核动力工程

Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor

Cao Liangzhi, Zou Xiaoyang, Liu Zhouyu, Wan Chenghui, Wu Hongchun

2021, 42(2): 1-15. doi: 10.13832/j.jnpe.2021.02.0001

Abstract(1815) PDF(305)

Abstract:
Numerical reactor based on high fidelity model and method is with the characteristics of high precision and high resolution, but the inherent uncertainty of nuclear data and other parameters will seriously affect the uncertainty of numerical reactor analysis results. Based on the review of the research progresses of numerical reactors and its uncertainty quantification in the world, this paper focuses on the research progresses in NECP Laboratory of Xi’an Jiaotong University in recent years, including the development of one-step high fidelity numerical reactor program NECP-X, the generation of nuclear data covariance database, the uncertainty propagation method based on deterministic method and sampling method, and the uncertainty quantification in transient calculation. An advanced sampling method COST is proposed. Based on the high fidelity numerical reactor program, the uncertainty propagation of covariance of various nuclear parameters in the steady-state and transient analysis of the reactor core is quantified for the first time, which is of great significance for engineering application of numerical reactors.

Experimental Study on Effect of Diffusion Angle on Turbulence in Venturi Tube

Shentu Yunqi, Song Yuchen, Yin Junlian, Yuan Hong, Wang Dezhong

2021, 42(2): 16-22. doi: 10.13832/j.jnpe.2021.02.0016

Abstract(311)

Abstract:
In order to study the effect of the diffusion angle on the turbulence in the venturi tube, in this work, the stereo-particle image velocimetry is used to measure the transient velocity of the diffusion part of venturi tube, which diffusion angle is 10°, 12.5°, 15°and 20°, respectively, and the time-averaged velocity is obtained. Based on the transient velocity of the diffusion part of the venturi tube, the distribution of turbulent kinetic energy in the diffusion section is obtained by statistical analysis of the instantaneous velocity field. It is shown that the average velocity in the diffusion section of the venturi tubes of different structures presented an axisymmetric single peak distribution in the direction of the section diameter, and the turbulent kinetic energy presented an axisymmetric double peak distribution in the section diameter. Flow separation can be observed in all experimental condition. With the increasing of the diffusion angle, the largest turbulent kinetic energy increases, but the radial width of the main flow zone does not change, and the radial width of the separation flow zone increases, but the effect on the proportion of the separation flow zone is small, and the high turbulent kinetic energy zone widens. With the increasing of Reynolds number, the peak value of turbulent kinetic energy increases, which is mainly caused by the axial Reynolds stress, and the proportion of separated flow area decreased slightly, but the distribution of main flow area and separated flow area changes little. This study provides an experimental basis for studying the flow field of the venturi tube with high Reynolds number at different angles.

Conceptual Design of a Reduced Moderation Long-Life Heavy Water Cooled Thorium Small Modular Reactor

Sun Qizheng, Wang Lianjie, Zhang Tengfei, Li Xiangyang, Liu Xiaojing

2021, 42(2): 23-28. doi: 10.13832/j.jnpe.2021.02.0023

Abstract(208)

Abstract:
This paper proposes a conceptual design for an innovative reactor, reduced moderation small modular reactor (RMSMR), based on the technology of conventional PWRs. In this paper we firstly establish a two-dimensional model to analyze the influences of fuel types, coolant types and other parameters of PWR and RMSMR. Then a three-dimensional design scheme is proposed and a preliminary neutronic and thermal-hydraulic analysis is carried out. The results show that the power distribution can be flattened by adopting a three-zone 232Th-233U dioxide fuel configuration, a negative void coefficient can be ensured by using uranium-thorium dioxide fuel, and the conversion effect can be enhanced by utilizing blanket regions. RMSMR utilizes the small scattering cross section of heavy water to yield an epithermal-to-fast neutron spectrum, which is also beneficial for attaining a large conversion ratio, a long core lifetime and reducing the burnup reactivity swing. It is concluded that RMSMR can sustain the power generation of 100 MWe for 6 years without refueling, which is of referential value to the design of new generation reactors.

Study on Separation Characteristics and Mechanism of a New Swirl Vane Separator

Qian Yalan, Yang Lingfang, Zhang Tingting, Yin Junlian, Wang Dezhong

2021, 42(2): 29-34. doi: 10.13832/j.jnpe.2021.02.0029

Abstract(221)

Abstract:
In order to obtain the bubble separation behavior in the swirl vane separator of the molten salt reactor quantitatively, the liquid flow field distribution inside the separator was obtained by computational fluid dynamics method. Based on the bubble separation model that coupling the vortex model and interphase force, the numerical calculation program of bubble motion was developed to quickly predict the critical separation diameter of the bubble in the molten salt medium in the separator. By quantifying the force acting on the bubble, the separation mechanism of the bubble has been demonstrated. The analysis indicates that the interphase force of the bubble closely relates to the position of different radius in the separator. The virtual mass force and drag force acting on the bubble in the axial direction determine the bubble separation length. When pressure gradient force, lift force, drag force as well as virtual mass force acting on the bubble reach a balance in the radial direction, it can no longer centripetally move into the air core and be captured and separated.

