核动力工程

Research on Adsorption and Energy Storage of Refrigerants R1234yf and R32 in MOF-74

Zhang Cheng, Yan Xiao, Peng Shinian, Yuan Dewen, Liu Wenxing

2022, 43(3): 1-6. doi: 10.13832/j.jnpe.2022.03.0001

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Abstract:
The heat absorption of circulating working medium can be improved by using the mutual transformation of thermal energy and surface energy in the process of adsorption and separation of fluid molecules on the solid surface of nano-porous materials. In this paper, molecular simulation (molecular dynamics and Grand Canonical Monte Carlo) and adsorption theory are used to study the adsorption and energy storage of R1234yf and R32 in MOF-74. In the adsorption of pure working medium, it is found that the adsorption capacity of R32 in MOF is higher than that of R1234yf. The adsorption capacity of refrigerant in Zn-MOF-74 is larger than that in Co-MOF-74, and the pressure required for R1234yf to reach saturation adsorption is lower than that required for R32 to reach saturation in the corresponding adsorbent. In the mixed working medium adsorption, the adsorption capacity of R1234yf is higher than that of R32. With the increase of temperature, the adsorption capacity of R1234yf shows a gradually increasing trend, while that of R32 gradually decreases. The energy storage calculation shows that the higher the mass fraction of M-MOF-74 (M = Co, Zn) particles, the more heat energy needed to be absorbed for the phase transition of the mixed working medium.

Numerical Simulation of Lead-bismuth Alloy Solidification in Lead-water Reaction

Liu Dalin, Liu Xiaojing, Huang Yanping, Gong Houjun

2022, 43(3): 7-14. doi: 10.13832/j.jnpe.2022.03.0007

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Abstract:
In order to study the solidification mechanism of lead-bismuth alloy in the reaction between lead-bismuth alloy and water caused by steam generator tube rupture (SGTR), in this paper, by coupling the VOF model, the Realizable k-ε turbulence model, the solidification heat transfer model, and using the FLUENT software, a two-dimensional simulation model of the reaction process of lead-bismuth alloy and water is established, and the model is compared and verified with the results of the existing reaction experiments. Then, based on the enthalpy method, the enthalpy equation of solidification heat transfer characteristics which can directly describe the solidification phenomenon of lead-bismuth alloy is established, the factors and conditions affecting the solidification of lead-bismuth alloy are studied by controlling the model variables. Finally, the model is applied to the scene with complex structure. The results show that the temperature difference between lead-bismuth alloy and water, the initial speed of water jet and the diameter of water injection are the main factors affecting the solidification of lead-bismuth alloy. The model proposed in this paper has high reliability and can simulate the solidification phenomenon of lead-bismuth alloy under actual working conditions. The mechanistic and phenomenological conclusions obtained in this study can provide theoretical support for the safety analysis of lead-based fast reactor.

Development and Verification of Shielding Database Generation Module in NECP-Atlas

Zu Tiejun, Xu Ning, Cao Liangzhi, Wu Hongchun

2022, 43(3): 15-20. doi: 10.13832/j.jnpe.2022.03.0015

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Abstract:
The shielding database generation module Shield_calc is developed in the nuclear data processing program NECP-Atlas. This module firstly uses NECP-Atlas to generate a database of fine-group neutron and photon cross sections in MATXS format which is irrelevant to the problem; then uses the ultrafine group method and Bondarenko iterative method to perform resonance self-shielding calculation to obtain effective self-shielding cross-sections; Finally, based on the one-dimensional reactor model, NECP-Hydraa is used for transportation calculation to obtain the typical weight spectrum of the applied reactor type, and the fine-group shielding database is merged into the wide-group shielding database NECL-SHILED. Using Shield_calc module, based on the same evaluation nuclear database ENDF/B-Ⅶ.0 as BUGLE-B7, NECL-SHILED with 47 groups of neutrons and 20 groups of photons is generated and compared with BUGLE-B7. The numerical results show that the calculation results of NECL-SHILD and BUGLE-B7 are in good agreement, which verifies that the Shield_calc module has high accuracy.

Design and Heat Transfer Performance Analysis of Ultra-High Temperature Lithium Heat Pipe

Hu Chongju, Yu Dali, He Meisheng, Li Taosheng, Yu Jie

2022, 43(3): 21-27. doi: 10.13832/j.jnpe.2022.03.0021

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Abstract:
Ultra-high temperature lithium (Li) heat pipe cooled nuclear reactor has broad application prospects in the areas of deep-sea nuclear power and deep-space exploration due to its silence, small size and other advantages. In order to master the heat transfer characteristics of ultra-high temperature lithium heat pipe, the design of ultra-high temperature lithium heat pipe is carried out, and the Python program of ultra-high temperature lithium heat pipe is developed based on the thermal resistance grid method. On this basis, the heat transport performance of lithium heat pipe is analyzed. By comparing with other existing models and experimental data, the accuracy of the model developed in this paper is verified, and the ultra-high temperature lithium heat pipe designed in this paper is checked by using this program, and the influence of heat pipe structure on the transition time of ultra-high temperature lithium heat pipe to reach a new stable state under variable power condition is analyzed. The results show that the ultra-high temperature lithium heat pipe designed in this paper meets the design requirements; increasing the wall thickness and wick thickness increase the transition time; increasing the length of the condensation section is beneficial to reduce the transition time. The research in this paper provides the basis for the optimization design and safety analysis of the heat pipe reactor.

Characteristic Analyses of “177 Core” of HPR1000

Li Dongsheng

2022, 43(3): 28-32. doi: 10.13832/j.jnpe.2022.03.0028

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Abstract:
The most remarkable technical feature of HPR1000 PWR nuclear power plant is that the reactor adopts a core composed of 177 fuel assemblies (hereinafter referred to as “177 core”), which has complete independent intellectual property right. In order to deeply analyze its characteristics, this paper introduces the main technical features of “177 core”, the number of fuel assemblies and control rod assemblies is compared with the core composed of 157 fuel assemblies (hereinafter referred to as “157 core”); the similarities and differences between two typical reactor core loading schemes (“177-A core” and “177-B core”) are described and evaluated. The results show that “177 core” has more advantages in safety and economy compared with “157 core”; the first cycle loading arrangements of the two typical cores have their own advantages; in terms of burnable poison material selection, “177-B core” is better than “177-A core”. Finally, possible optimization suggestions for the HPR1000 reactor core in the future are given from five aspects, such as canceling the central position control rod assembly of the core, setting the radial metal reflector of the core, implementing no neutron source startup, loading self-developed fuel assemblies in batches and optimizing the length of the active core.

