Advance Search

2016 Vol. 37, No. 4

Display Method:
Experimental Study on Flow Excursion of Two Phase Natural Circulation under Low Pressure in Narrow Rectangular Channel
Zhou tao, Qi Shi, Song mingqiang, Chen baixu, Huang yanpin, Xiao zejun
2016, 37(4): 1-5. doi: 10.13832/j.jnpe.2016.04.0001
Abstract(10) PDF(0)
Abstract:
Based on the natural circulation experiment platform, the Flow Excursion of natural circulation in the narrow rectangular channel has been experimentally investigated on the natural circulation experiment platform. When the heating power comes up to a certain range, the natural circulation flow excursion occurs in the system, accompanied by the flow oscillation. The stability of the whole system increases with the increasing of the inlet sub-cooling, the pressure and the channel size. The stability curve of the system is drawn according to the sub-cooling and the phase change. It is discovered that there are 4 stages of flow excursion in the narrow rectangular channel, and the experience formula is proposed for the initial power of natural circulation flow excursion in the narrow rectangular channel.
Theoretical Study on Critical Heat Flux in Vertical Narrow Tube with Axial Power Step
Liu Ye, Zhao Dawei, Liu Wenxing, Zan Yuanfeng
2016, 37(4): 6-10. doi: 10.13832/j.jnpe.2016.04.0006
Abstract(10) PDF(0)
Abstract:
In the present study, the critical heat flux(CHF) experiments in vertical narrow tube with power step are conducted to verify the CHF predictions by non-uniform heating CHF model.The boiling crisis position, CHF, qN,CHF, and average heat flux, qN,AVE, predicted by non-uniform heating CHF model, show a reasonable agreement with the CHF experimental results. The CHFs for different power steps are calculated using this non-uniform heating CHF model to analyze the effects of power step ratio, Rh, power step length, Lh, and the power step position on the boiling crisis: The qN,CHF approximately linearly increases with an increase of Rh. While the qN,AVE decreases with the increasing Rh. As Lh increases, qN,CHF decreases significantly and qN,AVE approaches to CHF under uniform heating condition, qU,CHF. When the position of the power step moves to the upstream, the boiling crisis position will depart from the end of power step. As boiling crisis is trigged at the outlet of tube, the qN,CHF and qN,AVE are both approximately equal to qU,CHF.
Theoretical Investigation on Critical Heat Flux at Inclined Heater Surface in Low Pressure
Guo Rui, Liu Xiaojing, Cheng Xu, Yu Hongxing
2016, 37(4): 11-14. doi: 10.13832/j.jnpe.2016.04.0011
Abstract:
In-vessel retention(IVR) of molten core debris via water cooling at the external surface of the reactor vessel is an important severe accident management feature of advanced passive plants. For this concept, it is important to keep the heat load on the vessel wall surface lower than the critical heat flux at any position of the lower head. Based on the existing bubble crowding model in the literatures, a theoretical CHF model is developed suitable for the inclined heater surface in low pressure. In the new model, the effect of orientation on bubble velocity and bubbly layer thickness is taken into consideration. A new method is introduced to calculate the steam quality of both layers. Comparison with the present experimental data shows that the new model satisfies the prediction accuracy. The maximum deviation between the model prediction and experimental data is less than 10%.
Effect of Velocity on Steam Condensation with Non Condensable Gas
Zhou Shan, Han Liyong, Zhao Wei, Yang Lin
2016, 37(4): 15-18. doi: 10.13832/j.jnpe.2016.04.0015
Abstract(12) PDF(0)
Abstract:
Condensation on the containment structures during an accident is one of the critical thermal-hydraulic phenomena that would affect the pressure in the containment vessel. Experiment facility was set up to investigate the steam condensation with non condensable gas on a cold plate,focusing on the conditions in the containment. In the experiment, pressure range is 0.11 ~ 0.5 MPa;mass fraction of air in the bulk flow is 29%~78%; the wall sub-cooling is 26~60℃. Gas velocity varies from 0.4 to 1.9 m/s. The experimental result agrees with empirical correlations when the gas velocity is up to 0.9 m/s. Gas velocity becomes one of the primary factors influencing the steam condensation in non-condensable gas with higher gas velocity. The promotion effect on the heat transfer of increasing velocity is more significant when the mass fraction of non-condensable gas is lower. The transition criterion from free to mixed convection for steam-gas flow condensing should be different from that of the single phase heat transfer.
