核动力工程

Research and Development Progress and Application Prospect of Nuclear Fuels for Commercial Pressurized Water Reactors

Jiao Yongjun, Yu Junchong, Zhou Yi, Li Yuanming, Chen Ping, Duan Zhengang

2022, 43(6): 1-7. doi: 10.13832/j.jnpe.2022.06.0001

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Abstract:
Pressurized water reactor (PWR) is the main reactor type of nuclear power plant at present, and nuclear fuel is the energy source of reactor and the main source of fission products, which is related to the economy and safety of nuclear power plant. This paper summarizes and evaluates the performance characteristics, technical status and prospects of doped UO₂ fuel, high density fuel, micro-capsulated fuel and metal fuel currently developed for commercial PWR applications. Among these fuels, the large grain fuel has high technology maturity and is expected to be commercially available in PWR in the near future; the corrosion and oxidation of high density fuel and metal fuel in high temperature water and the behavior under accidents still need to be studied and solved; micro-encapsulated fuel with extremely safety is more suitable for small reactors for special purposes. Research and development of advanced fuel assemblies, design criteria and high fidelity performance analysis technology shall be carried out in coordination to maximize the reliability and high burnup advantages of new fuels.

Numerical Study of Spacer Effects on Convective Heat Transfer at Low Flow Rates

Ding Guanqun, Xiao Yao, Gao Xinli, Liu Bo, Li Junlong, Gu Hanyang

2022, 43(6): 8-14. doi: 10.13832/j.jnpe.2022.06.0008

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Abstract:
A numerical study is carried out for the thermal-hydraulic characteristics of a circular tube with spacer at low flow rates and high heat flux. The convective heat transfer of single-phase water in a smooth circular tube at low flow rates and the spacer effects are calibrated by the empirical correlation and experimental data. A CFD method based on the SST k-ω model is established. The simulation results show that the heat transfer characteristics downstream of the spacer depend on the buoyancy lift parameter. For the forced convection zone and mixed convection heat transfer decreasing zone, the heat transfer downstream of the spacer is always enhanced and the Nussel number decays exponentially. For the mixed convection heat transfer recovery zone and natural convection zone, due to the coupling effect of flow field and heat transfer, the heat transfer downstream of the spacer deteriorates and the Nussel number oscillates with damping. The influence range of heat transfer at the downstream of the spacer first increases and then decreases with the increase of buoyancy lift parameters. The larger the spacer blocking ratio is, the more severe the heat transfer oscillation is, and the worse the heat transfer caused by the spacer is. This study can provide a reference for the design of core in a low-flow core.

Stage Progress and Analysis of International Standard Problem 51 “Open Test”

Zhang Xueyan, Yang Jun, Wang Shiqi, Deng Chengcheng, Li Yuquan, Zhang Peng, He Dandan

2022, 43(6): 15-23. doi: 10.13832/j.jnpe.2022.06.0015

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Abstract:
In order to verify the ability of the thermal-hydraulic code to simulate the behavior of passive nuclear reactor under the condition of loss of coolant accident, and to assess the ability of optimal estimation codes to predict specific experiments, based on China's large-scale global effect test bench-Advanced Core Cooling System Mechanism Experiment (ACME) facility, the Organization for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) organized the International Standard Problem No. 51 (ISP-51) project. A preliminary comparative analysis is made on the code calculation results submitted at the current open test stage. The results show that for the loss of coolant accident with a 2-inch small break in the same cold pipe section, the simulation results of the thermal-hydraulic optimal estimation code RELAP5 used by Huazhong University of Science and Technology and Polytechnic University of Catalonia are in good agreement with the experimental data in terms of triggering time and flow value of the passive safety system. In the simulation results of TRACE code used by Polytechnic University of Madrid and Spanish NFQ Company, the triggering time of each passive safety system is delayed, and the gas-liquid flow rate of ADS1~3 is significantly higher than the experimental value, which may be related to the selection of different critical flow models and the setting of valves and pipelines. This project has set a precedent for non-member countries of OECD to initiate and take charge of international standard problem projects. It also helps relevant scientific research teams in China to become more familiar with the operation, organization and management of international nuclear science and technology research cooperation projects, and to undertake more work in international nuclear science and technology cooperation.

Study on Slug Characteristics of Air-water Two-phase Flow in Horizontal Narrow Rectangular Channel

Liu Antai, Cheng Linhai, Gu Haifeng, Yan Changqi, Meng Zhaoming, Gong Suijun

2022, 43(6): 24-29. doi: 10.13832/j.jnpe.2022.06.0024

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Abstract:
As the key parameters of slug flow, slug velocity and liquid film thickness are of great significance in heat transfer analysis and mechanical analysis. In this paper, the characteristics of slug in a horizontal narrow rectangular channel with 1.9 mm×68 mm are studied by using high-speed camera and PCB liquid film sensor. For liquid Reynolds number ${{Re}}_{\text{l}}\text{ < 2500}$) , the rectangular channel is a laminar area; For Re_l≥2500, the rectangular channel is a turbulent area. Based on the gas-liquid two-phase mixing velocity, the predictive relation of slug velocity is fitted respectively. The results show that the laminar area distribution coefficient (C₀) can be calculated by Ishii equation and the drift velocity is 0; the turbulent area C₀ is 1.0. For slug Reynolds number Re_b<3100, the liquid film thickness (δ_b) at the bottom of slug increases with the increase of capillary number; For Re_b≥3100, δ_b shows volatility. Existing δ_b predictive relation is not applicable to narrow rectangular channels. Considering the influence of channel aspect ratio, a new δ_b predictive relation is proposed to verify 210 data in the literature, and the prediction errors are all within ±20%.

