of tubes were conducted. Based on the normal distribution, a burst pressure prediction model and a defect depth prediction model with uncertainty are established. The direct cumulative and simplified statistical methods are used to calculate the burst pressure of the tubes with uncertainty. The results show that a more conservative prediction value of the burst pressure can be obtained by using the direct accumulation method to include the material properties, the burst pressure prediction model, and the 95th percentile worst-case scenario of eddy current testing. And, the simplified statistical method is used to take into account the material properties, the burst pressure prediction model and the uncertainty of eddy current detection, which can effectively reduce the over-conservative problem caused by direct accumulation error.

Study on Methodology and Requirements of Equipment Availability Qualification under Severe Accident in Nuclear Power Plant

Jin Xin, Tang Hui, Han Jiwei, Xu Jiaoshen, Zhu Zengpei, Zheng Xiaowei

2024, 45(6): 153-158. doi: 10.13832/j.jnpe.2024.06.0153

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In the design and construction of nuclear power plant, the equipment availability qualification is carried out to ensure that the equipment necessary for severe accident is capable to perform its intended safety functions under severe accident conditions. The requirements related to equipment availability qualification under severe accident proposed by the present codes and standards are studied in this paper. With consideration of the phenomena and processes of severe accidents, a set of systematic and operable methodology and requirements for the equipment availability qualification under severe accident is put forward, including the basic principles, selecting method of equipment, requirements for the environmental conditions determination, and equipment availability qualification processes, and typical example is also provided.

Research on Nuclear Signal Generator Based on Signal Characteristics of Pulsed Neutron Detector

Luo Tingfang, Bao Chao, Gao Zhiyu, Wang Li, Sun Qi

2024, 45(6): 159-165. doi: 10.13832/j.jnpe.2024.06.0159

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Pulsed neutron detector converts neutron fluence rate into random weak current pulse signal. Due to the particularity of this signal, nuclear measurement equipment generally requires reactor tests to verify the actual detection performance. Because of the research method by reactor test costs a lot and has time limit, this paper, based on a typical pulsed neutron detector such as boron-coated proportional counting tube, studies a pulsed detector signal model and its nuclear signal generator implementation scheme. The characteristics of each key part are verified through simulation. The verification results show that: the proposed nuclear signal generator scheme can generate a sequence of time intervals satisfying the exponential distribution, the single current pulse shape is similar to the detector and the amplitude can vary randomly in a uniform distribution.

Mechanism Study and Control of ⁴¹Ar Source Term in EPR Unit Coolant

Li Xiaoning, Ma Boyang, He Weihua, Tao Jiantang, Su Xing, Zong Guopeng

2024, 45(6): 166-171. doi: 10.13832/j.jnpe.2024.06.0166

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In order to control the Argon-41 (⁴¹Ar) radiation source term in European Pressurized Reactor (EPR) coolant, this research pioneers the development of a transport model for ⁴¹Ar in EPR coolant and elucidates the pattern of ⁴¹Ar specific activity variation under different conditions. Through the model analysis, it is found that the specific activity of ⁴¹Ar in the coolant of a unit is abnormal, and it is the first time to detect the formation of a stable gaseous radiation hotspot within the gas-liquid separator of the EPR's hydrogen refueling station. The model is used to trace the specific path that the environmental air carries ⁴⁰Ar into the coolant, and measures are taken to effectively control the radiation source term of ⁴¹Ar and reduce the radiation risk of the unit.