Experimental Study on Flow Instability in Parallel Twin Rectangular Channels under Motion Conditions
Tang Yu, Chen Bingde, Xiong Wanyu, Huang Yanping, Xu Jianjun
2016, 37(4): 19-23. doi: 10.13832/j.jnpe.2016.04.0019
Abstract(12) PDF(0)
Abstract:
An experiment has been made on the flow instability under station, inclination,fluctuation and rolling conditions in a 50 mm×2 mm parallel rectangular channel at the pressure of3MPa~8MPa. First of all, the effects of motion conditions on phenomena of flow oscillation have been studied in time domain and frequency domain. Then, by comparing the results under station condition, the effects of motion conditions on flow instability boundary were obtained. Finally, an empirical correlation of instability boundaries was correlated based on the experimental results.
Research on Shape Optimization of RPV Opening of High Temperature Gas-Cooled Reactor Using Boundary Element Method
Bo Wen, Li Zhengcao, Wang Haitao, Shi Li
2016, 37(4): 24-27. doi: 10.13832/j.jnpe.2016.04.0024
Abstract(13) PDF(0)
Abstract:
Stress concentration is one of the fields that attract major concerns in the structure integrity analysis of the reactor pressure vessel(RPV) of the High Temperature Gas-cooled Reactor(HTGR). The concentrated stresses depend on the opening shape. In this paper, the boundary element method(BEM) is used for shape optimization of the HTR-PM RPV large opening. The BEM requires only boundary meshing and thus only boundary meshes move during the opening shape optimization process, avoiding the issue of volume mesh distortion. By using the shape optimization algorithm based on adaptive growth, stresses can be effectively reduced with normal growth happens in the areas of stress concentrations. Numerical tests show that optimized opening shape with low stresses can be readily obtained using the presented method.
Study on Stiffness Analysis Model of PWR Holddown Springs
Jin Yuan
2016, 37(4): 28-33. doi: 10.13832/j.jnpe.2016.04.0028
Abstract:
PWR fuel assembly holddown spring systems generally consists of three to five spring leafs. To analyze the stiffness of the holddown spring system, commercial FEM code is usually used for the evaluation currently. But the optimization capability of the main current commercial FEM code is weak, during the research and development progress, it usually cannot screen a better performance design product. This paper established the holddown spring stiffness analysis model from the perspective of theoretical analysis. Using the established theoretical analysis model, the paper gives three main factors impacting the holddown spring stiffness: leaf thickness, leaf width and leaf height, and also give the analysis on how these factors affecting the holddown spring stiffness.
Pressure for Hydrostatic Test of Pressure Vessel and Its Advantages and Disadvantages
Zhang Jingcai, Hu Youming
2016, 37(4): 34-38. doi: 10.13832/j.jnpe.2016.04.0034
Abstract(11) PDF(0)
Abstract:
Hydrostatic test pressure and its advantages and disadvantages have been discussed in detail. The paper indicated that the hydrostatic test pressure is about 40%~45% of the plastic instability pressure; Hydrostatic test stress criteria allowable pressure is about 50%~75% of the plastic instability pressure; Hydrostatic test may improve and increase the loading capability of the pressure vessel and decrease the failure possibility by hydrostatic test mechanism; Hydrostatic test is the useful and efficiency examination method and inspection technology for the leakage and strength of the pressure vessels; Periodic hydrostatic test during operation may establish the maximum defect in the pressure vessel or the maximum loading so that to provide basis and data for continuous operation, but the pressure of the periodic hydrostatic test does not exceed the first hydrostatic test pressure. The overpressure test and the warm pre-stressing test and harmful effects of pressure test need to be researched in depth.
Scale Analysis of Hydraulic Test of Tube Support Plate in Steam Generator Based on Computational Fluid Dynamics
Li Yong, Zhu Haiyan, Wen Bo, Zhao Erlei, Zan Yuanfeng, Zhuo Wenbin, Li Pengzhou
2016, 37(4): 39-43. doi: 10.13832/j.jnpe.2016.04.0039
Abstract:
Various analysis focused on hydraulic performance of TSP model with different scale and structure was carried out by CFD calculation. The comparison of calculation results show that the characteristics, which are not nearly influenced by the wall of flow channel, such as flow resistance, flow field and mass flow distribution in the regular hexagon TSP models with structures of 1/3rd and 1/2nd orifice, are basically consistent with the ones in the prototype. The structure can be applied in the design of TSP model for the hydraulic test.