Investigations on Supercritical CO2 Critical Flow through Mini Tubes

Wang Junfeng, Wang Yangle, Zhou Yuan, Huang Yanping

2021, 42(2): 35-38. doi: 10.13832/j.jnpe.2021.02.0035

Abstract(148)

Abstract:
In this paper, an experimental supercritical CO2 critical flow investigation was carried out. The effect of upstream stagnation parameters and length to diameter ratio on the critical flow rate was obtained. A prediction correlation of mass flow rate equation was proposed based on the experimental data. The comparisons of correlation prediction and experimental data from open literature show good agreement. The deviation is within the boundary of ±15%. It shows that the present correlation can predict the supercritical CO2 critical flow rate.

Development of 3-D Parallel First Collision Source Module

ang Chao, Yu Tao, Deng Li, Cheng Tangpei

2021, 42(2): 39-42. doi: 10.13832/j.jnpe.2021.02.0039

Abstract(100)

Abstract:
The computational efficiency of the first collision method is low for large-scale mesh problems, and meanwhile there is a large calculation error in the near source region. In order to solve these problems, 3-D parallel first collision source module is developed in this paper. The spatial domain partition algorithm is adopted to improve the computational efficiency, and the adaptive source region mesh refinement technique is performed to decrease the calculation error in the near source region. The results show that the parallel computing improves the computational efficiency. The parallel efficiency can reach 66.89% when 140 cores are used. The adaptive source region mesh refinement technique reduces the calculation error in the near source region.

Research on Ring Reactivity-Equivalent Physical Transformation Method for Homogenization of Double-Heterogeneous Systems

Lou Lei, Peng Xingjie, Chai Xiaoming, Yao Dong, Li Mancang, Yu Yingrui, Wang Lianjie

2021, 42(2): 43-49. doi: 10.13832/j.jnpe.2021.02.0043

Abstract(144) PDF(53)

Abstract:
Because of the double heterogeneity of the dispersed particle-type fuel and the burnable poison, the traditional volumetric homogenization method (VHM) on double-heterogeneous (DH) systems will bring unacceptable calculation deviation. The reactivity-equivalent physical transformation (RPT) method applied to the DH systems with particle-type dispersed fuel has the features of simplicity and high calculation accuracy. In this paper, the traditional RPT and IRPT methods have been analyzed and verified, and it suggests that for some DH systems with dispersed particle-type burnable poison with large absorption cross section, these two methods are with relatively large calculation deviation. Ring RPT (RRPT) method and Two-step ring RPT (TRRPT) method are proposed to deal with the DH systems of one and two types of particles. Results of depletion calculations for different types of dispersed particle-type fuel and burnable poisons and the comparison with Mote Carlo results of grain models prove the validity of RRPT method and TRRPT method. The preliminary conclusion can be obtained that Ring RPT method has wider scope of usage and higher accuracy than the traditional RPT method.

Fine CFD Simulation Study of Whole Core for HPR1000 Reactor Vessel

Bi Shumao, Liu Yu, Liu Luguo, Xu Youyou, Deng Jian, Miao Yifei, Wu Lingyan

2021, 42(2): 50-54. doi: 10.13832/j.jnpe.2021.02.0050

Abstract(327)

Abstract:
Fine and large scale CFD simulation of the whole reactor core is an important methodology in the research and design of HPR1000 and digital reactor. Based on a series of reasonable simplification of the core, the geometry model of the whole core were built for HPR1000. Meanwhile, the fuel assembly is discretized in group and the CFD model of the whole core is obtained. Besides, the distribution of flow field and the parameters of thermal hydraulics are obtained by the fine and large scale CFD simulation for the whole core. The results verify the rationality of design parameters of reactor core of HPR1000 and provide reference to the optimal design and safe operation of reactor. The results show that: due to the combined action of the 1/4 symmetrical structure of the HPR1000 reactor core and the 1/3 symmetrical structure of the coolant inlet and outlet of "three in and three out", the flow distribution factor of the core first increases and then decreases in the radial direction, and the maximum flow is not in the center of the core; the maximum temperature of the fuel assembly on the cross section of the inlet nozzle is about 331.2℃, and the temperature distribution is asymmetrical. The maximum temperature region is not in the center of the core, which is similar to the trend of the core flow distribution factor. It is the result of the joint action of the core power distribution and the coolant flow distribution.

Analysis of Enhanced Convection Heat Transfer of Coaxial Cross Triple Twisted Tape for Single Phase Fluid

Liu Xiaoya, Zhang Yongfa, Yang Zimu, Ding Ming

2021, 42(2): 55-59. doi: 10.13832/j.jnpe.2021.02.0055

Abstract(143)

Abstract:
A large number of heat exchangers are used in the passive safety system of nuclear power plants. Under the condition of natural circulation, the flow of the fluid in the heat exchanger is weak, that may be in laminar flow and leads to the low heat transfer coefficient. In order to enhance the heat transfer, the new twisted tape named coaxial cross triple twisted tape is presented. The heat transfer effect of the water in the tube fitted CCTTT with different twist ratio (y = 2,3,4,∞) is studied by numerical simulation. When the Reynolds number changes from 40 to 1200, the results show that CCTTT can effectively enhance the heat transfer of single phase laminar flow in the tubes. Moreover, with the decrease of twist ratio, the heat transfer effect and the performance evaluation criteria (PEC) increased. The PEC number of CCTTT with twist ratio of 4.0 is 2.02, while that of CCTTT with twist ratio of 2.0 is 2.64. Compared with other working fluids, at the same twist ratio, the PEC number of CCTTT increases with the improve of Prandtl number.