Transient Simulation Study of Primary Loop Nitrogen-typed Pressurization System

Yan Xinlong, Li Yi, Luo Hanyu, Tian Ye

2022, 43(3): 33-37. doi: 10.13832/j.jnpe.2022.03.0033

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Abstract:
Based on the fundamental of nitrogen pressure stabilizing, the lumped parameter method is adopted to develop the nitrogen pressure stabilizing system transient simulation model, which breaks through the limitations of the existing independent stabilizer model, realizes the direct coupling between the primary loop system and the nitrogen stabilizer, and verifies the program with the test data of the nitrogen pressure stabilizing system of the floating nuclear power plant. On this basis, a design method of nitrogen pressure stabilizing system based on sensitivity analysis is proposed. Compared with the existing design method, this design method can obtain the optimal configuration scheme of nitrogen pressure stabilizing system. At the same time, through the adaptive design, it can ensure that the pressure of the nitrogen pressure stabilizing system does not exceed the range of the temperature and pressure limit curve of the primary loop system during the start-up process.

Thermal-Hydraulic Performance Analysis of Horizontal Spiral Tube Steam Generator for European Lead-Cooled Fast Reactor

Zhang Wei, Li Jingsong, Shi Huilie, Qiao Pengrui, Wang Cong, Zhang Tianqing, He Yingzhao

2022, 43(3): 38-45. doi: 10.13832/j.jnpe.2022.03.0038

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Abstract:
Taking the horizontal spiral tube steam generator (HST-SG) of European lead-cooled system (ELSY) as the research object, combined with its structural and operating parameters, this paper selects an appropriate heat transfer resistance model and develops a one-dimensional steady-state thermal hydraulic calculation program. Firstly, the program is used to check and calculate ELSY HST-SG to verify the accuracy of the program calculation. Then, combined with the calculation results, the thermal and hydraulic performance of ELSY HST-SG is analyzed in detail, and the comparative analysis and research is carried out for different operation parameters. The analysis results show that the selection of various parameters of ELSY HST-SG is reasonable, the thermal and hydraulic performance is excellent, and the purpose of compact structure is achieved. Therefore, this program can be used for design development and performance analysis of ELSY HST-SG.

Research on Distribution Characteristics of Time-averaged Flow Field Downstream of Spacer Grid under Pulsating Flow

Li Xing, Wang Qianglong, Tan Sichao, Qiu Jinrong, You Ximing

2022, 43(3): 46-52. doi: 10.13832/j.jnpe.2022.03.0046

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Abstract:
Studying the evolution characteristics of the instantaneous flow field downstream of the spacer grid in the rod bundle channel under flow fluctuation is of great significance to reveal the flow and heat transfer mechanism in the fuel assembly under ocean conditions. In this paper, the structure of the spatio-temporal evolution flow field downstream of the spacer grid in the rod bundle channel under pulsating flow is obtained by using the particle image velocimetry (PIV) technique, and the effects of pulsating parameters (pulsating period and pulsating amplitude) on the velocity distribution and turbulence characteristics downstream of the spacer grid are analyzed. The results show that there is little difference between the time-averaged velocity downstream of the spacer grid under pulsating flow and that under steady flow, and it basically does not change with the variation of pulsating amplitude and pulsating period; The root mean square of transverse velocity and axial velocity downstream of the spacer grid under pulsating flow are significantly different from that under steady flow, and show different trends with the change of pulsating parameters. The results of this paper help to reveal the transient characteristics of fuel assemblies under unsteady conditions and lay a foundation for the design and optimization of fuel assemblies.

Research on Intelligent Optimization Method for Core Flow Zoning of Lead-bismuth Reactor

Ling Yufan, Dai Shengqi, Zhao Pengcheng, Zhu Enping, Wang Jifeng, Tang Huan

2022, 43(3): 53-57. doi: 10.13832/j.jnpe.2022.03.0053

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Abstract:
Core flow zoning is an important means to achieve core outlet temperature flattening. Reasonable zoning can improve the reactor safety and economy. In this paper, the artificial intelligence optimization algorithm is coupled with the single channel model, and the calculation model of reactor core flow zoning is constructed. The convergence analysis of genetic algorithm, differential evolution algorithm and quantum genetic algorithm in reactor flow zoning is carried out respectively. According to the obtained optimal algorithm, taking the power distribution at the beginning of life cycle as the sample data and the maximum power of each fuel assembly throughout the life cycle as the sample data, the comparative analysis of two different flow zoning schemes is carried out based on the small long-life natural circulation lead-bismuth fast reactor SPALLER -100. The results show that among the three intelligent optimization algorithms, the quantum genetic algorithm has the best convergence on the reactor flow zoning problem, and can quickly search the optimal zoning results; The maximum outlet temperature of the fuel assembly based on the power distribution at the beginning of the life cycle exceeds the thermal safety limit of the reactor, while the maximum outlet temperature of the fuel assembly based on the maximum power of each fuel assembly during the entire life by 140 K and remains below the thermal safety limit; The optimal number of zones for SPALLER-100 reactor is 5, and increasing the number of zones has little effect on improving the thermal safety performance of the reactor.