Research of Scanning Acoustic Microscope Thickness Measurement Method for Zirconium Alloy Multilayered Composite Material
Ren Junbo, Tang Yueming, Xu Guiping, Wang Xuequan, Cai Yukun
2016, 37(4): 44-47. doi: 10.13832/j.jnpe.2016.04.0044
Abstract(13) PDF(0)
Abstract:
The zirconium cladding material is a kind of multilayer structure. The outer layer is zirconium alloy cladding and the middle layer is some kind of metal powder rolling products. The thickness of the zirconium cladding is about 0.3mm to 0.6mm. To satisfy the quality control requirement of the rolling technology, accurate thickness of the compound layer needs to be measured. The traditional ultrasonic thickness measurement method can not meet the requirements because of its poor sensitivity and resolution ratio. In this paper, SAM thickness measurement technology is used to measure the thickness of the zirconium cladding, and the thickness mechanism is analyzed theoretically. The experimental verification is carried out. The experiment shows that SAM is available for the thickness measurement of the cladding with thickness of 0.3mm to 0.6mm.
Study on Thermal Mechanical Fatigue Performance of 316LN Stainless Steel
He Kun, Zhou Jun, Luo Qiang, Cheng Yong, Ren Liping, Zhu Yonghui
2016, 37(4): 48-52. doi: 10.13832/j.jnpe.2016.04.0048
Abstract(11) PDF(0)
Abstract:
Thermal mechanical fatigue behavior of 316 LN stainless steel is studied by thermal mechanical fatigue test, the fatigue data and curve are obtained. The result shows that 316 LN stainless steel has a hardening and then softening process during the whole life. The hysteresis loop is fusiform, which means a well plastic deformability. With the increasing of the temperature, the deformability is enhanced and the fatigue characteristic is more distinct. Under the operation condition of the surge line(temperature is less or equal to 320℃), the strain has the primary effect on the fatigue life. The fatigue life under the 120~320℃ and 120~230℃ conditions is longer than the low cycle fatigue life under constant temperature 350 ℃, which means that the traditional test data of low cycle fatigue under high temperature is too conservative to assess the thermal mechanical fatigue life of the surge line.
Study on Automatic Measure Method of Rod Drop Time and Its Implementation
Li Guoyong, Jin Yuan, Zheng Gao, Xu Mingzhou
2016, 37(4): 53-57. doi: 10.13832/j.jnpe.2016.04.0053
Abstract:
Based on the analysis of the rod drop process of rod control cluster assemble(RCCA)and the action of control rod drive mechanism gripper and the response of the primary coil of the rod position detector, this paper presents the automatic judgement, the analysis of rod drop wave and the automatic calculation of the rod drop time, and performs the design with the virtual instrument technology. The relevant design is validated by the real rod drop wave, and the test result shows that it can well and truly capture the rod drop wave and automatically calculate the rod drop time. The production has been applied in Hainan Changjiang Nuclear Power Plant and Fujian Fuqing NPP unit 3 and unit 4.
Mitigation and Treatment of Accidents in HPR1000 SGTR
Xing Ji, Yu Pei, Li Jun
2016, 37(4): 58-62. doi: 10.13832/j.jnpe.2016.04.0058
Abstract(11) PDF(0)
Abstract:
HPR1000 provides mitigation measures for accidents in SGTR in the design. It can prevent SG overfill and mitigate SGTR accident by many means, such as decreasing the shutoff head of the safety injection pump and providing functions of fast cooling, automatic isolation and automatic control of auxiliary feedwater, and post-accident discharge of the blowdown system. The origin and mitigation measures for HPR1000 SGTR accidents are analyzed, and the process of accident treatment is described. Effectiveness of the mitigation measures to prevent SG overfill is verified, to satisfy the acceptance criteria for the radiological consequence of HPR1000.
Study and Simulation of Fuzzy Controller of Pressure of Pressurizer in PWRs
Qian Hong, Song Liang, Zhou Lei, Fang Zhenlu
2016, 37(4): 63-67. doi: 10.13832/j.jnpe.2016.04.0063
Abstract(10) PDF(0)
Abstract:
To improve the control performance of pressurizer pressure in PWRs, according to different controlled characteristics of pressure rising and reducing, the fuzzy adaptive PID control which can timely adjust parameter of controllers is considered, and the pressure control system using two fuzzy controllers is designed. Simulation on MATLAB/simulink demonstrates that the control performance of this pressurizer pressure control system using fuzzy controllers is obviously improved.
Study on Releasing Critical Current of Movable Armature of Control Rod Drive Mechanism
Li Yuezhong, Zhao Maomao, Wei Qiaoyuan, Zhang Fei, Ran Xiaobing, Duan Yuangang, Dai Changnian
2016, 37(4): 68-70. doi: 10.13832/j.jnpe.2016.04.0068
Abstract:
Based on the working principle of the releasing critical current of CRDM movable armature, academic analysis and experimentation were combined to study the relationship between releasing critical current and electromagnetic force. Some influencing factors were analyzed by investigation tests about the non-confirmation of movable armature releasing critical current. In conclusion, a new proper acceptance criterion and some improving advice of CRDM movable armature releasing critical current were proposed.