Development and Application of Full Range Closed-Loop Testing System for DCS Transformation of Daya Bay Nuclear Power Plant

Li Minggang, Wang Lei

2024, 45(6): 172-177. doi: 10.13832/j.jnpe.2024.06.0172

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Large-scale transformation projects of distributed control system (DCS) of in-service nuclear power plants are usually characterized by complex compatibility requirements, multiple transformation construction constraints, short construction period, and high quality requirements. Daya Bay Nuclear Power Plant is the first in-service nuclear power plant in China to carry out digital transformation of DCS, and it faces great challenges as there is no precedent in the country to draw reference from. Therefore, a full-range closed-loop testing method of DCS based on hardware-in-the-loop simulation technology is proposed. The simulation model of nuclear power plant process system is introduced into the testing environment of DCS system, and the closed-loop control of DCS on the process system is realized in the testing environment, so that the on-site commissioning, operation and maintenance test of nuclear power plant can be verified in advance in DCS supplier's factory. At present, it has been applied to the 30-year instrument and control system transformation project of Daya Bay Nuclear Power Plant. After full-range closed-loop verification, the DCS of Unit 2 of Daya Bay Nuclear Power Plant has been transformed within the overhaul period, with a total instrument and control transformation period of 98 days and successfully put into operation at the first test run. Practical results show that this technology can be applied to DCS transformation testing, realizing the goals of reducing the risk of on-site implementation, reducing the number of on-site debugging programes, and saving the duration of on-site implementation.

Research on Optimization Control of Nuclear Power Plant Coordination System Based on ESO-MPC

Guo Yongfei, Zhang Rongbin, Yao Zhiyuan, Lang Yukai, Zhao Jiayu

2024, 45(6): 178-184. doi: 10.13832/j.jnpe.2024.06.0178

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The regulation characteristics of nuclear island and conventional island are quite different, so it is necessary to coordinate the synchronous control of nuclear island and conventional island to achieve better control effect. It is of great significance to study the optimal control strategy of coordinated control system. In this paper, a model predictive control (MPC) algorithm based on extended state observer (ESO) is proposed to solve the problem that the coordinated control system of nuclear power plant is prone to disturbance. The proposed method accurately estimates the external disturbance by using ESO, and then integrates the disturbance estimation value into the rolling optimization process of MPC to realize the adaptive correction of the prediction model, thereby obtaining the required optimization control rate. In the simulation experiment, the algorithm proposed in this paper was compared with the performance of proportional integral differential control and multivariable model predictive controller, and the results showed that the algorithm proposed in this paper had good performance. In the scenario of unit load setting disturbance, the mean square error of main steam pressure and unit load by the proposed algorithm is 0.06 and 0.02 respectively, which is significantly better than the other two algorithms. The algorithm proposed in this paper can enable the coordinated control system of nuclear power units to achieve precise control performance in the presence of external disturbances.

Research on Multi-objective Optimization of Parameters of Small Pressurized Water Reactor Nuclear Steam Supply Control System

Li Zheng, Chen Chuqi, Zeng Wenjie, Li Ruokun

2024, 45(6): 185-191. doi: 10.13832/j.jnpe.2024.06.0185

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In nuclear reactors, the conventional PI controller parameter tuning method is cumbersome and complex, with strong dependence on human experience, which makes it difficult to achieve the simultaneous co-optimization of multiple PI controller parameters in the reactor. To study this problem, a small pressurised water reactor nuclear steam supply control system is established with the reactor coolant average temperature and steam pressure controller parameters as the optimization objectives, and a Non Dominated Sorting Genetic Algorithm-II (NSGA-II) is used to achieve the parameter optimization of the nuclear steam supply control system. The results show that the optimized control system effectively reduces the overshoot and response time of the controlled objects, improves the control performance of the control system, and at the same time lessens the dependence on human experience and achieves the intelligence of the parameter tuning process.

Analysis of the Initiating Event of TOPAZ-II Space Nuclear Reactor Power System

Yang Jialin, Ding Hongchun, He Fang, Zhang Haochun, Zhao Yulan

2024, 45(6): 192-196. doi: 10.13832/j.jnpe.2024.06.0192

Abstract(15) HTML (4) PDF(1)

Abstract:
To comprehensively evaluate the safety of space nuclear reactors and provide technical support for their design and operation, based on probabilistic safety assessment (PSA) method, this study takes TOPAZ-Ⅱ space nuclear reactor power system as the research object, and a study on its initiating events (IEs) is carried out. A list of 15 IEs covering all mission phases of the TOPAZ-Ⅱ is obtained through two means: operational experience feedback and Failure Mode and Effects Analysis (FMEA). These IEs are further classified into 6 groups according to their system response process. Research has shown that due to the uniqueness of design features, operating environment, and mission phases, there are significant differences in the IE of space nuclear reactor power system compared to ground nuclear power facilities. The IE list and its grouping method determined for TOPAZ-Ⅱspace nuclear reactor power system in this study have laid the preliminary research foundation for the PSA of space nuclear reactors.

