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2016 Vol. 37, No. 3

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Overview of Research and Development on Underground Nuclear Power Plant
Niu Xinqiang, Luo Qi, Zhao Xin, Zhang Wenqi
2016, 37(3): 1-5. doi: 10.13832/j.jnpe.2016.03.0001
Abstract:
In cooperate with Nuclear Power Institute of China, Changjiang Institute of Survey, Planning, Design and Research has completed the demonstration concept design research of underground nuclear power plant and proposed the underground nuclear power plant with independent intellectual property rights in China for the first time, and preliminarily formed complete sets of underground nuclear power plant technology. CUP600’s technical and economic parameters have reached the standards of the third generation nuclear power plant. This paper introduces the overall goal, technique route and research process of the research project of the underground nuclear power technology, and describes the setting of the key advisory project of underground nuclear power plant in China academy of engineering in detail with emphasis on the summary of the original and breakthrough achievements in the technology research stage.
Overall Design of Chinese Underground Nuclear Power Plant
Niu Xinqiang, Li Xiang, Li Qing, Zhao Xin, Li Manchang, Liu Haibo
2016, 37(3): 6-9. doi: 10.13832/j.jnpe.2016.03.0006
Abstract(12) PDF(0)
Abstract:
On the basis of the independent research and development achievements, in cooperation with Changjiang Institute of Survey, Planning, Design and Research, Nuclear Power Institute of China(NPIC) finished the conceptual design of Chinese underground nuclear power plant, and proposed the design scheme for the first Chinese underground nuclear power plant called CUP600 with independent intellectual property rights. This paper summarizes and illustrates the overall design issues of the CUP600, such as overall technical requirements, main technical selections and main technical parameters.
Theoretical and Experimental Investigation on Flow-Resistance Performance of Natural Circulation System
Wu Lei, Jia Haijun, Liu Yang, Zhang Tao, Ma Jizhong, Yang Xingtuan
2016, 37(3): 10-15. doi: 10.13832/j.jnpe.2016.03.0010
Abstract(11) PDF(0)
Abstract:
Experiments prove that the relations between the natural circulation capacity G and the heating power Q, and the total flow resistance Δpf and the natural circulation capacity G are: G~Qm and Δpf~Gq, respectively. Power indexes of m and q are defined as the natural circulation flow rate-heating power characteristic number and the natural circulation flow-resistance characteristic number, respectively. In consideration of the effects of local resistance and frictional resistance, the analytical relations about thermal parameters are obtained by using approximation and fitting approach, and they also have good agreement with the experimental results. It is indicated that the natural circulation flow rate-heating power characteristic number m has a close relation with the flow resistance characteristics, and it is a function of Reynolds number Re and the ratio of the local resistance coefficient and the frictional resistance coefficient Rn in the natural circulation system. Its value decreases with the increasing of Rn. In addition, the relation between the natural circulation flow-resistance characteristic number q and the natural circulation flow rate-heating power characteristic number m is: q =(1-m)/m, which is validated by the experiment.
Extending and Verification of RELAP5 Code for Liquid Fueled Molten Salt Reactor
Shi Chengbin, Cheng Maosong, Liu Guimin
2016, 37(3): 16-20. doi: 10.13832/j.jnpe.2016.03.0016
Abstract:
In order to analyze the liquid fueled molten salt reactor using RELAP5 code, models in RELAP5 code need to be extended. This paper attempts to add new point kinetic model of liquid fueled reactor and thermo-hydraulics model with internal heat source based on the original RELAP5 models, then the code is verified using MSRE experimental datum. The results indicate that the extended RELAP5 code can be applied to model and analyze the liquid fueled molten salt reactor.
Theoretical Research on DNB-type Critical Heat Flux in Uniform Heating Narrow Rectangular Tube
Zhao Dawei, Liu Wenxing, Xiong Wanyu, Yang Zumao, Huang Yanping
2016, 37(3): 21-25. doi: 10.13832/j.jnpe.2016.03.0021
Abstract(15) PDF(0)
Abstract:
Based on liquid sublayer dryout mechanism, a critical heat flux(CHF) model has been developed for the prediction of DNB-type CHF. In the model, the forces exerted on the vapor blanket are analyzed to calculate the boiling crisis parameters including vapor blanket velocity, sublayer thickness with modified lift force coefficient. The predictions of present CHF model are verified by CHF experimental results, which are obtained in two narrow rectangular test sections with different heating length. The deviations between the CHF predictions and CHF experiment results are within ±15% error band, which show more reasonable accuracy than using Bowring correlation and Bettis correlation.