Analysis and Research on the Influence of Closed Loop Structure on Main Pump Performance Based on Source Term Method

Wang Xiaofang, Li Jialing, Lu yeming, Liu Haoran, Li Yingyue, Wang Hui, Cui Huiru

2022, 43(6): 30-36. doi: 10.13832/j.jnpe.2022.06.0030

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The traditional main pump flow analysis platform is mostly a simplified open flow path, which is quite different from the real closed loop operation conditions. In order to explore the flow characteristics and mechanism of the main pump in the real loop, this paper takes the full-channel hydraulic model of the main pump with seal ring clearance as the research object, and uses the source term method for steady-state and transient calculation and analysis. The steady-state calculation results show that the vortex flow pattern is formed in the closed circulation loop, which leads to pre-swirl at the inlet of the main pump and inflow distortion, resulting in the increase of turbulent kinetic energy and uneven energy distribution. The transient calculation results show that compared with the open flow path, the inflow distortion of the closed loop brings changes in the characteristics of flow field pressure, velocity, turbulent kinetic energy and pressure fluctuation, which leads to the decrease of pump head and efficiency and the increase of radial force and axial force. The analysis of the flow performance of the main pump based on the closed loop structure is closer to the real flow.

Numerical Simulation of Three-Phase Coupling for High-Temperature Lithium Heat Pipe

Mao Shang, Zhou Tao, Liu Wenbin, Wei Dong, Xue Chunhui

2022, 43(6): 37-42. doi: 10.13832/j.jnpe.2022.06.0037

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Abstract:
To study the heat transfer mechanism of lithium heat pipes and to improve their application in small reactors, a solid-liquid-gas three-phase coupled model of the tube wall, the wick and the vapor chamber inside the tube is developed using COMSOL Multiphysics software. The results show that when the heat flow in the evaporation section increases from 13.9 kW to 20.8 kW, the pipe wall temperature, steam temperature, steam pressure, liquid pressure inside the wick, and liquid axial velocity increase with the increase of heating power, while the steam axial velocity first increases and then decreases with the increase of heating power. In the steady-state operation, the pipe wall temperature decreases step by step, while the vapor temperature and pressure remain basically unchanged, indicating that the lithium heat pipe has good isothermality.

CFD Calculation and Analysis of Boiling Heat Transfer of Full-length Fuel Bundle with Bent Rod

Ren Bing, Gan Fujun, Yang Ping

2022, 43(6): 43-50. doi: 10.13832/j.jnpe.2022.06.0043

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Abstract:
The boiling heat transfer in a 5 × 5 full-length fuel bundle containing a bent fuel rod is simulated by computational fluid dynamics (CFD) analysis. The calculation is based on the Eulerian-Eulerian two-fluid model and the improved wall boiling model, and the calculation method is verified based on the test data in the PSBT benchmark task. The calculated average void fraction of the cross section is in good agreement with the test data, which indicates the reliability of the existing calculation methods. Based on the calculation results, the effects of bent rod on the flow field, temperature field, void fraction and other key parameters in the bundle channel are investigated. The research results show that the existence of bent fuel rod has no significant impact on the cross section transverse flow, fluid temperature, void fraction, etc., but the increase of the surface temperature of bent fuel rod will also lead to bubble aggregation, which increases the risk of critical heat flux (CHF).

Research and Platform Development of Multi-physical Coupling Scheme Based on Unified Framework

Qian Guanhua, Yu Tao, Yang Tao, Zhao Yanan, Zhao Pengcheng

2022, 43(6): 51-60. doi: 10.13832/j.jnpe.2022.06.0051

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In order to realize the multi-physics, multi-process and high-fidelity numerical calculation of the reactor and capture the more real physical behavior in the reactor core, in this paper, the coupling scheme of multiple physical programs is deeply studied, and based on the upper monitoring architecture, serial computing mode, and the explicit coupling scheme of grid one-to-one mapping, a unified framework-based multi-physical coupling platform is built by relying on the open source integration platform SALOME, the common platform interface ICoCo, the three-dimensional core neutronics program ADPRES, and the system thermal hydraulic program RELAP5. The verification of NEACRP-L-335PWR PWR rod ejection benchmark task shows that the calculation results of the coupling platform are in good agreement with the benchmark task, and the coupling platform is more accurate in power peak capture, which can release part of the safety margin; the calculation results of the parameters at the end of the transient are also accurate enough, which proves that the coupling platform can carry out more precise and in-depth numerical calculation and safety analysis of the reactor multi-physical and multi-process coupling conditions.

Study on Drag Force Model of High-Temperature Particle Under Steam Entrainment

Peng Cheng, Deng Jian

2022, 43(6): 61-65. doi: 10.13832/j.jnpe.2022.06.0061

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The settling process of high-temperature melt in coolant is related to the triggering of steam explosion and the development of follow-up process, and affects the design and implementation of serious accident mitigation measures. Based on the entrainment of the steam film on the surface of high-temperature particles, a semi-empirical relation for predicting the drag force coefficient of particles in the coolant during the coarse mixing stage is constructed by means of theoretical modeling and experimental fitting, which can be expressed as a function of particle Froude number (Fr_p) and entrainment Reynolds number (Re_α). By comparing with the experimental data of boiling motion in the falling coolant of high-temperature steel ball, it is verified that the entrainment effect of steam in the initial stage of coarse mixing is the main factor of particle settling. In addition, the change of settling velocity is affected by the diameter of high-temperature particles. The smaller the diameter of the particles is, the closer the settling characteristics are to the "cold particles", which is mainly related to the decrease of steam entrainment.

