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2017 Vol. 38, No. 1

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Investigation of Correlation for Forced Convective Heat Transfer to Supercritical Carbon Dioxide Flowing in a Vertical Tube
Liu Shenghui, Huang Yanping, Liu Guangxu, Wang Junfeng, Zan Yuanfeng, Lang Xuemei, Huang Jun
2017, 38(1): 1-5. doi: 10.13832/j.jnpe.2017.01.0001
Abstract(16) PDF(0)
Abstract:
An experimental investigation of turbulent heat transfer in vertical upward and downward supercritical carbon dioxide flow was conducted in a heated bare tube with an inner diameter of 10 mm.In the experimental conditions,it shows that the phenomenon of buoyancy and acceleration is obvious.For the upward cases,as the buoyancy effect increasing the forced convective heat transfer of carbon dioxide will transient to mixed convective heat transfer,then free convective heat transfer,in which the heat transfer capacity is weakened firstly,then resumed and strengthened.The acceleration often weakens the heat transfer capacity both for upward flow and downward flow.A new heat transfer correlation for vertical upward flow of supercritical carbon dioxide was developed,by which 95.03% of the predicted Nusselt numbers agree with the experimental data within ±30%.
Experimental Study on Transient Characteristics of Passive Containment Cooling System
Cheng Cheng, Wen Qinglong, Lu Donghua, Wu Xiaohang, Niu Wenhua, Wei Shuhong
2017, 38(1): 6-9. doi: 10.13832/j.jnpe.2017.01.0006
Abstract(20) PDF(0)
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The transient characteristics of containment loop heat pipe cooling system are studied,and the characteristics of natural circulation flow rate,containment pressure and flow instability of the heat pipe cooling systems are analyzed by liquid head start-up based on the experimental data.The results show that the start-up of containment cooling system works well.In accident conditions,such as LOCA or MSLB,the heat of the containment vessel can be successfully transferred out.In the start-up stage,containment pressure drops rapidly,which is good for containment pressure inhibition.Flow instability of the containment cooling system happens under some conditions,which relevant to cooling-tank temperature,containment pressure and input heat power.Flow instability has no obvious effect on the trend of pressure in the containment.
Research of Material Irradiation Fine Neutron Flux Spectrum in HFETR
Liu Hongqian, Liu Shuiqing, Xiang Yuxin, Wang Hao, MA Liyong
2017, 38(1): 10-12. doi: 10.13832/j.jnpe.2017.01.0010
Abstract(14) PDF(0)
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The paper calculated and compared the fine neutron flux spectrum of the material at the typical irradiation channels of HFETR using MCNP code,including axial and radial distribution of neutron fluxes,and the 172-group neutron energy spectrum.It indicates that the distributions and characteristics of neutron energy spectrum in different channels filled with the same material,and in the same channel filled with different materials are almost identical,that is,the neutron energy spectrum with high energy is similar to fission neutron spectrum,the spectrums with middle energy and low energy are similar to Fermi spectrum distribution and Maxwell distribution separately.
Effect of Ocean Conditions on Transient Flowing Characteristics in Condensering Hot Well for Nuclear Power Plant
Li Yong, LiN Yuansheng, Guo Zhandong, Li Shumin, WEi Zhiguo, Kong Xiaming
2017, 38(1): 13-19. doi: 10.13832/j.jnpe.2017.01.0013
Abstract(15) PDF(0)
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During the marine nuclear power plant operation in harsh ocean environment,the large rocking motion of ship hull can induce condensate sloshing and pressure fluctuation in the condenser hot well,which has harmful effects on the operation safety of the condensate system.Hence,the transient flow and pressure fluctuation characteristics in the condenser hot well of the floating nuclear power plant are investigated by CFD numerical method under three rolling motion and three pitching motion.The results show that the condensation water pressure in the condenser hot well fluctuates periodically due to the rocking motion.The periods of pressure fluctuation and rolling motion are identical,whereas the fluctuation amplitude increases with the decreasing of the rolling motion period.Under the effect of rocking motion with small period,the inlet pressure of the condensate system falls below the minimal anti-cavitation pressure,which leads to potential cavitation hazard of the condensate system.The violent sloshing of condensation water level as well as the transient variation of mass force effecting on condensation water under rocking motion are two important factors for the transient thermal hydraulic characteristics in the condenser hot well.