Safety Characteristics Analysis of Helium-Xenon Cooled Reactor System under LOCA

Liao Haoyang, Ming Yang, Zhao Fulong, Lu Ruibo, Wei Ruixuan, Gao Puzhen, Tan Sichao, Tian Ruifeng

2024, 45(6): 197-205. doi: 10.13832/j.jnpe.2024.06.0197

Abstract(22) HTML (3) PDF(3)

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In order to avoid the high risk and harm caused by the loss of coolant accident (LOCA) in the reactor system, the developed LOCA analysis code for helium-xenon cooled reactor system was used to simulate a variety of LOCA transient conditions, and the system transient characteristics, volume filling influence characteristics, load following failure influence characteristics, break location influence characteristics and break size influence characteristics were analyzed. The results show that the system pressure and mass flow rate decrease rapidly after the LOCA. Volume filling can alleviate LOCA, reducing the rate of mass flow rate decline and the rate of reactor outlet temperature rise by 77.15% and 90.27%, respectively. Both the constant load and the break at high pressure have a negative impact on LOCA, resulting in an increase in the rate of mass flow rate decline by 13.85% and 79.83% respectively, and an increase in the rate of reactor outlet temperature rise by 15.84% and 96.06% respectively. The decrease rates of system pressure and mass flow rate rise with the increase of the break size. Especially between 15 mm and 30 mm of the break size, the flow rate decreases and the reactor outlet temperature increases significantly by 258.84% and 595.91%, respectively.

Study on Influence of Material Physical Properties Change on Containment Performance under Severe Accident

Liu Jing, Liu Baojun, Zhang Chunlong, Wei Wei, Liu Yu

2024, 45(6): 206-212. doi: 10.13832/j.jnpe.2024.06.0206

Abstract(15) HTML (3) PDF(1)

Abstract:
As the last barrier of PWR nuclear power plant, the integrity of containment in severe accident condition depends not only on the occurrence of severe accident phenomenon, but also on the performance characteristics of containment. At present, in the performance analysis of HPR1000, only the physical characteristics of materials at normal temperature are considered, which cannot reflect the influence of temperature rise and pressure rise of the containment itself under severe accidents. In this paper, according to the response of containment under severe accident and considering the change of material properties under accident, the influence of material properties on containment performance under accident high temperature is analyzed, and the difference of containment performance between normal temperature and accident high temperature is compared. The difference of severe accident risk at different temperatures is analyzed, and the influence on early massive radioactive release frequency, massive radioactive release frequency and severe accident management is evaluated. The analysis results show that the performance of the containment decreases with the increase of the temperature in severe accidents, but the vulnerability of containment is the equipment hatch as before. Compared with the two curves of the containment failure probability distribution at normal temperature and high temperature, due to the large free volume of HPR1000 containment, the loads generated by DCH and AICC will not threaten the integrity of the containment. The original operation setpoint of containment filtration and exhaust system is not challenged.

