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2017 Vol. 38, No. 6

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Study on Calculation Difference of the SCWR Core Steady State Analysis Code
Yang Ping, Wang Lianjie, MinG Zhedong, Zhao Wenbo, SuN Wei, Xu Yang, Li Haibo
2017, 38(6): 1-4. doi: 10.13832/j.jnpe.2017.06.0004
Abstract(14) PDF(0)
Abstract:
A coupled neutronics/thermal-hydraulics(N/T) three dimensional code system SNTA is developed for SCWR core steady state analysis.This paper studies the calculation difference between the SNTA code and the SRAC code. By using the impacts exclusive method, it is confirmed that the calculation difference between these two code is caused by the different feedback of the reaction cross-section. The reaction cross-section data and the energy group structure of the SRAC code differs from the SNTA code, and the density coefficient of reactivity calculated by the SRAC code is higher, which means the feedback of the density and power distribution is bigger and the axial power distribution varies rapidly.Compare to the SRAC code, the SNTA is more suitable for the SCWR core steady state analysis by coupling neutronics and thermal-hydraulics.
Experimental Research on the Parameter Affecting on the Characteristics of Secondary Side Residual Heat Removal System
Xi Zhao, XiE Feng, Gong Houjun, Yu Shimo, SuN Doucheng, Xiong Wanyu, Zan Yuanfeng
2017, 38(6): 5-8. doi: 10.13832/j.jnpe.2017.06.0005
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The experimental research has been carried out to investigate the effect of parameters on the characteristics of the secondary side residual heat removal system in for Hua Long Yi Hao using the Emergency Secondary Passive Residual heat removal system Integral Test facility(ESPRIT). The ESPRIT test facility, experimental working conditions and results are presented in this paper. The experimental results reveal that the PRS is capable of removing the core residual heat in station blackout scenario without staff intervention in 72 hours. The 0.8%FP core heat can be removed with the single PRS series or under the 40%-160% heat exchanger heat transfer area. The liquid level in the steam generator has no significant effect on the system pressure and steam temperature.
Integrating Reliability of Passive Residual Heat Removal System for AP1000 into Probabilistic Safety Assessment
Guo Haikuan, Zhao Xinwen, CAi Qi, Zhang Yongfa, Huang Liqin
2017, 38(6): 9-13. doi: 10.13832/j.jnpe.2017.06.0009
Abstract:
The paper studied the reliability of AP1000 passive residual heat removal system, and various importance and sensibility index were used to analyze the hardware reliability of the system. Importance and sensibility order for various failure mode of equipment leading to system failure is obtained to provide references to optimize the system and improve the hardware reliability. The hardware and physical process reliability of the system is integrated into the probabilistic safety assessment. The result was found that the evaluation of the reliability of the passive residual heat removal system not only analyzed the hardware reliability but also considered the physical process reliability, and the hardware and physical process reliability should be integrated into the probabilistic safety assessment to evaluate the passive residual heat removal system reliability comprehensively.
Optimization of Design on Passive Containment Cooling System by using Relap5 Code
BAi Jinhua, Zhao Bo
2017, 38(6): 14-17. doi: 10.13832/j.jnpe.2017.06.0014
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The performance of three conceptual designs with two types of passive containment cooling system(PCS) was studied by using Relap5 code. The evaluation of the performance of different design scheme was conducted in terms of startup, the stability of the operation process, the heat removal capacity in long-term balanced condition. The results show that the startup of the open loop PCS will cost the least time and its heat removal capacity of the in-containment heat exchanger will be the best one if the same heat exchangers are used in both open loop PCS and closed loop PCS. The stability of the operation process for closed loop PCS is the best. The closed loop PCS may be improved by introducing a double-heat transfer area heat exchanger. The performance of the improved closed loop PCS has the advantages of stable operation and good heat removal capacity.
Experimental Research on Heat Transfer of Supercritical Water in Square Annular Channel
Zhu Haiyan, Yan Xiao, Li Yongliang, Huang Yanping, Xiao Zejun
2017, 38(6): 18-22. doi: 10.13832/j.jnpe.2017.06.0018
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The experimental research of the heat transfer in square annular channel with helix rib has been carried out under various thermo-hydraulic parameters in this paper. The experimental parameters are as follows, the system pressure P: 23~25 MPa; the mass flow velocity G: 600~1000 kg/(m2·s); the heating flux q: 300~800 k W/m2; and the wrapped pitch is 160 mm.The experimental results are described in detail and the effects of pressure, mass flow and heat flux on the heat transfer are also analyzed. The comparison between the square annular channel with helix rib and bare square annular channel results indicates that the helix rib can enhance the heat transfer evidently. Based on the experimental data, a satisfactory correlation of heat transfer is given.
