核动力工程

PLIF Technology and Its Application in Researches of Nuclear Reactor Thermal-hydraulics

Tan Sichao, Wei Tianyi, Yu Xiaoyong, Xie Guanhui, Li Xupeng, Tian Ruifeng

2024, 45(1): 1-10. doi: 10.13832/j.jnpe.2024.01.0001

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Abstract:
The Nuclear Reactor Engineering Laboratory of Harbin Engineering University (HEU-NUREL) has long been committed to the exploration of planar laser-induced fluorescence (PLIF) technology and its application in reactor thermal hydraulic study. As a non-invasive advanced measurement method, PLIF technology can achieve qualitative and quantitative measurements of the full-plane of the physical field, providing benchmark experimental data for numerical model validation. This paper will comprehensively display the latest achievements and progress of HEU-NUREL in thermal hydraulic research of nuclear reactors based on PLIF technology, focus on the practical path and measurement effect of PLIF technology in concentration measurement, temperature field analysis, two-phase distribution, and technical exploration, and describe the the technical features and unique contributions of PLIF technology in different fields of thermal and hydraulic study, aiming to promote PLIF technology to better serve the reactor system design and safety analysis.

Research and Engineering Application of Xenon Dynamic-Flux Measurement Method for M310 Unit Based on Bamboo-C Code

Fang He, Wei Luo, Wan Chenghui, Gao Yiyuan, Li Zaipeng

2024, 45(1): 11-18. doi: 10.13832/j.jnpe.2024.01.0011

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Abstract:
In order to optimize the current technical requirements of M310 unit, i.e. to wait for 24 h after the power increases to the specified power level to perform in-core flux measurement test, so as to further improve the economy of PWR nuclear power plant, a new method for core flux measurement under dynamic xenon conditions is developed based on Bamboo-C, an advanced PWR core physical analysis software independently developed by Xi'an Jiaotong University. Based on the dynamic xenon flux measurement data, the measured values of core power distribution under the condition of balanced xenon were predicted, and the engineering application was verified in the overhaul of Tianwan No.5 unit. The verification results show that the dynamic xenon flux measurement method proposed in this paper can shorten the waiting time of xenon balance to 2 h at a specified power level, and has high prediction accuracy for key physical quantities in the core.

Validation and Verification of COSINE Code Based on Rod Bundle Heat Transfer Experiment

Duan Bingqi, Zhao Meng, Zhang Hao, Chai Xiang, Yang Yanhua

2024, 45(1): 19-26. doi: 10.13832/j.jnpe.2024.01.0019

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Abstract:
To further improve the stability and accuracy of software calculation, and to confirm, evaluate and improve the important model in the software COSINE, the system safety analysis code cosFlow in the COSINE thermal-hydraulic software package was used to model and calculate the thermal-hydraulic physical process during the core reflooding stage of large-break loss-of-coolant accident (LOCA) in a nuclear power plant. The calculation modeling was based on the Rod Bundle Heat Transfer (RBHT) experiment, and the results of the experiment were used to examine the system safety analysis code. The calculation results show that the change trend of the wall temperature of the rod bundle is basically consistent with the experimental data, indicating that cosFlow can accurately analyze the progress of quench front in the large-break LOCA. However, the progress speed of the quench front is faster in the early stage and slower in the later stage compared to the experimental results of RBHT. It is speculated that this discrepancy may be attributed to the significant axial temperature gradient in the heating rod as well as a lack of axial heat conduction module in the original code. Therefore, the future code development and research will focus on improving the thermal-hydraulic transfer model of the progress of quench front.

Scaling Analysis on Core Makeup System of Small Reactor under Marine Conditions

Tang Jilin, Liu Yusheng, Tan Sichao, Li Dongyang, Wang Shuguang, Qiu Liqing

2024, 45(1): 27-33. doi: 10.13832/j.jnpe.2024.01.0027

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Abstract:
In order to solve the test verification problem of passive core makeup system (PCMS) design under marine conditions, this study takes the branch of core makeup tank (CMT) as an example to identify the thermal hydraulic phenomena of PCMS under marine conditions. Based on the hierarchical two-tiered scaling (H2TS) analysis method, scaling analysis is conducted for the identified key thermal hydraulic phenomena, and the similarity criteria for scaled PCMS design are obtained. The results show that there are various complex thermal hydraulic phenomena and their coupling processes in PCMS, among which the thermal hydraulic phenomena of CMT branch are the most representative, and its natural circulation process is most significantly affected by ocean conditions. The ocean conditions lead to additional inertia force on the fluid in PCMS system, and the consistent additional acceleration is a necessary condition for the experimental model to reproduce the influence of ocean conditions. The scaling design of core makeup system under ocean conditions should follow similarity criteria such as resistance number, Richardson number, condensation number and acceleration. The test model designed based on these similarity criteria can reproduce the key thermal hydraulic phenomena and their coupling effects in the prototype with reasonable distortion level.

Research on Correction Method of Pressure Spike at Gas-Liquid Interface Crossing Cells Based on RELAP5

Wang Xiaohu, Zhao Pinghui, Ye Taohong, Xiong Yan, Tan Chao, Chen Yunlong

2024, 45(1): 34-40. doi: 10.13832/j.jnpe.2024.01.0034

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Abstract:
In order to reduce or eliminate the non-physical pressure spike of gas-liquid interface crossing cell edges in vertical stratified flow calculation of RELAP5 and to improve the stability of the calculation, this study investigates the causes of pressure spikes in the vertical stratified flow from the perspective of momentum equations, and proposes a method to reduce the pressure spikes by directly correcting the momentum equations. It is found that the liquid phase unsteady term is the main cause of such pressure spikes. Based on this, a numerical method for correcting the liquid phase unsteady term is developed and validated with a vertical pipe filling and discharging problem and a manometer problem. The validation results show that the correction method can achieve the effect of reducing the non-physical pressure spikes in the process of the gas-water interface crossing cell edges, and therefore, it is beneficial to improve the stability of the code calculation.