Experimental Investigation of Counter-Current Flow in Downcomer of CAP1400

Fei Likai, Zhang Peng, He D, an, Yuan Haowei, Hu Fuquan, Cui Lei, Shen Feng, Liu Lifang

2021, 42(2): 60-64. doi: 10.13832/j.jnpe.2021.02.0060

Abstract(325)

Abstract:
In order to study the phenomenon of counter-current flow in the downcomer of the CAP1400 reactor under real working conditions, a test bench with the height and diameter ratio of 1:1 of the downcomer of the pressure vessel and a 60° slice was built using CAP1400 as a prototype. The test working medium was air and water, and the two-phase gas-liquid flow and ECC bypass phenomenon with different mass flow rate and gas flow rate were experimentally studied. The results show that the gas-liquid reverse flow (CCFL) phenomenon is obvious with the increasing of gas injection rate, and when the gas volume reaches 4 kg/s or more, the injection water cannot be in the core completely. In addition, comparing the experimental results with the theoretical formula of CCCL, Kutateladze empirical correlation and UPTF empirical correlation could not accurately describe the CCFL phenomena in the downcomer of CAP1400. Therefore, based on the experimental data, the new correlations were fitted. The comparison of the test data with and without U-type reflector shows that the U-type reflector can reduce ECC bypass water and enhance the injection ability.

Flow Instability in Parallel Narrow Rectangular Channel under Different Radial Heat Distribution

Wang Yanlin, Zhou Lei, Zan Yuanfeng, Yang Zumao, Yan Xiao

2021, 42(2): 65-68. doi: 10.13832/j.jnpe.2021.02.0065

Abstract(136)

Abstract:
The flow instability in the parallel channels under different radial heat distribution was investigated in the following condition: the system pressure is 0.89~1.32 MPa, the mass flux is 500-750 kg/(m2?s) and the inlet temperature is 58.5℃~132.3℃ with the upflow deionized water. The test section consists of 5 parallel tube channels with the dimension of 1400 mm×Φ8 mm×2 mm. The characteristics of the flow instability in the parallel channels and the flow instability boundary was studied based on the experimental results. The results show that the distribution of the radial heating heat flux has no obvious effect on the unstable boundary of the parallel multi-channel flow, when dealing with the parameters of unsteady flow in the hot channel. For the parallel multi-channel structure with different heat flux distribution, the flow instability boundary is the same as that of the parallel multi-channel structure with complete symmetric heating.

Study on Physical Characteristics of Mid-Beryllium Control Rod of HFETR

Kang Changhu, Liu Shuiqing, Chen Qibing, Xiang Yuxin, Zou Peng, Song Jiyang, Song Yuge, Liu Wenbin

2021, 42(2): 69-71. doi: 10.13832/j.jnpe.2021.02.0069

Abstract(96)

Abstract:
In this paper, the physical characteristic of the mid-beryllium control rod of HFETR is studied. Firstly, the few group cross section of every assembly is calculated by CELL code. Secondly, the core physical calculations of the setting up control rod cell and mid-beryllium control rod cell are carried out respectively, and the axial thermal flux distribution, the isotope yield and the control rod worth are compared. The result shows that the mid-beryllium control rod can be used for the reactor reactivity control and can improve the reactor’s safety and economy.

Numerical Analysis of Density Wave Instability Phenomena of Supercritical Water in Two Parallel Channels

Zang Jinguang, Yan Xiao, Huang Yanping

2021, 42(2): 72-76. doi: 10.13832/j.jnpe.2021.02.0072

Abstract(183)

Abstract:
The flow instability of supercritical water is an important aspect of thermal-hydraulic design of Supercritical Water-cooled Reactor (SCWR). The validated numerical simulation method helps to obtain the internal mechanism of flow instability, a wider instability boundary, and further deepen the understanding of supercritical water flow instability. Based on the experimental data of supercritical water instability that has been carried out, this paper carried out the verification of the calculation method; on this basis, using the system analysis program as a tool, the flow instability phenomena of supercritical water in parallel channels were systematically studied. In this paper, the inflection point of the influence of the supercritical water inlet temperature is obtained; the instable threshold power increase with the increase of the operating pressure at the same inlet temperature and the working fluid under the supercritical condition has better stability characteristics; the suitability of the dimensionless criterion is evaluated under supercritical conditions.

Formation Mechanism for Turbulence Induced Disturbance Wave on Liquid Film of Annular Flow

He Hui, Ren Quanyao, Ye Tingpu, Wu Yao, Pan Liangming

2021, 42(2): 77-81. doi: 10.13832/j.jnpe.2021.02.0077

Abstract(177)

Abstract:
To explore the formation mechanism of the disturbance wave from the perspective of accurately predicting the properties of the liquid droplet/film mass transport and dryout scenario, the high-speed camera and conductivity-based techniques are employed to visualize and measure the evolution behaviors of the interfacial waves and the liquid film thickness respectively. The millisecond-scale ripple waves induced by turbulent gas as precursors experience a series of second-scale self-convolution and evolve the random distribution of disturbance waves, and the time interval between the disturbance waves can be predicted statistically by the gamma distribution with an order n which increases with the gas superficial velocities but appears to be independent of the liquid superficial velocity.