Research on the Application of IDHEAS Human Event Dependency Analysis Method

Zheng Tengjiao, Zhang Jiajia, Hou Jie, Xu Yunlong, Xu Qingqing

2024, 45(6): 213-219. doi: 10.13832/j.jnpe.2024.06.0213

Abstract(18) HTML (3) PDF(1)

Abstract:
Dependency is one of the most important issues in human reliability analysis (HRA). However, the existing dependency analysis methods have some problems, such as lack of basic data, insufficient cognitive theoretical foundation, and unclear selection principles of influencing factor levels, which lead to overly conservative analysis results and poor stability. For this reason, the Nuclear Regulatory Commission (NRC) of the United States established the IDHEAS dependency model based on the behavioral cognitive model of IDHEAS, and put forward the IDHEAS dependency analysis method (IDHEAS-DEP). In this paper, IDHEAS-DEP is systematically studied, the implementation process and key points are analyzed and summarized, and a typical category C human factor event group of Level 1 probabilistic safety analysis (PSA) is selected for example analysis, and a quantitative comparison is made with other dependency analysis methods. Theoretical research and case analysis show that IDHEAS-DEP can solve the problems of insufficient basic theory and conservative analysis results of existing dependency analysis methods to a certain extent, and the screening analysis of this method is more universal and practical in engineering. However, it still needs to solve the interface problem with other HRA methods, and the determination of the minimum joint human error probability is also the direction that this method needs to be improved in the future.

Study on Prediction and Diagnosis of Fuel Rod Break Size during Degassing Operation of PWR

Ye Yaoxin, Fu Pengtao

2024, 45(6): 220-225. doi: 10.13832/j.jnpe.2024.06.0220

Abstract(29) HTML (8) PDF(2)

Abstract:
The operation data of PWR nuclear power plant show that after the unit implements large flow degassing operation, the specific activity of fission products of primary coolant oscillates violently in a short time, which makes the fuel damage prediction method based on the average core state fission release-to-birth ratio (R/B) have prediction bias. Based on the parameters of degassing system and the mechanism of inert gas release in PWR nuclear power plant, this paper establishes a modified prediction and analysis model of inert gas release in degassing operation, gives the calculation method of degassing factor and inert gas release rate under degassing conditions, and optimizes the traditional prediction method of fuel rod break size based on R/B. The modified prediction method of degassing operation has been applied and verified in a PWR nuclear power plant. The maximum relative deviation of specific activities of six common inert gas nuclides predicted is 33.4%, and the others are less than 20%. The predicted fuel rod break size is large, which is consistent with the inspection results after shutdown.

Analysis and Optimization of Primary Circuit Pressure Drop Caused by RCP Inching for the First Time after Refueling Overhaul in Nuclear Power Plant

Li Jun, Wang Shuqiang

2024, 45(6): 226-231. doi: 10.13832/j.jnpe.2024.06.0226

Abstract(17) HTML (5) PDF(2)

Abstract:
After the refueling overhaul of pressurized water reactor nuclear power plant, the first inching of the reactor coolant pump (RCP) will cause a significant decrease in the primary circuit pressure. If the pressure drops below the operating range specified in the technical specifications, there is a risk of RCP cavitation and shaft seal damage, which threatens the safety of the core. In this paper, the mechanism that the primary pressure drops sharply when the RCP is inched for the first time after refueling overhaul in nuclear power plant is analyzed theoretically, and three optimization methods are put forward to control the pressure drop, including determining the timing and principles for shutting down the RCP, determining the starting sequence of the RCP, setting the initial pressure of the primary circuit and using the primary pressure regulating valve to automatically respond quickly. The practical results show that the optimized and improved control method for the inching of RCP can reduce the pressure drop in the primary circuit by 0.2 MPa. The optimized and improved control method of inching RCP established in this study can reduce the risk of cavitation and shaft seal damage of RCP.

Research on Environmentally Assisted Fatigue Analysis Method for Primary Circuit of Pressurized Water Reactor Units

Lyu Fangming, Jiang He, Tong He, Cao Guochang, Cao Hongsheng

2024, 45(6): 232-236. doi: 10.13832/j.jnpe.2024.06.0232

Abstract(19) HTML (5) PDF(3)

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Environmentally assisted fatigue (EAF) is an important aging mechanism for reactors with water as coolant, which has significant influences on the fatigue life of critical sensitive components in primary circuit of nuclear power plants. In this paper, the EAF screening process of critical sensitive components and the calculation formula of environmental fatigue correction factor (F_en) deriving from the United States are introduced. Meanwhile, the results of EAF key sensitive part screening for the license renewal units in the United States are statistically analyzed. Referring to the EAF analysis method of the United States, the critical sensitive components of the primary circuit of Units 1 and 2 in a domestic plant are selected. The processes, methods, and cases have important guiding and demonstration significance for EAF analysis in the domestic nuclear power plants.