Peak Temperature Prediction Method Based on A Theoretical Model of Equivalent Thermal Conductivity for Fully Ceramic Microencapsulated Fuel

Wang Mouhao, Bu Shanshan, Zhou Bing, Li Zhenzhong, Chen Deqi

2024, 45(1): 41-48. doi: 10.13832/j.jnpe.2024.01.0041

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Abstract:
In order to satisfy the engineering need for rapid prediction of peak temperatures of Fully Ceramic Microencapsulated (FCM) fuels, inverse calculations of peak fuel temperatures could be performed by substituting the equivalent thermal conductivity of the fuel into a simplified homogeneous thermal conductivity model. In this paper, based on the multi-annulus model developed in previous study, a theoretical model for calculating the equivalent thermal conductivity of FCM fuel is derived by starting from the basic thermal conductivity equation and taking the peak fuel temperature as a conserved quantity, and the multi-annulus model is further equivalent to a homogeneous model. Then the derived model is compared with some conventional equivalent thermal conductivity theoretical models. The results show that the developed theoretical model could be combined with the homogeneous model to effectively achieve the prediction of peak fuel temperatures. The developed theoretical method is suitable for predicting the peak temperature of FCM fuel elements with internal heat sources, because the deviation between the predicted peak temperature and the actual value is basically within 3%. The developed theoretical model prediction method based on equivalent thermal conductivity could realize the rapid prediction of the peak temperature of FCM fuel.

Numerical Study of Gas-Lead-Bismuth-Alloy Two-Phase Flow Characteristics in Vertical Circular Tube

Liu Zihua, Wang Di, Liang Ren, Cai Dechang, Ouyang Yong, Lin Zhikang, Tan Linhao

2024, 45(1): 49-54. doi: 10.13832/j.jnpe.2024.01.0049

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Abstract:
This study adopts a drift flux model based on force balance theory to predict the void fraction of gas-liquid lead-bismuth two-phase flow in the vertical circular pipe. The distribution of liquid-phase flow velocity, shear stress distribution and void fraction distribution of gas-liquid lead-bismuth two-phase flow at different flow path radii, Bankoff exponents and Galileo numbers are obtained through numerical calculations. The intrinsic relationship between the distribution parameters of the drift flux model and the above macroscopic parameters is analysed. The results show that the predicted void fractions and the evolution of the distribution parameters are in good agreement with the experimental results. The numerical method established in this study can be used to study the flow characteristics of gas-liquid lead-bismuth two-phase flow in a circular tube, and provide reference for the rapid calculation of key two-phase flow parameters such as void fraction in steam generator tube rupture(SGTR) accident of lead-bismuth fast reactor.

Study on Flow Distribution and Resistance Characteristics of Secondary Side of Intermediate Heat Exchanger

Song Guangdong, Jiang Lin, Qiu Binbin, Liu Yuning, Xing Shuai

2024, 45(1): 55-59. doi: 10.13832/j.jnpe.2024.01.0055

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Abstract:
Intermediate heat exchanger is an important equipment for heat exchange between liquid-sodium cooled fast reactor core and steam generator. Due to the large number of tube bundles, it is difficult to get accurate flow distribution and resistance characteristic parameters by theoretical calculation. In order to solve the above problems, an intermediate heat exchanger was taken as a prototype, and the flow distribution and resistance characteristics tests were carried out on a 1∶1 model. The flow distribution factors of different flow distribution structures and the resistance coefficients of local structures such as the central descending section of the secondary side and the heat exchange tube section were obtained. The results show that the conical plate and no flow-distribution structure will lead to the high outer flow. The orifice plate structure can reduce the outer flow distribution factor and make the flow distribution more uniform. The resistance coefficients of the central descending section and the heat exchange tube section of the secondary side decrease slightly with the increase of Reynolds number (Re). The resistance coefficient of the secondary side lower head from large to small is conical plate, orifice plate 2, orifice plate 1, no flow distribution structure, and the change of the total resistance coefficient also follows this law. The research results can provide data support for the optimization design of intermediate heat exchanger.

Experimental Study on Dryout Heat Flux of Homogeneous Debris Bed and Stratified Debris Bed

Zou Wenbin, Tong Lili, Cao Xuewu

2024, 45(1): 60-64. doi: 10.13832/j.jnpe.2024.01.0060

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Abstract:
The coolability of debris bed is a concern for reactor severe accident mitigation strategies, and the structure of debris bed has a significant impact on its coolability. In this paper, an experimental facility for the coolability of volumetrically heated debris bed was established, and the experimental study on the coolability of homogeneous, axial stratified, and radial stratified debris beds was carried out to compare and analyze the dryout phenomenon and heat flow flux for different debris bed types. The results show that under the condition of homogeneous debris bed, the larger gas-liquid flow resistance of small particles makes the permeability decrease, which leads to the smaller DHF of small particle homogeneous debris bed; under the condition of axial stratified debris bed, the DHF of axial stratified debris bed is much smaller than that of homogeneous debris bed composed of small particles at the top due to the large resistance caused by the decrease of porosity at the stratified interface; under the condition of radial stratified debris bed, due to the large gas-liquid flow resistance of small particle layer, steam will migrate and accumulate to large particle layer, resulting in DHF lower than that of the homogeneous bed composed of large particles.