Bubble Coalescence Characteristics in Micro-Scale Nucleate Boiling
Bi Jingliang, David M. Christopher, Xu Jianjun, Huang Yanping, Zan Yuanfeng
2016, 37(3): 26-30. doi: 10.13832/j.jnpe.2016.03.0026
Abstract(11) PDF(0)
Abstract:
Micro-scale bubble growth, coalescence and heat transfer characteristics are numerically investigated. Bubble interface is tracked with VOF(Volume of Fluid) method in CFD(Computational Fluid Dynamics) software. A 3 mm×2 mm×2 mm 3-D micro-channel is set up and 5 heating elements are arranged at the bottom surface. These elements can measure the local heat flux as well as provide the power for the bubble nucleation. Bubble growth and coalescence characteristics and heat flux distributions under bubbles are analyzed and compared with the experimental observations. The heat transfer mechanisms in nucleate boiling are better interpreted by developing the correlations of bubble dynamics and heat flux variations at different positions.
Study on Flow Similarity Laws between Water and Liquid Lead-Bismuth under Natural Circulation
Zheng Jie, Chen Zhao, Zhao Pengcheng, Chen Hongli
2016, 37(3): 31-33. doi: 10.13832/j.jnpe.2016.03.0031
Abstract:
The natural circulation lead-bismuth cooled reactor is an important direction for the reactor development. Understanding of the natural circulation flow characteristics of lead-bismuth is crucial to the development of natural circulation lead-bismuth cooled reactor. Experimental modeling similarity research is one of the most extensive method for this. Based on the similarity theory, this paper adopts a combination of theoretical derivation and numerical simulation to explore the possibility of water to simulate the flow characteristics of lead-bismuth. Final results showed that the flow characteristics of lead-bismuth can be simulated by water successfully under steady state and transient state.
Thermodynamic Analysis of Coupling Supercritical Carbon Dioxide Brayton Cycles
Huang Xiaoli, Wang Junfeng, Zang Jinguang
2016, 37(3): 34-38. doi: 10.13832/j.jnpe.2016.03.0034
Abstract:
Based on the first law of thermodynamics, a thermodynamic investigation was carried out on the coupling supercritical carbon dioxide Brayton cycles. The thermodynamic behavior and parameter limitations of the recompression cycle have been analyzed for a heat source system on the basis of equipment model and assumed initial conditions. A new cycle layout, named the cascaded partial flow cycle, was proposed for a high temperature difference heat source system. The thermodynamic behavior of the cascaded partial flow cycle was analyzed and evaluated. The applicable target of the recompression cycle and the cascaded partial flow cycle were proposed based on quantitative comparisons.
Analysis of Structures and Heat Transfer for Packed Beds
Li Rui, Ren Cheng, Yang Xingtuan, Wu Hao, Jiang Shengyao
2016, 37(3): 39-42. doi: 10.13832/j.jnpe.2016.03.0039
Abstract(13) PDF(0)
Abstract:
In order to analyze the structures of packed beds, the discrete element method(DEM) reveals great advantages for numerical simulation. There are only few particles at densest state in random packed beds. Porosity increases rapidly with the oscillation at near-wall region. More local microstructure information will be obtained with Voronoi tessellation. Packing characteristics at the bulk region is random and ordered at the wall region instead. With homogenization assumption at Voronoi tessellation, a heat transfer model is established which can take the conduction between adjacent spheres and thermal radiation heat exchange into account and the numerical results agrees well with the experimental data.
Suggested Spectral Accelerations for Seismic Margin Assessments of Nuclear Power Plants Based on Statistics
Wang Yushi, Li Xiaojun, Zhao Lei, Hou Chunlin
2016, 37(3): 43-46. doi: 10.13832/j.jnpe.2016.03.0043
Abstract(12) PDF(0)
Abstract:
Based on the statistics of 350 sets of strong-motion acceleration records on bedrock in Next Generation Attenuation(NGA) database of the US and 14 sets of strong-motion acceleration records on bedrock in Wenchuan MW7.9 earthquake and Lushan MW6.6 earthquake, normalized horizontal spectral accelerations on bedrock for seismic margin assessments of nuclear power plants were suggested. The effects of earthquake magnitude on frequency contents of strong-motions were adequately taken into account in this suggested spectral acceleration, which could help evaluate the effects of different seismic tectonic environments on normalized spectral acceleration inputs in the seismic margin assessments of different nuclear power plants. In comparison to the normalized spectral accelerations suggested in RG1.60, this suggested spectral acceleration could reflect the high frequency contents of strong-motions induced by medium-strong earthquakes in near field more reliably.