Numerical Simulation of Single Bubble Growth in Subcooled Boiling Water
Xu Chuan, Cheng Ning, Peng Changhong
2017, 38(6): 23-26. doi: 10.13832/j.jnpe.2017.06.0023
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The Volume of Fluid method(VOF) integrated in fluid dynamic program ANSYS FLUENT was adopted to calculate the process of single bubble growth in subcooled static water under two-dimensional condition. The phase change process could be simulated by adding extra mass source term and energy source term to the Navier-Stokes equations in the cells adjacent to vapor-liquid interface. The surface tension force, contact angle between bubble and wall and the evaporation of micro layer beneath the bubble had also been taken into consideration. The reliability of the calculation could be verified by comparing computational results with contrast experimental results. The numerical analysis could also obtain some physical parameters which are hard to be got from experiment directly, such as temperature fields and velocity fields.
Theoretical Study of Steady State Neutron Flux Density Re-Construction in ADS by Using Higher-Order Modes
XiE Jinsen, Chen Zhenping, Yu Tao, XiE Qin, Zeng Wenjie, Liu Zijing, HE Lihua
2017, 38(6): 27-30. doi: 10.13832/j.jnpe.2017.06.0027
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Neutron flux density in Accelerator Driven Subcritical reactor(ADS) is formed by superposition of fundamental and higher-order modes neutron flux density. Based on the bi-orthogonal properties of forward and ad-joint modes, modes expansion theory for steady state neutron flux density in ADS is established in this paper, and numerical studies on three dimensional four groups ADS diffusion problem are performed. Results indicate that both λ and prompt α modes can effectively re-construct the steady state neutron flux density of ADS, and neutron flux re-construction accuracy is improved by increasing expansion modes. Comparing to prompt α modes, λ modes are more appropriate for the steady state neutron flux density re-construction. Because of the symmetries of external neutron source and core pattern in this paper, only modes that have symmetrical properties have contributions to the steady state neutron flux density.
Research of Neutron Characteristics of HFETR Irradiation Channels
Liu Shuiqing, Liu Hongqian, Xiang Yuxin, Kang Zhanghu, Liang Guangyuan, Ran Zhongkang
2017, 38(6): 31-35. doi: 10.13832/j.jnpe.2017.06.0031
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This paper studied the neutron characteristics of the irradiation channels in High Flux Engineering Text Reactor(HFETR) by employing the MCNP programs as calculation tools, including the ratio of the neutron fluxes of E>1.0 MeV(φE(29)1.0 MeV) to the neutron fluxes of E>0.625 e V(φE(29)0.625 eV) within the different channels, the fast neutron flux ratio of the material farthest and nearest away from the core, and the optimum arranging height of the materials. The results indicate that the k within the different channels varies differently with the changing of the axial and radial locations. However the average varying extent is identical. For the fast neutron flux ratios of the material farthest and nearest away from the core, the 9# channel is bigger(1.43), while the G7 and K11 channels are smaller(1.21 and 1.18 respectively). Comprehensively speaking, the arrangement height of the materials can reach 500 mm for P15 channel, and for the 9# channel, it can reach 600 mm.
Simulation Investigation of Thermal-Mechanical Behaviors of MPS Defect Fuel Rod
Tang Changbing, Chen Ping, Zhou Yi, Chen Liang, Li Wei, Wang Lu
2017, 38(6): 36-41. doi: 10.13832/j.jnpe.2017.06.0036
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Manufacturing defects such as missing pellet surface(MPS) defect are inevitable in the manufacturing of pellets, which may result in over local stress on the cladding and cause cladding failure. In this paper, under the calculation frame of ABAQUS software, the related irradiation effects and thermal effects of fuel rods were introduced into the simulation based on vairous kinds of user defined subroutines, and then the thermal –mechanical behaviors of nuclear fuel rods were simulated. The sensitivity analysis of MPS defect sizes effect on thermal –mechanical behaviors of fuel rods were completed. The simulation results found that MPS defect causes higher center temperature, and the tensile stress and comprehensive stress appears on the cladding surface coinstantaneous. These effects would be more obvious with the increasing of MPS size. The defects with relatively bigger sizes should be concerned because these defects may threat the integrity of nuclear fuel rods during the operation period.