Development of Metal Lithium Coolant Thermophysical Properties Calculation Model and Code for Space Reactors
Li Huaqi, Guo Xiaoyu, Yang Ning, Zhu Lei, Ma Tengyue, Hu Pan, Tian Xiaoyan
2017, 38(1): 20-24. doi: 10.13832/j.jnpe.2017.01.0020
Abstract(12) PDF(0)
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Through literature analysis and theoretical research,a calculation model is established for the function of the metal lithium thermophysical properties and the temperature and pressure.A metal lithium coolant thermophysical properties calculation code SNPS_LITHIUM is developed by using FORTRAN.The SNPS_LITHIUM can be used to calculate the thermophysical properties of metal lithium solid,liquid and steam.
Development of Coupled 2D/1D Whole Core Transport Calculation Code Based on DRAGON
Gao Mingmin, Cao Xinrong, Zhu Chenglin, Wang Jiangmeng
2017, 38(1): 25-28. doi: 10.13832/j.jnpe.2017.01.0025
Abstract(10) PDF(0)
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DRAGON code can conduct 3D method of characteristics(MOC) heterogeneous core calculations by introducing a lot of approximations and multiple acceleration methods,but the computing memory and time are still unaffordable for present engineering application.The transport calculation code DRAGON_HEU which is developed based on the 2D MOC calculation module of DRAGON code employs a coupled planar MOC solution and axial diffusion solution scheme,2D full core heterogeneous pin MOC calculation and 1D homogeneous axial pin diffusion calculation are coupled through radial and axial leakage under the 3D coarse mesh finite difference(CMFD) global framework,in such way the whole core Pin-by-Pin calculation is achieved.DRAGON_HEU code is testified with the C5G7 3D extension benchmark.Numerical results demonstrate that DRAGON_HEU code can achieve good accuracy while saving a large amount of time.
Modeling Method for Random Arrays of Rod in Spent Fuel Dissolver
Li Hang, Zhou Qi, Zhu Qingfu
2017, 38(1): 29-31. doi: 10.13832/j.jnpe.2017.01.0029
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The model for random arrays of shearing rod in spent fuel dissolver was built using Matlab commercial mathematical software.The model verification was tested using Monte Carlo neutron transport program MCNP and MONK9a and then the criticality calculation with different fuel components and length shearing rod was done.The stability of the random model is also analyzed.
Research on Hot Moist Air Reflow of Forced Draft Cooling Tower
Chang Liang, Li Lujun, Li Haiyan
2017, 38(1): 32-35. doi: 10.13832/j.jnpe.2017.01.0032
Abstract(13) PDF(0)
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In this paper,the three dimension numerical simulation models(FLUENT) are built based on some power plant engineering.The path lines of the hot moist of the forced draft cooling towers air are simulated,and the hot moist reflow phenomenon are studied.The effects of the operation plans,the weather conditions and the cross wind on the distribution the reflow rate of the hot moist air are studied.The result indicates that the maximum reflux rate reaches 17%,leading to the temperature of dry bulb at the air inlet increasing by 0.77℃ and wet bulb increasing by 1.67℃.
Preliminary Analysis of AP1000 PCCS and Its Enhanced Performance
Li Le, Li Cheng, Zhang Yajun, Zhu Mingliang
2017, 38(1): 36-40. doi: 10.13832/j.jnpe.2017.01.0036
Abstract:
To analyze the operation features of the Passive Containment Cooling System(PCCS) for the Advanced Passive Pressurized Water Reactor,the one-volume model was developed to calculate the heat transfer of the containment inner gas mixture.The previous multi-volume code for the containment external cooling by evaporation was adopted and accordingly the coupled code was used to calculate the LOCA accident.The simulation results fitted well with WGOTHIC and another reference one.By discussing the containment operation performances,the containment output performance was concluded.Finally,a suppression scheme was provided to improve the PCCS output performances.A new scheme is proposed which can suppress about 10% of the containment pressure and can reduce the engineering investments and difficulties for PCCS.
Analysis of Effect of Marangoni Convection on Formation of Microbubble Emission Boiling
LEi Yiqi, Mo Zhengyu, SuN Licheng, Xie Guo, Liu Hongtao, Du Min
2017, 38(1): 41-45. doi: 10.13832/j.jnpe.2017.01.0041
Abstract(14) PDF(0)
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In view of the extremely high heat transfer capability of Microbubble Emission Boiling(MEB) and the situation of unknown to its formation mechanism,a numerical investigation was carried out to compare the effects of Marangoni and buoyancy convection on the formation of MEB.This can also help researchers to understand the mechanism of the phenomenon of MEB.It is indicated that Marangoni convection has a much higher effect than the natural convection in the region around the vapor film.At subcooling of 40 K and wall superheating of 80 K,the average velocities caused by Marangoni effect and gravity are 0.4 m/s and 0.01 m/s,respectively.All of these indicated that Marangoni convection plays a key role in triggering MEB and enhancing the heat transfer process involved.