Seismic Response Analysis Based on Dynamic Artificial Boundaries for Nuclear Power Engineering
Li Zhongcheng, Fan Hong, Li Jianbo
2016, 37(3): 47-50. doi: 10.13832/j.jnpe.2016.03.0047
Abstract:
Dynamic artificial boundary is a necessary means to using the 3D finite element technique to solve the dynamic problem of foundation. Based on the viscoelastic boundary and the transmitting boundary developed, the seismic response analysis of a building structure in a typical example of nuclear power engineering in the northeast of China is carried out, and the contrast analysis of two kinds of complex artificial foundation boundary model results with simplified homogeneous foundation model is implemented in order to explain the analysis differences based on the different methods in the simulation of dynamic effects for complex foundation.
Quantification Method for Nuclear Power Plant Seismic Risk
Wang Jinkai, Lin Modi
2016, 37(3): 51-53. doi: 10.13832/j.jnpe.2016.02.0051
Abstract(11) PDF(0)
Abstract:
In the seismic risk analysis for the nuclear power plant, the appropriate data processing method and analytical technique are required. In this study, seismic risk quantification software is developed independently for the first time in China. It uses Monte Carlo sampling method to simulate the seismic frequencies and component conditional failure probabilities, and integrates with accident sequences to assess the plant seismic risk. This method overcomes the shortcomings of the traditional probabilistic risk assessment modelling software in the seismic risk evaluation. Compared with the similar software abroad, this software is more reasonable in the uncertainty process.
Effects and Mechanism of Ti Substitution on the Ability of Anti-Disproportionation of Zirconium Cobalt–Hydrogen System
Zhang Guanghui, Sang Ge
2016, 37(3): 54-56. doi: 10.13832/j.jnpe.2016.03.0054
Abstract(17) PDF(0)
Abstract:
ZrCo alloys with Ti substitution were prepared via the arc-melting method, and then the products before and after hydrogenation were characterized by X-ray diffraction. Results showed that the crystal structure of ZrCo alloys substituted with Ti substitution formed cubic phase, and the lattice parameters of ZrCo alloys and these hydrides decreased with Ti substituted. And the kinetics of the hydrogen-induced disproportionation in the desorption mode for all these alloys was also investigated. Results demonstrated that the rate and extent of the disproportionation of ZrCo alloys decreased with the content of Ti substitution. It could be inferred that the effect of element substitution on the disproportionation of ZrCo alloys was caused by the radius change of hole sizes of hydrogen occupation sites.
Effect of Temperature on Pickling Reaction Kinetics of Zirconium Alloy
Liu Yunming, Li Chuanfeng, Chen Jiangang, Liu Lijian, Du Peinan, Qian fang, Wang luquan
2016, 37(3): 57-60. doi: 10.13832/j.jnpe.2016.03.0057
Abstract(14) PDF(0)
Abstract:
The effect of temperature on pickling rate of zirconium alloy and the heat release constant was studied in this paper by soaked method. The results show that temperature has exponential effect on pickling rate of zirconium alloy, the rate increases 2.0~2.5 times 10℃ at 20~60℃. The pickling frequency factor is about 1.18×1012 and the activation energy is about 70 k J/mol. The pickling reaction of zirconium alloy is exothermic. The reaction temperature increases with the pickling process in which the heat release constant is 760 k J/mol.
Research on Uniform Corrosion of Inconel 690 Alloy for Stream Generator in Nuclear Plant Water Environment
Dang Ying, Lin Zhenxia, Pan Xiaoqiang, Li Weijun
2016, 37(3): 61-65. doi: 10.13832/j.jnpe.2016.03.0061
Abstract(11) PDF(0)
Abstract:
Under the stimulated nuclear plant water environment, corrosion property and oxide film characteristics of three kinds of commercial 690 alloys for steam generators were studied in the flow-water corrosion loop. Meanwhile, the general corrosion rate and corrosion product release rate were also estimated using Chinese standard and American standard respectively. The results show that: under the stimulated nuclear plant water environment, 690 alloy tube exhibits very low corrosion rate and corrosion product release rate, and the corrosion performance of Japan Sumitomo 690 alloy tube is better than that of Baosteel 690 alloy tube.
Impact Wear Behavior of Stellite-6 of Grippers of CRDM
Zhou Jun, Chen Yong, Luo Qiang, Wang Kun, He Kun, Lin Zhenxia
2016, 37(3): 66-69. doi: 10.13832/j.jnpe.2016.03.0066
Abstract:
Simulating the water chemistry of the PWR primary loop, the impact wear tests of stellite-6 which used in CRDM were carried out at different temperature and different impact force by using a specially designed impact wear test device. Results showed that the effect of temperature was not obvious when the sample was under normal impact force. The size loss of the sample was approximate 1×10-8 mm, and the weight loss of the sample was approximate 2×10-6 mg, after 1 cycle impact. The primary wear mechanism of stellite-6 is plastic deformation and fatigue flack under high impact force. The size loss of the sample was approximate 1.3×10-7mm, 4.7×10-7 mm and 5.3×10-7mm, and the weight loss of the sample was approximate 7.5×10-6 mg, 4.17×10-5mg and 4.83×10-5mg at room temperature, 90℃ and 150℃ respectively.