Verification and Analysis of Fine Subchannel Rod Bowing Model

Zhou Shan, Jiang Li, Shan Jianqiang, Guo Junliang

2024, 45(1): 65-71. doi: 10.13832/j.jnpe.2024.01.0065

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Abstract:
In order to improve the prediction ability of the sub-channel code for the local parameter changes under rod-bowing condition, a fined sub-channel rod-bowing model was developed in this study. The model, based on a detailed division of the sub-channels, allows a comprehensive analysis of rod bowing. The prediction ability of the fined sub-channel rod-bowing model on the axial flow and cross flow of the sub-channel was validated by computational fluid dynamics (CFD). Based on the validity of the model by CFD, the analysis of the trends of axial and transverse flows in the rod-bowing section was conducted. The results show that the predictions of the changing trends of axial flow and cross flow by ATHAS and CFD are basically consistent, and the fined sub-channel rod bowing model can better predict the changing trend of axial flow and cross flow caused by rod bowing. Therefore, the fined sub-channel rod bowing model can predict the influence of bowing rod on local flow field, which provides a basis for the prediction of critical heat flux.

Development and Application of Core pin-by-pin Calculation of Sub-channel Analysis Code LINDEN

Xia Hang, Xu Rongshuan, Li Jinggang, Wu Yingwei, Wang Ting, Zhu Yuanbing, Wang Ke

2024, 45(1): 72-78. doi: 10.13832/j.jnpe.2024.01.0072

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Abstract:
To realize the pin by pin calculation function of the sub-channel analysis code LINDEN and improve the accuracy of core thermal hydraulic analysis by LINDEN code, an automatic modeling function has been developed for the full core pin-by-pin calculation, and the storage and solution problems of large sparse matrices have been solved. The full core pin-by-pin calculation and analysis of 157 assemblies were carried out. The calculation results indicate that LINDEN can perform full core pin-by-pin calculations, and the detailed distribution results of key thermal hydraulic parameters such as coolant temperature distribution, DNBR, mass flow rate distribution can be obtained. The power difference between adjacent assemblies affects the distribution of thermal hydraulic parameters within the assembly.

Study on the Passive Residual Heat Removal System with Heat Pipe in Pool Type Heating Reactor

Lyv Jun

2024, 45(1): 79-83. doi: 10.13832/j.jnpe.2024.01.0079

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Abstract:
In order to realize the passive safety design of pool type heating reactor, further improve its environmental friendliness, and meet the requirements of nuclear heating facilities close to public construction, this study introduces the experimental verification of passive residual heat removal by using low-temperature heat pipe according to the operating characteristics of pool type heating reactor at low temperature and normal pressure, and analyzes the key test parameters and gives some suggestions for selection. The research results show that the low-temperature heat pipe can realize the passive export of heat from the pool of pool type heating reactor to the ambient atmosphere, and provide important support for engineering design.

Research on High Temperature Oxidation Behavior of Zirconium Alloy for Fuel Element Based on MOOSE Platform

Wu Zhouzhi, Zhang Kun, Wang Yanpei, Yu Hongxing, Zhang Lin, He Liang, Tang Changbing

2024, 45(1): 84-89. doi: 10.13832/j.jnpe.2024.01.0084

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Abstract:
In order to establish a prediction method for the high-temperature oxidation behavior of the new N36 zirconium alloy and allow the autonomous fuel element performance analysis code FORWARD to be applied to the loss of coolant accident (LOCA) condition, the high-temperature steam oxidation test of the new N36 zirconium alloy was carried out in this study. The high-temperature oxidation model of N36 zirconium alloy was developed and validated, and the high-temperature oxidation behavior of N36 zirconium alloy under LOCA condition was predicted using the FORWARD code. The results show that the predicted oxidation weight gain of N36 zirconium alloy is in good agreement with the verification test results, and the predicted oxidation behavior of N36 zirconium alloy at high temperature under LOCA condition is reasonable. Therefore, the model and fuel element performance analysis code developed in this study can be used to predict the high temperature oxidation behavior of the new N36 zirconium alloy.

Thermal-shock Properties and Anticorrosion Behavior in Static LBE of Al₂O₃-TiO₂/FeCrAl Coating by Multi-Arc Ion Plating

Zhang Shunlin, Pan Dong, Yin Xing, Chen Yong, Zhao Haibo, Sun Lan, Wang Jun

2024, 45(1): 90-97. doi: 10.13832/j.jnpe.2024.01.0090

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Abstract:
This work aims to explore a method for preparing the surface coating of FeCrAl alloy, a cladding material in nuclear power industry. The Al₂O₃-TiO₂ coating with FeCrAl as the intermediate layer was prepared on FeCrAl alloy by multi-arc ion plating. The thermal-shock test was carried out to explore the thermal-shock resistance of the coating. The corrosion resistance of the coating was studied after the static lead-bismuth eutectic (LBE) corrosion test at 600℃ for 1000 h. The phase composition and micromorphologies of substrate and coating samples before and after LBE corrosion were characterized. The results showed that the Al₂O₃-TiO₂ prepared by multi-arc ion plating was amorphous. After 30 thermal shock tests, the coating did not crack or fall off. After the corrosion, the surface of FeCrAl substrate showed obvious dissolution corrosion. GIXRD results showed that Al₂O₃ crystallization occurred in the coating samples after corrosion. The Al₂O₃ structure on the surface shrank and pores appeared, while the inner layer of the coating remained dense. The sectional analysis showed that LBE did not penetrate into the coating. Therefore, the Al₂O₃-TiO₂/FeCrAl coating can effectively protect the substrate from LBE corrosion.