Study on CCFL Characteristics in Downcomer during Discharge Phase of LOCA with RELAP5 Code

Li Xiang, Sun Wan, Ding Shuhua, Huang Tao, Li Zhongchun, Pan Liangming

2022, 43(3): 58-65. doi: 10.13832/j.jnpe.2022.03.0058

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Abstract:
Under the loss-of-coolant accident (LOCA), the two-phase countercurrent in downcomer is extremely likely to cause the vapor-liquid counter-current flow limitation (CCFL), which is not conducive to the smooth entry of emergency coolant into the core, which greatly affects the safety performance of the nuclear reactor system. Based on RELAP5 code, the Wallis overflow relation is used to model the UPFT experimental device and calculate the water injection behavior in the downcomer during discharge phase of LOCA; The validity of the model is verified by comparing the water storage capacity of lower chamber, the pressure in the downcomer and the transient changes of steam flow at the break, and the distribution characteristics of vapor phase velocity field and liquid phase volume fraction in the downcomer are analyzed. The results show that the flow irregularity caused by the three-dimensional characteristics of the channel structure in the downcomer affects the characteristics of the vapor-liquid CCFL. With the increase of steam flow, the greater the pressure gradient and upward flow velocity gradient in the connection area between the break loop and the downcomer, the division method with fewer nodes is difficult to truly reflect the vapor-liquid overflow relationship in the local area of the downcomer channel; The cooling water injected in the circuit close to the break is more difficult to reach the lower chamber, while the cooling water in the circuit far from the break can easily enter the lower chamber; The superheated steam is cooled by the cooling water during the flow process, resulting in condensation, resulting in the steam flow at the outlet being less than the inlet steam flow, and the condensation effect decreases with the increase of the inlet steam flow. The model and method established in this study can be applied to the prediction of vapor-liquid CCFL in the downcomer channel during discharge phase of LOCA.

Research on the Key Influencing Factors of the Driving Force of the Primary Loop System of the Natural Circulation Lead-based Fast Reactor

Zhai Pengdi, Zhu Enping, Zhao Pengcheng, Wang Tianshi, Yu Tao

2022, 43(3): 66-73. doi: 10.13832/j.jnpe.2022.03.0066

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Abstract:
In order to deeply study the key factors affecting the driving force of the primary loop system of the natural circulation lead-based fast reactor, taking the natural circulation lead-based fast reactor SNCLFR-10 as the research object, a steady-state operation model describing the natural circulation of the primary loop of the reactor was constructed; The effects of heat transfer in cold/hot pool, nonlinear temperature distribution of heat source and heat sink and heat dissipation of reactor pressure vessel wall on natural circulation capacity are quantitatively analyzed theoretically, and relevant numerical simulation verification is carried out. The results show that the numerical simulation results are in good agreement with the theoretical calculation values in this study; the coupling effect of the three natural circulation capacity influencing mechanisms will reduce the natural circulation capacity of the SNCLFR-10 system, resulting in the relationship between natural circulation flow and power no longer satisfying the 1/3 power obtained by theory.

Transient Characteristics Analysis of 2000 kW High Temperature and High Pressure Test Loop of HFETR during Loss of Power Accident of Main Pump

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

2022, 43(3): 74-77. doi: 10.13832/j.jnpe.2022.03.0074

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Abstract:
In order to study the safety characteristics of 2000 kW high temperature and high pressure test loop in high flux engineering test reactor (HFETR) in the process of loss of power accident of main pump, the simulation model of test loop is established based on RELAP5 code, and the transient characteristics of loss of power accident of main pump are analyzed by using the verified model. The calculation results show that in the process of loss of power accident of main pump, the high-speed working condition of the main pump switches to the low-speed working condition of the two accident pumps, the flow rate decreases rapidly and finally stabilizes to half of the initial flow rate, and the fuel cladding reaches a peak temperature of 763 K at 4.34 s; Then, due to the continuous decline of power, the cladding temperature decreases continuously; The minimum departure from nucleate boiling ratio during the accident is greater than 1.3, which indicates that departure from nucleate boiling will not occur and meets the safety requirements.

Research on Multi-objective Optimization of Flow Distribution in Natural Circulation Reactor Whole Life-Cycle

Zhu Enping, Wang Ting, Zhao Pengcheng, Ling Yufan, Wang Tianshi, Yu Tao

2022, 43(3): 78-84. doi: 10.13832/j.jnpe.2022.03.0078

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Abstract:
The design of core flow distribution is the key content of core structure optimization of natural circulation reactor, which is highly valuable for improving core economy and safety. In this paper, the local optimal flow distribution calculation model is constructed based on the reactor closed parallel multi-channel model, and the existing flow distribution scheme is analyzed. Aiming at its limitations, a multi-objective comprehensive evaluation method based on the optimal time zone is proposed, which can realize the optimization calculation of the whole life-circle multi-objective reactor flow distribution. According to the proposed theory, combined with the TOPSIS comprehensive evaluation method, the maximum output power under natural circulation, the maximum temperature difference at the outlet during the reactor life-circle, and the standard deviation of the maximum temperature difference over time are used as attribute values to develop the optimization of core flow distribution scheme for a small long-life natural cycle lead-bismuth fast reactor SPALLER-100. The research results show that the flow distribution scheme of SPALLER-100 reactor core is the best based on the power distribution of 3182 d running time. Compared with the flow distribution scheme based on the power distribution at the beginning of life circle, the maximum temperature difference at the core outlet is reduced by 30 K, the standard deviation of the maximum temperature difference at the core outlet with time is reduced by 41%, and the maximum output power of the reactor natural cycle is increased by 2.35%.

Prediction of Vertical-Downward Two-Phase Flow Pattern based on PCA-GA-SVM

Qiao Shouxu, Zhong Wenyi, Tan Sichao, Li Xupeng, Hao Sijia

2022, 43(3): 85-93. doi: 10.13832/j.jnpe.2022.03.0085

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Abstract:
In order to improve the accuracy and timeliness of flow pattern identification under the condition of small samples, an optimized identification model integrating wavelet packet decomposition (WPD), principal component analysis (PCA), genetic algorithm (GA) and support vector machine (SVM) is proposed and successfully applied to the flow pattern recognition of vertical-downward two-phase flow. WPD is used to decompose and reconstruct the non-stationary conductivity fluctuation signal, extract the wavelet packet energy and construct the feature vector; The dimension of feature vector is reduced by PCA to reduce the complexity of feature input; At the same time, the key parameters penalty factor (C) and kernel function parameter (g) of SVM are determined by GA global iterative optimization. After verifying the identification effect of PCA-GA-SVM, it is compared with SVM, PCA-SVM and GA-SVM networks. The results show that the SVM network optimized by PCA and GA is significantly improved in terms of flow pattern identification accuracy and timeliness. The overall prediction accuracy of bubble flow, slug flow, stirred flow and annular flow reaches 94.87%, and the time consumed is only 3.95 s, which can meet the needs of on-line identification.