Dynamic Analysis of Control Rod Drive Mechanism under Stepping Loads
Chen Cheng, Liu Jinyang, Xu Yantao, Xie Yongcheng
2016, 37(4): 71-76. doi: 10.13832/j.jnpe.2016.04.0071
Abstract:
In order to calculate the dynamic response of the control rod drive mechanism(CRDM) under the stepping loads, a three-dimensional CRDM finite element model is established based on the theory of flexible multi-body dynamics, in which the coupling of rigid-body motion and elastic deformation is considered. The modal reduction method is used to reduce the number of deformation variables in the equations, so that such model is more efficient than the nonlinear finite element method in calculating the motion curve, impact load and stress response. In addition, the present model is more accurate than the conventional finite element method using the rod element.Not only the motion curve, but also the stress distribution of each component can be accurately obtained to analyze the weak position.
Diagnose and Treatment of Charging Pump Cavitation in Nuclear Power Stations
Yang Zhang, Wang Hexu, Jiang Yanlong, Sun Chengbin
2016, 37(4): 77-80. doi: 10.13832/j.jnpe.2016.04.0077
Abstract:
The typical vibration characteristics, cause and solutions of the cavitation problem of the horizontal-12stages-single suction-bag type-centrifugal charging pump(RHM100-205.12) has been analyzed. There were abnormal vibration fluctuation and fluctuation value beyond the alarm limitation phenomenon when charging pump runs on in-charge condition. It was found that there was typical vibratory shock signal in time domain; broadband noise, blade passing frequency and harmonic component and so on in frequency domain especially by charging pump’s entrance side in vertical direction. It was diagnosed that cavitation corrosion fault happened near the first vane wheel,which was approved by strip inspection during overhaul. The lower cavitation safety margin k of charging pump RHM was the main cavitation cause, which should be increased in order to avoid the cavitation corrosion fault. It has proved that the charging pump’s cavitation can be influenced by changing the cap between first impeller and vane by experiment study infield. The cavitation has been improved effectively through expanding the cap by 1 mm. Also it has proved that the cavitation of the charging pump RHM can be diagnosed simply by monitoring its vibration characteristics near the pump entrance side in the vertical direction and the cavitation can be improved effectively through expanding the cap.
Numerical Simulation of Wall Condensation Phenomenon of Steam in Containment Based on Code HYDRAGON
Hou Bingxu, Yu Jiyang, Jiang Guangming, Chen Bin
2016, 37(4): 81-86. doi: 10.13832/j.jnpe.2016.04.0081
Abstract(10) PDF(0)
Abstract:
In order to simulate the wall condensation phenomenon of steam in containment during a severe accident in a nuclear power plant, a wall condensation model is added in code HYDRAGON, a computational fluid dynamics(CFD) code specialized for the containment hydrogen analysis. The model, which is established on the analogy theory between heat and mass transference, provides boundary condition, mass source and heat source for Navier-Stokes equations. In order to validate the model and the code, a TOSQAN experiment is selected from published literatures as a test case to be compared with the simulation results. The study manifests that the computation results agree well with the experiment data. An analysis of the results also shows that the wall condensation phenomenon works in two aspects. On one hand, wall condensation decreases the amount of steam and mitigates the increase of pressure in the containment. Meanwhile, the percentage of non-condensable gases(e.g. hydrogen) increases. On the other hand, the near-wall convective heat transfer caused by condensation intensifies the flow of gas in the system. It is unfavorable for the formation of a stable stratification zone of hydrogen at the top of containment and can reduce the hazard of hydrogen explosion.
Optimization Scheme for Bypass Valve in Unit 1 of Yang Jiang Nuclear Power Plant
Liu Daoguang, Li Xianming, Zhang Xiaolei, Yu Hang
2016, 37(4): 87-89. doi: 10.13832/j.jnpe.2016.04.0087
Abstract:
After the replacement of FISHER by MASONEILAN to regulate the feedwater valve in Unit 1 of Yang Jiang Nuclear Power Plant, the parameters of the feed water bypass control system does not match the actual situation, and it directly results in the steam generator water level fluctuations, and thus brings the reactor trip. The control logic ARE407/408/409RG(small valve control function curve) curve function in ARE system is optimized, and the steam load and the discharge of water is matched by optimizing the control function to compensate for the inconsistent of field devices and control logic. Comparison of optimized and reference units showed that for the steam generator level disturbance at the switchover point for the small and large valves during the normal power rising and during the valve switchover test for 30%FP platform, the test result in Unit1 of Yang Jiang Nuclear Power Plant is significantly better than that for the reference unit.