Research on Reactor Safety Analysis Method due to Thermal System Change of Secondary Circuit
Zhang Zhao, Tang Qi, WEi Yanhui, SHen Rongfa, SHeng Guolong, Yan Junjie, Zhong Daotong
2017, 38(1): 46-50. doi: 10.13832/j.jnpe.2017.01.0046
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Based on the quantitative analysis method of thermal economy for the secondary circuit of M310 PWR nuclear power plant,a reactor safety analysis method is proposed.With this method,the accident of two high-pressure reheater trains out of service is analyzed and calculated.This accident is also assessed on the full-scope simulator.The results by the method carried out in the paper agree well with those of the simulation,which indicates that the reactor safety analysis method is correct and valid.
Risk Evaluation for Induced Steam Generator Tube Rupture
Yang Jian, Zhu Wentao
2017, 38(1): 51-55. doi: 10.13832/j.jnpe.2017.01.0051
Abstract(11) PDF(0)
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In the progression of a severe accident sequence,the induced rupture of SG tubes is of most concern in level 2 PSA.The key factors of these phenomena are identified firstly,and then modeled in the Accident Progression Event Tree(APET) by the Risk Oriented Accident Analysis Method(ROAAM).The probabilities of important nodes in these event trees are determined by the combination of hydrogen-thermal analysis and parameter sample methods.Finally,the conditional probabilities and frequencies of induced SGTR during severe accident phase are calculated.The design characteristics for reducing the risk of induced SGTR are also discussed in this paper.
Creep-Fatigue Analysis and Evaluation of Thorium-Based Molten Salt Reactor Loop Pipe
Lu Xifeng, Wang Xinjun, Ai Honglei, Lyu Yongbo, HE Feng
2017, 38(1): 56-59. doi: 10.13832/j.jnpe.2017.01.0056
Abstract(20) PDF(0)
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The operation temperature of loop pipe is higher than 500℃ in the thorium-based molten salt reactor(TMSR).The pipe evaluation is based on rules for high temperature reactors.The stress,strain,deformation and creep-fatigue shall be evaluated for satisfying the structural integrity.The traditional method of ANSYS evaluating the creep-fatigue is the finite element analysis,but it is more complex and time-consuming and boring labor.The PIPESTRESS is used to analyze the loop pipe of TMSR.The results show that the stress indices are calculated and the creep-fatigue fast evaluation of the TMSR loop pipe is done by PIPESTRESS.The analysis method is a more effective and practical method used for the evaluation of the TMSR loop pipe.
Study on Calculation of Artificial Seismic Accelerograms
Wu Wanjun, Huang Xuan, SHen Pingchuan
2017, 38(1): 60-66. doi: 10.13832/j.jnpe.2017.01.0060
Abstract(15) PDF(0)
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The requirements of artificial accelerograms in nuclear power plant seismic design are reviewed,and comparison study is made on the variety of artificial accelerogams generation approach,the main difference about these approaches lies in how to obtain the initial accelerogams,including based on the real earthquake record which has a spectrum similar to design spectrum and the relation between response spectrum and power spectrum density.Using the later approach,a computer code is implemented for the calculation artificial accelerograms,and the calculation for specify design response spectrum is performed.The result shows that the time history has good configuration and statistic parameter to real earthquake record and is highly compatible to the design spectrum.The computer code is suitable for the artificial accelerograms calculation.
Multi-Physics Analysis of Metallic Fuel Deformation by Numerical Fuel Rod Code YUAN
FEi Jingran, Si Shengyi, CHen Qichang
2017, 38(1): 67-71. doi: 10.13832/j.jnpe.2017.01.0067
Abstract(20) PDF(0)
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Focusing on the notable radiation swelling deformation of metallic pellet,a tightly coupled multi-physics numerical model for fuel rod has been proposed and employed in a Numerical Fuel Rod code YUAN,which integrates neutronics,thermal-hydraulics and mechanics,and etc.based on unique floating finite element grids,to model the coupled fuel rod behavior of thermal conduction,mechanical deformation,neutron transport and isotopes depletion of fuel pellet/cladding.Typical metallic fuel is simulated by YUAN,and the deformation-caused multi-physical effects on thermal conduction and neutronics is analyzed.