Magnetic Barkhausen Noise Study of Proton-Irradiated Domestic RPV Steel
Qian Wangjie, Liu Xiangbing, Xu Chaoliang, Wang Haitao, Zhang Guodong, Xu Zhong, Zheng Kai
2017, 38(6): 42-46. doi: 10.13832/j.jnpe.2017.06.0042
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Specimens of domestic nuclear reactor pressure vessel(RPV) steel were irradiated using 240 ke V protons with fluence ranging from 0.25 to 5×1017 cm-2(0.05 to 1.0 dpa) at room temperature. The measurement of the Magnetic Barkhausen Noise(MBN) and Vickers microhardness tests were conducted to explore the relationship between the MBN and irradiation fluence, mechanical properties. The results showed that the MBN signal was very sensitive to irradiation-induced defects. With the increase of irradiation damage the MBN signal initially decreased rapidly and reached the minimum at about 0.22 dpa. Then it increased and finally had no obvious change in the range of 0.3 to 1 dpa. These changes may be mainly attributed to the interaction between magnetic domain wall and irradiation-induced defects. The Vickers microhardness test indicated that there exists a remarkable radiation hardening tendency after proton irradiation. Meanwhile, a significant linear correlation between the MBN and Vickers microhardness was obtained.
Sensitivity of 58Co and 60Co for Activated Corrosion Products Source Terms in Pressurized Water Reactor
Hu Wenchao, Han Jingru, Li Tieping, Zhao Chuanqi, JinG Jianping, Zhang Chunming
2017, 38(6): 47-50. doi: 10.13832/j.jnpe.2017.06.0047
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Taking a pressurized water nuclear power plant as an example, this research uses CORA code to analyze the effect of material construction, the cobalt content of steam generator tubes, purification efficiency, core power and the concentration of Li OH on the corrosion production of 58 Co and 60 Co in the primary circuit. The results show that these factors can effectively reduce the activity concentration of activated corrosion products sources terms through the control of cobalt content of the steam generator tubes, increasing the concentration of Li OH, improving the coolant purification efficiency and reducing the power level. The research can provide guidance for the radiation dose control of nuclear power plants.
Transient Simulation on Reactor Core Melt and Lower Support Plate Ablation in In-Vessel Retention
Chen Xuyi, Zhang Xiaoying, Wang Biao, Xu Junying, Zhang Lei, Zhang Huiyong, Zhan Dekui
2017, 38(6): 51-56. doi: 10.13832/j.jnpe.2017.06.0051
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To precisely understand the accident process of reactor core melt in In-vessel retention(IVR) condition, 3 dimensional transient thermal conduction analysis with moving boundary is performed on quarter reactor core model. The decline of decay power and water level in reactor pressure vessel(RPV), and the radial distribution of assemblies of different material is considered. Convective heat transfer on rod surface and coolant interface in computed with empirical correlation of natural convection of saturated steam vapor/water. Radiation heat transfer with 16 neighboring rod is considered. Also, the ablation caused by continuously accumulation of molten corium on lower support plate(LSP) is simulated. The impingement heat transfer of the falling corium and the molten pool formed in LSP ablation cavity is taken into account. The simulation gives the ablation process on the surface of LSP as well as temperature history and molten proportion of the reactor core, which shows agreement with reference. Simulation shows: the melt process of reactor core accelerated in the accident process of 2600 s, when coolant in RPV dry up 65% of the core mass has molten at 8000 second. LSP is completely penetrated in 6000 s, the ablation of LSP is mainly focused on an annular region of radius 700 mm.
Design of Data Acquisition Method of Temperature Probe for Primary Circuit in Nuclear Power Plants
Liu Jianguang, Liu Ruhuan, Huang Yun
2017, 38(6): 57-60. doi: 10.13832/j.jnpe.2017.06.0057
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The traditional data acquisition method for reactor coolant system temperature measurement channel calibration test(RCP63) is artificial acquisition method. It is with low processing efficiency, prone to human error, poor archiving and other issues. For these problems, this paper studies the DCS(digital instrument control system) data acquisition characteristics of nuclear power plants, and uses the database data in the DCS(digital instrument control system) for data acquisition. The method had been successfully applied to the unit 3 of Hongyanhe Nuclear Power Plant, which increased the work efficiency and reduced the potential human errors.
Research on Control Characteristics of Small Nuclear Reactor Pressurizer
Zhang Yilin, Shi Bo, Zhao Fuyu
2017, 38(6): 61-65. doi: 10.13832/j.jnpe.2017.06.0061
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The smaller the volume of the pressurizer, the violent the change of the pressure and water level of the pressurizer caused by same surge flow in the transient condition of nuclear power plant. This phenomenon enhances the coupling of the pressure and the water level in pressurizer, and.causes the frequent action of the heaters and spray valves, and thus degrades the system stability or even results in unstable operation. Two-region non-equilibrium model of pressurizer is built using Matlab in this paper, and the coupling equation between main coolant system and pressurizer is derived. At last, the decoupled controller transfer function was calculated using the method of diagonal matrix, and the pressurizer decoupled controller is designed utilizing the frequency domain method.