Experimental Study on Edge Blockage Accidents and Central Blockage Accidents in a Rectangular Channel

Yuan Dongdong, Deng Jian, Tan Sichao, Zhu Jiahong, Li Chengwei, Qiao Shouxu

2022, 43(6): 66-72. doi: 10.13832/j.jnpe.2022.06.0066

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Abstract:
To obtain the evolution law of the flow field under the rectangular channel blockage accident, the particle image velocimetry (PIV) technology is adopted to carry out a full-flow-field visual experimental study on the vertical narrow rectangular channel blockage accident with a gap of 3 mm to compare and analyze the difference of the flow field structure under the condition of edge blockage and central blockage at 70% blockage rate. It is found that the flow field structure of edge blockage is more complex than that of central blockage, the edge blockage condition will form a larger backflow area due to channel blockage, and a wall separation vortex will be formed at high Reynolds number (Re); the edge blockage is more unfavorable to the removal of fluid heat, and the vorticity of the vortex in the backflow area is smaller, the fluidity of the vortex is worse, and the motion frequency is more complex; At the same time, the variation trend of vortex structure in edge blockage and central blockage with Re is also completely opposite.

Verification of COSINE Reflooding Model and Sensitivity Analysis of Parameters

Li Xuelin, Zhang Hao, Yang Yanhua

2022, 43(6): 73-78. doi: 10.13832/j.jnpe.2022.06.0073

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Reflooding is an important stage after the PWR large break accident. In order to evaluate the calculation capability of the system program at this stage, it is necessary to select a variety of heat transfer models to reproduce the accident and analyze the sensitivity response of the parameters. In this paper, the PWR Loss of Fluid Test (LOFT) bench is modeled, and the calculation results of different heat transfer models in the COSINE program are compared with the experimental data to verify the accuracy of the model; At the same time, the parameter sensitivity calculation in the reflooding stage is carried out, and the parameters which have the greatest influence on the second peak cladding temperature (PCT) are identified. The calculation shows that the heat transfer model of COSINE program can well predict the reflooding phenomenon, and the sensitivity parameters that have great influence on the calculation results include UO₂ volume heat capacity, droplet diameter, droplet interphase heat transfer coefficient and heat transfer coefficient of film boiling wall against vapor phase.

Research on Fast Prediction of Key Parameters of Containment Based on Time Series Deep Learning Model

Feng Qianyi, Guo Zhangpeng, Li Zhongchun, Zhang Jiayu, Zhao Houjian, Ruan Yanghui, Yu Yu

2022, 43(6): 79-84. doi: 10.13832/j.jnpe.2022.06.0079

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Abstract:
The Main Steam Line Break (MSLB) accident threatens the safe operation of nuclear power plant. In this paper, the time-dependent transient response of key safety parameters of passive containment cooling system (PCCS) in nuclear power plant under MSLB accident is predicted based on time series deep learning model. The transient safety parameters are taken as the research objects. The data are preprocessed by linear normalization and feature label segmentation, trained by short-term data sets, and the time series deep learning model of single parameter and multi parameter coordination is established by using long short-term memory (LSTM) and recurrent neural network (RNN); long-term untrained data sets are predicted by a multi-parameter coordination model. The research shows that the prediction based on time series deep learning model is applicable under the same accident and different working conditions; it is feasible to predict long-term data based on short-term training data; the prediction accuracy of the single-parameter model or multi-parameter coordination model using LSTM is higher than that of RNN. The deep learning model based on LSTM can effectively, accurately and quickly predict the transient safety parameter response characteristics of PCCS under MSLB accidents, and can provide a fast prediction and analysis model for accident safety analysis.

Parameters Sensitivity Analysis and Optimization of High-Temperature Heat Pipe for Heat Pipe Reactor

Tian Zhixing, Wang Chenglong, Guo Kailun, Zhang Dalin, Tian Wenxi, Qiu Suizheng, Su Guanghui

2022, 43(6): 85-92. doi: 10.13832/j.jnpe.2022.06.0085

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Abstract:
The design of high-temperature heat pipe for heat pipe reactor is a multi-objective optimization problem with constraints. This paper aims to achieve rapid multi-objective design optimization of high-temperature heat pipe. For high-temperature heat pipe, main line, channel, wire mesh, sintering and other wicks are considered. In this paper, based on the improved thermal resistance network method, the non-dominant genetic algorithm II is used to optimize the thermal resistance and capillary mass flow. The results show that the performance of the heat pipe is related to the working medium and the wick. The working medium potassium is better for round and rectangular main lines, and the working medium sodium is better for triangular grooves and sintered fibers; For thermal resistance in sodium heat pipe, the ranking is in the order of circular main line, wire mesh, rectangular groove, sintered particles, sintered fiber, triangular groove, circular main line and rectangular main line; for flow, the ranking is in the order of circular main line, wire mesh, sintered particles, rectangular groove, rectangular main line, circular main line, triangular groove, sintered fiber. In the range of 800~950 K, the increase of working temperature results in the reduction of thermal resistance by more than 89.9% except for the annular main line, flow increase by more than 320.8%. In the annular main line, the thermal resistance is reduced by 93.5%, but the flow is reduced by 8.8%. This study can provide reference for the design optimization of high-temperature heat pipe of nuclear reactor, and for improving the performance of high-temperature heat pipe.