Applicability Research of RELAP5 for Steam Generator Tube Rupture Accident of AP1000 NPP
Fang Jun, Wu Nan, Qi Ting
2016, 37(3): 70-74. doi: 10.13832/j.jnpe.2016.03.0070
Abstract(12) PDF(0)
Abstract:
Steam generator tube rupture(SGTR) accident is a significant design basis accident which shall be analyzed in PWR nuclear power plant safety analysis reports. Thermo-hydraulic analysis of SGTR accident for licensing has been done using a specific code entitled LOFTTR2. To validate the applicability of RELAP5 for AP1000 SGTR accident, a RELAP5 model for AP1000 is established and used to assess this accident. Results show that both the accident progression and the doses result of RELAP5 calculation correspond well with that in Safety Analysis Report. Thus, RELAP5 and RELAP5 model of AP1000 can apply to other SGTR type analysis.
Investigation on Uncertainty Quantification Method in Realistic LOCA Analysis
Lin Zhikang, Wang Ting, Lin Jianshu, Liang Ren, Lu Xianghui
2016, 37(3): 75-79. doi: 10.13832/j.jnpe.2016.03.0075
Abstract(10) PDF(0)
Abstract:
In this paper, we introduce many kinds of uncertainty quantification methods used in realistic LBLOCA analysis, and quantify the uncertainties of the output of large break loss of coolant accident based on the best estimate thermal hydraulic system code CATHARE GB LBLOCA model of CPR1000 nuclear power plant. Furthermore, we compare and identify the differences in many areas such as the process of input and output parameters and analysis procedure, and recognize that the sensitivity study method is most conservative because of the existing of the artificial conservative assumption, and the normal distribution test is required in the conventional parameter statistic method, Owen factor method and Bootstrap method, but not in Wilks method, and Wilks method can achieve more realistic results.
Uncertainty and Sensitivity Analysis of Numerical Simulation Results of Hydrogen Recombiner Case
Hou Bingxu, Yu Jiyang, Zhong Xianping, Jiang Guangming, Zou Zhiqiang
2016, 37(3): 80-86. doi: 10.13832/j.jnpe.2016.03.0080
Abstract(17) PDF(0)
Abstract:
In the containment hydrogen analysis, there is uncertainty in the computation results because uncertainty exists in the input parameters. It is of great importance for the safety issue to study the variation ranges of the computation results as well as the contribution of each input parameter to the uncertainty of the results. In order to numerically simulate the hydrogen recombiner case in the containment, the recombiner model is firstly added to the computational fluid dynamics code HYDRAGON. Deflagration-detonation-transformation index, total explosion energy and related instance variables, are selected as objects of the study. Numerical simulations are performed after sampling for certain input parameters. The uncertainty of the computational results is studied with nonparametric statistical method and the variation ranges of the results are provided. Furthermore, the sensitivity between the computational results and the input parameters are investigated. The input parameters, whose uncertainty has larger influence on the computational results, are also screened out.
Effect of Blade Inlet Position on Flow Characteristics of Nuclear Reactor Coolant Pump under Gas-Liquid Two-Phase Condition
Fu Qiang, Xing Shubing, Zhu Rongsheng, Li Tianbin, Wang Xiuli
2016, 37(3): 87-93. doi: 10.13832/j.jnpe.2016.03.0087
Abstract(10) PDF(0)
Abstract:
In order to study the effect of blade inlet position on flow characteristics of nuclear reactor coolant pump under gas-1iquid two-phase condition, three different schemes of inlet position had been designed and simulated on steady and unsteady stage under gas-liquid two-phase condition. The analysis of the results shows that the blade inlet of the nuclear coolant pump protrusive properly, can be good to keep the stability of the pressure boundary in the LOCA(Lost of Coolant Accident) event. But the protrusion can exacerbate the degree of blade twist, and lead to a large number of bubbles accumulating on the suction surface. The blade inlet offsets backwards, easy to cause strong jet trails and large amplitude pressure pulsation at the outlet of the impeller. By the comparative analysis, the scheme B turns out to be the optimal solution. The external characteristics of the optimal model prototype, based on the simulation, was tested on different inlet gas rate conditions. The rustle shows that predicted curve is basically consistent with the experimental one under 0% inlet gas operation condition and the performance meets the design requirements. With the increasing of the gas rate, large deviation appears in the experimental values and simulation values. This is mainly due to the limited test conditions, and the error of simulation results with the experimental results.