Analysis of Heat Transfer Characteristics of Natural Convection Condensation of Steam Containing Air under Different Tube Bundle Arrangements

Liu Le, Chen Wenzhen, Wang Jue, Wang Cong, Hu Chen

2022, 43(3): 94-100. doi: 10.13832/j.jnpe.2022.03.0094

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Abstract:
In order to study the effect of tube bundle arrangement of heat exchanger in passive containment cooling system (PCCS) on the condensation heat transfer characteristics of steam containing air under natural convection conditions, the condensation heat transfer process in channels of single tube, single-row to five-row tube bundles is numerically studied by coupling gas component transport equation and condensation model. It is found that there is a “inhibiting effect” that reduces the condensation heat transfer capacity due to the interference of high-concentration air layers between tube bundles, and a “suction effect” that strengthens the condensation capacity on the wall due to the transverse flow of gas caused by water vapor wall condensation. The influence of such two effects on condensation heat transfer under different tube bundle structures is analyzed. The results show that with the increase of the number of rows of tube bundles, the influence of the two effects on the condensation heat transfer gradually increases, leading to the increase of non-uniformity of circumferential local condensation heat transfer capacity of condensing tubes. Among them, the maximum circumferential local condensation heat transfer coefficient (HTC) of five-row tube bundles is 2.3 times that of single tube, and the minimum is only 44.7% of that of single tube. In the double-row, three-row and four-row tube bundles, the condensation heat transfer capacity of the regular quadrilateral arrangement is better than that of the regular triangle arrangement, while in the five-row tube bundle, the condensation heat transfer capacity of the regular triangle arrangement is stronger. This study can provide a technical reference for the optimization of the tube bundle arrangement of the PCCS heat exchanger.

Design and Analysis of Irradiation Test Scheme for Fast Reactor Fuel Slug with HFETR

Zhang Liang, Yang Wenhua, Zhao Wenbin, Sun Sheng, Jin Shuai, Lei Jin

2022, 43(3): 101-106. doi: 10.13832/j.jnpe.2022.03.0101

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Abstract:
Considering the irradiation test index and fuel test safety, the operation requirements of high flux engineering test reactor (HFETR), the pressure difference fluctuation in the test section and other factors, in this paper, based on HFETR, the design and analysis of the fast reactor fuel slug irradiation test plan were carried out, and the key parameters such as the thickness of the lead-bismuth alloy layer, the structure of the cooling water channel, the structure of the thimble plug, and the flow rate of the cooling water were determined, and a high linear power density irradiation test plan with the maximum temperature of the hot rod cladding at (490±60)℃ was obtained. The results show that when the maximum linear power density of the hot rod is 68~85 kW/m, the temperature of the cladding and the fuel core meets the requirements of the irradiation test and has a margin; In the range of 200~300 kPa core pressure difference, the calculated CFX value of the flow in the test section under the same pressure difference is about 9%~11% smaller than the test value; The flow share of the narrow channel outside the test section is 7.3%, which is significantly lower than the flow area share of the channel, meeting the cooling requirements of fuel slugs when the linear power density is 85 kW/m. The irradiation test plan proposed in this paper can provide a reference for the high linear power density irradiation test of fast reactor fuel rods.

Thermal Shock Behavior Analysis of SiC Composite Cladding

Liu Shichao, Pang Hua, Zhou Yi, Li Yuanming, He Liang, Zhang Kun, Tu Teng

2022, 43(3): 107-112. doi: 10.13832/j.jnpe.2022.03.0107

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Abstract:
In order to solve the problems of poor convergence and insufficient research on thermal shock performance in the simulation of thermal shock behavior of SiC composite cladding, this paper simulates the internal stress state of double-layer SiC composite cladding under Loss of Coolant Accident (LOCA), uses the COMSOL software of multi-physical field coupling to numerically simulate the thermal shock behavior of SiC composite cladding, and analyzes the effects of thickness ratio, thermal shock temperature and end plug on the thermal shock resistance of SiC composite cladding. The results show that the circumferential stress produced by thermal shock increases with the increase of the thickness ratio of chemical vapor infiltration layer (CVI layer) to chemical vapor deposition layer (CVD layer); When the thickness ratio of CVI layer to CVD layer is 9:1, the circumferential tensile stress of SiC composite cladding during thermal shock can reach 113 MPa; The circumferential stress produced by thermal shock increases with the increase of thermal shock temperature difference. When the thermal shock temperature is 1200 K, the circumferential stress is 112.7 MPa; During thermal shock, there is obvious stress concentration at the end plug, and its radial stress is up to 22.3 MPa, which is higher than the bonding strength reported in the literature (20~25 MPa), which is the main reason for the failure of the end plug connection.

Effect of Zinc-injected Water Chemistry on Corrosion Behavior of Zr-Sn-Nb Alloy

Yin Zhaohui, Lai Xuping, Tang Min, Chen Zirui, Zhao Yongfu, Deng Ping, Yang Hong, Zhang Gen, Xiong Jing, Gong Bin

2022, 43(3): 113-117. doi: 10.13832/j.jnpe.2022.03.0113

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Abstract:
The corrosion behavior of Zr-Sn-Nb in the simulated PWR primary circuit zinc-injected aqueous chemical environment was investigated by autoclave corrosion test to analyze the surrosion and oxide film morphology of Zr-Sn-Nb alloy in zinc-free and zinc-added aqueous chemical environment. The results show that the surrosion curve of Zr-Sn-Nb alloy turns when it is corroded in zinc-free and zinc-added aqueous chemical environment for 150 days. Zinc addition has no significant effect on the surrosion, corrosion kinetics law, oxide film morphology, oxide film phase, oxide film thickness, hydride distribution and hydrogen absorption concentration of Zr-Sn-Nb alloy.