Stress Corrosion Cracking of 316NG Stainless Steel Weld Joint in High-Temperature High-Pressure Water Containing Chloride
Wang Jiamei, HE Kun, Wang Li, Zhang Lefu
2017, 38(1): 72-76. doi: 10.13832/j.jnpe.2017.01.0072
Abstract(16) PDF(0)
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The effect of applied potential on stress corrosion cracking(SCC) behavior of stainless steel 316 NG weld joint was studied by slow strain rate tensile(SSRT) tests and high-temperature electrochemistry test in high-temperature high-pressure water containing high concentration chloride.The results revealed that the SCC susceptibility increases dramatically with the applied potential than the potential above a critical potential which exists between +50~+100mV(vs.SHE).The SCC susceptibility is low and no obvious intergranular stress corrosion cracks but only a little transgranular stress corrosion cracks can be found when the applied potential below this critical potential which corresponds to oxygenated water chemistry.When electrode potential is more than this critical potential,the weld joint exhibits significant SCC appearing large area intergranular stress corrosion cracks.Besides,in Ar and low corrosion(electrode potential ≤50mV) environments,the weld joint cracking is plastic fracture determined by the mechanical properties,which related to the hardness distribution along the joint.In high corrosion(electrode potential >50mV) environments,the weld joint cracking is brittle fracture determined by the corrosion resistance and apparently the weld and heating affected zone exhibit higher SCC susceptibility than the base metal.
Research on Welding Deformation Control between AP1000 RCP Casing and Steam Generator
HE Guangqing, Li Cuicui, Yang Dongsheng
2017, 38(1): 77-81. doi: 10.13832/j.jnpe.2017.01.0077
Abstract(14) PDF(0)
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The weld joint between reactor coolant pump(RCP) casing and steam generator is important pressure boundary,and the nozzle position tolerance of RCP casing will impact consecutive assembly of the RCP and pipe.This paper gives the example of the domestic first equipment fabrication,aiming at the welding deformation and easily nonconformance condition,and proposes some control measures avoiding the exceeding position tolerance due to serious weld deformation,such as RCP casing assembly alignment,weld preparation design and process control.The data shows that:the four as-built RCP casings according to the welding process,the discharge angle,angle of pitch between the main flange surface and the steam generator datum A and axes direction welding shrinkage is acceptable,and meet design requirements.
Improment of High Pressure Hydrogen Injection System in Sanmen Nuclear Power Plant
Wang Xu
2017, 38(1): 82-84. doi: 10.13832/j.jnpe.2017.01.0082
Abstract(12) PDF(0)
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During nuclear power plant operation,the coolant in the primary loop will decompose to hydrogen and oxygen under radiation condition.Oxygen will aggravate the corrosion of stainless steel in the primary loop,reducing the equipment reliability and increasing the radioactive activation products.Normally nuclear power plants use the primary loop hydrogen injection to inhibit the radiation decomposition of the primary coolant.Traditional nuclear power plant uses the chemical and volume control tank to reduce the pressure of coolant and inject low pressure hydrogen.AP1000 use the high-pressure hydrogen injection technology,because there is no chemical and volume control tank in AP1000 nuclear power plant.This paper introduces the current high-pressure hydrogen injection scheme used by AP1000,analyzes the possible problems during operation,and suggests the improved scheme.
Feasibility Study on Large-Loaded Hydraulic Snubber Seal Life Extension Used in Primary Equipment Supports of Nuclear Power Plants
Xie Honghu, Yang Jinchun
2017, 38(1): 85-87. doi: 10.13832/j.jnpe.2017.01.0085
Abstract(12) PDF(0)
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Based on the variation of compression set and service time,the projected seal life of the Lisega large loaded hydraulic snubber uninstalled from Mc Guire nuclear power station Unit 2 after maintenance free service for over 20 years are calculated.The results show that,taking the large loaded hydraulic snubber provided by Lisega as an example,the projected seal life for external seal and a seal within the fluid boundary is much longer than the replacing cycle for these seals regulated in nuclear power stations in operation or under construction.