Research on Detectable Rate Optimization of Nuclear Grade Pipe Welds

Wu Xiang, Cui Cong, Wu Zhisheng, Cai Dingyang, Zhao Qianli, Gan Yiran, Su Yingbin, Xiao Yunfei

2024, 45(6): 237-241. doi: 10.13832/j.jnpe.2024.06.0237

Abstract(14) HTML (2) PDF(1)

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The problem of undetectable welds in nuclear grade pipelines not only makes the weld detectable rate fail to meet the requirements of current standards, but also affects the state monitoring of welds during their life cycle, which is not conducive to the safe operation of reactor coolant system. Taking the pre-service inspection of nuclear grade pipeline welds in floating nuclear power plants as an example, the reasons for the undetectable welds are analyzed in detail by using descriptive statistics. The analysis results show that the imaging obstacle of weld detection is the main factor causing the weld to be undetectable, followed by the inaccessible factor of weld detection. The reasons are related to design, installation, structural functionality and equipment structural characteristics. The proposed optimization measures can improve the weld detectable rate from 65.5% to not less than 74%, which effectively improves the weld detectable rate and ensures the safety of the system operation.

Optimization Research on the Operating Diagram of HPR1000 Units

Cui Huaiming, Cai Zhiyun

2024, 45(6): 242-247. doi: 10.13832/j.jnpe.2024.06.0242

Abstract(27) HTML (4) PDF(4)

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Aiming at the narrow area of the operating diagram of HPR1000 nuclear power unit under specific conditions, this paper makes an optimization analysis and research on the operating diagram from the aspects of increasing the operating pressure of the residual heat removal (RHR) system, lowering the lower limit of the start-up pressure of reactor coolant pump, enlarging the allowable temperature difference between the two ends of the pressurizer surge line, and lowering the limit of the sub-cooling of the reactor coolant. And then, an optimized operating diagram for HPR1000 is proposed. The comparative analysis before and after optimization shows that the optimized operating diagram has a wider allowable operation range, improved operation efficiency and enhanced operation reliability.

Investigation on Hybrid Discontinuous Galerkin Method Based on First-Order Neutron Transport Equation

Sun Qizheng, Liu Xiaojing, Zhang Tengfei

2024, 45(6): 248-253. doi: 10.13832/j.jnpe.2024.06.0248

Abstract(19) HTML (6) PDF(2)

Abstract:
The development of advanced reactor designs imposes higher demands on neutronic numerical methods. To achieve accurate and efficient simulation of complex problems, this paper introduces a hybrid discontinuous Galerkin (HDG) method based on the first-order hyperbolic neutron transport equation (NTE). The method decouples the original equation into independent equations for each angular direction using the discrete-ordinates (S_N) method in angular space. In spatial discretization, this paper employs an upwind scheme that results in a blocked-lower-triangular global matrix coupling system, making it well-suited for complex, geometrically heterogeneous neutron transport scenarios with a large number of meshes. The study evaluates the performance of the proposed HDG method using the TAKEDA1 benchmark and a heterogeneous assembly problem. The results demonstrate that the HDG method achieves stable convergence for the aforementioned problems, with a maximum error between the effective multiplication coefficient k_eff and the reference solution of 108 pcm (1pcm = 10⁻⁵). In addition, compared with the traditional second-order even-parity method, the first-order HDG method is more efficient in spatial scanning, and the acceleration ratio is about 2 times in the above examples. Therefore, the proposed HDG method can provide an alternative solution for complex reactor problems.