Consideration on Improvement of Failed Fuel Separator Roller in High Temperature Gas-Cooled Reactor

Zhang Yan, Xu Guangduo, Jin Dongjie, Zhao Binbin, Ma Lanqing

2022, 43(3): 118-122. doi: 10.13832/j.jnpe.2022.03.0118

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Abstract:
Since the existing high-temperature gas-cooled reactor failed fuel separator roller adopts a straight-toothed rack, and the redirecting teeth are embedded on the rack, the fuel balls will collide when passing through the redirecting teeth, which is easy to cause the fuel balls to be damaged, and cause the device to vibrate; At the same time, the structure can not ensure that all unqualified fuel balls are sorted out, and the detection rate is uncontrollable. In this paper, the roller of the failed fuel separator is improved by adopting a spiral V-shaped groove self-disturbing roller; the structure principle of the roller is described, the curved surface equation of the V-shaped groove of the roller is given, and the scanning trajectory of the fuel ball is simulated. The improvement results show that the fuel balls can adjust their posture by themselves, which can achieve the purpose of stable operation, high detection rate and avoiding collision. It is worthy to be popularized and applied in this reactor type.

Study of Residual Stress Distribution of Large Ring-type Dissimilar Metal Weld in Reactor Pressure Vessel

Fu Qiang, Min Yuansheng, Liu Chuan, Li Meifu, Li Yuguang

2022, 43(3): 123-128. doi: 10.13832/j.jnpe.2022.03.0123

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Abstract:
To obtain the residual stress distribution of the large ring-type dissimilar metal weld in the reactor pressure vessel(RPV), and guide the structural design and manufacturing process optimization of the reactor pressure vessel, the dissimilar metal welding structure simulators of nickel base alloy and low alloy steel, which can represent the welding structure form of products, are designed and manufactured. The longitudinal residual stress in the welding structure simulators is tested by contour method, and the transverse and longitudinal residual stress of the welding structure simulators are simulated and calculated by finite element method. The residual stress distribution characteristics of the whole dissimilar metal welding joint are obtained. The results show that the longitudinal residual stress in the weld area is tensile stress, the peak stress reaches about 500 MPa, and the surface stress is greater than the internal stress, and the peak stress occurs at 3 mm and 24 mm from the lower surface; The distribution trend of transverse residual stress in the weld area from the upper surface to the lower surface is tensile stress - compressive Stress - tensile stress. The peak value of compressive transverse residual stress reaches −300 MPa and appears about 18 mm away from the lower surface. The research in this paper can provide theoretical guidance for the design of welding structures.

Analysis on the Improved Control Rod Drop Behavior among Deformation Channels Based on the Rigid-flexible Coupling

Yue Ti, Zheng Lele, Zhu Fawen, Wang Haoyu, Yuan Pan, Sun Yu, Deng Shuang

2022, 43(3): 129-134. doi: 10.13832/j.jnpe.2022.03.0129

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Abstract:
The fuel assembly is prone to bend after long-term irradiation in the core, which affects the safe drop of the control rod. Therefore, it is urgent to study the influence mechanism of the control rod drop behavior under the deformation channel. By means of numerical simulation, the law of rod drop behavior after bending deformation of guide tube is analyzed and studied, the rigid-flexible coupling method is used to calculate the rod drop behaviors in the straight, C-shaped and S-shaped guide tubes respectively, analyze the variation of the entire rod drop journey, velocity, acceleration, and collision force along the journey with time, and the effects of straight and two different deformation channels on rod drop behavior are compared. The results show that the rigid flexible coupling method can better simulate the rod drop behavior under the deformation channel. The C-shaped rod drop does not get stuck, and the S-shaped rod drop is stuck at the second bend. This study will help to provide a basis for judging the ultimate bending threshold of the problem of rod drop sticking caused by bending deformation, and provide a reference for engineering design.

Research on Core Power Control of Small Lead-based Reactor Based on Ant Colony Algorithm

Li Jinyang, Liu Yinuo, Zeng Wenjie, Hu Yang, Yu Tao

2022, 43(3): 135-138. doi: 10.13832/j.jnpe.2022.03.0135

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Abstract:
Small lead-based reactor has a wide range of applications and complex and changeable operating conditions. It is difficult to achieve good control of core power by using traditional control methods. In order to solve the problem that the parameters of the traditional linear quadratic Gaussian control (LQG)/loop transmission recovery technology (LTR) controller cannot be adjusted online, the core state space model is established by using the perturbation theory, an LQG/LTR controller based on ant colony algorithm is designed, the core power control system of small lead-based reactor is established, the parameters of LQG/LTR controller are adjusted online, and the dynamic simulation of the core is carried out. The results show that the LQG/LTR controller based on ant colony algorithm is easier to be stable and the amplitude of change is smaller than that of the traditional LQG/LTR controller.

Selection and Experimental Research of the Passive Start-Up Neutron Detector in Nuclear Power Plant

Jiang Tianzhi, Li Biao, Zhang Yun, Wang Yinli, Li Wenping, Huang Youjun, Shen Feng, Sun Congjian

2022, 43(3): 139-143. doi: 10.13832/j.jnpe.2022.03.0139

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Abstract:
In order to monitor the reactor core during the first cycle charging, shutdown and start-up of nuclear power plants, domestic and foreign nuclear power plants generally introduce two primary neutron source components in the core, but the primary neutron sources are imported from the United States, which has the problem of import limitations. In order to solve this problem, it is studied to cancel the primary neutron source component in the first cycle and use the neutrons produced by the spontaneous fission of the fuel assembly as the starting neutron source. The neutron intensity produced by spontaneous fission of fuel assembly is much lower than that of primary neutron source. In view of the above situation, it is necessary to use a higher sensitivity detector to monitor the neutron fluence rate outside the reactor. Based on the analysis of the basic principles of various high-sensitivity detectors, this paper gives suggestions on the selection of high-sensitivity neutron detectors, and verifies their performance by experiments. The experiment results show that even when the γ dose rate is greater than 0.1 Gy/h, the ³He proportional counter tube can be set with an appropriate discriminating voltage, and can effectively discriminate γ noise. The maximum γ dose rate verified by the experiment is 1.0 Gy/h.