Research on Availability of CIS Digital Module as Monitoring and Radiation Alarm Equipment under Condition of Nuclear Accidents
Xu Shoulong, Zou Shuliang, Wu Zhao, Luo Zhiping, Huang Youjun, Song Luying
2017, 38(1): 88-94. doi: 10.13832/j.jnpe.2017.01.0088
Abstract:
The availability of CMOS image sensor(CIS) digital module has been analyzed and studied as a monitoring and radiation warning equipment in the nuclear accident conditions.The radiation damage and the interference mechanism were studied based on the analysis of the color and the gray level information of three kinds of CIS modules in real time.The availability of the CIS digital module as a monitoring and radiation warning equipment is studied.The study proves that the CIS digital module can not only maintain a better monitoring picture quality,but also can reflect the intensity of the radiation field of the monitored environment in a short time radiation environment with low dose rate.Therefore,it is concluded that the CIS digital module can be used as a real time monitoring and the dark image noise is related to the intensity of the radiation field in the monitored environment.
The Research on Flywheel and Water-Lubricated Bearing of the AP1000 Reactor Coolant Pump
Wang Lilai
2017, 38(1): 95-98. doi: 10.13832/j.jnpe.2017.01.0095
Abstract(17) PDF(0)
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Based the theoretical derivation and the canned motor RCP design,operation and troubleshooting experience,the submerged flywheel friction losses is analyzed and calculated.According to the Double-Acting Pivoted Pad Thrust Bearing tribology design principles,the impact of the submerged flywheel on the AP1000 reactor coolant pump performance is analyzed.
Analysis of PSI Defect Problem of CPR1000 Reactor Pressure Vessel
Zhang Jin, Wan Zhijian, Li Jiakang, Dong Yiling, Deng Xiaoyun, Liu Pan
2017, 38(1): 99-103. doi: 10.13832/j.jnpe.2017.01.0099
Abstract(11) PDF(0)
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During the pre-service inspection of a certain CPR1000 reactor pressure vessel(RPV),unacceptable defect is found in the weld between outlet nozzle and flange-nozzle shell which complies with the requirements of ultrasonic testing(UT) during manufacturing.To solve the problem,the comparison of UT technique employed during manufacturing and PSI has been demonstrated.It is proved that the immediate cause of problem is that the acceptance criteria of manufacture according to RCC-M code has been applied to UT technique of PSI,and the root cause is the differences of the sizing method and the grouping requirements for defects between manufacturing and PSI.On account of PSI is intended to act as a zero point for in-service inspection,the view point is presented that unacceptable defects found during PSI,if possible,should be re-examined and accepted in accordance with the testing method and acceptance criteria of manufacturing.
Analysis of Deviation of Medium Pressure Control Valve Position Feedback and Instruction Run Over the Limit in Nuclear Power Plant Turbine Fast Cut Back
Gao Jianqiang, Bian Yan, Fang Lijun, Rao Wan
2017, 38(1): 104-107. doi: 10.13832/j.jnpe.2017.01.0104
Abstract(13) PDF(0)
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In the 1000 MW nuclear power units FCB(fast cut back) test,the medium pressure control valve can not be closed in time,the deviation of the valve position feedback and instruction run over the limit ocurred,which caused the turbine trip protection action.Through the analysis,it was found that the mismatch of the aerodynamic moment and the spring moment was the cause of the problem.The three-dimensional flow field was numerically simulated by FLUENT,and the change regulation of aerodynamic moment was obtained.The conclusion shows that the aerodynamic moment first increased and then decreased with the increasing of opening degrees and the maximum value was reached when the opening degree was around 30°.At this point,if the aerodynamic moment did not match the spring moment,the medium pressure valve would close slowly and even clamped,leading to the deviation of the valve position feedback and instruction run over the limit.
Application of Main Pump Lifting Device in Nuclear Power Plant NCC&CFT
Xu Jianhui, Wang Yuxu, Wu Hangquan
2017, 38(1): 108-109. doi: 10.13832/j.jnpe.2017.01.0108
Abstract(13) PDF(0)
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CPR1000 nuclear power plant primary circle washing test(NCC) and cool function test(CFT) set and commissioning is introduced.Without the nuclear power plant main pump motor,a test on use of main pump rotor lift device shows that it is safe and reliable.This lift device can be used in CPR and EPR nuclear power plants to reduce the project risks.