Failure Probability Assessment of Passive Residual Heat Removal System for NPPs
Tang Huapeng, Zhang Zhizhu, Li Haibo, Zhang Kai, Deng Chunrui, Ran Xu
2017, 38(6): 66-71. doi: 10.13832/j.jnpe.2017.06.0066
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Passive safety systems are increasingly implemented in the design of advanced nuclear power plants, in order to improve the safety of reactors. Up to now, no systematic and mature methodology has been presented addressing the evaluation for passive thermal-hydraulic systems. Also its failure probability is generally not considered in PSA. This paper presents the quantification result of failure probability about natural circulation functional carried out for the Passive Residual Heat Removal System of Secondary Side(PRS) of PWR. The analysis took overall approach reported in the Reliability Methods for Passive System(RMPS, European Commission) project as reference. The importance indicators of parameters affecting its performances have been evaluated based on statistics methodology and T-H simulation. At last, natural circulation functional failure probability is calculated using Monte-Carlo method and also with Response Surface method. The assessment shows the functional failure probability of PRS is 2.14×10-3. The method adopted in this paper can be used in Passive safety systems design optimization.
Strategy for Loss of Coolant Accident in ACP100+
Zeng Wei, Song Danrong, Chen Zhi, Zhu Li, Liu Songtao
2017, 38(6): 72-75. doi: 10.13832/j.jnpe.2017.06.0072
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Based on the scheme of the ACP100+ small modular reactor such as the inner PRZ and steel containment, the strategy for the emergency core cooling, pressure controlling in the containment and long-term cooling are proposed. According to the elementary calculation and analysis, the accumulator can be eliminated for the ACP100+ and the safety injection system is simplified. The release of mass and energy can be controlled more easily due to the smaller size of break based on pressure restrained pool and containment pool above the small steel containment. And the long-term residual heat can the taken out through the pool above the containment based on the natural convection and the inherent safety can be enhanced.
Probabilistic Safety Assessment of Off-Site Consequence of Xi’an Pulsed Reactor Nuclear Accident
Tang Xiuhuan, Shen Zhiyuan, Wang Baosheng, Zhu Lei, Yang Ning
2017, 38(6): 76-80. doi: 10.13832/j.jnpe.2017.06.0076
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The emphasis of this paper lies in the quantitative estimation of off-site risk to the public from Xi’an Pulsed Reactor(XAPR). Off-site consequence model for XAPR nuclear accident was established, and the meteorology data acquired from XAPR site was used as the input parameter with probabilistic theory. The off-site risk to the public of XAPR’s radioactivity release was preliminarily analyzed in application of probabilistic safety assessment at 100 m point of XAPR boundary in radioactivity. The results demonstrate that the conditional probabilities of effective dose exceeding 1 m Sv and 10 m Sv are about 0.652% and 0.0750% respectively in the case of radioactivity release spectrum. The overall frequencies with which individual effective dose 10 m Sv is exceeded is less than 2.20×10-9 a-1, meanwhile the other overall frequency with which individual cancer fatality risk is exceeded is not greater than 1.89×10-6 a-1. Accordingly the draft safety goal of XAPR in this research is met from this quantitative risk. Our initial assessment leads to conclude that the off-site risk to the public from XAPR is extremely low and then the high safety characteristic of XAPR is proved.
Optimal Control of Nuclear Power Plant Steam Generator Based on GMFAC
Huang Wei, Yang Shuangshuang
2017, 38(6): 81-86. doi: 10.13832/j.jnpe.2017.06.0081
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According to the nonlinear control system of steam generator water level, large lag and the "false water level" caused by load changes and other issues, based on the model free adaptive control(MFAC) theory, an improved model free adaptive control(GMFAC) theory which is based on high"universal model" is proposed, and the relevant controller is designed to control the water level of steam generator.For the model free adaptive control parameter optimization problem,A swarm intelligence optimization algorithm based on animal behavior — artificial fish swarm algorithm(AFSA) is proposed.In order to avoid the local optimum and improve the convergence rate, an improved AFSA algorithm(PSO-AFSA) is proposed.In order to improve the accuracy of the algorithm and to improve the accuracy of the algorithm, a reference particle swarm optimization(PSO) algorithm is defined to improve the accuracy of the algorithm.The simulation results show that the GMFAC has better performance and disturbance rejection ability after optimization of the artificial fish swarm algorithm.
Design and Optimization of Cam Curve for Main-Auxiliary Model Constant Spring Hangers
Liu Karen, HE Mengfu, Han Lang, Tang Feng, Yan Liang
2017, 38(6): 87-91. doi: 10.13832/j.jnpe.2017.06.0087
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The constant hangers are usually used in nuclear power stations. Fitting relation is researched for the main-auxiliary constant hangers. The structure is designed based on the curve differential equation of cam after the analysis of the characteristics of the constant. The designed cam theoretically satisfies the work demand of the constant hangers, providing the direction for the optimization of the curve.