Optimal Control of Stretch-out Operation for CPR1000 Nuclear Power Unit

Yang Zhang, Song Yinlei, Tian Wei

2022, 43(3): 144-150. doi: 10.13832/j.jnpe.2022.03.0144

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Abstract:
Stretch-out operation (SO) is an important means of flexible operation for PWR nuclear power unit. It is of great significance to study how to improve the safety and economy of the unit under SO mode. For the case of fluctuation of important parameters such as primary loop average temperature, core thermal power, core axial power deviation and temperature regulating rod (R rod) position of a Chinese improved three loop pressurized water reactor (CPR1000) nuclear power unit under SO mode, relevant research shows that the main reason is that the high-pressure regulating valve of the steam turbine of the CPR1000 nuclear power unit operates in the steep region of the flow characteristic curve, which causes the valve opening to fluctuate under the influence of external disturbances, and induces fluctuations of important parameters such as main steam flow, average temperature of primary loop, etc. Combined with the operation characteristics of the nuclear power unit, strategies such as optimizing the flow characteristic curve of the high-pressure regulating valve and optimizing the main steam flow limit are proposed to improve the safety and economy of the unit during SO. The effectiveness of the strategy is verified by several engineering cases of stretch-out operation of several CPR1000 nuclear power units.

Numerical Analysis of Leakage Characteristics of Valve Sealing Structure

Tian Xiaoshuai, Zhang Donglin, Yang Yong, Tang Yueming, Tan Shushi, Xie Tong

2022, 43(3): 151-157. doi: 10.13832/j.jnpe.2022.03.0151

Abstract(343) HTML (61) PDF(55)

Abstract:
Taking the sealing structure of safety valve in nuclear reactor system as the research object, a three-dimensional rough surface model of sealing structure based on porous medium theory is established in this paper. The calculation formula of leakage rate of sealing structure is deduced by Darcy's law. The effects of roughness, autocorrelation length and sealing specific pressure on surface characteristics, as well as the effects of roughness and sealing surface contact width on leakage rate were studied. The results show that the relationship between the roughness and the sealing performance is not linear, and it is limited to only use the roughness as the evaluation index of the sealing performance. When the roughness is certain, the different autocorrelation length will also affect the porosity and permeability of the sealing interface, thus affecting the sealing performance of the safety valve. The decrease of sealing specific pressure leads to the increase of contact height, which makes the porosity between valve seat and disc increase rapidly, resulting in the enhancement of leakage characteristics of sealing structure. The increase of roughness makes the leakage rate increase nonlinearly, and the increase of sealing surface contact width makes the leakage rate decrease linearly.

Analysis and Optimization of Flow Field in an Axial Flow Lead-Bismuth Pump

Zhang Shuanglei, Li Liangxing, Song Liming

2022, 43(3): 158-164. doi: 10.13832/j.jnpe.2022.03.0158

Abstract(555) HTML (107) PDF(106)

Abstract:
The lead-bismuth main pump is the key conveying equipment of the primary circuit in a lead- bismuth cooled fast reactor, and its safe operation is essential to the safety of the lead- bismuth cooled fast reactor. The flow characteristics of liquid lead-bismuth alloy in the pump have an important impact on the long-term safe operation of the pump. In order to study the flow field in an axial flow lead-bismuth pump, the impeller model of the main pump is constructed through Workbench/BladeGen software, and the flow field in the pump is simulated and analyzed in ANSYS CFX software. According to the numerical simulation results, the thickness of the guide vane is improved, and the wing outlet angle of the moving vane is optimized, thereby improving the flow field in the pump. The research results show that too fast angle change near the vane pattern outlet of lead- bismuth pump will lead to uneven vane pressure distribution and local high pressure, which may lead to more serious erosion. After optimizing the thickness of the guide vanes and the outlet flow angle of the moving vane, the overall trace of the flow field in the pump is relatively stable and the flow velocity of lead- bismuth at the exit of the guide vane maintained at about 1.8 m/s.

Numerical Simulation of Condensation in Supercritical CO₂ Compressor Based on Equilibrium Condensation Model

Chen Laijie, Lu Chuan, Shen Xin, Yi Jingwei, Li Yang, Ouyang Hua, Du Zhaohui

2022, 43(3): 165-172. doi: 10.13832/j.jnpe.2022.03.0165

Abstract(219) HTML (23) PDF(29)

Abstract:
Supercritical carbon dioxide (sCO₂) Brayton cycle is one of the main solutions of the Generation IV nuclear reactor energy conversion system. During the actual operation, sCO₂ in the compressor may condense. As a result, the efficiency is reduced and the operation stability is affected. In this paper, the equilibrium condensation numerical model of sCO₂ was established with the Span-Wagner model, and the numerical simulation of the sCO₂compressor was carried out. The main condensation regions, causes and influences of the condensation of sCO₂were analyzed. The results show that the condensation of SCO₂is mainly affected by the flow velocity. The condensation of sCO₂occurs in the area above 50% blade height on the suction surface of the leading edge of the main blade and near the pressure surface in the leading edge gap. The former area is caused by local acceleration of sCO₂, and the latter area is caused by tip clearance leakage; Under the given working condition, the condensation area is very small, which does not extend to the whole channel, the condensed sCO₂ is very little, and no two-phase flow is formed, which has little influence on the operation of the compressor.

Testing of Influence of Start-Stop on the Reliability and Drag Torque of the Thrust Bearing

Zhang Jianxin, Zhang Donghui, Gu Jipin, Guo Xiaoxian, Chen Shuming, Liu Xiaojun

2022, 43(3): 173-178. doi: 10.13832/j.jnpe.2022.03.0173

Abstract(257) HTML (41) PDF(22)

Abstract:
In order to test the effect of repeated start-stop on the reliability and drag torque of the thrust bearing of the primary circuit main pump in sodium cooled fast reactor (SFR), a reliability statistical scheme was designed by using the dispersion coefficient method suitable for small samples, three sets of Babbitt thrust pads and one upper combination bearing prototype were manufactured, a test bench was designed and built, the change of starting drag torque with shutdown loading time was tested, repeated start-stop tests were carried out by simulating the real situation of the thrust bearing. The results show that the starting drag torque increases with the increase of shutdown loading time. The repeated start-stop has little effect on the wear life of the thrust pads, and the reliability of the thrust bearing start-stop for 125 times without failure exceeds 0.99996 when the confidence level is 0.9. The repeated start-stop has effect on the drag torque of the thrust bearing, and the drag torque of the thrust bearing shows a slow upward trend with the increase of starting times. It is proved that the effect of the start-stop times must be factored in the design of starting capacity of main pump motor. This study can provide a reference for the design of starting capacity of main pump motor.