Research on Extraction of Mo from Simulated Low Enriched Uranium Fuel Solution of Medical Isotope Reactor Based on Spherical Alumina

Wang Haijun, Sun Zhizhong, Zhang Jinsong, Chen Yunming, Luo Ning, Wu Jianrong, Geng Zisheng

2024, 45(1): 98-105. doi: 10.13832/j.jnpe.2024.01.0098

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Abstract:
In order to improve the situation that alumina column is easily blocked during the extraction of ⁹⁹Mo in Medical Isotope Test Reactor and further improve the adsorption effect of ⁹⁹Mo under low enriched uranium condition, the synthesis of spherical alumina was studied. Alumina microspheres were prepared by sol-gel-oil column molding, and their adsorption behavior and dynamic adsorption conditions were investigated. The study on the extraction of Mo by spherical alumina in simulated low enriched uranium fuel solution was carried out. The results show that the prepared spherical alumina has larger particle size and specific surface area, which can effectively alleviate the column plugging problem and improve the adsorption capacity of Mo. The adsorption process of Mo was consistent with the quasi-second-order kinetic model and the Freundlich adsorption isotherm model. In the simulated solution of low enriched uranium fuel, the recovery rate of spherical alumina to Mo is 87.4%, and the impurities also meet the requirements of pharmacopoeia. Therefore, the prepared spherical alumina is expected to be applied to the extraction process of ⁹⁹Mo in medical isotope test reactor.

Design and Experimental Study of Dynamic Vibration Absorber for Small Branch Pipe of Complex Pipeline of Nuclear Power Unit

Ren Zhiying, Ren Jiaxin, Liu Tianyan, Li Zhen, He Mingyuan, Liang Shengtao

2024, 45(1): 106-114. doi: 10.13832/j.jnpe.2024.01.0106

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Abstract:
The long-term irregular vibration of small branches in the complex pipelines of nuclear power units can lead to fatigue fracture and damage, causing leaks and even nuclear safety accidents. Currently, reinforcement or vibration reduction methods are commonly used, but the effectiveness is limited, and reducing the vibration peak at certain frequencies is challenging. To address these issues, this paper proposes the design of a structurally simple, easy-to-install, and adjustable-quality dynamic absorber. By simulating actual vibration conditions, the dynamic modeling and analysis of small branches with and without dynamic absorbers are conducted. Using the acceleration transfer rate as a reference, the relevant parameters of the dynamic absorber are designed to obtain the optimal mass, stiffness, and damping of the dynamic absorber. Finally, the physical prototype is produced and experimentally verified. The results show that the designed dynamic absorber in this paper reduces the three-axis vibration of small branches at resonance frequencies by more than 60% (exceeding the engineering requirement of 5 dB). Fine-tuning the number of mass blocks expands the adjustable range of the absorption frequency band, meeting the requirements of different operating conditions. Adding metal rubber material further enhances the absorption effect of the dynamic absorber, improving the damping performance by more than 10% compared to the structure without damping. This research provides an effective method for vibration reduction in the small branches of complex pipelines in nuclear power units.

Study on Tensile Behavior of Small Size Specimen of A508-III Steel Based on Finite Element Aided Testing Method

Li Yihan, Li Shuai, Li Junwan, Xin Shengmin, Ning Guangsheng, Zhong Weihua, Yang Wen

2024, 45(1): 115-122. doi: 10.13832/j.jnpe.2024.01.0115

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Abstract:
In order to overcome the influence of size effect brought about by the reduction in the size of tensile specimen, the tensile properties have been investigated for small size specimens of A508-III steel in this paper. The small size specimens of A508-III steel with varying grain sizes and specimen thicknesses were prepared by altering the heat treatment temperature, and their tensile tests were conducted at room temperature. The effects of grain size and thickness on the tensile mechanical properties of small size specimens were analyzed and the underlying mechanism of the potential tensile size effect was revealed. The results indicate that, due to the changes in the thickness and grain size of small size specimens, the tensile mechanical property exhibits an obvious size effect. A Swift mechanical constitutive model considering size effects was developed by introducing influence parameters that comprehensively characterize the specimen feature size and grain size effects. The ductile damage evolution parameters of the specimens were determined using the finite element aided testing (FAT) method, and the error between numerical and experimental results was less than 3%. Based on the established mechanical constitutive model, the tensile mechanical properties of the small size specimens were predicted using the finite element method, and the normalized models of their yield strength and tensile strength were constructed, so as to provide reference for the engineering application of small size specimens.

Analysis and Optimization of Key Factors Affecting the Expansion Quality of Threaded Sleeve and Guide Pipe

He Fuchun, Fu Chunming, Tang Dewen, Huo Shaoyong, Chen Kai, Wang Shirui, Zong Benyang

2024, 45(1): 123-129. doi: 10.13832/j.jnpe.2024.01.0123

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Abstract:
This paper aims to explore the effects of the loading distance of the expansion rod and the friction coefficient between two pipes on the expansion quality between the threaded sleeve and the guide pipe in the support frame of nuclear fuel assembly. Based on the established equal scale numerical model of the threaded sleeve and guide pipe and the designed two-factor four-level orthogonal test, the finite element analysis method is used to numerically analyze different test combinations under given conditions. The test results show that the loading distance of the expansion rod and the friction coefficient between two pipes have a significant influence on the quality of the expansion joint. In order to improve the optimization efficiency, combined with Kriging model and particle swarm optimization algorithm, with the maximum thinning rate ≤10% as the constraint, the optimal solution is carried out within the given loading distance of expansion rod and the friction coefficient between two pipes, and the obtained loading distance of expansion rod is 28.42 mm and the friction coefficient is 0.14. The research in this paper provides a theoretical reference for improving the quality of expansion joint between the threaded sleeve and the guide pipe and improving the expansion joint process.