Scaling of AP Containment Wall Heat Removal
Li Cheng, Li Le, Zhang Yajun, Li Junming
2016, 37(3): 94-98. doi: 10.13832/j.jnpe.2016.03.0094
Abstract(10) PDF(0)
Abstract:
To achieve the experimental scheme study of the AP containment, H2TS method was used to carry out the system-level scaling and the result showed the prominent distortion of area-volume ratio. Based on the basic criteria of volume-power ratio, the Bottom-Up scaling approach was used to analyze the interdependent relations for the coupled heat transfer on the containment wall. The П group on AP containment heat removal was then obtained. Thermal conduction on the containment wall was taken as an example for scaling analyses. Schemes were then given to offset the scaling distortions. Relations of structure dimensions by scaling-down were finally listed. The current scaling technique could be applicable to the scaling of large scale and large space thermal-hydraulic phenomena.
Development and Application of the Irradiation Facility in HFETR
Sun Sheng, Yang Wenhua, Tong Mingyan, Huang Gang
2016, 37(3): 99-102. doi: 10.13832/j.jnpe.2016.03.0099
Abstract(10) PDF(0)
Abstract:
With the demand for life extension and the design life improvement for the in-service and new nuclear power plants, the neutron flux becomes higher during the irradiation of structural materials used in the nuclear reactors, leading to a sharp increase in material irradiation time. Correspondingly, the research period increases and cannot meet the project progress requirements. In the paper, an irradiation facility used in the high flux area of the High Flux Engineering Test Reactor(HFETR) was successfully developed according to the characteristics of the HFETR, and the problem of the material temperature control in the Φ63mm irradiation channel was also solved. The irradiation facility can greatly shorten the material irradiation test cycle and has been successfully applied in the engineering test.
Study on Design and Test of Air Valve Sealing in Nuclear Power Plants
Shen Wei, Zhang Qiangsheng, Wang Zhiqiang, Deng Dong
2016, 37(3): 103-105. doi: 10.13832/j.jnpe.2016.03.0103
Abstract(12) PDF(0)
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Aiming at the risk of internal and external leakage for the damper in the nuclear power plants, some design measures for solving the internal and external leakage have been proposed. By the leakage tests before and after the seismic test, the seal design rationality of the damper is demonstrated. Based on the analysis of the difference between the internal and external leakage test data of two set of the dampers, the results are indicated that the leakage of the above two sets of the dampers is in compliance with the technical specification.
Geometry Design of Printed Circuit Heat Exchanger Based on Flow and Heat Transfer Correlation
Liu Shenghui, Huang Yanping, Lang Xuemei, Zhao Dawei, Wang Junfeng, Liu Guangxu, Zang Jinguang
2016, 37(3): 106-109. doi: 10.13832/j.jnpe.2016.03.0106
Abstract:
The printed circuit heat exchanger is a kind of compact heat exchanger, which geometry design method is studied based on the flow and heat transfer correlation. After the analysis of the design input, an analysis program was developed with the help of a mathematical software Matlab in this paper. Compared with the results of FLUENT15.0, the reliability of this program was confirmed.
Analysis of Adaptability of AP1000 Passive Safety Systems Commissioning to Nuclear Safety Laws and Guidelines
Qiu Fengxiang, Mazhongjie, Liu Jiahe, Sun Jingyi, Liu Chi
2016, 37(3): 110-115. doi: 10.13832/j.jnpe.2016.03.0110
Abstract(10) PDF(0)
Abstract:
Firstly the Nuclear safety laws and guidelines was introduced briefly, and then the requirement of safety systems in nuclear safety laws and guidelines was decrebied, and the adaptability of passive core cooling system and passive containment cooling system commissioning to laws and guidelines was analysied. The results show that the commissioning of AP1000 passive safety systems are adptive to the laws and guidelines of nuclear safety, but the sequence of the related tests should be optimized. When the laws and guidelines are updated in the future, these commissioning of passive safety systems and these tests cannot be performed on site due to test conditions should be taken into account.