Degradation Trend Prediction of Nuclear-level Electric Valve Based on Hilbert-Huang Transform and BP Neural Network

Liu Jie, Zhang Lin, Wang Yunsheng, Yan Xiao, Zhan Li, Ou Zhu

2022, 43(3): 179-184. doi: 10.13832/j.jnpe.2022.03.0179

Abstract(221) HTML (161) PDF(27)

Abstract:
Due to the harsh service environment of nuclear-level electric valves, degradation and failure are easy to occur. Therefore, in order to accurately predict the performance degradation trend of nuclear-level electric valves, this study adopts a method based on Hilbert-Huang transform (HHT) and BP neural network (BPNN) combined method (HHT-BPNN) to predict the degradation state of nuclear-level electric valve. In this paper, the vibration signal of a nuclear-level electric valve reliability test is used to predict the degradation trend of the electric valve. The results show that the method can accurately predict the three degradation states of the nuclear-level electric valve, and the relative error is within the acceptable range. The analysis and research results show that HHT can effectively extract the degradation information of the signal, and BPNN can accurately predict the degradation trend of nuclear-level electric valves. The HHT-BPNN prediction method can effectively solve the difficulty of predicting the performance degradation of nuclear-level electric valves.

Research on Optimization of Post-72h Procedures for Passive Nuclear Power Plant

Shi Jin, Guo Donghai

2022, 43(3): 185-189. doi: 10.13832/j.jnpe.2022.03.0185

Abstract(153) HTML (79) PDF(27)

Abstract:
Post-72h procedures are special procedures for passive nuclear power plants. In order to evaluate the necessity and sufficiency of post-72 h procedures for passive nuclear power plants, and optimize and improve their weak links, this study analyzes the overall logic of the design basis and operating procedure system for passive nuclear power plants in demonstration projects. Based on the content and structure of the post-72 h procedures, this study proposes a general method for evaluating the necessity and sufficiency of the post-72 h procedures, the possible risk items in the process of procedure implementation are identified by using the logical block diagram of procedure implementation, and some optimization suggestions are put forward according to the rules of the demonstration project. The analysis results show that the post-72 h procedures are sufficient and necessary for the safe operation of passive nuclear power plants; their expressions are related to the procedure architecture; Clarifying the priority of procedure operation path and reducing procedure jump can improve the efficiency of procedure implementation. The relevant optimization suggestions can provide technical reference for the long-term safe operation of passive nuclear power plants after accidents.

Improvement and Application of Water Chemistry Control Process during Shutdown of One 3^rd Generation PWR Unit

Wang Zhu, Zhou Jia, Chu Jianwei, Li Xiaoning

2022, 43(3): 190-195. doi: 10.13832/j.jnpe.2022.03.0190

Abstract(361) HTML (87) PDF(61)

Abstract:
In order to effectively reduce the corrosion rate of reactor coolant system (RCP) materials of pressurized water reactor (PWR) units and effectively remove activated corrosion products, so as to reduce the exposure dose of workers to ensure the smooth progress of unit overhaul, a 3^rd Generation PWR unit uses enriched boric acid (EBA) for reactivity control, while taking advantage of its significant advantage of water chemistry control of RCP system coolant during power operation. At the same time, during the first overhaul of the unit, improvement measures are taken for the shutdown water chemistry control process (including the conversion from alkaline environment to acid environment, the conversion from reducing environment to oxidizing environment, the repeated addition of hydrogen peroxide to the primary loop during forced oxidation to maintain oxidizability, the maximum flow purification of the mixed bed of chemical and volumetric system, etc.), and the purpose of reducing the radiation dose of the unit and the radiation dose of workers is realized in the downward stage of unit shutdown.

Research on an Adaptive Method for Predicting the Remaining Life of the Roller Screw Pair of the Control Rod Drive Mechanism

Li Lin, Zhang Liming, Jiao Meng, Zhang Shuai, Wang Benmeng

2022, 43(3): 196-201. doi: 10.13832/j.jnpe.2022.03.0196

Abstract(229) HTML (49) PDF(33)

Abstract:
Aiming at the problem of how to select effective health indicators and reasonably construct the prediction model in the remaining useful life (RUL) prediction of roller screw pair of reactor control rod drive mechanism (CRDM), a new RUL prediction model of roller screw pair is proposed. The negative logarithmic likelihood probability (NLLP) index based on the generated topology mapping algorithm (GTM) is used as the health index of the roller screw pair, and the K-means clustering algorithm is used to divide the NLLP index. Using historical data and online monitoring data, an adaptive prediction model based on Markov model and least mean square algorithm (LMS) is constructed, and the remaining life is predicted according to the set threshold. Through experimental verification, the results show that the health status indicators selected in this paper can effectively reflect the equipment status. The prediction accuracy of the proposed adaptive prediction model is higher than that of the general prediction model, which provides a basis for the reasonable construction of RUL prediction model.

Research on Two-Dimensional Ultrasonic Contour Imaging Algorithm for Weld between Nozzle to Shell of Reactor Pressure Vessel

Zhou Lusheng, Deng Jingshan, Liu Yizhou, Sun Maorong

2022, 43(3): 202-206. doi: 10.13832/j.jnpe.2022.03.0202

Abstract(181) HTML (32) PDF(22)

Abstract:
The visual display of ultrasonic signal position of complex geometric welds in nuclear power plants has important reference value for defect judgment. The inner surface of the nozzle of reactor pressure vessel (RPV) in nuclear power plant usually has a certain inclination angle. By using the traditional rectangular B-scan imaging algorithm, there are some outstanding problems in ultrasonic B/C scan imaging of welds between nozzle and shell, such as unintuitive display and inaccurate defect location. In this paper, the algorithm of tangent connection between straight line and straight line, arc and elliptical arc through arc is proposed. The library function of drawing straight line and arc is used to draw the contour, and the ultrasonic signal is displayed in the contour to form a B-scan image with contour. A saddle shaped C-scan image is formed by calculating the threshold of the A-scan signal passing through the gate line in the contour. The validity and practicability of the algorithm are verified by the on-site ultrasonic scanning data of the RPV nozzle and shell weld in the nuclear power plant.