Research and Design of Digital First Loop Hydrostatic Test Overpressure Protection Device
Shi Xiaowei, Li Jiurui, Yang Xiaoqi, Qiu Jianwen, Zhu Yuandong
2016, 37(4): 90-93. doi: 10.13832/j.jnpe.2016.04.0090
Abstract:
Considering the problem of serious aging, high cost for maintenance and upgrade of the original first loop hydrostatic test overpressure protection device of the Daya Bay nuclear power plant, a new digital overpressure protection device for the nuclear plant is designed to replace the old one, which is based on the Siemens S7-200 PLC system, and combined with the analog input module EM231, analog output module EM232 and human-machine module TD400 C, the new device can achieve dynamic real-time monitoring, analyzing and processing of the pressure data and simultaneously displaying of all the data, and output the control signal after logic operations on the test data during the hydrostatic test. Meanwhile a new offline test platform for the overpressure protection device is built to do offline testing and calibration. The new designed overpressure protection device is with advantages of more flexible, portable, high-precision, smart and easy maintenance. And the new designed equipment has been successfully used in Daya Bay and other nuclear plants for more than two years, to make a great contribution to shorten the overhaul period.
Research on Key Technology of LOCA Qualification Test of Residual Heat Removal Pump Motor
Huang Wenyou, Shuai Jianyun
2016, 37(4): 94-98. doi: 10.13832/j.jnpe.2016.04.0094
Abstract(11) PDF(0)
Abstract:
For the LOCA qualification test of residual heat removal pump motor of CPR1000 unit, the key technology are to realize the thermal shock in a LOCA test vessel of a test facility and the smooth running of the motor. A great flow rate saturated steam source with pressure 1MPa was introduced to the facility, which resulted in the critical steam flow during the shock process of the test. Combining the source with reasonable system design of the facility led to the shock about 12 seconds. A hydraulic dynamometer was employed as the motor load. The motor, installed inside the vessel, was connected with the dynamometer outside the vessel by an intermediate shaft. The position of the motor shaft and the intermediate shaft compared with that of dynamometer shaft changed continuously due to different temperature between inside and outside the vessel and different material. To reduce the vibration of the drive system, the intermediate shaft was fixed on the vessel wall, and larger deflection couplings and the pre-adjustment installation were adopted.During the test, the motor ran smoothly. The test for the full-scale motor with full load was conducted successfully.
Investigation on Spike of Source Range for RPN and Its Modification
Yang Wei, Li Fei, Lai Houjing
2016, 37(4): 99-101. doi: 10.13832/j.jnpe.2016.04.0099
Abstract:
Considering the abnormal spike of source count for RPN system present on several units of various nuclear projects of China General Nuclear Power Group, even causing the trip of the unit, it is considered as the very serious issue for the power increasing of the unit. In order to avoid this phenomenon, a professional site team is built to do the investigation and propose the modification. Taking units 2 and 3 of project B as examples in this paper, good experiences are implemented in the erection. Long time operation from commissioning, start-up and commercial operation verified that no abnormal spike count is observed.
Study on Overall Arrangement of Underground Nuclear Power Plant Based on Concept Design Site
Yu Fei, Li Maohua, Zhang Tao, Su Yi, Tang Yongtao
2016, 37(4): 102-107. doi: 10.13832/j.jnpe.2016.04.0102
Abstract:
Based on the landform and geological conditions of selected concept design site in the mountainous and hilly region, the suitable underground nuclear island form and overall arrangement basic model are selected. Considering the benchmark flood level, amount of earthwork excavation engineering and water head of circulating water, and so on, the ground plateau elevation for the first platform is determined, and based on the thickness of the coverage and the reactor cavity height, the ground plateau elevation of the second platform is determined. In accordance with the principle of overall arrangement, referring to the standards and regulations, the general arrangement of underground nuclear power plant CUP600 is studied to determine the main plant technical indicators of the layout scheme.