Study on Leak Mechanism and Leakage Rate Prediction Model of Reactor Containment Sealing Structure
Huang Xiaoming, Li Jun, Xu Guoliang, Cheng Zhuo, Lyu Xiangkui
2016, 37(3): 116-121. doi: 10.13832/j.jnpe.2016.03.0116
Abstract(11) PDF(0)
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To find a better way to control the leakage characteristics of the reactor containment static sealing structure, the present work proposed a new leakage rate prediction model. In this model, the micro-mechanism of the static sealing is described based on the porous media seepage theory, and Hertz contact theory was applied to solve the change of the interface microscopic structure with the interface stress. Finally, the leakage rate can be calculated independently with any experimental data. The new model was applied to predict the leakage rate of a DN20 shut-off valve. Good agreement between the predicted outcomes and test data verified the validity of model.
Ultrasonic Inspection for Zirconium Alloy Nuclear Fuel Cladding Tubes
Xia Jianwen, Han Cheng
2016, 37(3): 122-126. doi: 10.13832/j.jnpe.2016.03.0122
Abstract(17) PDF(0)
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The cladding tube is the main component of the nuclear fuel assembly, and as the first protective barrier, its quality is very important for the safe operation of nuclear power plants. After the completion of cladding tubes, a non-destructive testing is required, in which the ultrasonic inspection is a primary method. This paper introduces the ultrasonic flaw testing method and techniques of the zirconium alloy nuclear fuel cladding tubes for pressurized water reactor(PWR), which used in automatic ultrasonic inspection equipment, and discusses the detector response to the longitudinal and transverse artificial defects of different length, width, depth and angle. Its actual shape and size are measured by metallographic anatomical analysis for some typical defects to confirm the flaw detection results. Consider its quality and cost control, the acceptance rules are proposed for different defects. The application shows that the existing detection method and process can inspect the fine defects about 10μm at different locations of the cladding tube. Due to the influence of the defect type and orientation, the echo amplitude obtained by the detector is not completely true to the actual size and nature of the defects. In order to ensure the quality of the cladding tube, it is necessary to take appropriate and more strict measures to control the different defects in the actual inspection of the tube manufacturing process.
Analysis for Downstream Effect(ex-core) of Containment Sump Strainer
Zhang Wei
2016, 37(3): 127-130. doi: 10.13832/j.jnpe.2016.03.0127
Abstract:
During the generation of the thin bed on the containment sump Strainer, the particular debris and the fiber debris with the small size will pass though the strainer and go into the Safety Injection System(RIS)/ Containment Spray System(EAS) flow path. The debris effect on the downstream equipment should be evaluated according to the requirement of the nuclear supervisory authorities. The general method which recommended by evaluation document WCAP-16406 is used in QINSHAN Phase II Extension Project, the calculation and analysis for the blockage and wear evaluations on the downstream effect(ex-core) in PWR are performed.
Handling of Flooding of Circulating Pump Pit in Ling’ao Phase Ⅱ Nuclear Power Station
Chen Jianrui, Qin Fei, Zhang Xihui
2016, 37(3): 131-133. doi: 10.13832/j.jnpe.2016.03.0131
Abstract(15) PDF(0)
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During the power test(COC53) for the 125 V DC power supply system(LBA) of Unit 3 of Ling’ao PhaseⅡNuclear Power Station, due to the loss of electricity, the circulating water pump(3CRF001PO) sewage pump pit cannot start the automatic drainage. The power failed for more than 6 h, and the water entered in the lower radial bearing oil chamber of the water circulation pump. The reasons for this problem is analyzed, and the power supply for the sewage pump is rebuilt, to avoid the risk of equipment flooding caused by the single loss of power.
Valve Erosion Damage Analysis under Approach of Numeral Simulation
He Ziang, Zhang Wei, Yin Kaiju, Tang Wu, Chen Yong, Hong Xiaofeng
2016, 37(3): 134-137. doi: 10.13832/j.jnpe.2016.03.0134
Abstract(12) PDF(0)
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Under the approach of numeral simulation, this paper firstly aims to carry out the simulating calculation work of value under different operating conditions and value lift, and then analyzes the erosion damage of the internal components of the value, which include value core, cage and seat. From this research, we can find that the extent and the location of the erosion are directly affected by the opening degree of the value. When the distance between the value core and seat is deliberately 20 mm, the upstream face of the value core erodes most seriously. When opening degree is small(2-20), the side face and conical surface are scoured the most, and the erosion degree is inversely proportional to its opening degree. When the opening degree is larger(40mm), the cage orifice is eroded the most. Under the same flowing condition, the opening degree of the value is inversely proportional to the degree of erosion, and the conical surface of the value core and the cage outlet is vulnerable to erosion.