Theoretical Study of Bubble Departure Diameter in Subcooled Flow Boiling

in Di, Xiong Jinbiao, Cheng Xu

2021, 42(2): 82-87. doi: 10.13832/j.jnpe.2021.02.0082

Abstract(170)

Abstract:
Bubble departure diameter model is an important sub-model for wall boiling calculation. In order to predict the wall heat transfer accurately, this paper models the bubble departure diameter in subcooled flow boiling by using the force balance method with a new improved bubble growth model. Bubble growth model considers the contribution of microlayer, superheated layer and subcooled liquid layer at the top of bubble to bubble growth. It is validated through the comparison with the saturated boiling and subcooled boiling experiments, and the prediction results show good agreement with the experimental data. Three databases for subcooled flow boiling are selected to validate the bubble departure diameter model, and the predictive accuracy is in acceptable error range.

Experimental Study of Single-Phase Flow Characteristics in a Narrow Rectangular Channel with Vertical upward Flow

Song Mingliang, Ma Jian, Huang Yanping, Zhang Yu, Lin Zhenxia

2021, 42(2): 88-92. doi: 10.13832/j.jnpe.2021.02.0088

Abstract(153)

Abstract:
In this paper, a vertical upward narrow gap rectangular channel is taken as the research object, and the experimental study of single-phase flow characteristics under constant heat flux density of heating conditions is carried out. According to the measured temperature, flow rate, pressure drop and heat flux density, the experimental data of non-isothermal friction factors including laminar, transitional and turbulent flows in a certain range of working conditions are obtained, and the existing empirical correlations are compared and evaluated with the experimental data. The result shows that the predictions of laminar flow friction factor proposed by Kays and Clark and the predictions of turbulent flow friction factors proposed by Blasius, Techo and Moody respectively have a good agreement with the relevant experimental data.

Research on Design Method of HPR1000 Emergency Residual Heat Removal System Based on System Engineering Method

Chen Guocai, Li Feng, Tang Huapeng, Qiu Zhifang, DengJian

2021, 42(2): 93-98. doi: 10.13832/j.jnpe.2021.02.0093

Abstract(123)

Abstract:
Based on the system engineering, the system design is conducted for the emergency core residual removal system of HPR1000 nuclear power plant. Considering the requirements of safety, the economic efficiency and technology maturity, the systematic evaluation index system is constructed with the objective of optimizing the engineering application and overall technical index of the nuclear power plant. The optimal emergency residual heat removal system is obtained using AHP method. The results show that it is a optimal solution to eliminate the turbine auxiliary feedwater and extend the function of PRS to deal with the operation events and design benchmark accidents.

Study on Design of Semi-Physical Simulation System for Xi’an Pulsed Reactor

Zhang Liang, Yuan Jianxin, Zhao Wei, Wang Baosheng, Zhang Qiang, Zhu Guangning, Yang Ning, Chen Lixin, Jiang Xinbiao

2021, 42(2): 99-104. doi: 10.13832/j.jnpe.2021.02.0099

Abstract(155)

Abstract:
A semi-physical simulation system of Xi’an Pulsed Reactor is developed for the purpose of providing the analog signal source, verifying the power regulation method and personnel training for the digital instrument and control system of Xi’an pulse reactor under construction. The design philosophy of semi-physical simulation system is proposed and the system frame is presented. The core physical simulation model is improved and the real-time core simulation program is coded by MATLAB. The human-machine interface is programmed by the software Kingview, and the rod control and rod position measurement are carried out by programmable controller S7-200. A series of hardware equipments, including the control rod driving mechanism simulator, the signal generator and the manual operation board, are developed. And the internal communication of the system is established. The change of the core power when the control rod is raised or lowered, and the pulse parameters after the pulsed rod is emitted are simulated with the semi-physical simulation system. The results are in good agreement with the experimental data on the reactor. The output of the signal generator is measured and the consequence is in accordance with the expectation. It is shown that the semi-physical simulation system can realize the design purpose and has good performance.

Improvement and Test Verification of Reactor-Core Neutron and Temperature Detector

Ma Xiaoyu, Deng Tao

2021, 42(2): 105-109. doi: 10.13832/j.jnpe.2021.02.0105

Abstract(217)

Abstract:
The reactor-core neutron and temperature detector is an integrated detector integrating rhodium self powered neutron detector and thermocouple thermometer. The detector can simultaneously measure the neutron fluence rate of the core and the outlet temperature of the fuel assembly. This paper focuses on the design improvement and process improvement plan in the process of detector development, which would significantly improve its reliability under accident conditions. It is verified through experiments that the improved scheme can meet the requirements of the detector technical specifications.

Development and Application of Fuzzy Switching Control System for Nuclear Reactor Core

Jiang Qingfeng, Zeng Wenjie

2021, 42(2): 110-114. doi: 10.13832/j.jnpe.2021.02.0110

Abstract(190)

Abstract:
In order to make use of the advantages of different types of controllers, based on the fuzzy multi-model of reactor core, a fuzzy switching controller is designed by using PID controller and fuzzy controller with the T-S fuzzy rules. Taking the pressurized water reactor core of the Three Mile Island nuclear power plant as an example, a fuzzy switching control system for reactor core power is developed and the simulation studies are carried out. The results show that the designed core fuzzy switching controller is more suitable for the core power control under the core reactivity step disturbance and core inlet coolant temperature step disturbance than the traditional PID controller.