Intelligent Optimization of Lead-bismuth Reactor Core Based on Radial Basis Function Surrogate Model and Niche Genetic Algorithm

Li Qiong, Liu Zijing, Wang Weijia, Zhao Pengcheng, Yu Tao, Chang Haotong

2022, 43(6): 93-100. doi: 10.13832/j.jnpe.2022.06.0093

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Abstract:
In order to solve the complex nonlinear multi-dimensional optimization problem under the influence of multi-factor coupling of lead-bismuth reactor, an intelligent optimization method for reactor core was constructed based on radial basis function (RBF) surrogate model prediction, orthogonal Latin hypercube sampling (OLHS) and niche genetic algorithm optimization. A design optimization platform for lead-bismuth reactor was developed, which included the functions of sampling, Monte Carlo program coupling treatment, and core parameter prediction and optimization. The scheme optimization verification was carried out with the minimum fuel loading of the core as the optimization objective. The results show that the RBF surrogate model can accurately and quickly predict the core characteristic parameters of the lead-bismuth reactor. Compared with the calculated values of the Monte Carlo program, the relative error of the predicted core effective multiplication factor k eff is within ± 0.1%. This intelligent optimization method is feasible for lead-bismuth reactor core optimization, which can find the optimal target scheme under the constraint of multi-factor co-variation, and greatly reduce the search calculation time of the design scheme. Therefore, the intelligent optimization method established in this study can provide new ideas for the optimization design of multi-physics, multi-variable and multi-constraint coupling effects of lead-bismuth reactor.

Corrosion Behavior of Alumina-forming Austenitic Heat Resistant Steel in Supercritical Carbon Dioxide

Ma Zhaodandan, Cong Shuo, Chen Yong, Guo Xianglong, Zhang Ruiqian, Liu Zhu, Zhang Xian

2022, 43(6): 101-107. doi: 10.13832/j.jnpe.2022.06.0101

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The corrosion behavior of Alloy 20Cr-25Ni and a new structural material alumina-forming austenitic heat resistant steel (AFA steel) in supercritical carbon dioxide (S-CO₂) environment at 600℃/20 MPa was studied. The morphology, composition, and structure of oxide films of the two alloys were analyzed. It is found that the weight gain curves of the two alloys were in accord with the “parabola” law. The weight of corrosion products increased slowly; it was only 2.11 mg/dm² after 1000 h. By comparison, the coarse oxide products appeared on the surface of Alloy 20Cr-25Ni and increased with the extension of corrosion time, while oxide film of AFA steel remained dense and continuous. Through the analysis of the cross-sectional morphology of the oxide film, it is found that the Alloy 20Cr-25Ni has a two-layer oxide film structure after corrosion, which is mainly composed of Fe₃O₄ and FeCr₂O₄ oxide layer and a small amount of spinel. However, there are three layers of oxide film structure in AFA steel, the middle and inner layers are Cr₂O₃ and Al₂O₃ oxide films respectively, and the outer layer is distributed with a discontinuous FeCr₂O₄ spinel oxide. Due to the formation of dense Al₂O₃ oxide film, the corrosion resistance of AFA steel is greatly improved.

Effect of Working Medium Pressure and Creep on the Stress Corrosion Cracking Behavior of Cold Deformed 310S Stainless Steel in Supercritical Water Environment

Su Haozhan, Wang Peng, Zhang Lefu

2022, 43(6): 108-116. doi: 10.13832/j.jnpe.2022.06.0108

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310S stainless steel is a good candidate cladding material for supercritical water-cooled reactor. In order to enrich the research on stress corrosion behavior of 310S stainless steel in supercritical water environment, especially the data on crack growth rate, this study used the method of online monitoring crack growth to measure the crack growth rate of 310S stainless steel with different cold deformation under various working conditions, and analyzed the effect of working medium pressure, high-temperature creep and other factors on the 310S cracking behavior. The results show that: The pressure change of supercritical water or high-temperature steam has a limited effect on the cracking behavior of 310S stainless steel at 500℃, and cold deformation promotes the crack growth of the material. The high-temperature creep behavior of materials plays an important role in accelerating the process of stress corrosion cracking in supercritical water, especially for materials under high cold deformation and high load. This study enriches the data of stress corrosion crack growth rate of 310S in supercritical water environment, and proves that improving the creep resistance of materials is one of the important means to optimize the service performance of cladding materials, and large cold deformation shall be avoided in the process of cladding design and manufacturing.

Study on Flow Blockage Accident of HFETR Multi-layer Annular Fuel Assembly

Liu Wenbin, Deng Caiyu, Song Jiyang, Xiang Yuxin, Kang Changhu, Liu Chang, Song Yuge, Guo Yufei

2022, 43(6): 117-121. doi: 10.13832/j.jnpe.2022.06.0117

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The fuel assembly of High Flux Engineering Test Reactor (HFETR) adopts multi-layer annular narrow slot flow channel design to improve the heat transfer capacity. However, it should be noted that the narrow slot flow channel is more likely to be blocked. In this paper, the HFETR fuel assembly model is established based on the RELAP5 program. The comparison between the calculated value and the experimental value shows that the model is reasonable and accurate. Based on the model, the transient characteristics and influencing factors of hot box fuel assembly under flow blockage accident are studied. The results show that: ① When the flow blockage ratio is greater than 0.5, the peak temperature of fuel cladding and core increases significantly with the increase of flow blockage ratio; ② Even if a single flow channel is blocked completely, due to the cooling of the surrounding flow channels, the maximum peak temperature of the fuel cladding is only 218.6 ℃, which can ensure the integrity of the fuel cladding; ③ The flow rate and other parameters fluctuate greatly at the initial stage of a single flow channel full blockage accident, but the main parameters of the fuel assembly are basically stable 15 s after the accident.