Positioning Error Analysis and Optimization of Floating Nuclear Power Plant Reactor Refuelling Based on Small Displacement Torsors

Wang Bingyan, Chen Shuhua, An Yanbo, Dong Dailin, Zhan Hui

2022, 43(3): 207-213. doi: 10.13832/j.jnpe.2022.03.0207

Abstract(130) HTML (40) PDF(20)

Abstract:
Fuel assembly loading and unloading is an important operation of nuclear power plant reactor refueling maintenance. Because of its special operating environment, the loading and unloading positioning accuracy of floating nuclear power plant is higher. Based on the tolerance modeling method of small displacement torsor (SDT), this paper analyzes the guidance and positioning error of reactor loading and unloading in floating nuclear power plant. Using the method of rigid body dynamic coordinate system transformation, the expression of loading and unloading positioning error is obtained; the simulation calculation is carried out using MATLAB program. The limit inclination of reactor loading and unloading under marine conditions is analyzed, and the relationship between the maximum inclination and wave parameters is given. The key parameters such as the guide clearance of the fuel assembly are optimized, and the results are compared with the experimental data, which are in good agreement.

Research on Corrosion Behavior of Stainless Steel Cladding Welding Simulator in Nuclear Power Plant

Wang Yuxin, Hu Yuefei, Gao Yu, Guo Chengxiang, Zuo Jinghui

2022, 43(3): 214-219. doi: 10.13832/j.jnpe.2022.03.0214

Abstract(301) HTML (35) PDF(33)

Abstract:
The corrosion behavior of the welding simulators of three stainless steel cladding materials S32205, S32101 and S30403 in spent fuel pool of nuclear power plant was studied by soaking them for 6 months under the conditions of H₃BO₃ concentration of 2500 mg/L, SO₄²⁻concentration of 1500 mg/L, Cl⁻ concentration of 5%, pH value of 5.0, temperature of 80 ℃ and saturated oxygen. The results show that a large number of chlorine induced stress corrosion cracks appear near the welding joints and gaps of S30403 welding simulator; S32101 welding simulator has corrosion pits, especially near the welding joints and gaps; S32205 welding simulator has the lightest corrosion, and no corrosion pits and cracks are found on the surface of the test piece. The research shows that the corrosion resistance law of the three material simulator is: S32205>S32101>S30403. S32205 has good comprehensive mechanical properties and corrosion resistance, is an ideal improved pool cladding material.

Research on Structural Response Characteristics of Research Reactor Fuel Assembly During Collision

Liu Menglong, Wang Haoyu, Zhou Yi, Zhu Fawen, Yuan Pan, Huang Shan, Deng Shuang

2022, 43(3): 220-225. doi: 10.13832/j.jnpe.2022.03.0220

Abstract(691) HTML (93) PDF(34)

Abstract:
The structural response characteristics of typical research reactor fuel assembly collision process are numerically simulated. The stress response characteristics of the fuel assembly, the bearing characteristics of the positioning structure of the fuel element and the protective effect of the filling cushion material on the collision of the fuel assembly during the collision are analyzed. It is found that there is an obvious stress concentration near the end of the fuel element during the collision, and increasing the cladding length at the end of the fuel element can avoid the effect of stress concentration on the core. The load borne by the end positioning structure of the fuel element is significantly greater than that in the middle, and strengthening the strength of the end positioning structure can improve the bearing capacity of the fuel assembly; the cushion material has a good energy absorption and protection function for the fuel assembly during the collision, and can smooth the change of kinetic energy in the collision and the structural vibration after the collision.

Research on Percolation Models of UO₂ Fission Gas Release

Li Wenjie, Qi Feipeng, Sun Dan, Xin Yong, Li Quan, Li Yuanming

2022, 43(3): 226-231. doi: 10.13832/j.jnpe.2022.03.0226

Abstract(907) HTML (110) PDF(37)

Abstract:
In order to analyze the dynamic process of intermittent release of fission gas caused by UO₂ fuel grain boundary bubble connection, so as to solve the problem that the radial release share predicted by the current diffusion model is inconsistent with the experimental measurement, a two-dimensional percolation model is used to simulate the evolution of UO₂ fuel grain boundary bubble network and the release process connected with the free space in the fuel rod. The results show that the percolation model predicts that the share of fission gas released along the radial direction of the pellet shows a local peak in the middle part of the pellet, and advances to the outside of the pellet with time, which is qualitatively consistent with the radial fission gas distribution phenomenon observed in the irradiation test under different burnups. Therefore, the percolation model established in this study can explain the radial non-monotonic distribution of fission gas release share of UO₂ fuel that could not be predicted by the previous percolation model from the mechanism.

Research on Use of a Graded Approach in Application of Emergency Preparedness and Response Requirements for Research Reactors

Yu Hong, Cheng Shisi, Liu Ting

2022, 43(3): 232-236. doi: 10.13832/j.jnpe.2022.03.0232

Abstract(625) HTML (59) PDF(28)

Abstract:
The current emergency management of research reactors in China dose not have different requirements for emergency preparedness and response of different types of research reactors. A graded approach is a good means of properly applying these requirements in light of the potential hazards associated with the reactor. According to the steps of the graded approach, based on China's current research reactor safety classification criteria, the power related emergency threat classification criteria of the International Atomic Energy Agency (IAEA) and the graded approach for applying IAEA emergency preparedness and response requirements, this paper puts forward the emergency management classification criteria of Chinese research reactors and the requirements for emergency state classification and emergency planning areas of research reactors with different emergency management categories, which provides a basis for simplifying the scope, extent and level of the content and details of emergency plans for low-power research reactor operating organizations, and for establishing an emergency management system for China’s research reactors that is commensurate with the hazard assessment results for different types of research reactors.

2022 Vol. 43, No. 3