Experimental Study on the Impact of Heat Pipe Failure on High-Temperature Heat Pipe Bundles and Matrix

Wang Jinyuan, Li Panxiao, Wang Chenglong, Zhang Zeqin, Tian Wenxi, Qiu Suizheng, Su Guanghui

2024, 45(6): 254-262. doi: 10.13832/j.jnpe.2024.06.0254

Abstract(40) HTML (4) PDF(6)

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In order to verify the feasibility of small heat pipe reactor, an experimental device of high temperature heat pipe bundle is designed in this study. The experimental device simulates the failure of heat pipe by pulling out the heat pipe, and simulates the core fuel rod by electric heating rod to explore the influence of heat pipe failure on heat pipe bundle, fuel rod and core matrix. It is found that the most direct impact of heat pipe failure is a sudden local temperature rise in the nearby matrix. Under the power of 4.2 kW, the average increase of matrix temperature in the vicinity of single heat pipe failure is about 70℃, and the average increase of matrix temperature in the vicinity of double heat pipe failure is about 120℃. The failure of a single heat pipe has a minor impact on the remaining normal heat pipes, with an average temperature rise of 15℃ in the evaporation section of the normal heat pipes. The average temperature increase of fuel element in the vicinity of double heat pipe failure is about 66℃. The experimental data of high temperature heat pipe bundle under heat pipe failure obtained in this study can provide data support for the modeling and simulation of heat pipe reactor.

Design and Experimental Study of High Temperature Flowing Liquid Metal Corrosion Device

Song Xiaoyong, Pang Yongqiang, Meng Xiancai, Tian Shujian, Zhang Dehao, Li Xu

2024, 45(6): 263-270. doi: 10.13832/j.jnpe.2024.06.0263

Abstract(16) HTML (9) PDF(1)

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Aiming at the compatibility problem of high-temperature flowing liquid metal on structural materials, especially the corrosion problem, in the first wall of liquid lithium and liquid metal blanket components in the nuclear fusion reactor, a high-temperature flowing liquid metal corrosion experimental device is designed, and three-dimensional numerical simulation and analysis of the flow and heat transfer characteristics of the liquid metal are carried out by using the software ANSYS. The simulation and test results show that the experimental device can realize the conditions of liquid lithium temperature (300-600℃) and flow rate (< 0.2 m/s) in the first wall and blanket structure, and is qualified to study the corrosion characteristics of dynamic liquid lithium and structural materials at high temperature. Meanwhile, the corrosion behavior of domestically produced low-activation ferrite/martensite steel (9Cr-0.4Mo-0.3Y steel) in 0.2 m/s liquid Li at 550℃ for 1000 hours (h) is preliminarily studied. The results show that 9Cr-0.4Mo-0.3Y steel experiences obvious intergranular corrosion and pitting corrosion, and the surface hardness of the sample is reduced to different degrees due to non-uniform corrosion. The XRD analysis reveals that there is no phase transformation on the corroded surface of 9Cr-0.4Mo-0.3Y steel. The 03-1049#FeNi peak is detected on the sample' surface due to the dissolution and migration of Ni element from the 304 stainless steel vessel.

Study on Corrosion Behavior of High-corrosion Resistance AFAs in Supercritical Water

Gao Yang, Guo Xianglong, Jiang Yufan, Wu Jianwen, Tang Rui, Huang Yanping, Zhang Lefu

2024, 45(6): 271-279. doi: 10.13832/j.jnpe.2024.06.0271

Abstract(12) HTML (2) PDF(1)