Discuss on Off-site Consequence Evaluation Criteria after Severe Accident for Zhangzhou NPP

Xing Ji, Shi Xueyao, Huang Shuming

2022, 43(6): 122-127. doi: 10.13832/j.jnpe.2022.06.0122

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In order to quantitatively evaluate the basic objective of “offsite protective actions that can reduce the radiological consequences technically are limited or even can be cancelled”, this paper proposes the design objective of “no evacuation is required beyond 3 km, no concealment and iodine taking are required beyond 5 km” after severe accidents from the perspective of simplifying off-site emergency after accidents. Combined with the site conditions of Zhangzhou NPP, a set of radioactive consequence evaluation criteria for Zhangzhou NPP are derived. Through the analysis of the typical severe accident process and radioactive release process of HPR1000, the results show that the HPR1000 reactor type of Zhangzhou NPP meets the radiological consequence evaluation criteria proposed in this paper, and can achieve the goal of “no evacuation is required beyond 3 km, no concealment and iodine taking are required beyond 5 km” after a severe accident.

Experimental Research on Aerosol Condensation and Retention in Narrow Cracks of Steel Containments

Wang Shanpu, Tong Lili, Cao Xuewu

2022, 43(6): 128-132. doi: 10.13832/j.jnpe.2022.06.0128

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Currently, the effect of narrow cracks on aerosol retention is not taken into account in the radioactivity assessment of containment in a nuclear power plant. However, compared with the conventional size, the high surface/volume ratio of the narrow crack has considerable retention of aerosol leakage, and the assessment results are too conservative. Through the aerosol leakage experiment in the rectangular straight channel, the aerosol retention efficiency in the narrow crack of the steel containment with a crack height of about 100 μm is obtained, and it is observed that the entrance area of narrow crack channel is the main particle deposition area. Besides, by establishing and maintaining a certain temperature gradient in the narrow crack flow direction, the aerosol leakage process in the containment narrow crack is simulated when the containment passive cooling system is put into operation. The results show that the narrow crack has a good retention effect on submicron aerosol, and the steam condensation introduced by temperature gradient can significantly improve the retention efficiency of aerosol to about 91%, and reduce the leakage area.

Design Optimization of Retention Coil for 2 MWt Liquid Fuel Thorium-based Molten Salt Experimental Reactor

Wang Chengcai, Xu Hui, Zhang Xueyong, Yang Zhaodong, Yan Guizhen, Wang Yanhui, Yang Zhongwei

2022, 43(6): 133-138. doi: 10.13832/j.jnpe.2022.06.0133

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The retention coil of liquid fuel thorium-based molten salt experimental reactor is an important nuclear safety equipment in the tail gas treatment system of molten salt reactor, which provides a closed environment for the decay of short-lived nuclides in the tail gas, removes the decay heat at the same time, and reduces the temperature of the iodine adsorption bed and the activated carbon adsorption bed downstream it. This paper introduces the problems existing in the original design scheme of the retention coil of the 2 MWt liquid fuel thorium-based molten salt experimental reactor provided by the owner and the new design scheme after structural optimization. The optimized design scheme of retention coil has the advantages of good heat dissipation, few leakage points (few welds), easy welding, easy inspection, monitoring and maintenance, less materials, etc., and has good reliability and economy.

Research on High Temperature Mechanical Properties of Main Pipeline Materials

Li Pengzhou, Li Yilei, Yao Di, Luo Jiacheng, Sun Lei, Qiao Hongwei

2022, 43(6): 139-145. doi: 10.13832/j.jnpe.2022.06.0139

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In order to provide material property parameters for leak before break (LBB) design of nuclear power plant, it is necessary to measure the mechanical properties of base metal and welding material of main pipeline at high temperature, including the high-temperature dynamic mechanical properties of materials in seismic environment. Based on universal servo material testing machine and high-speed material testing machine, the static-dynamic tensile mechanical properties, crack growth rate and static-dynamic fracture toughness of nitrogen controlled 00Cr17Ni12Mo2 stainless steel and OK Tigrod 316L welding material for main pipeline base metal of nuclear power plant at high temperature (350℃) are measured. Compared with the normal temperature mechanical properties of the main pipeline base metal and welding materials, the static-dynamic tensile mechanical properties of the two materials at 350℃ and the static-dynamic fracture toughness of OK Tigrod 316L at 350℃ have significantly decreased compared with the normal temperature, and the crack growth resistance of the two materials at 350℃ has slightly decreased compared with the normal temperature. The research results can provide experimental technology and material parameter support for LBB design of nuclear power plant pipelines.