Research on High-fidelity Dynamic Model Updating Technology of Nuclear Power Valve Piping System

Xue Ruiyuan, Zhang Yongnan, Zhang Xiheng, Yu Shurong, Meng Xiaoqiao, Yu Jianping

2024, 45(1): 130-138. doi: 10.13832/j.jnpe.2024.01.0130

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Abstract:
The high-fidelity dynamic model is the basis for seismic margin analysis, seismic performance monitoring and continuous improvement of seismic design of in-service nuclear power valve piping system. However, its establishment process is faced with the problems of great difficulty and low efficiency. Based on the high-fidelity dynamic model of a practical valve piping system, the anti-resonance frequency characteristics of various piping systems are studied. The results show that the origin anti-resonance frequency and the resonance frequency appear alternately. In the low frequency region, the theoretical values of the origin anti-resonance frequency of various piping systems match the measured values well. At the connection position between the valve and the pipeline, different structural parameters have close sensitivity to the origin anti-resonance frequency and resonance frequency, so the measuring points should be preferentially arranged at these positions in practical engineering. Based on this theory, a finite element model updating method combining Gaussian Radial Basis Function-Response Surface Method and Improved Adaptive Genetic Algorithm is proposed with resonance frequency and anti-resonance frequency as the objective function. The example shows that the proposed method overcomes the difficulties of insufficient test data and slow convergence speed, and can identify the unknown structural parameters in nuclear valve piping with high precision and high efficiency, as well as establishes its high-fidelity dynamic model. The research results provide the possibility for further improvement of the economy and safety of seismic design of nuclear power valve piping system.

Dynamic Characteristics Analysis and Steady-State Operation Scheme Design of Gas-Cooled Microreactor

Qiu Leilei, Fan Sui, Wei Xinyu, Liao Shengyong, Sun Peiwei

2024, 45(1): 139-144. doi: 10.13832/j.jnpe.2024.01.0139

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Abstract:
The gas-cooled microreactor is a highly integrated system, in which the equipment is highly coupled and mutually constrained. The dynamic simulation model of a 5 MW gas-cooled microreactor is established through theoretical derivation, and the dynamic characteristics of the system are analyzed. On this basis, the steady-state operation scheme design method of the gas-cooled microreactor is proposed. Compared with the design value, the relative error of steady-state calculation results is less than 1%. The transient calculation results of the model are consistent with the experimental results. The maximum error of flow is less than 0.53 kg/s, and the maximum error of pressure is less than 0.027 MPa. The steady-state operation scheme is as follows: at the load level of 50%FP~100%FP (full power), the helium temperature at the reactor outlet and the rotor speed remain unchanged as the load decreases; at the load level of 0%FP~50%FP, the reactor outlet helium temperature and rotor speed linearly decrease with the decrease of the load; the helium temperature at the outlet of precooler and intercooler is constant at 0%FP~100%FP load level. The proposed steady-state operation scheme ensures the safe, stable and efficient operation of the gas-cooled microreactor under a wide range of variable operating conditions.

A Method for Judgment of Inclusion Failure of Reactor Compartment Based on Radiation Monitoring

Lin Xiaoling, Wang Lin

2024, 45(1): 145-148. doi: 10.13832/j.jnpe.2024.01.0145

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The inclusion failure of the reactor compartment is an important technical condition for judging the upgrade of emergency state and an important technical basis for emergency decision. In this paper, a method using compartment radiation level as the basis for judgment of the inclusion failure of reactor compartment is proposed. The monitorable quantities (¹³¹I nuclide activity concentration and γ dose rate of compartment) used to judge inclusion failure are determined, the transfer relationship between compartment radiation level and reactor leakage rate is established, and the calculation method of criterion value is given.

Study on Response and Consequence of Loss of Coolant Accident in Integrated Small Reactor

Cai Wei, Yue Zhidong, Wei Ting

2024, 45(1): 149-155. doi: 10.13832/j.jnpe.2024.01.0149

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Abstract:
The overall analytic model including the core, primary loop and containment was established base on RELAP5 code to fully analyze the Loss of Coolant Accident (LOCA) for the integrated small modular reactor (or small reactor for short). The transient response of the core and containment as well as the radiological consequence for the environment were calculated and analyzed. The results showed that the accident consequences met the acceptance criteria stipulated in the small reactor safety review principle. Besides, the improved scheme of close-fitting containment and the optional parameter configuration were proposed. The calculation results show that with the improved scheme, the pressure balance of the primary loop and containment can be achieved very shortly and the break flow is terminated earlier, so the loss of primary coolant and the radiological release are reduced. The core safety is enhanced, and the dose consequence of the accident is decreased. Meanwhile, the long-term cooling of containment can be ensured by the pool outside the containment. The research results can provide references for the engineering application and containment design of the integrated small reactor.