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To address the inapplicability of traditional stainless steel in the high-temperature, high-pressure and highly corrosive service conditions of supercritical-water cooled reactor core, a novel alumina-forming austenitic stainless steels (AFAs) was designed and prepared, and its corrosion behavior in supercritical water at 600℃/25 MPa was studied by autoclave immersion test. The morphology, composition and structural characteristics of the oxide scale on the surface of AFAs after corrosion in supercritical water were studied by using a variety of advanced micro-analysis techniques to explore the corrosion resistance mechanism of the alloy in supercritical water and the scale-forming behavior of alumina. The results show that the AFAs offered excellent corrosion resistance by forming a continuous alumina scale in supercritical-water at 600℃, with a corrosion weight gain of less than 10 mg/dm² after 1000 h of exposure, which is better than that of C276 and 310-ODS alloys corroded under the same conditions reported in the literature. The alumina scale is densely and tightly combined with the alloy matrix without obvious separations, effectively hindering the direct contact between the oxidizing media and the alloy matrix, and inhibiting the external diffusion of Fe in the alloy, thus providing excellent protection for the alloy. However, the Laves in the AFAs can affect the uniformity of the localized alumina scale, leading to the formation of MnCr₂O₄ particles. Therefore, AFAs need to maintain continuous formation of alumina scale while strictly controlling the Laves phase level to satisfy the requirements of supercritical water-cooled reactor applications. The results of this paper can provide data and theoretical support for the development and design of aluminum-forming austenitic stainless steels for supercritical water-cooled reactors.

Optimization of Oxygen Control Strategy for Corrosion Mitigation in Lead-Bismuth Cooled Fast Reactors

Chen Jiajie, Wang Shiwei, He Hui, Liu Xiaojing, Xiong Jinbiao

2024, 45(6): 280-289. doi: 10.13832/j.jnpe.2024.06.0280

Abstract(15) HTML (4) PDF(1)

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To obtain the optimal oxygen content control strategy for mitigating corrosion of fuel cladding in lead-bismuth fast reactors (LBFRs), this study constructed a T91 oxidation/corrosion model to analyze the evolution of the fuel element cladding interface. On this basis, taking the thickness of oxide layer as the constraint condition of the optimization problem, the whale optimization algorithm (WOA) was used to optimize the oxygen content control strategy, and the "low-medium-high-low" cyclic fluctuation oxygen content control strategy was obtained. Furthermore, this study simulated and compared the distribution of the oxide layer on the surface of fuel elements under fixed oxygen-dominated condition and optimized oxygen control strategy. The results indicated that under the optimized oxygen control strategy, the fuel element cladding did not trigger dissolution corrosion, and the overall thickness of the surface oxide layer was significantly reduced compared to the fixed oxygen-dominated condition, with an average thickness reduction of 95.6% for the magnetite layer and 44.2% for the spinel layer. The optimal oxygen content control strategy constructed in this paper can provide a reference for mitigating corrosion of cladding in lead-bismuth fast reactors.

A Preliminary Study on Aerosol Transport Characteristics in Containment Based on DPM Method

Wang Yuqing, Weng Yanyun, Ni Muyi, Deng Lilin, Tan Yi, Zhang Minghao

2024, 45(6): 290-296. doi: 10.13832/j.jnpe.2024.06.0290

Abstract(22) HTML (3) PDF(4)

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Aerosols are the main leakage pathway of radioactive materials in nuclear power plants during normal operation and severe accidents. In this thesis, a numerical simulation study is carried out based on the Discrete Phase Model (DPM) of the Euler-Lagrange method in FLUENT for the aerosol transport characteristics in the containment of a lead-based reactor. The simulation results show that the number of aerosol small-size particles in the static flow field is uniformly distributed by Brownian force on different walls, and the number of large-size particles is unevenly distributed by gravity on different walls. Meanwhile, the applicability and accuracy of the modeling approach in the containment environment are verified by using the settling phase of the PHEBUS FPT0 experiment as a benchmark condition. Finally, based on a two-dimensional axisymmetric simplified model of a typical containment, the aerosol migration and settling processes from the core surface to the containment are simulated under the normal operating conditions of the reactor. It is found that under normal operating conditions, 0.1 μm aerosols will migrate widely with the flow line and be captured near the junction of the upper and vertical containment walls, while 3 μm and 10 μm aerosols will be retained near the lower containment wall or settle to the lower containment wall. The preliminary conclusions based on this modeling study can provide reference for the subsequent aerosol force analysis experiments and aerosol migration experiments in containment.

Current Issue 2024 Vol. 45, No. 6

Current Issue
2024 Vol. 45, No. 6