Pre-Conceptual Design and Analysis of a SCWR-FQT Loop Test Section Based on CSR1000
Zhang Liang, Wang Hai, Tong Mingyan, SuN Sheng, Yang Wenhua, Si Junping
2017, 38(6): 92-98. doi: 10.13832/j.jnpe.2017.06.0092
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Two preliminary conceptual designs of test section for fuel qualification test loop based on China’s supercritical water-cooled reactor(CSR1000) fuel element is proposed, which are the 2×2 assembly design and the 3×3 assembly design. The MCNP code and the CFX code are used to proceed the neutron, the thermal-hydraulic analysis and the preliminary evaluation of different designs. The results show that the two designs are engineering feasible and meet the requirements of fuel qualification test, but there are significant differences in performance. The fuel rod power of the 2×2 assembly is 23.6 25.3 k W and the average power is 24.3 k W, while these values for 3×3 design are 15.9~26.7 k W and 21.4 k W, respectively. The radial power peak factor of fuel assembly for 3×3 design is 1.25, which is not conducive to fuel assembly power flattening, limiting the average power of the assembly. The preliminary thermal-hydraulic analyses with wireless fuel assembly indicate that the outlet coolant temperature of the two designs exceeds the quasi-critical temperature of the pressure of 25 MPa, and the fuel pellet temperature and the fuel cladding outer surface temperature are lower than the design limits, allowing certain safety margins.
Simulation Study on Fission Chamber Wide-Range Electronics
Luo Tingfang, Zhu Hongliang, Liu Lixin
2017, 38(6): 99-102. doi: 10.13832/j.jnpe.2017.06.0099
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Using computer simulation technology, the impact of frequency band and noise in the fission chamber wide-range electronics is modeled and simulated. The analysis results show that the pulse counting mode is mainly influenced by the frequency band, and the upper limit of counting rate is about tenth of the frequency band under 10 MHz; the Campbell mode is influenced by the frequency band and noise level at the same time, 300 k Hz frequency band has an excellent signal to noise ratio, and lower noise level can effectively decreases the lower limit of counting rate.
Study on Classification Method for NPP Equipment Reliability Based on AP913
Qin Feng, Zhu Guixia, Qiao Zhen, Guo Longzhang
2017, 38(6): 103-106. doi: 10.13832/j.jnpe.2017.06.0103
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In order to allocate NPP equipment management and maintenance resources reasonably, improve the reliability of equipment, ensure the NPP safe and stable economic operation, the classification management has to be carried out for the equipment of NPPs. This paper studied the reliability of equipment classification method based on AP913 equipment reliability management system, and eventually developed a new classification process which was different from the classical RCM and Streamlined RCM. Based on this method, the 205 systems of equipment classification work has been completed in a domestic nuclear power plant successfully.
Sensitivity Analysis for Passive Residual Heat Removal System of AP1000 Based on Grey Correlation Degree
Qi Shi, Zhou Tao, Li Bing, Li Yu, Jiang Guangming, Yu Hongxing
2017, 38(6): 107-112. doi: 10.13832/j.jnpe.2017.06.0107
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To analyze the behavior of passive residual heat removal system(PRHRS) of AP1000 under loss of main feedwater accident, a thermal hydraulic model is built by RELAP5. Based on the results of RELAP5, the grey correlation degree is used to analyze the importance degree of influence factors. The results show that when the resolution is 0.1, the influence factors of reactor coolant outlet and fuel cladding maximum temperature can be well distinguished. Reactor coolant outlet maximum temperature directly affects the fuel cladding maximum temperature. The correlation coefficient between them is great. The initial reactor power has greatest effect on the reactor coolant outlet maximum temperature and the fuel cladding maximum temperature. And the sequences of influence factors are temperature of IRWST, height of ascending pipe and initial pressure. Correspondingly, the diameter and resistance coefficient of PRHRS-HX have less effect.
Research on Reverse-Heating for Test of Heat Exchanger of Spent Fuel Pool in EPR Units
Liu Zhenyong, Liu Chunlei, Ruan Hongqiao, Lin Zhihang
2017, 38(6): 113-116. doi: 10.13832/j.jnpe.2017.06.0113
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Based on the research on the test of heat exchanger in the spent fuel pool(SFP) of European Pressurized Reactor(EPR), this paper provides a method to test the exchanger through reverse-heating of PTR exchanger from RCP via the trains of RCP→RHR→RRI→PTR systems to heat up the spent fuel pool. Moreover, by the means of modeling the heat exchanger of SFP, the required test condition is able to be transposed from extreme accident to a relatively achievable test condition.