A Method for Predicting Configuration of Corium Pool in Lower Plenum of Reactor Vessel
Liu Lili, Yu Hongxing, Chen Liang, Deng Jian, Zhang Hang
2016, 37(3): 138-141. doi: 10.13832/j.jnpe.2016.03.0138
Abstract(10) PDF(0)
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A method is designed to predict the configuration of the corium pool in the lower plenum of reactor vessel. It takes into account both thermo-chemical reaction of corium and the influence of different paths of corium relocated from core region into the lower plenum. The prediction by this method is consistent with MASCA experimental results and is applied to investigate the corium configuration of a reactor which is subjected to a hypothetical station blackout(SBO) accident. It is shown that for the sideward relocation, the configuration shows a three-lay corium pool. The composition in each layer is calculated based on the thermo-chemical data of the materials. The method is designed to predict the integrity of the reactor vessel under IVR condition.
Top Design Study on Fast Assessment System for Nuclear Accident Emergency Core Damage
Liu Yuanyuan, Zhang Shaojun, Jin Hongbo, Fu Jie, Lin Quanyi, Yue Huiguo
2016, 37(3): 142-145. doi: 10.13832/j.jnpe.2016.03.0142
Abstract(13) PDF(0)
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In this paper, based on the investigation results of domestic and foreign countries for core damage assessment methods, we proposed a suitable core damage assessment method for current nuclear power plants of operation and under-construction in China, namely CDAG and IAEA TECDOC-955 approach, and a detailed top design of the system. It is believed that this design will play a significant role for the final top design of the fast assessment system for nuclear accident emergency core damage.
Optimizing Analysis of Thermodynamic Performance Optimizing Analysis of Stirling Conversion System for Space Nuclear Power Installation
Zhang Haochun, Feng Zhiyuan, Cai Shuyi, Ji Yu, Zhang Yining, Zhao Guangbo
2016, 37(3): 146-151. doi: 10.13832/j.jnpe.2016.03.0146
Abstract(11) PDF(0)
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Advanced Stirling convertor designed by Sunpower was investigated as the object of the study, and a theoretical model of a Stirling cycle was established. Based on the operation condition of the nuclear power plant in the space, the theory of finite time thermodynamics was adopted to investigate the system performance, considering the effect of variable heat absorption temperature on the cycle. Moreover, applicability of the system optimization based on the entropy theory and the entransy theory was discussed. In addition, results between the ecological optimization criterion and the output power optimization criteria were compared. It was found that cycle efficiency and output power could be effectively improved through increasing the cycle endothermic temperature. In addition, the output power could be optimized by using entropy production rate and entransy loss rate. The output power optimization criterion can be applicable at lower exothermic temperature condition, while the ecological optimization criterion is more reasonable at higher exothermic temperature condition.
Numerical Simulation of Premixed H2/O2 Combustion
Liu Yinhe, Zhang Yun, Zhao Zhenxing, Lin Mingsen
2016, 37(3): 152-157. doi: 10.13832/j.jnpe.2016.03.0152
Abstract(12) PDF(0)
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A mathematical model was established to simulate the combustion of the premixed hydrogen and oxygen in this paper,and the influences of the ignition position,the reaction mechanism and elementary reaction kinetic parameters on the combustion process were analyzed. The results show that buoyancy has significant influence on the flame propagation. The top ignition method is more beneficial for controllable and safe eliminating of hydrogen. The amounts of consumed hydrogen are nearly equal when the flame propagates the same distance for the two ignition positions. Additionally, the predicted hydrogen combustion processes according to Marinov and Warnatz mechanisms show good consistency with each other. The HO2 formation/consumption reactions are the limiting reactions in the Marinov and Warnatz mechanisms, and their reaction rates greatly affect the total reaction rate.
Experimental Research on FCI Process of High Superheated Molten Metals
Lu Qi, Chen Deqi, Song Jiaban, Pan Ruian, Tang Tao, Pan Liangming
2016, 37(3): 158-162. doi: 10.13832/j.jnpe.2016.03.0158
Abstract(10) PDF(0)
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An experimental investigation is carried out to study the FCI process as the high superheated molten metals contact with the subcooled water during the severe accident of nuclear reactors. In this study, the characteristics of FCI processes with different molten metals and initial temperatures are analyzed, which are beneficial for further studying of the complex boiling processes and heat transfer mechanisms on the high superheated surfaces of molten metals. The metal samples include aluminum, lead and bismuth, and the initial temperatures include 700℃, 750℃, 800℃, 850℃, 900℃ and 1000℃. According to this study, the effect of molten metal property on FCI process and the effect of initial temperature on the amount of vapor generated on the surface of molten metal are discussed. Moreover, the factors affecting the fast heat transfer between the high temperature molten metals and the subcooled water are discussed.