Research on Active Disturbance Rejection Control of Once-through Steam Generator

Cheng Yuyu, Ma Zhicai, Wang Mingyang, Zhang Yue

2022, 43(6): 146-150. doi: 10.13832/j.jnpe.2022.06.0146

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The once-through steam generator (OTSG) has small water capacity and heat storage capacity, and its mathematical model is uncertain and nonlinear. When there are disturbances and load changes, the steam pressure fluctuates greatly, which adversely affects the system equipment. The conventional proportional-integral-derivative (PID) control has some disadvantages, such as overshoot, poor anti-disturbance performance, etc., which is difficult to meet the system performance requirements. To solve the above problems, the active disturbance rejection control (ADRC) is used to control the steam pressure of OTSG. However, because there are many parameters to be tuned in ADRC, the shuffled frog-leaping algorithm (SFLA) is improved and optimized in this paper, which is used to optimize the parameters of ADRC, and a simulation model is established for simulation test. The results show that the ADRC with improved SFLA for parameter self-tuning can realize fast tracking control of OTSG without overshoot, reduce the control error of steam pressure, and has good anti-disturbance ability.

Review Strategy of Application for Extension of Validity Period of Operation License for Research Reactor

Yang Zhe

2022, 43(6): 151-154. doi: 10.13832/j.jnpe.2022.06.0151

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The Ministry of Ecology and Environment Order No. 8 "Regulations on Safety Permit Procedures for Nuclear Power Plants, Research Reactors and Nuclear Fuel Cycle Facilities" (hereinafter referred to as the "Regulations") has made new clear provisions on the extension of operation licenses for nuclear power plants, research reactors and nuclear fuel cycle facilities. In order to promote the establishment of the aging management standard system of research reactors in China, the development history of life extension review strategy of research reactors in China is analyzed in this paper. Combined with several key issues in the review of the application for extension of the validity period of the operation license of research reactors such as high-flux engineering test reactor, the review strategy is proposed, which gives priority to periodic safety review, focuses on aging management and takes into account the review of technical specifications and differences. The research results provide practical experience and theoretical guidance for the establishment of regulations and standards for research reactor aging management in China.

Study on Core Power Control of Small Lead-Bismuth Cooled Fast Reactor

Sun Aodi, Sun Peiwei, Wei Xinyu

2022, 43(6): 155-161. doi: 10.13832/j.jnpe.2022.06.0155

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Lead-bismuth cooled fast reactors are mostly used for isolated power supply in remote and special environments such as oceans and mountains. The characteristics of users require that small lead-bismuth cooled fast reactors have good load following ability. In this paper, three different control methods are applied to the core power control of lead-bismuth cooled fast reactor, and the stability and set value tracking ability of the controller are tested by introducing noise, dead zone and time lag. The results show that the proportional integral derivative (PID) controller is difficult to achieve good control effect, so other links are often added in industrial applications to ensure the stability of the PID controller. Both active disturbance rejection control (ADRC) and H_∞ robust controller have good anti-noise ability, and can independently achieve good control effect, but good anti-noise ability must sacrifice a certain sensitivity. The comparative analysis of the three kinds of controllers shows that because the simulation calculation simplifies the actual object, the controller designed under such conditions shall be designed with more conservative parameters.

Research on Proportional Complex Integral Control Strategy for Rod Position Detector Power Supply System

Gao Longjiang, Xu Qiwei, Yu Tianda, Tang Jiankai, Luo Lingyan, Li Qingzhao, Huang Siyu

2022, 43(6): 162-167. doi: 10.13832/j.jnpe.2022.06.0162

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Abstract:
Single-phase inverter works in voltage source mode in rod position detection system of nuclear power plant. The load rod position detector is usually equivalent to a inductive load, but inductive load connected to the output of the inverter cannot output a stable sinusoidal voltage. In this paper, a proportional complex integral (PCI) control strategy is proposed for the power supply system of single-phase inverter rod position detector, and the parameter design method of the controller is given. First, based on the frequency domain model of the system, the analytical expressions of the parameters of the current loop proportional integral (PI) controller and the voltage loop PCI controller are derived, and the influence of the load parameters on the system stability is considered. Then, the expected open loop cut-off frequency and phase margin of current loop and voltage loop are set, and the PI and PCI controller parameters are analytically calculated by this method. Finally, through MATLAB simulation, it is verified that the proposed method can obtain stable output voltage and output current of single-phase inverter under inductive load, and the output current total harmonic distortion (THD) is less than 0.3%. This method can provide guidance for the power supply system control of single-phase inverter rod position detector.

Verification and Research of LBB Pipe Crack Leakage Rate Calculation Software PICLES

He Feng, Wu Wanjun, Ma Ruoqun, Fang Yonggang, Ai Honglei, Wang Xinjun, Sun Yingxue

2022, 43(6): 168-173. doi: 10.13832/j.jnpe.2022.06.0168

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Abstract:
The calculation of pipe crack leakage rate is the key technology in leak-before-break (LBB) analysis. By comparing with the effective software and the experimental results, the domestic independently developed leakage rate calculation software PICLES is verified and studied. Compared with the international similar software (PICEP and SI-PICEP) with mature engineering applications, there is a small difference between PICLES and its calculation results. Compared with the experimental results of pipe crack leakage rate, the leakage rate calculated by PICLES is −80.23%~−43.79% different from that calculated by PICLES, and the leakage crack length calculated by PICLES is 21.84%~79.07% different from the measured crack length, which shows that the LBB analysis of supercooled water pipeline using PICLES is highly conservative. Therefore, PICLES can be used for LBB analysis in practical projects.