Analysis and Design of Solved Method on Main Feed Water Regulating Valve Frequent Fault of Qinshan 2th Nuclear Power Plant
Zhang Yulong, Li Yanglong
2017, 38(6): 117-121. doi: 10.13832/j.jnpe.2017.06.0117
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The fault of main feed water regulating valve of CP600 nuclear power units occurs frequently. This paper analyzes several fault conditions frequently occurred on main feed water regulating valves, and studies the stuck of the main feed water regulating valve of CP600 simulator. Three control schemes are compared and the optimal control scheme for the hot standby condition is proposed.
Research of Target PSD Generation in Seismic Qualification Tests of Nuclear Equipment
Jin Ting, Zhang Guihe, Xu Xiao
2017, 38(6): 122-124. doi: 10.13832/j.jnpe.2017.06.0122
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In order to avoid the distortion of power spectral density of the artificial time history in seismic qualification test, the power spectral density envelope should be verified. On the basis of the relationship among floor response spectrum, Fourier amplitude spectrum and power spectral density, the method for generating target power spectral density is studied. Respectively using Kaul formula and SRP3.7.1 appendix A method, fast Fourier transform method are introduced to retrieve the power spectral density of floor spectrum and artificial acceleration. The basic assumptions and uncertainties of different methods and the differences of the shape and the peak value of the obtained power spectral density are comparatively analyzed. The method based on response acceleration and its fast Fourier transfer is recommended as the target power spectral density generating methodology.
Research on Sintering Process of UO2-Er2O3 Fuel Pellets
Liu Yu, Yang Jing
2017, 38(6): 125-128. doi: 10.13832/j.jnpe.2017.06.0125
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The paper introduces the research on sintering process of UO2-Er2O3 burnable poisonous fuel pellets. The experiments show that the main performance of the pellets, such as: integrity, density and grain size meet the requirements in the following sintering process: the density of green pellets should be among 55%~60%T.D., the temperature is 1700~1750℃, and sintering in hydrogen atmosphere for 2~3 h.
Effect of Heat Treatment on Tensile Properties and Microstructure of New Domestic Zirconium Alloys
Chen Le, Yang Zhongbo, Qiu Jun, Liang Bo, Li Weijun, Hong Xiaofeng, Wang Lian
2017, 38(6): 129-133. doi: 10.13832/j.jnpe.2017.06.0129
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The tensile properties of stress relieved(SR) and recrystallized(RX) alloys, with two different compositions, namely SZA4(Zr-0.8 Sn-0.25 Nb-0.35 Fe-0.1 Cr-0.05 Ge) and SZA6(Zr-0.5 Sn-0.5 Nb-0.3 Fe-0.015 Si) at room temperature(RT) and 385 ℃were studied by MTS material test machine, the fracture morphology of alloys was observed by scanning electron microscopy(SEM) and the microstructure was observed by transmission electronic microscopy(TEM). The results showed that each alloy was excellent on mechanical properties, SZA4-450℃was the strongest while SZA4-560℃ and SZA6-560℃ had the best elongation.The fracture mode of alloys was ductile, according with lots of dimples in the fracture morphology, second phase particles(SPPs) which were also observed by SEM, distributing in the base uniformly and dispersely. There were two kinds of SPPs which were observed by TEM in SZA-4 alloy, one was Zr(Nb Fe Cr)2 with a smaller size and the other was Zr(Nb Fe Cr Ge)2 with a larger size, both of them had a HCP structure. In SZA-6,(Zr Nb)2 Fe SPPs with a FCC structure and Zr(Nb Fe)2 SPPs with a HCP structure was found. The effect of heat treatment and composition on alloys was analyzed, showing that the heat treatment played a dominant role.
Study on Quantitative Analysis Technique of Uranium in Uranium Mixture
Song Qiang, Zheng Lingling, Qiao Hongbo, Liao Zhihai
2017, 38(6): 134-136. doi: 10.13832/j.jnpe.2017.06.0134
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Determination conditions of uranium mixture such as dissolve method of uranium, determination method and interference of matrix were studied. Interference on the determination of uranium was eliminated by coordinating matrix elements in solution with NH4 F. The quantitative analysis method of uranium in uranium mixture was built by precision experiments. Relative standard uncertainty of this method was below 0.2% by determination of simulation samples.
Study on Performance Sensitivity of 1Cr15Co14Mo5VN Stainless Steel
Wang Li, Liu Xiao, Li Weijun
2017, 38(6): 137-141. doi: 10.13832/j.jnpe.2017.06.0137
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Based on the results of metallographic examinations as well as relevant verifying tests, performance tests and influencing factor analysis of the materials sampled from different batches of the products, this paper reveals the characteristics of the 1Cr15Co14Mo5VN ductile-toughness, which is very sensitive to the heat treatment process. The study also shows the sensitivity of the products manufactured with 1Cr15Co14Mo5VN to surface defects and stress concentration due to the lower crack resistance of the material. Therefore, it is suggested to improve the product design and the material. Analysis indicates that the fracture toughness and the ductile-toughness of 1Cr15Co14Mo5VN can be increased by measures such as adding Ni and Nb, using double vacuum smelting, and increasing the solution temperature.