Research on General Layout of Nuclear Island of Underground Nuclear Power Plant
Xiao Yunfei, Tang Yongtao, Su Yingbin, Wang Shuai, Su Rongfu, Zhang Zhiguo
2016, 37(4): 108-112. doi: 10.13832/j.jnpe.2016.04.0108
Abstract:
The underground nuclear power plant takes the medium scale mountain terrain in hilly ground as its base, and the general layout of the nuclear island building is of "Terrace horizontal buried" style. Based on the layout scheme that locating the underground building of the nuclear island in L cavern and the ground building on the outside of the massif slop, the concept design of the general layout of the ground building of nuclear island as well as the equipment transport and personnel access, pipe and electrical cable connecting access of the underground building is conducted. The consequence by putting the traditional ground nuclear buildings under the ground is analyzed, and the analysis show that the layout scheme of the nuclear island building of the underground nuclear power plant is reasonable and feasible.
Design of Long-Distance Fuel Transfer System of Underground Nuclear Power Plant WENG Song-feng
Weng Songfeng, Chen Shuhua, Luo Ying, Huang Xindong, Zhan Hui, Tan Wenjie, Li Xiang, Su Yi
2016, 37(4): 113-115. doi: 10.13832/j.jnpe.2016.04.0113
Abstract:
For distribution change of underground nuclear power plant, the distance between reactor building and fuel build is increasing remarkably, so different section fuel transfer canal is designed to installation and maintenance, flexible drive and master/slave servo drive can suit to long-distance and front-back movement. This paper introduces the above system design and analyses influence.
Analysis of Surrounding Rock Stability of Underground Nuclear Power Plant Caverns
Zhang Zhiguo, Zhou Shuda, Chen Rui, Han Qianlong, Li Qing, Li Song
2016, 37(4): 116-120. doi: 10.13832/j.jnpe.2016.04.0116
Abstract:
Considering the characteristics of the underground nuclear power Caverns, the stability of the surrounding rocks in the excavation and earthquake is analyzed by the methods of project survey, comparison and numerical calculation. The results show that we have the experience of same size project as the underground nuclear power caverns in the underground project practice in China; and the nuclear power cavities are with strong seismic capacity. The safety and stability of the underground nuclear power caverns can be effectively guaranteed by the site selection, rational layout and suitable supporting.
Study on Start-up Characteristics of Passive Containment Cooling System for Underground Nuclear Power Plant
Li Feng, Zhang Shu, Zhang Dan, Ming Zhedong, Li Manchang, Yu Fei
2016, 37(4): 121-124. doi: 10.13832/j.jnpe.2016.04.0121
Abstract(14) PDF(0)
Abstract:
Passive containment cooling system(PCCS) is used to remove the heat in the containment. The mechanism is the natural circulation driven by the density difference between the downward-pipe and the upward-pipe. In the underground nuclear power plant, the difference of water level between heat exchanger and tank is 180 m, which leads to large pressure head to increase the capability of nature circulation. In this paper, the start-up process of PCCS is studied. It is found that the nature circulation may be not established if the initial temperature of the heat-removal tank is too high or the system layout is not reasonable. Meanwhile, flash distillation may occur in the backward of upward-pipe, resulting in the natural circulation oscillation and excurse.
Preliminary Analysis of Probabilistic Safety Assessment for Underground Nuclear Power Plant
Zhang Hang, Deng Chunrui, Kong Xiangcheng, Zou Zhiqiang, Zhang Dan, Wu Lingjun, Su Yi
2016, 37(4): 125-129. doi: 10.13832/j.jnpe.2016.04.0125
Abstract:
Based on the natural safety advantages of underground nuclear reactors such as the bounding rock and soil, considering depressurization cavern, isolation gate, filtration and venting system, and compared with that of the normal ground NPPs, the large release frequency(LRF) for underground NPPs is analyzed with probabilistic safety assessment(PSA) method. Analysis results show that the LRF of underground NPPs is at least two orders of magnitude lower than that of the ground NPPs, so the goal of practical elimination of large release by design could be realized.
Economic Analysis of Underground Nuclear Power Plants
Liu Haibo, Su Yi, Zhao Xin, Zhang Tao, Yu Fei, Li Xiang, Xu Yang
2016, 37(4): 130-132. doi: 10.13832/j.jnpe.2016.04.0130
Abstract:
Economy is an important factor for the construction of underground nuclear power plants. It is usually considered that the cost of underground engineering is worse than that of the nuclear power plant. But the conceptual design of the underground nuclear power plant shows that the addition static investment of the underground nuclear power plant is less than 12% compared to the same size and same reactor type nuclear power plant on the ground, which is in the acceptable range. Since the cost for the decommission of the underground nuclear power plant by burring is only 30% of the immediate decommissioning, less people and cost for the nuclear security is required comparing with ground nuclear power plant, and it is possible to eliminate the contingency plan area, the overall economy of the underground nuclear power plant is better than that of the ground nuclear power plant.