Design and Verification on Nuclear Safety Class Digital Instrument Control System Based on FPGA

Ma Xiaoyu, Huang Xiaojin, Wang Dong

2021, 42(2): 115-120. doi: 10.13832/j.jnpe.2021.02.0115

Abstract(194)

Abstract:
The common cause failure(CCF) due to the software design of the digital safety class instrument and control (I&C) system based on micro-processor technology for NPP is resolved by the diversity measures implemented based on FPGA, which could prevent the failure of the reactor protection system (RPS) when the Anticipated Transients Without Scram (ATWS) occurs. This paper presents the design of overall system, platform hardware and platform logic implemented with FPGA technology. The design is further validated for its capability and performance by RPS system of a nuclear power plant, demonstrating that the DCS system in this study is consistent to the requirement of digital safety I&C for NPP on feasibility, applicability and reliability.

Study on Temperature Field of Control Rod Drive Mechanism Based on Natural Convection

Li Wei, Peng Hang, Yu Ti, a, Fu Guozhong, Wu Hao, Tang Yuan, Tang Jiankai

2021, 42(2): 121-124. doi: 10.13832/j.jnpe.2021.02.0121

Abstract(186)

Abstract:
After the improvement of the temperature resistance level of the control rod drive mechanism (CRDM) of the reactor, in order to analyze whether it is feasible to eliminate the forced ventilation system of CRDM, a calculation model of the airflow field intended for single CRDM is established to determine the calculation method and boundary condition, which is conducted by fitting with the experiment results. The temperature field and the airflow field of the full-scale CRDM clusters are calculated, and the temperature field effect of the full-scale CRDM clusters is obtained based on the natural flow field. The research results reveal that the natural convection could take the heat generated by the CRDM clusters out of the reactor pit without pit fever effect. Moreover, as the CRDM clusters output power increases, the natural circulation intensity increases as well. The natural convection could cool down the CRDM coils. Thus, the forced ventilation system could be eliminated with the improvement of the CRDM temperature resistance level.

Research on Accuracy of Hydraulic Performance Calculation under Full Working Conditions of a Nuclear Main Pump Based on Detached Eddy Simulation

Wang Xiuyong, Liu Zhiyuan, Liu Yuning, Li Yibin

2021, 42(2): 125-130. doi: 10.13832/j.jnpe.2021.02.0125

Abstract(212)

Abstract:
In order to study the effect of the detached eddy simulation model on the accuracy of hydraulic performance prediction for a nuclear main pump, with the hexahedral structured grid, the detached-eddy simulation (DES), delayed detached-eddy simulation (DDES) and improved delayed detached-eddy simulation (IDDES) based on SST κ-ω, were used to carry out the unsteady numerical simulation under full working conditions, and the calculation results were compared with that of RNG κ-ε model. The calculation accuracy of the four turbulence models was comprehensively evaluated from the aspects of relative calculation error and its dispersion. The results show that under full working conditions, the comprehensive calculation accuracy of each detached eddy simulation model is much higher than that of the RNG κ-ε model. The calculation accuracy of DDES and IDDES is almost the same under full working conditions, and is higher than that of DES model. The calculation accuracy of DDES near the design operation point is obviously better than that of IDDES. In comparison, the DDES model is more suitable for the performance prediction of a nuclear main pump.

Thermal-Fluid-Solid Coupling Characteristics of a Circular Tube in Two Degrees of Freedom Flow-Induced Vibration

Ding Lin, He Haoyu, Yang Zuomei, Zhang Li, Yang Lin

2021, 42(2): 131-136. doi: 10.13832/j.jnpe.2021.02.0131

Abstract(98)

Abstract:
This paper studies the fluid-induced vibration-heat transfer coupling characteristics of a single circular tube with constant temperature, and the variation rules of the cylinder vibration response, temperature field, average Nusselt number and maximum local Nusselt number are analyzed. The results show that when Reynolds number Re=100, mass ratio mr=1.37, and reduced velocity Ur=5, the cylinder vibrates in a clockwise "8" shape, and the transverse amplitude is much larger than the downstream amplitude. The vortex shedding mode is 2S, and the vortex shedding will cause the local Nusselt number to increase abruptly near the rear stagnation point. In this study, the Nusselt number of the vibrating cylinder was significantly larger than that of the fixed cylinder, and the average Nusselt number increased by 5.73%.

Design and Simulation Verification of Boron Concentration Algorithm of Boron Meter

Zheng Junwei

2021, 42(2): 137-143. doi: 10.13832/j.jnpe.2021.02.0137

Abstract(176)

Abstract:
In order to solve the problem of the deviation of the total boron concentration of the coolant of nuclear island primary circuit measured by the off-line boron meter (OFBM) from the chemical titration boron concentration in the primary coolant in pressurized water reactor nuclear power plant exceeds the standard, the boron concentration algorithm of the OFBM was analyzed, and the impact factors and causes of the boron concentration deviation were analyzed. The analysis results show that the main impact factors of the boron concentration deviation were the 10B concentration and the 10B abundance of the nuclear island primary circuit coolant; the main reason of the boron concentration deviation generated was the OFBM boron concentration algorithm ignoring the 10B abundance change in the primary circuit coolant of the nuclear island. A new boron concentration algorithm capable of tracking the 10B abundance of the coolant of the nuclear island primary circuit was designed at the same time, and the 10B concentration calculation function was implemented in this new algorithm. The simulation model of the new boron concentration algorithm was built finally, and the calculation accuracy of the new boron concentration algorithm was verified based on the calibration data of OFBM of two M310 units. The simulation results show that the calculation accuracy of the new boron concentration algorithm meets the OFBM design specifications.