Study on Electrochemical Pitting Corrosion of 316NG and 321SS for Primary Pipes in Simulated Marine Atmosphere Environment
Shu Ming, Wang Conglin, Chen Yong, Xu Qi, WEi Guangqiang, Zhao Yuxiang, Wang Hao
2017, 38(6): 142-146. doi: 10.13832/j.jnpe.2017.06.0142
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The pitting corrosion of 316 NG stainless steel and 321 stainless steel in the 3.5% Na Cl solution was studied by using the electrochemical polarization curves and electrochemical critical pitting temperature test. The results showed that the pitting potential of 316 NG was markedly higher than that of 321 at all the test temperatures(room temperature, 40℃, 60℃ and 80℃). The pitting potential of 316 NG and 321 dropped sharply when the test temperature increased, which demonstrated that the pitting resistance became worse. The critical pitting temperature(CPT) of 316 NG and 321 were 20.1℃ and 3.9℃ respectively. From the perspective of electrochemistry, the pitting corrosion resistance of 316 NG was apparently better than that of 321 in the simulated marine environment. The particles of Ti N or Ti C could be observed in the corrosion pits by using scanning electron microscopy(SEM), which led to a decrease of the pitting resistance.
Design and Research on Two-Phsse Numerical Calculation for Small Modular Pressurized Reactor Cavity Injection System
Li Haoxiang, Zhu Dahuan, Li Songyu, Li Quan, Zeng Wei, Guo Yun
2017, 38(6): 147-151. doi: 10.13832/j.jnpe.2017.06.0147
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Residual heat from melting core located in the bottom of the reactor vessel could be removed by cavity injection system(CIS). The consequence of severe accident can be mitigated. The method of two phase numerical calculation which could be used to predict the under head critical heat flux(CHF) has been researched in this paper. Through the research better geometry which could bring high CHF can be found. The CHF curves of flat ellipsoid and spherical has distinct difference. Despite the maximum value is similar, the CHF curves of flat ellipsoid geometry is larger than that of spherical geometry during the central section(30°~60°). The flat ellipsoid geometry could enhance the safety margin of the cavity injection system.
Study on Transition Strategy of Mode-C from Base Load to Load Follow
Gong Zhaohu, Liu Tongxian, Li Qing, Jiang Zhumin, Li Tianya, Zhou Jinman, Cai Yun
2017, 38(6): 152-156. doi: 10.13832/j.jnpe.2017.06.0152
Abstract(15) PDF(0)
Abstract:
As a new operation and control mode, Mode-C is similar to Mode-G during base load operation, but similar to MSHIM during load follow operation, thus combining the advantages of both. In order to meet the demand of reactivity control, it is easy to lose the axial offset control during the transition from base load to load follow near the end of cycles. In the load change process, the control rod motion, flux redistribution and Xenon feedback, which have effects on reactivity and axial offset control, are tightly coupled, so it is difficult for analysis and calculation by separation of variables. Therefore, this paper takes the typical daily load follow transition of a double loop PWR as an example to make direct simulation study on three strategies, namely, temporary boron change, pre boron change and base load with deep K-bank insertion. The results show that only the strategy of base load with deep K-bank insertion can quickly and smoothly transit to load follow condition, and this strategy also has the advantages of longer cycle life, less boron change and no need to modify the reference axial power difference during transition.
Development of Self-reliant Subchannel Analysis Code CORTH
Liu Yu, Tan Zhanglu, Pan Junjie, Wang Xiaoyu, Xu Liangjian, Deng Jian
2017, 38(6): 157-162. doi: 10.13832/j.jnpe.2017.06.0157
Abstract(11) PDF(0)
Abstract:
CORTH is a subchannel code self-reliant developed by Nuclear Power Institute of China(NPIC) based on four equation model with slip ratio to perform thermal-hydraulic analysis of reactor cores or experimental facility with heating rod bundles. The development of CORTH adopted modular design and object-oriented programming method, and specially provided with graphic user interface for input deck preparing and output analysis. CORTH has been executed software testing by independent third organization to ensure the readability and standardization of coding. Testing data in NPPs, international benchmark and AP1000 nominal condition were used to carry out verifications and validations of CORTH. The results show the good accuracy of CORTH that meets the requirement of engineering design and analysis.