Development and Application of Preventive Maintenance Strategy Template in Pressurized Water Reactor Nuclear Power Plants
Zhang Sheng, Zhang Tao, Chen Yu, Jiang Hong, He Shanhong, Mo Chunni
2017, 38(1): 110-115. doi: 10.13832/j.jnpe.2017.01.0110
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Based on the maintenance template database and its application technology in foreign countries,and combined with the actual development in domestic NPPs,this paper establishes the maintenance strategies template for domestic NPP equipment,presents the development process of maintenance template for domestic pressurized water reactor(PWR) nuclear power plants,and also demonstrates the application of maintenance template in the maintenance program development and optimization combined with cases.
Research on Improvement of Torque Jitter Fault in EAS013/014VB of Nuclear Power Plants
Zhang Huaqi, Yang Jiansong, JiA Yong, Yuan Fengwu
2017, 38(1): 116-119. doi: 10.13832/j.jnpe.2017.01.0116
Abstract(15) PDF(0)
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This paper applies the qualitative analysis and experimental verification to the torque hopping problem of EAS013/014 VB electric equipment of Nuclear Power Plant by applying the root cause analysis method.Valve start,the inertia is too large to make the remote transmission speed of 120 r/min caused by the motor worm move in short time,so that the torque switch protection,the performance of torque jitter.Through the analysis of the principle of the electric device and field experiments,it is concluded that improper selection of the electric device,which leads to the torque jump of the electric device,has important reference significance for the selection of the electric device with the remote control operation.
Research on Multi-Objective Optimization Method for Maintenance Decision of Nuclear Power Plants
Lyu Yan, Liu Jingquan, Zeng Yuyun
2017, 38(1): 120-125. doi: 10.13832/j.jnpe.2017.01.0120
Abstract(14) PDF(0)
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This paper proposes a new kind of maintenance model which characteristic is in the system level,incompletely maintenance and sequential maintenance.The model introduces age reduction factor measuring the maintenance effect and obtains the system reliability by the adjacency matrix method and the structural function matrix method.At the same time,the influence of the replacement of each component on the system reliability and total cost is taken account,and the formula for calculating the reliability and the total cost is derived.Then,using the NSGA-II method to make the multi-objective optimization of the model,obtaining the multi-objective maintenance plan about a system for the whole life of a nuclear power plant and making suggestions for maintenance decision making.Finally,an example is presented to verify the calculation steps,the economic analysis and multi-objective optimization of the maintenance program.
Separation Model and Experimental Research on Purification Column in Liquid Waste Treatment System of Nuclear Power Plants
Zheng Wei, Li Qing, LiN Peng, Liu Xiajie, Lu Jie, Lyu Yonghong
2017, 38(1): 126-130. doi: 10.13832/j.jnpe.2017.01.0126
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Separation level of non-volatile radioactive substances was analyzed by theoretical method,and specially designed purification tower was tested.The results showed that the experimental result was unanimous of which was gained by theoretical method.So the theoretical method and calculation model were rational.The theoretical model was applied on the evaporation equipment in the liquid waste treatment TEU system of a nuclear power plant,and the results showed that 3 sieve plates with wire mesh demister was rational and it could satisfy the requirement of nuclear power plant with a high safety margin.
Research of MCR Inleakage Tracer Gas Test of AP1000 Nuclear Power Plant
Yang Yang
2017, 38(1): 131-134. doi: 10.13832/j.jnpe.2017.01.0131
Abstract(14) PDF(0)
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Combining with the sealed design of AP1000 MCR and passive design of MCR ventilation systems,this paper explores the method and theory which is suitable for AP1000 main control room inleakage tracer gas test and anticipates the test results and difficulties.Results show that the constant injection test is the preferred method for AP1000 main control room inleakage tracer gas test.The constant injection test based on mature theories and with strong operability and high precision,can well meet the AP1000 passive design verification requirements.
Research of High Positional Accuracy of Manipulator Crane for Floating Nuclear Power Plants
An Yanbo, Luo Ying, Liu Cong, Tan Yue, Huang Hui, Chen Shuhua
2017, 38(1): 135-138. doi: 10.13832/j.jnpe.2017.01.0135
Abstract(16) PDF(0)
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This paper builds a mathematics model of the positioning accuracy of the manipulator crane for floating nuclear power plants according to the manipulating process,and particularly analyzes every part relevant to positioning accuracy.On that basis,a kind of high positioning accuracy technology is provided:the technology to combine the open-loop and the close-loop compensating.This technology can increase the positioning accuracy of floating nuclear power plant manipulator evidently,and satisfies the high positioning accuracy of floating nuclear power plant manipulator crane.