Development of Ultrasonic Inspection System for Control Rod Guide Split Pins in Nuclear Power Plants

Ma Guanbing, Ding Hui, Wang Weiqiang, Yan Jingli, Ma Chao, Wang Bin

2021, 42(2): 144-147. doi: 10.13832/j.jnpe.2021.02.0144

Abstract(230)

Abstract:
The control rod guide split pin is a key connecting part of the nuclear power plant reactor internals. Stress corrosion cracking is an important potential danger that affects the quality of the split pin after long-term operation. In order to achieve an efficient and accurate inspection of the control rod guide split pin, a multi-chip integrated card-shaped ultrasonic probe is designed in this paper, and a system suitable for automatic underwater non-destructive testing is established. The results show that the split pin inspection system is with high detection efficiency and task adaptability, which provides a basis for field application inspection.

Research on PA Ultrasonic Inspection Technology in RPV

Hu Chenxu, Sun Jiawei, Li Bingqian

2021, 42(2): 148-152. doi: 10.13832/j.jnpe.2021.02.0148

Abstract(156)

Abstract:
By building the RPV model, the phase array (PA) ultrasonic detection technology was developed to replace the conventional ultrasonic detection technology, and then the detection and quantitative capability of the two detection technologies were compared and evaluated. The verification tests show that the detection and quantification capability of PA ultrasonic detection technology meets the standard requirement. At the same time, it can greatly shorten the inspection time and have considerable economic benefits.

Study on Key Parameters of the Selection of Steam Piping Traps in Nuclear Power Plant

Liu Chongyu, Duan Zhengqiang, Wu Wei

2021, 42(2): 153-156. doi: 10.13832/j.jnpe.2021.02.0153

Abstract(125)

Abstract:
In order to gain quickly the quantity of the drain of the nuclear power plant steam pipe, by calculating the quantity of the drain of the steam pipe in the start period and normal operation period, the heat dissipation picture of steam pipes with various piping specification are analyzed and drawn. By the heat dissipation picture, the quantity of the drain of the steam pipe in nuclear power plant can be obtained quickly when selecting the steam traps. To extend the life and reduce the cost of traps, this paper analyzes the influence of trap’s different safety factors (k) on its performance, and it recommends the suitable safety factor of traps used in the steam pipe of the nuclear power plant. It can improve the utilization efficiency of steam and economic benefits, and prolong the life of traps and reduce the cost of manufacturing of traps.

Application of Bayesian Estimation Method in Parameter Estimation of Alpha Factor Model

An Jin, Yan Lin

2021, 42(2): 157-160. doi: 10.13832/j.jnpe.2021.02.0157

Abstract(171)

Abstract:
The probabilistic safety analysis (PSA) results of the nuclear power plant shows that common cause failures (CCF) plays an important role in the system reliability. In China, generic data are often used for common cause failure data in PSA. It is difficult to reflect the operation characteristics of the nuclear power units. The alpha factor model is the most commonly used model in PSA to model the common cause failure due to its simplified parameter estimation form and accurate calculation results. However it is difficult to obtain reasonable statistical values with the classical estimation algorithm due to the rarity of common cause failure events. So the paper introduces the Bayesian estimation algorithm. The method can combine the prior information and sample information to obtain a better estimation without a large sample. The problem of lack of common cause failure events in nuclear power plants and unreasonable calculation results by classical estimation method are solved effectively. The method is suitable for the parameter estimation of common cause failure model in nuclear power plants.

Simulation Research on Dynamic Characteristics and Small Break Fault of Once-through Steam Generator

Xu Yu, Huangfu Zeyu, Xu Jianqun, Tian Peiyu, Li Chunmei, Cheng Xiang, Yan Siwei, Liao Xianwei

2021, 42(2): 161-167. doi: 10.13832/j.jnpe.2021.02.0161

Abstract(368)

Abstract:
Taking a once-through steam generator designed by Babcock & Wilcox as the study object, a graphical model was established, based on the basic module of supporting platform APROS. On this basis, the steady-state and dynamic simulation were carried out under varying operating conditions. According to the results, the enthalpy of the primary inlet and the pressure of the secondary outlet had the greatest influence on its steady-state characteristics, and the temperature of the primary inlet had the greatest influence on its dynamic characteristics. The impact of the position and degree of the breakage on the operating characteristics of the once-through steam generator when the heat exchange tube ruptures was studied further. The results show that it has the largest influence on the operation of once-through steam generator when the rupture takes place in the position near by the primary side’s inlet. And the more the leakage mass flow, the more influence on the operation characteristics of the once-through steam generator.

Research of UO2-(Zr0.8Ca0.2)O1.8 FCC Solid Solution Spheres Preparation

Long Dijun, Zeng Cheng, Lu Zhangxian, Li Jia

2021, 42(2): 168-172. doi: 10.13832/j.jnpe.2021.02.0168

Abstract(155)

Abstract:
In order to develop the single - phase ceramic fuel with good radiation resistance, the UO2-(Zr0.8Ca0.2)O1.8 microspheres with different uranium contents (30mol%, 50mol%, 70mol%) were prepared by Inner-Sol-Gel technique. The technique is with five processes including sol preparation, casting and gelation, washing, drying, and sintering. The parameters of this technique were fixed on the basis of analysis and experiment. The UO2-(Zr0.8Ca0.2)O1.8 spheres were analyzed by XRD. The results show that UO2-(Zr0.8Ca0.2)O1.8 microspheres with different uranium contents (30mol%, 50mol%, 70mol%) were FCC solid solution.