A Simplified Model to Simulate AP1000 Containment Pressure Response during Design Basis Accidents
Wang Guodong, Tang Weijian, Wang Zhe, Zhang Jingyu, Zhang Di, Ni Chenxiao, Wei Shengjie, Wang Zhangli, Hu Benxue
2016, 37(3): 163-168. doi: 10.13832/j.jnpe.2016.03.0163
Abstract:
A simplified model is developed to simulate the AP1000 containment pressurization under design basis accidents(DBAs). The containment response has been performed by WGOTHIC code for comparison purpose. It shows a good agreement between the model and WGOTHIC code prediction results on containment heat and mass transfer process.
Investigation on Numerical Simulation for Hot Gas Mixing Structure of HTR-PM by Considering Leakage Flow
Zhou Yangping, Hao Pengfei, Li Fu, Shi Lei, He Feng
2016, 37(3): 169-172. doi: 10.13832/j.jnpe.2016.03.0169
Abstract(11) PDF(0)
Abstract:
The numerical simulation for the design of the hot gas mixing structure of Pebble-bed Module High Temperature gas-cooled Reactor(HTR-PM) are carried out with consideration of the leakage flow out of the reactor core. According to the profiles of temperature, pressure drop and flow velocity, the thermal mixing is mainly produced by the secondary flow such as the whirlpools which is perpendicular or parallel to the main flow direction. In addition, the pressure drop is mainly caused by the pressure loss leading by the sudden change of flow area and direction which is also the main reason of production of secondary flow. The numerical simulation indicates that the design of the hot gas mixing structure at HTR-PM reactor outlet can fulfill the requirement of high thermal mixing performance and low pressure drop by taking into account the leakage flow out of reactor with the condition of rated flow.
Uncertainty Analysis of Advanced Pressurized Water Reactor Fuel Assembly
Hao Chen, Zhao Qiang, Li Fu, Yu Yan, Zhang Chunyan, Zhang Huiyan
2016, 37(3): 173-180. doi: 10.13832/j.jnpe.2016.03.0173
Abstract(16) PDF(0)
Abstract:
General perturbation theory(GPT) is applied to study the contribution of the nuclear data to the uncertainty of the macroscopic cross section(XS) of fuel assembly of AP1000. Through the comparison and analysis of the contribution of the different uncertainty sources, the correlation matrix of macroscopic XS has been gained. At the same time, the sensitivity analysis and stepwise comparison methods had been used to study the mechanism of the contribution of nuclear data to the uncertainty of macroscopic parameters of fuel assembly of AP1000 under different operation conditions. The final numerical results indicate that the uncertainty of macroscopic XS propagated from nuclear data is, to some extent, constant. To the point the major contributors to macroscopic XS uncertainties are 235U nubar, 238U capture, 238U inelastic scattering and H-1 inelastic scattering. Also the numerical results include that the uncertainty of kinf and the macroscopic cross section increases with the temperature rises. The uncertainty of kinf increases with the decreasing of the enrichment of fissile material and the existence of burnable poison. At the same time, the uncertainties of fast group cross section are much larger than the thermal cross section uncertainties. Specially, the 238U(n,γ) and 238U(n,n’) should be received significant attention and made further evaluation and improvement.
Technical-Economic Study on High Temperature Reactor for Combined Heat and Power Generation
Wang Yongfu, Sun Yuliang
2016, 37(3): 181-184. doi: 10.13832/j.jnpe.2016.03.0181
Abstract(13) PDF(0)
Abstract:
The market prospect of the combined heat and power generation in China was analyzed, and the feasibility of high temperature reactor for combined heat and power generation(HTR-CHP) was studied in technical and economic aspects. The results show that HTR-CHP can balance the operation safety and heating efficiency, and guarantee that the radiation effect on the surrounding people and the environment is sufficiently low. Through commercialization development, the economy of HTR can be improved remarkably, and has economic advantage over gas-fired plants, which reveals that the high temperature reactor has the potential to replace the gas-fired plants to supply combined heat and power.
A Shielding Assembly for Residual Stress Measurement of Radioactive Monitoring Sample by Neutron Diffraction
Gao Jianbo, Li Meijuan, Wei Guohai, Liu Xiaolong, Li Yuqing, Liu Yuntao, Chen Dongfeng
2016, 37(3): 185-188. doi: 10.13832/j.jnpe.2016.03.0185
Abstract:
In nuclear industry some components will be radioactive from neutron induced activity. Radiation safety should be taken into consideration when the sample is measured. Considering the neutron diffraction technique and radiation safety, a method for measuring the radioactive sample is proposed and a shielding assembly is designed and fabricated. Using this assembly, residual stress of the radioactive sample can be measured within a safe radiation environment.