Eddy Current Effect Analysis of a New Self-sensing Rod Position Detector for Reactor

Zhang Yixuan, Xu Qiwei, Tang Jiankai, Liu Yanting, Huang Siyu, Luo Lingyan

2024, 45(1): 156-163. doi: 10.13832/j.jnpe.2024.01.0156

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Abstract:
Currently, coded rod position detectors have many problems such as a large number of coils, complex structure, rough measurement accuracy and poor reliability, which seriously hinder the development of reactor miniaturization. Therefore, we propose a new self-sensing rod position detector in this paper, which employs a double-winding structure with four equal-length A and B coils, where the two sets of coils are independent of each other, and the linear variation of the coil inductance with the motion of the driving rod enables continuous rod position measurement. By quantifying the magnetic skin effect on the rod, the mathematical model of self-sensing rod position detector based on eddy current effect is derived. The finite element simulation model is built to verify the accuracy of the mathematical model, and the influence of the temperature rise of the rod and the key structure parameters of the detection coil on the measurement accuracy of the detector is studied. It is found that the inductance variation of the coil under different temperatures is influenced by the relative permeability and conductivity of the rod. The increase of coil turns is conducive to the increase of inductance variation, and the increase of coil spacing first increases and then decreases the inductance sensitivity. And then the detector structure is optimized considering the above change rules. The prototype is tested and verified. The results show that both coils have an inductance resolution of 0.14 mH/10 mm, and the displacement identification accuracy of 10 mm can be achieved. This study can provide reference for the application of self-sensing rod position detectors in small nuclear reactors.

Time-varying Reliability Evaluation Method of Steam Generator Heat Transfer Tubes Considering Fretting Wear

Xue Yingcheng, Wu Zonghui, He Jian

2024, 45(1): 164-170. doi: 10.13832/j.jnpe.2024.01.0164

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Abstract:
In order to investigate the impact of fretting wear on the reliability of steam generator heat transfer tubes under impact, an evaluation method for the safety performance of steam generator heat transfer tubes was established. By fitting the distribution form of the wear coefficient of the heat transfer tube, a time-varying wear model of the heat transfer tube was established. Using the multiplier dimensionality reduction method to obtain the fractional moment of the limit state of the heat transfer tube, and using the NM (Nelder Mead) simplex algorithm to optimize the maximum entropy parameter to calculate the failure probability of the heat transfer tube. Based on the time discrete method, the time-varying reliability of steam generator heat transfer tubes under impact and fretting wear is studied. The results indicate that when the reliability index greater than 2 is taken as the acceptable standard of structural reliability, the time-varying reliability of the steam generator heat transfer tubes subjected to impact cannot meet the requirements in the 10th year of wear.

Research on Control Strategy and Energy Storage Capacity Allocation under NAC and Frequency Regulation

Qian Hong, Wang Guoping, Li Baolong

2024, 45(1): 171-177. doi: 10.13832/j.jnpe.2024.01.0171

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For the participation of nuclear power in primary frequency regulation, this study proposes a frequency regulation mode with integration of nuclear power and chemical storage, and the control strategy of frequency regulation with integrated nuclear power and chemical storage is designed. Through the analysis of battery life, combined with the big data analysis of annual power grid frequency, a solution model aiming at the lowest energy storage cost is established, and the optimal energy storage capacity is obtained by using the improved particle swarm optimization algorithm. Based on the frequency statistics of a power grid for one year, the actual operating parameters of the nuclear power plant are taken as the example for calculation and verification. The results show that the proposed scheme cannot only enable the nuclear power plant for primary frequency regulation, but also greatly reduce the number of unit load changes. Therefore, this study can ensure the safety of unit operation while realizing the participation of nuclear power plants in primary frequency regulation.

Development of Test Platform for LBE Aerosol Kinetics and Preliminary Parameter Measurement

Wang Yuqing, Deng Lilin, Ni Muyi, Wu Jiewei, Tan Yi, Jing Futing, Xia Mingming, Tian Chao

2024, 45(1): 178-185. doi: 10.13832/j.jnpe.2024.01.0178

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Abstract:
In Lead-Bismuth Fast Reactors, the primary coolant liquid lead-bismuth eutectic (LBE) is subject to neutron irradiation, which generates the radioactive nuclide polonium (²¹⁰Po). Given the volatility of ²¹⁰Po, it is essential to thoroughly investigate its migration and diffusion behavior. Aerosols are the primary carriers of volatile radioactive nuclides. Drawing on domestic and international experience in the design and operation of reactor aerosol experimental platforms, this paper presents the development of an LBE aerosol kinetics experimental platform. By combining scanning electrical mobility and optical scattering methods, a broad-spectrum measurement of LBE aerosol particle count and size distribution was achieved. The results indicate that the particle size distribution of LBE aerosols is mainly at the nano-level. Preliminary LBE aerosol kinetics analysis was conducted through data processing of the measurement results, thereby providing key parameters for subsequent safety analysis of radioactive aerosols in Lead-Bismuth Fast Reactors.

The pH and Iodine Diffusion Model for IRWST after LOCA Accident of HPR1000

Wang Chengyu, Lu Changdong, Guo Shaoqiang, Chen Yichen, Zhou Wentao, Jiang Pingting

2024, 45(1): 186-193. doi: 10.13832/j.jnpe.2024.01.0186

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Abstract:
A passive pH adjustment basket is set in the HPR1000 pit to control the pH of the in-containment refueling water storage tank (IRWST) after the loss of coolant accident (LOCA) by adding alkaline additives, so as to reduce the concentration of gaseous iodine in the containment. Predicting the pH and iodine concentration after an accident is critical for accident source term and radioactive analysis. In this paper, aiming at the recirculation water flow after LOCA, combined with the gas-liquid distribution of iodine and the two-film theory and the relationship between iodine form and pH, a macro transient model is established to realize the calculation of IRWST transient pH, substance concentration and gas-liquid two-phase iodine concentration in containment after the accident. By comparing with the results of Visual MINTEQ software, the pH calculation part of the model is verified. The condition parameters are substituted into the model to analyze the influencing factors. The results can correctly show the relationship between pH and iodine concentration, which proves that the model has the ability to predict pH and iodine concentration after accident.