Qinshan-I Reactor Shielding Simulation and Sensitivity Analysis Based on JMCT Monte Carlo Code

Deng Li, Li Rui, Ding Qianxue, Qiu Youheng, Li Gang, Fu Yuanguang, Shi Dunfu, Liu Peng

2021, 42(2): 173-179. doi: 10.13832/j.jnpe.2021.02.0173

Abstract(324)

Abstract:
Qinshan-I reactor has been operated for over thirty years. A great amount of data is obtained from the design, building, operation, life-extension and decommission and can provide the valued reference for the design, building, operation and maintenance of the nuclear power plants with new generation reactors, such as HPR1000. In this paper, the test verification for JMCT Monte Carlo particle transport code is carried out based on the test data of Qinshan-I base mental section, weld section and main pipe. The simulated results show that JMCT code is with high precision in radiation shielding simulation. The resolution of JMCT is better than the commercial code for about 1~2 order.

Research on Contact Shape Function and Contact Model of Planetary Roller Screw Drive Pair

Liu Jia, Peng Hang, Luo Ying, Zhang Yixiong, Zhu Zihao, Yan Dapeng, Deng Qiang

2021, 42(2): 180-182. doi: 10.13832/j.jnpe.2021.02.0180

Abstract(86)

Abstract:
Based on the static contact state of planetary roller screw drive pair, a contact characteristic model and contact force analysis method are proposed by the shape function analysis of the drive pair and Hertz contact theory. The results of finite element analysis show that the method is reasonable, and the contact profile shape function in the contact surface perpendicular to the radial direction satisfies the quadratic function characteristics, and the Hertz contact theory can be used to study the static contact characteristics of the drive pair.

Study on Online Monitoring System Technology for Safety Instruments in Nuclear Power Plants

Zhao Yao, Huo Yujia, Wang Jun, Yu Junhui

2021, 42(2): 183-187. doi: 10.13832/j.jnpe.2021.02.0183

Abstract(137)

Abstract:
At present, in the nuclear power plants in China, the periodic calibration is often adopted to manage the drift of safety instruments, however it is too conservative and uneconomical. Thus, this paper studies the technology of online monitoring system for safety instruments. First of all, the actual drift data of the safety instruments is analyzed, which defines the main types of the drift of safety instruments in nuclear power plants, and proves the feasibility of applying online monitoring to safety instruments. Secondly, through the analysis and research of related regulations and standards, the basic requirements of online monitoring technology for safety instruments in nuclear power plants were clarified. Finally, the data analysis research of the online monitoring system technology was carried out, an equivalent average algorithm was proposed for redundant instruments, the non-redundant instruments algorithm was analyzed, and the multivariate state estimation mode (MSET) method was verified by modeling based on the actual data of nuclear power plants, which proved the feasibility of this method in nuclear power plants.

Preliminary Study of Influencing Factors of Power-to-Mass Ratio of Heat Pipe Reactor Power System Based on Simple Open Brayton Cycl

Wang Jinyu, Yu Hongxing, Zhang zhuohua, Ma Yugao, Chai Xiaoming, Chen Wei, Yi Jingwei, Zeng Chang, Su Dongchuan, Xiao Cong

2021, 42(2): 188-192. doi: 10.13832/j.jnpe.2021.02.0188

Abstract(210)

Abstract:
The heat pipe reactor system based on the simple open Brayton cycle is with the characteristics of simple structure, inherent safety, and low risk of radioactive leakage, and is a potential leading technology option to meet the demand for more practical and innovative nuclear reactor designs capable of providing clean energy. The power-to-mass ratio of the heat pipe reactor is often used as an important indicator in the evaluation of engineering design. In this paper, we established a model for the power-to-mass ratio of the heat pipe reactor to explore the effect of various key parameters on the power-to-mass ratio. The analysis of a design like the Westinghouse eVinci? heat pipe (HP) reactor shows that the power-to-mass ratio varies versus temperature difference in the heat transfer path, including difference between the heat pipe and the core structure, and that between the heat pipe and the heat exchanger. The optimal power-to-mass ratio is positively correlated with the maximum operating temperature limit of the core structure material. In the future, more detailed modeling of compressors, turbines, heat pipes could be carried out to improve the accuracy of models.

Research and Design of Flow Measurement Tests of Reactor Coolant System

Huang Zongren, Wang Mingli, Li Feng

2021, 42(2): 193-196. doi: 10.13832/j.jnpe.2021.02.0193

Abstract(177)

Abstract:
The reactor coolant pump inlet and outlet differential pressure gauge was set in HPR1000 to measure the loop flow of the reactor coolant system, and the elbow flowmeter set in the second generation progressive nuclear power unit was eliminated. The change of the flow measurement means directly affects the implementation of the flow measurement tests of RCS system. By studying the operation characteristics of the main pump and the resistance characteristics of the system, a test method based on the electric power of the main pump to measure the flow of RCS system was proposed. Combined with theoretical analysis results and engineering practice experience, the test method and acceptance criteria of the reactor coolant idling flow test were given. The research shows that the main pump electric power method can measure the flow of RCS system, and the reactor coolant idling flow can be verified by the speed change of the main pump during the idling process.