PIV Experiment Research of Lateral Flow Field in 2×2 Rod Bundles Channel with Mixing Vane Grids
Zhou Mengjun, Mao Huihui, Feng Ya, Yang Lixin
2016, 37(4): 133-137. doi: 10.13832/j.jnpe.2016.04.0133
Abstract(14) PDF(0)
Abstract:
Based on the scale-up model of 2×2 rod bundle channel, an experimental study of effects of mixing vane on lateral flow field in rod bundle channels of fuel assembly was performed with particle image velocimetry(PIV). The development of lateral flow velocity at different places after the fluid passing through mixing vanes was investigated. The effect of Reynolds number on the lateral flow field was gained by comparing the changes of the flow fields when bending angles equal 20°,25°, 30°, 35°, 40° and 45°. The characteristics of flow fields under four typical Reynolds numbers, which were 33000, 36000, 40000 and 45000, were gained. The results show that both of Reynolds number and bending angle can significantly influence the lateral flow in the rod bundle channel.
Experimental Research of Heated Upward Mixed Convective Heat Transfer
Chen Yuzhou, Yang Chunsheng, Zhao Minfu, Bi Keming, Du Kaiwen
2016, 37(4): 138-141. doi: 10.13832/j.jnpe.2016.04.0138
Abstract(13) PDF(0)
Abstract:
In the heated upward mixed convection, the distributions of velocity and shear-stress vary significantly due to the existing of buoyancy force, leading to the laminarization of boundary layer and the delay of the onset of fully developed turbulent convection, and the heat transfer coefficient experiences a process from the deterioration, recovery and enhancement. The present investigation deals with the experiments in tubes of different diameters for the forced convection of pure steam and the forced convection and natural circulation of supercritical water. The research indicates that as the buoyancy force increases, the heat transfer deteriorates and reaches a minimum at a certain value of buoyancy force, then it turns to recovery and enhancement finally. The present investigation presents a correlation of heat transfer coefficients based on the experimental data for heated upward flow.
Study on Pressurizer Characteristics and Stability of Load Tracking Transient and Reactivity Accident in Small Nuclear Heating Reactors
Xie Fei, Li Jingcai, Xie Heng
2016, 37(4): 142-147. doi: 10.13832/j.jnpe.2016.04.0142
Abstract(10) PDF(0)
Abstract:
Taking the Moroccan desalination projects(NHR-10) developed by Tsinghua University as the prototype, the small integrated natural circulation nuclear heating reactor model was established. By the study of the load tracking transient condition and reactivity accident, the pressurizer characteristics and stability of the performance of nuclear heating reactor were analyzed.The results indicate that the subcooling of outlet in integrated nuclear heating reactor is small enough to ensure that boiling does not occur during the transient process of reactor, which is good to PWR. The upper gas space volume has effect on the pressure trends of the reactor during the transient process. Initial subcooling of outlet has little effect on the transient process and final result.Integrated nuclear heating reactor has good stability in PWR run mode.
Simulation of Bubble Movement with Lattice Boltzmann Method
Chen Sixu, Li Longjian, Hu Anjie, Huang Yanping
2016, 37(4): 148-153. doi: 10.13832/j.jnpe.2016.04.0148
Abstract(10) PDF(0)
Abstract:
A multiphase Lattice Boltzmann free-energy model with two particle distribution functions is adopted in order to simulate a few kinds of bubble movements. The model itself is analyzed as well. Few problems were discovered during our simulations. Furthermore, an assumption was proposed to explain how buoyancy acts in the model according to the analysis of pressure field during the bubble rising. Based on the assumption, a proper method to introduce the buoyancy into Lattice Boltzmann free-energy model was proposed as well.
Development and Verification and Validation of MOC Module in Advanced Neutronics Lattice Code KYLIN-2
Chai Xiaoming, Tu Xiaolan, Lu Wei, Lu Zongjian, Yao Dong, Li Qing, Wu Wenbin
2016, 37(4): 154-159. doi: 10.13832/j.jnpe.2016.04.0154
Abstract(10) PDF(0)
Abstract:
In order to simulate the complex fuel assembly in the advanced nuclear reactor, the method of characteristics(MOC) calculation module is developed in neutronics lattice code KYLIN-2. The MOC module adopts cyclic trajectory layout and angle quadrature set for special geometrical shapes, and uses ray prolongation method for other general geometrical shapes. The generalized coarse mesh finite differential(GCMFD) acceleration method is adopted in the MOC module. The numerical results show that the MOC module in KYLIN-2 code can calculate 2D neutronics problems accurately, and can be used in future reactor design.