Research on Static Characteristics of High Sensitivity Built-in Capacitance Control Rod Position Sensor

Li Yanlin, Qin Benke, Bo Hanliang

2024, 45(1): 194-200. doi: 10.13832/j.jnpe.2024.01.0194

Abstract(150) HTML (43) PDF(28)

Abstract:
The built-in capacitance control rod position sensor can be applied to the control rod hydraulic drive system (CRHDS). Aiming at the four-helix electrode capacitance control rod position sensor, the static characteristics of control rod position measurement by the method of double capacitance values are studied in this paper. Through the principle experiment of the sensor, the calculation model of the sensor is established, and the numerical simulation results are verified by experiments. Assuming that the measured rod is eccentric as a whole, the functional relationship between two groups of capacitance values and rod position of control rod is established by numerical simulation method. The error analysis of the function is conducted, and the applicability of the function to different deflection directions and tilt states of the measured rod is analyzed. The results show that according to the rod position measurement method of double capacitance values, the maximum measuring error of rod positions can be limited within ± 6 mm. The sensitivity of the sensor can be improved to the order of 0.05 pF/mm, which can meet the actual requirements of control rod position measurement.

Numerical Analysis of Transient Process of HPR1000 Reactor Coolant Pump Shaft Jamming Accident Condition

Pan Jun, Li Yibin, Qu Zehui, Guo Yanlei, Yang Congxin, Wang Xiuyong

2024, 45(1): 201-209. doi: 10.13832/j.jnpe.2024.01.0201

Abstract(119) HTML (31) PDF(38)

Abstract:
In order to reveal the pipeline transient mechanism under the shaft jamming accident condition of the reactor coolant pump (RCP), a simplified fluid domain model of the coolant system of the three-loop reactor was established by matching the relationship between the resistance characteristics of the reactor coolant pump and the pipeline of the primary system. Based on the computational fluid dynamics (CFD) method, the actual transient flow process and the real-time change rule of parameters in the reactor coolant system under shaft jamming accident condition were reproduced, and the accident safety evaluation method of reactor coolant system under shaft jamming accident condition was established. The transient changes of main pipeline pressure, wall load of transition bend and pressure of three typical heat transfer tubes with radius of curvature were analyzed under shaft jamming accident condition. The results show that: in the process of the shaft jamming accident, the flow in the accident loop decreases to 0 m³/h and then increases in reverse, and reverse flow occurs. The pressure and wall load of the accident loop and other loops will change dramatically after the shaft jamming accident, and the change degree of the accident loop is greater. The pressure oscillation law of the heat transfer tubes with different curvature radii is basically the same, and the peak pressure of the monitoring point increases gradually along the direction from the inlet to the outlet of each heat transfer tube.

Research on Dynamic Characteristics of PWR Nuclear Power Unit under Variable Power Loads Based on APROS

Wang Yue, Yuan Tianrun, Zhuang Yaping, Zhang Jin, Zhou Yuanyuan, Han Xiaoqu

2024, 45(1): 210-217. doi: 10.13832/j.jnpe.2024.01.0210

Abstract(159) HTML (26) PDF(27)

Abstract:
A dynamic model of the million kilowatt PWR unit including the main equipment of the primary and secondary circuits was built based on the modular modeling method by using the simulation software APROS. The steady-state and dynamic processes were simulated and verified. The dynamic characteristics of the main parameters of the system under the different rates of linear load shedding and different step load shedding were studied. The results show that when the step load reduction is less than or equal to 2% full power (FP), the average temperature (T_avg) fluctuation in the primary circuit is small, which cannot cause the action of control bar; when the step load reduction is greater than 2%FP and less than or equal to 5%FP, the T_avg fluctuation causes the control bar to act but quickly returns to the temperature dead zone. Eventually, the stabilized T_avg ended up higher than the initial temperature. In the process of linear load change, the maximum pressure change of the pressurizer can reach 0.3 MPa. Since the specific volume of coolant is positively related to the temperature, the changing trend of the pressurizer water level is consistent with that of the average temperature of the loop. The purpose of this study is to provide theoretical reference for the flexible operation of PWR nuclear power plant.

Nonlinear Ultrasonic Detection Method of Collinear Wave Mixing for Thermal Damage in High Temperature Pipeline

Jiao Jingpin, Li Zhiqiang, Sun Junjun, Wan Guorong, Li Ji, He Cunfu, Wu Bin

2024, 45(1): 218-224. doi: 10.13832/j.jnpe.2024.01.0218

Abstract(83) HTML (33) PDF(20)

Abstract:
Aiming at the need of safe operation of power plant, the nonlinear ultrasonic detection method of thermal damage of high temperature pipeline by collinear mixing is studied. The excitation conditions of significant mixing effect were determined by frequency sweep experiment, and nonlinear ultrasonic testing experiments were carried out on four Super304H pipes with different thermal damage degrees. The detection signals were analyzed by bispectrum to investigate the distribution of bispectrum of sum frequency component in phase interval, and the phase distribution range of nonlinear response caused by thermal damage was determined. The results show that the nonlinear acoustic coefficients extracted in turn have a good correlation with the thermal damage degree of the specimen, which can be used to characterize the thermal damage of the pipeline. The research work provides a feasible scheme for thermal damage detection of high temperature pipelines in power plants.