Simulation and Analysis of New-Type Fuel Assembly Top Connection Structure
Wang Haoyu, Pu Zengping, Zhu Fawen, Chen Ping, Gu Mingfei, Liu Yanghua
2017, 38(6): 163-166. doi: 10.13832/j.jnpe.2017.06.0163
Abstract(17) PDF(0)
Abstract:
The function of fuel assembly top connection is to provide structure continuity and ensure the structure stability of fuel assemblies, so it plays an important role in the safety and performance of fuel assemblies. In order to analyze and evaluate a new-type fuel assembly top connection structure, considering contact nonlinear, material nonlinear and large deformation, a numerical simulation model is established in this paper by using ABAQUS. By comparison of the simulation result and the experimental result, the established finite element model is considered to be reasonable, and can be used to analyze and evaluate this new-type top connection structure. Finally, the tensile process of the top connection is analyzed in detail, and it is concluded that the friction force is the key factor for its tensile capacity.
Synthetic Evaluation of RCCA Dropping and Damping Process in CF Series Fuel Assembly
Guo Xiaoming, MA Chao, Chen Ping, Xiao Zhong, Pu Zengping, Qin Mian
2017, 38(6): 167-169. doi: 10.13832/j.jnpe.2017.06.0167
Abstract(12) PDF(0)
Abstract:
In the process of rod cluster control assemblies(RCCA) dropping in PWRs, the guide thimble of the fuel assembly supplies a tunnel for the control rod, and the dashpot structure which is designed below the guide thimble, in order to decrease the impact force and ensure the safety of the fuel assembly. In this paper, with the CIGAL and SAM program, the analysis about the RCCA dropping and damping process is carried out to obtain the drop time and the impact force on the top nozzle in the series CF fuel assembly which includes CF2 and CF3. The calculation results demonstrate that: from CF2 to CF3, with the increasing of the screw hole length and the decreasing of the guide thimble inner diameter, the drop time gets longer, the max impact force on the top nozzle gets smaller, the damping time in the dashpot gets longer, and the pressure in the dashpot get larger.
Method Study on Mechanical Performance Analysis for Control Rod Used in HPR1000
Qin Mian, Pu Zengping, Chen Ping, Ru Jun, Li Yun, Li Hua, Liu Yanghua
2017, 38(6): 170-174. doi: 10.13832/j.jnpe.2017.06.0170
Abstract(11) PDF(0)
Abstract:
In this paper, the calculation method for mechanical performance analysis of the control rod has been developed base on the structural analysis. The performance parameters of the cluster control rod in HPR1000 such as the absorber temperature, internal pressure and cladding stress have been analyzed using this method to prove that its suitability and reliability.
Analysis of Measures for Safe Operation of PWR Fuel Assemblies
Ru Jun, Xiao Zhong, Zhu Fawen, Zhang Lin, Qin Mian, Liu Yanghua
2017, 38(6): 175-179. doi: 10.13832/j.jnpe.2017.06.0175
Abstract(11) PDF(0)
Abstract:
Analysis on fuel assembly design, manufacture, operation and post irradiation examination are performed, and the improvement of current safe operating insurance is proposed. In particular, the proposal of strengthen disassembly inspection, pool side inspection and hot cell analysis of fuel assembly is put forward. Suggestions are also proposed for the fuel assembly development on self-reliance, including preventing abrasion at lower part of the fuel rod, improving the design of anti-debris and strengthening the strength of the component structure.
Numerical Simulation of Irradiation-Thermal-Mechanical Coupling Behaviors in Nuclear Fuel Rods
Tang Changbing, Jiao Yongjun, Chen Ping, Li Yuanming, Zhou Yi
2017, 38(6): 180-184. doi: 10.13832/j.jnpe.2017.06.0180
Abstract(17) PDF(0)
Abstract:
The in pile behaviors of nuclear fuel rods are very complex under the reactor operation conditions, accurate and reliable fuel performance prediction is very important not only for the reactor safety calculation and fuel designing but also for the fuel performance assessment. In this simulation, the thermal effects and irradiation effects of UO2 material and zirconium cladding material were considered and gap heat transfer(gas heat conduction, radiation heat transfer and contact heat conduction) were considered. The subroutines were created to include above fuel models into the 3 D finite calculation and then preliminary established the simulation method for the nuclear fuel rods’ thermal-mechanical behavior.
Numerical Analysis of UN Fuel Performance
Tu Teng, Li Wenjie, Li Wei, Gao Shixin, Chen Ping
2017, 38(6): 185-188. doi: 10.13832/j.jnpe.2017.06.0185
Abstract(15) PDF(0)
Abstract:
Uranium nitride is an important candidate for Accident Tolerant Fuels(ATF) due to its high thermal conductivity and high heavy metal(HM) density. This paper introduced an enhancement of FUPAC by incorporating UN material property correlations and irradiation models. Some preliminary analyses of UN fuel performance at steady-states were also performed, and the results showed that UN fuel has good performance in aspects like fuel temperature, fission gas release, fuel rod pressure and cladding strain. Finally, future research and development works for UN fuel are discussed in this paper.