Analysis on Validity of Test Data of Primary Circuit Pressure Boundary Leakage Rate

Zhu Wei, Hou Qinmai, Cai Ning, Xu Sai, Zhu Jie, Li Qiuqi

2022, 43(6): 174-179. doi: 10.13832/j.jnpe.2022.06.0174

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Abstract:
The primary circuit pressure boundary gas leakage rate test of a research reactor is interfered by small volume, equipment cooling water and other factors, and there are extreme sample points in the test data. On the basis of not changing the test method, the effectiveness of the test results is ensured by using the significance test and variance analysis of the regression model. The test method draws lessons from the test method of containment leakage rate of PWR. In this paper, three commonly used methods for calculating the leakage rate of the containment, including ANSI/ANS-56.8, RCC-G-1988, and ПНАЭГ-10-021-90, are used for calculation and analysis. The results show that the calculation results of the three methods are basically the same. By analyzing the independence, normality and variance property of residuals in linear regression model, the influence of regression diagnosis on the calculation results is discussed. At the same time, for the problems of heterogeneity of variance, autocorrelation and extreme sample points found in regression diagnosis, the final results are modified by combining the residual weighted least square method and removing extreme sample points, and the reliability of the results is improved. The analysis method in this paper has been applied to nuclear safety review.

Optimization Technology and Application of Preventive Maintenance Program Dominated by Equipment Classification

Yang Lifei, Ma Yijin, Zhang Guanghui, Wang Wei, Tang Kun, Wang Shuwen, Liu Xiaolei, Wu You, Liu Haixiao, Wu Tao

2022, 43(6): 180-186. doi: 10.13832/j.jnpe.2022.06.0180

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Abstract:
The technical concept and guidance of preventive maintenance (PM) program optimization dominated by equipment classification are put forward, and the optimization technical route and cases of PM program of C/S/E/R equipment at all levels are given. It has been applied to the lean program of PM program of several CPR1000 nuclear power units in China from 2018 to 2020, which improved and strengthened the PM program of key important equipment and reduced excessive maintenance, so that the maintenance resources can be allocated reasonably. By the end of 2020, the total value of the PM program equivalent of 16 CPR1000 units has been reduced by more than 20%. It is conservatively estimated that the maintenance cost savings of PM is more than 50 million yuan. Practice has proved that the application of equipment classification based PM program optimization technology can not only ensure the reliability maintenance level of key and important equipment, but also effectively reduce the operation and maintenance costs of nuclear power units, which can be used as a reference for other nuclear power plants to optimize PM program.

Numerical Study on Oxygen Transport Characteristics in Liquid Lead-Bismuth Eutectic Circuit

Liang Ruixian, Yang Lingfeng, Wang Yifeng, Li Xiaobo, Niu Fenglei

2022, 43(6): 187-194. doi: 10.13832/j.jnpe.2022.06.0187

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Abstract:
In order to study the oxygen transport characteristics of the cover gas in a gaseous oxygen control device of liquid lead-bismuth eutectic (LBE) coolant system--expansion tank, the numerical calculation of oxygen transport was carried out by ANSYS Fluent software using computational fluid dynamics (CFD). According to the flow characteristics of the cover gas and the characteristics of low oxygen partial pressure in the gas mixture, the gas-phase space of expansion tank was simplified, the gas-liquid interface was regarded as the free surface boundary with constant oxygen concentration, and the oxygen concentration in liquid LBE after mass transfer between gas and liquid LBE was calculated by component transport model. The results show that the mass transfer coefficient increases with the increase of liquid LBE inlet velocity. When the inlet velocity of liquid LBE increases, the intensity of gas-liquid convection in the expansion tank increases, which is beneficial to enhance the oxygen transport in the expansion tank. The higher the LBE temperature in the expansion tank, the greater the average mass transfer coefficient of oxygen transport. When the liquid LBE inlet velocity is constant, the average mass transfer coefficient can be expressed as an increasing function of temperature. Within the saturated oxygen concentration threshold, the inlet oxygen concentration and the oxygen concentration at the gas-liquid interface do not affect the mass transfer coefficient of the expansion tank, which is beneficial for oxygen concentration control of liquid LBE circuit. In this study, the operating conditions for the liquid LBE circuit to be in a reasonable range of oxygen concentration were obtained quantitatively, which provided data reference for experiment and system design.

Design and Analysis of Personnel Airlock New Transmission Mechanism Development for Nuclear Power Plant

Zhang Feng, Qin Junwei, Xie Honghu, Liu Xiaohua, Chen Zhao, Yang Jinchun

2022, 43(6): 195-200. doi: 10.13832/j.jnpe.2022.06.0195

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Abstract:
Using digital design method, a new transmission mechanism of personnel airlock is developed and the prototype modular design is completed. Using FMEA (failure mode and effects analysis) method, the weak links of the new transmission mechanism are found out, that is, the transmission chain of the locking mechanism, the door swing module and the bolt plug-in module after the earthquake. Respectively, using ADAMS dynamic simulation analysis software, the locking function and motion performance of the locking mechanism transmission chain are simulated and analyzed. Using ABAQUS finite element analysis software, the operability of the door swing module and the effectiveness of the bolt plug-in module after earthquake are analyzed. The results show that the transmission chain of the locking mechanism is properly selected, the door swing module meets the operational requirements, and the bolt plug-in module meets the effectiveness requirements. The transmission chain and structure of the new transmission mechanism developed are reasonable and meet the requirements of the third generation nuclear power.