Online Maintenance Optimization of HPR1000

Ding Xiaochuan, Li Wenjing, Feng Churan, Yang Xiaoyan

2024, 45(1): 225-229. doi: 10.13832/j.jnpe.2024.01.0225

Abstract(180) HTML (49) PDF(18)

Abstract:
Carrying out online maintenance to reduce the maintenance items during outage is an important means to optimize the outage duration and improve the unit economy. In foreign countries, real-time configuration risk calculation tools have been used to calculate Risk Informed Completion Time (RICT), and online maintenance has been achieved through this tool. In this paper, the feasibility of online maintenance of HPR1000 is demonstrated by the risk-informed method without using real-time risk calculation tools. The research shows that in order to carry out on-line maintenance, it is necessary to permanently adjust the frontstop completion time (FSCT) of one safety system train inoperable from 3 days to 7 days. Thanks to the overall design balance and optimization of HPR1000, the importance of single equipment decreases, and the risk increment caused by this change is limited compared with the risk threshold. After perfecting and applying the real-time configuration risk calculation tool in the future, the flexibility of online maintenance scheme can be improved, and the maintenance period can be further extended to reduce the schedule pressure of maintenance.

Study on Transient Hydraulic Load of Reactor Coolant System under the Condition of Reactor Coolant Pump Rotor Seizure

Cui Huaiming, Tan Xin, Wang Yan, Kuang Chengxiao, Su Shu

2024, 45(1): 230-236. doi: 10.13832/j.jnpe.2024.01.0230

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Abstract:
In order to truly reflect the transient internal flow transition process and hydraulic load impact of the reactor coolant system under accident conditions, a high-precision three-dimensional closed system transient flow calculation method was established for the HPR1000 reactor and its primary system, and the pressure wave oscillation law and transient hydraulic load characteristics of the pipeline of the reactor and primary system during the transition process were obtained. The results show that: in the end of the reactor coolant pump rotor seizure, the flow rate at the reactor coolant pump outlet decreased to 81.3% of that in stable operation. During the transition process of the rotor seizure, the maximum pressure peak value in the system pipeline is located at the inlet section of the reactor coolant pump, which is 16.00 MPa; the minimum pressure valley value is located at the outlet section of the reactor coolant pump, which is 15.01 MPa. Finally, the pressure of each monitoring point in the system tends to the reference pressure of 15.50 MPa. Under the dual influence of the piping layout of reactor coolant system and the rotor seizure accident of reactor coolant pump, the fluid velocity of each section shows obvious uneven distribution, and obvious eddy currents occurs. The variation rule of hydraulic load on each wall of the system is consistent with the variation rule of system pressure pulsation. The maximum load force peak is located at the W3 wall at the first elbow of the transition section, which is 3.163×10⁶ N; the minimum load force valley value is located at the W12 wall of the elbow at the inlet of the reactor pressure vessel, which is 9.125×10⁵ N. This numerical prediction method can provide technical support for the design and safety assessment of the reactor coolant pipeline under the condition of the reactor coolant pump rotor seizure.

Research on Operation Characteristics of Heat Pipe Reactor Coupled with Open-Air Brayton Cycle

Liu Jiusong, Liu Chengmin, Yi Jingwei, Li Yi, Li Siguang

2024, 45(1): 237-245. doi: 10.13832/j.jnpe.2024.01.0237

Abstract(126) HTML (29) PDF(34)

Abstract:
In order to explore the operation characteristics of nuclear power conversion system with open-air Brayton cycle coupled with heat pipe reactor when the core power and load change, the system simulation model is established based on Modelica language, including the sub-models of the reactor core, the heat pipe and the Brayton cycle, and the accuracy of each model is verified. The transient simulation and analysis of loss of load (LOL) and power increment and reduction processes are carried out by using the established model. The calculation results show that in the transient process, the change of load or core power will lead to the change of rotating speed, and it is necessary to control the turbine flow through the bypass control valve to restore the rotating speed to stability. Under LOL condition, the core temperature will drop, and the reactivity feedback will increase the core power by 2.3% and the maximum fuel temperature by 1.7 K. During the reactor power increment and reduction, the peak normalized core power caused by reactivity feedback is 102.6% and 100.7% respectively. The results of this paper provide a reference for the safety analysis of the heat pipe reactor coupled with open-air Brayton cycle.

Research on Coincidence Detection Efficiency Based on Positron Decay Nuclides

Zhuo Xianglin, Qing Xianguo, Yang Zhenlei, Bao Chao, Jiang Tianzhi, Li Jin, Lu Jiawei

2024, 45(1): 246-252. doi: 10.13832/j.jnpe.2024.01.0246

Abstract(127) HTML (21) PDF(18)

Abstract:
In order to reduce the measurement lower limit of the primary pressure boundary leakage monitoring, the coincidence detection efficiency based on positron decay nuclides has been studied. In order to improve the coincidence detection efficiency, the particle transport process in the coincidence detection device has been simulated and analyzed by Monte Carlo code Geant4, and the effects of coincidence detection device structure and detector properties on the coincidence detection efficiency have been studied. The research results show that: adding β⁺ absorbing layers on both sides of the filter paper can significantly improve the coincidence detection efficiency, and the optimal coincidence detection efficiency is obtained when 0.5mm Al or 0.2mm Fe is used as β⁺ absorbing layer; due to the difference in energy resolution of different types of detectors, different types of detectors have different energy window coefficients for reaching the optimal detection efficiency. The optimal energy window coefficients of NaI(Tl), BGO and LaBr₃(Ce) are 14%, 26% and 7% respectively. The results of this research can provide a reference for the structural design of the detection device and the coincidence judgment logical design of the primary pressure boundary leakage monitoring system with ¹⁸F nuclide as the radioactive tracer.

2024 Vol. 45, No. 1