Preliminary Neutronics Analysis for Fluorine-Salt-Cooled Pebble-Bed Reactor

Li Zhifeng, Zhao Changyou, Zhang Guangchun, Han Song, Huang Jie

2021, 42(2): 197-201. doi: 10.13832/j.jnpe.2021.02.0197

Abstract(176)

Abstract:
In order to obtain the key neutronics parameters of the pebble-bed fluoride-salt-cooled high-temperature reactor (PB-FHR), the random packing method is used to obtain the coordinates of the three layers of isotropic coated particles (TRISO) particles, and the discrete element method is adopted to the coordinates of the fuel pebbles within the active core. Then the neutron transport calculated is conducted by the Monte Carlo code to analyze the effects of the fuel random distribution on the core neutronics parameters. The calculation results showed that the effects of the random distribution on the multiplication coefficient, group cross section and active core fuel pebble power can be neglected. Based on the above results, the nuclear design of the pebble-bed fluoride-salt-cooled high-temperature reactor can be simplified.

Research on Hybrid Monte-Carlo-Deterministic and Weight-Window Mesh-Coarsening Method Based on NECP-MCX

Zheng Qi, Shen Wei, He Qingming, Li Jie, Cao Liangzhi

2021, 42(2): 202-207. doi: 10.13832/j.jnpe.2021.02.0202

Abstract(190)

Abstract:
In the Monte Carlo simulation of a deep-penetration problem, only a small number of particles can penetrate the shielding layer and reach the target region, resulting in a very low computational efficiency. In order to solve the deep-penetration problem, the hybrid Monte-Carlo-Deterministic method is studied based on the consistent adjoint driven importance sampling method (CADIS). The hybrid method can automatically generate the input parameters required for deterministic SN calculation from the constructive solid geometry in the Monte-Carlo modeling. The adjoint SN calculation is used to generate consistent source biasing and weight-window parameters for forward calculation of the Monte-Carlo method. On the other hand, the mesh-based weight window applied in the hybrid method will encounter memory bottleneck in large-scale problems. A new structure of nested mesh is developed for mesh coarsening to save the memory of the weight-window parameters. The coarse mesh does not affect the variance reduction effect of the importance sampling. Based on NECP-MCX code system, the hybrid method and mesh coarsening method are implemented. The numerical results of HBR-2 benchmark show that the figure of merit (FOM) of the hybrid method is up to two orders of magnitude higher than that of MCNP. The weight window mesh can be reduced by 226 times without affecting the accuracy and efficiency of the final results.

Preliminary Research on a Multi-Physics Coupling Platform for Heat Pipe Reactors

Li Xiangyue, Xiao Wei, Zhang Tengfei, Li Peijie, Liu Xiaojing

2021, 42(2): 208-212. doi: 10.13832/j.jnpe.2021.02.0208

Abstract(284)

Abstract:
In order to achieve high-precision and high-fidelity numerical simulations of nuclear energy system and to explore the true physical process therein, a coupled three-dimensional high-fidelity calculation platform named MPCH for neutronics/thermal-conduction/stress-analysis calculation has been developed in this work. Therefore, MPCH can be used to perform a series of multi-physics coupled calculation of neutron transport, thermal conduction and thermal expansion. MPCH is constructed based on Picard iteration by integrating open-source codes: OpenMC, Nektar++ and SfePy. In this paper, the new space heat pipe reactor KRUSTY is calculated and analyzed under MPCH. The calculation results of multi-physics coupling show that MPCH can effectively predict the effective multiplication factor change, power distribution, temperature distribution and thermal expansion of the KRUSTY reactor. At the power of 4 kW, the local temperature difference of the whole reactor is 21.6K. The thermal stress leads to the deformation rate of 2.47%. Moreover, the effect of neutronics/thermal-conduction/stress-analysis coupling tends to make a more uniform temperature distribution of the core. This multi-physical coupling calculation program plays an important role in the design, development and verification of new reactors.

Development of Advanced Neutronics Code SCAP-N for Reactor Core High Fidelity Simulation

Peng Lianghui, Tang Chuntao, Yang Weiyan

2021, 42(2): 213-218. doi: 10.13832/j.jnpe.2021.02.0213

Abstract(226)

Abstract:
Reactor core neutronics calculation is important for the reactor design and analysis. In order to improve the resolution and accuracy of the reactor core neutron transport simulation, the advanced neutronics code SCAP-N was developed. In this code, firstly, the core was divided into layers according to the axial characteristics. Secondly, the two dimensional neutron transport calculation was carried out for each layer. Then the homogenized cross sections of all cells were obtained using the super homogenization method. Finally, the three dimensional pin-by-pin transport calculation was carried out to obtain the core effective multiplication factor and the pin power distribution. In order to improve the calculation efficiency, the MPI/OPENMP hybrid parallel method was adopted. The VERA benchmark problems and the AP1000 reactor start-up physical testing problem were used to verify the code. Numerical tests show that the pin-by-pin transport technology adopted by SCAP-N can improve the accuracy of core neutronics calculation compared with the commercial nuclear design system, and SCAP-N has higher computational efficiency compared with other high fidelity neutronics codes, which can further improve the economy and operation flexibility of nuclear power plants.

2021 Vol. 42, No. 2