Numerical Simulation of Gas-liquid Two-phase Separation in Vortex Separator

Zhang Zekai, Zhang Tingting, Yin Shasha, Yin Junlian, Wang Dezhong

2022, 43(6): 201-208. doi: 10.13832/j.jnpe.2022.06.0201

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Abstract:
For gas-liquid two-phase separation, traditional separators are either too large in volume or low in swirling intensity. So a new type of cyclone separator is proposed. Utilizing the reverse flow of vortex diode to form a high-strength swirling flow, a branch pipe is added above the swirling chamber. For the two-phase flow entering from the tangential inlet, due to the density difference and swirling flow, the gas phase will gather in the center and flow out of the separator from the upper branch pipe due to buoyancy, and the liquid phase will be distributed around the separator from the lower branch pipe due to gravity, thus realizing the separation of the two phases. The separators with different swirling chamber sizes and outlet shapes are calculated by numerical simulation. The simulation results show that under the working condition that the inlet flow is 0.5 t/h and the inlet air content is 1%~5%, the pressure at the control underflow port is the same as that at the inlet, the pressure difference between the overflow port and the inlet is 80~90 kPa, and the separation efficiency of the separator for bubbles with a diameter of 50 μm and 100 μm can be above 90%.

Study on the Adsorption of Co(Ⅱ) and Mn(Ⅱ) in Simulated Wastewater by ZIF-67

Zhou Yipeng, Men Jinfeng, Wang Xiaowei, Du Zhihui, Liang Chengqiang, Jia Mingchun

2022, 43(6): 209-216. doi: 10.13832/j.jnpe.2022.06.0209

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Abstract:
In order to develop an adsorbent with high efficiency and selectivity for long-lived activation products in radioactive liquid waste, metal-organic frameworks (MOFs) material ZIF-67 was prepared at room temperature, and the thermal stability test and structural characterization of the material were carried out. The influence of the initial pH values, adsorption time and initial concentration of the solution on the adsorption of Co(II) and Mn(II) by ZIF-67 was investigated for the first time. The results show that ZIF-67 is a microporous material with good hydrothermal stability. Under the conditions of pH=6.0, temperature of 30℃, and initial concentration of 500 mg/L, the saturated adsorption capacities of ZIF-67 for Co(II) and Mn(II) reached 305.63 mg/g and 197.43 mg/g, respectively. ZIF-67 has adsorption selectivity for Co(II), Mn(II) and Ni(II) in mixed metal ion solution. Therefore, ZIF-67 has a good application prospect in the treatment of activation products in actual radioactive liquid waste.

Conceptual Design Study of Ultra-high Flux Fast Neutron Research Reactor Fuel

Li Wenjie, Xia Bangyang, Yu Hongxing, Jiao Yongjun, Li Quan, Sun Dan, Wu Yu

2022, 43(6): 217-221. doi: 10.13832/j.jnpe.2022.06.0217

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Abstract:
Increasing neutron fluence rate is the development trend of high flux research reactor, which can greatly accelerate the R&D process of reactor materials. However, if the neutron fluence rate is increased to 10¹⁶ cm⁻²·s⁻¹, the peak power density will be several times higher than that of the existing research reactor, which will bring many challenges to the reactor and nuclear fuel design. For this reason, this paper semi quantitatively analyzes the impact of increasing neutron fluence rate on the performance of nuclear fuel from neutronics, heat transfer, fuel material behavior in the reactor and other aspects, and proposes design measures to meet the challenges of ultra-high flux and power density, providing guidance for the development of ultra-high flux fast neutron research reactor fuel design.

Requirement Analysis on Ultra-high Flux Fast Neutron Research Reactors

Liao Wei, Xia Bangyang, Yu Hongxing, Li Wenjie, Mu Keliang, Zhang Fengshou

2022, 43(6): 222-226. doi: 10.13832/j.jnpe.2022.06.0222

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Abstract:
Achieving the ultra-high fast neutron flux is an important development direction of the world's advanced research reactors, which is of great significance for accelerating the innovative development of fuels and materials for the fourth generation advanced nuclear power system. From the aspects of the irradiation test of structural materials and nuclear fuels in advanced nuclear reactor and the production of long-reaction-chain transplutonium element, this paper preliminarily analyzes the necessity of building ultra-high flux fast neutron research reactor in China. On this basis, the core maximum neutron fluence rate and its coolant of the ultra-high flux fast neutron research reactor are determined, and the main parameters of the reactor and the coolant flow scheme are given as follows: the thermal power of the reactor is 200MW, the coolant is lead-bismuth alloy, and the maximum neutron fluence rate is more than 10¹⁶ cm⁻²·s⁻¹.

Steady and Transient Thermodynamic Performance Analysis of UN-FeCrAl Fuel Element under High Burnup

He Liang, Qiu Bowen, Wu Zhouzhi, Zhang Kun, Chen Ping, Gao Shixin, Hu chao, Xing Shuo, Fan Hang, Wang Yanpei

2022, 43(6): 227-231. doi: 10.13832/j.jnpe.2022.06.0227

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Abstract:
The UN-FeCrAl fuel element is one of the main schemes for high burnup application of accident tolerant fuel, and the thermodynamic property of the fuel element under high burnup needs to be evaluated. The steady and transient thermodynamic property of UN-FeCrAl fuel element under 68000 MW·d·t⁻¹(U) was predicted by FUPAC software, and the analysis results showed there is a good thermodynamic property of UN-FeCrAl fuel element under steady condition. Under transient conditions, the maximum pellet center temperature of UN fuel is only 862℃, which can meet the requirements of pellet temperature design, but the maximum transient stress of FeCrAl cladding will reach 459 MPa, and the transient strain will increase by 0.23% compared with steady-state strain, which may make FeCrAl cladding face the risk of transient stress and transient strain criterion exceeding the limit. Therefore, the subsequent research shall focus on the transient stress and transient strain of FeCrAl cladding.

2022 Vol. 43, No. 6