Study on Optimization Methods of Nuclear Reactor Radiation Shielding Design Using Genetic Algorithm
Ying Dongchuan, Xiao Feng, Zhang Hongyue, Lyu Huanwen, Tan Yi, Liu Jiajia, Jing Futing, Tang Songqian
2016, 37(4): 160-164. doi: 10.13832/j.jnpe.2016.04.0160
Abstract(13) PDF(0)
Abstract:
Based on the genetic algorithm, this paper focuses on the optimization of the nuclear reactor shielding design and conducts the studies about the single-objective and multi-objectives optimization problem of the reactor shielding design. The methods, developed in this paper, have been validated by the shielding of the Savannah nuclear ship. The effectiveness and correctness of the single-objective and multi-objectives optimization methods for nuclear reactor shielding have been fully demonstrated. A new technique has been provided for the shielding optimization of the reactor design in the future.
Research of Reliability Assurance Schemes for MICON Platform-Based Network Communication Protocol
Liu Zhaohui, Chen Zhi, Wu Zhiqiang, Liu Yao, Yang Xiaohua
2016, 37(4): 165-169. doi: 10.13832/j.jnpe.2016.04.0165
Abstract:
This paper proposed some design measures and assurance schemes based on the system structure characteristics and the system requirement of data transmission on the MICON platform which developed independently in China. The results of the field test show that the improved heartbeat mechanism, the bit inversion mechanism and the dual CRC checkout mechanism can be implemented successfully and enhance the system reliability. Meanwhile, these reliability ensuring measures can satisfy the system real-time demands.
ICS-AAS Measuring Lithium Concentration in Water Containing Boric Acid
Lin Qinghu
2016, 37(4): 170-172. doi: 10.13832/j.jnpe.2016.04.0170
Abstract(10) PDF(0)
Abstract:
The boron-lithium coordination solution was used to control the primary circuit water chemistry conditions of pressurized water reactor nuclear power plants. Under the existence of the boric acid in the primary circuit water, the lithium concentration recovery significantly decreased. In this study, we developed an ion chromatography suppressor-atomic absorption spectroscopy(ICS-AAS) online analysis system. Under the optimized conditions, the recovery of standard addition and relative standard deviations were 100% and 0.37(n=6). The system effectively eliminates the interference of boric acid, which demonstrates that the present system can be applied in the lithium concentration analysis in the primary circuit water of nuclear power plants. The system has the advantage of simple structure, convenient operation, low cost, no pollution by the boric acid in flame atomic absorption spectroscopy, and "zero" reagent consumption.
Analysis of Effect of Importance Sampling to Monte Carlo Simulation Efficiency
Lyu Jingbin, Guo Weiqun, Liu Baobao
2016, 37(4): 173-176. doi: 10.13832/j.jnpe.2016.04.0173
Abstract:
Based on a kind of radioactive aerosol spectrum monitor, the effect of importance sampling to Monte Carlo computing efficiency was analyzed in the paper. In order to transform the sampling method of polar angle from uniform sampling to logarithm sampling, one bias parameter was introduced in the analog simulation, and the weighted value of particle was adjusted finally to obtain correct results. In this way, the importance sampling method was implemented in the software. The importance sampling method was fully implemented in Geant4, and the effect of bias parameter on the computing efficiency was analyzed too. Simulation results indicated that the efficiency of simulation was largely improved by the importance sampling method, and the value of bias parameter should not be too large in practice.
Optimization of Technical Specifications Based on Probabilistic Risk Assessment
Cao Yong
2016, 37(4): 177-180. doi: 10.13832/j.jnpe.2016.04.0177
Abstract:
As a risk analysis method, the probabilistic safety assessment(PSA) has been widely used in the optimization of NPP Technical Specifications. Technical specifications based on PSA analysis have been developed and implemented in several NPPs in American. But there is no experience for the optimization of technical specifications for heavy water reactors in the world. In this paper, No.2 shutdown system test frequency was analyzed as an example, to study the applicability of PWR guidelines and practice in the heavy water reactors.
Application of PDMS in Process System Design of Radioactivity Waste Treatment Center
Li Ming, Ma Xingjun, Chen Li, Ma Zhenqin, Xiong Wei, Gao Feng
2016, 37(4): 181-184. doi: 10.13832/j.jnpe.2016.04.0181
Abstract:
A 3D factory design software PDMS was introduced in this paper, and its application in the process system design is presented with the example of modeling a radioactive waste treatment center. The advantages of PDMS code for the process system design are also summarized.