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2015 Vol. 36, No. 1

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Transient Numerical Investigation of Direct Vessel Injection for AP1000 under Pressurized Thermal Shock
Qin Mian, Yu Tao, Yu Deyong, Li Zhifeng, Lu Li
2015, 36(1): 1-8. doi: 10.13832/j.jnpe.2015.01.0001
Abstract(28) PDF(0)
Abstract:
The transient temperature and the transient mixing characteristics of RPV downcomer in AP1000 under four kinds of operation condition(CMT Hot Functional Test, CMT Injection and ADS, CMT Recirc with ACC / IRWST Injection and Normal RHR Operation) has been investigated using the computational fluid dynamics(CFD) method combined with the conjugate heat transfer computing method. Results show that the coolant mixed strongly at the intersection of DVI-Nozzle and the inner wall surface of RPV under each operating condition. The coolant from the cold leg decide how the distribution of fluid temperature in downcomer of RPV changed. And three typical location of DVI-Nozzle experience a large step change in temperature under each operation condition.
Method of the Modeling and Abstraction of Typical Thermal Experiment Unit for Sub-Critical Fuel Components
PENG Jingfeng, HUANG Yanping, XU Jianjun, DUAN Shilin, XIAO Zejun
2015, 36(1): 9-13. doi: 10.13832/j.jnpe.2015.01.0009
Abstract(25) PDF(0)
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The sub-critical module-fuel component is the core component of the sub-critical blanket in fusion–fission hybrid energy system. The thermal safety experiment for module-fuel component is the foundation and necessary step of obtaining thermal design principle, developing structure design and safety analysis. In this paper, the necessary experiment modeling technique of thermal safety experiment is researched. The analysis for the particular thermal-hydraulics structure is developed. The related calculation, by computational fluid dynamic(CFD) method, is developed to confirm the parameter of object. The structure abstraction of typical thermal experiment unit is obtained. This work supplies the basis of following experiment body design.
Experimental and Numerical Investigations of Natural Convection in a Vertical Annulus
ZHANG Sheng, GU Hanyang, CHEN Yuqing, ZHOU Xiaojia, LIU Gang
2015, 36(1): 14-17. doi: 10.13832/j.jnpe.2015.01.0014
Abstract(26) PDF(0)
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Experimental investigation of natural convection in a long vertical annulus closed at the top and opened at the bottom is presented. Numerical simulation is performed using the restricted domain model and extended domain model. The outer wall and top wall of inner cylinder are insulated. The outer wall of outer cylinder is cooled with the following speeds: 2.9, 5.7, 8.6m/s. And the corresponding Grashof number is 9.8×106, 3.9×106, and 8.3×105. The discrepancy between the numerical and experimental results is small when Grashof number is less than 106 and becomes bigger when Grashof number exceeds 106.
Numerical Investigation of Binary Droplet Collision in Steam Separator
ZHANG Di, LUO Qi, HUANG Wei, WANG Kan
2015, 36(1): 18-22. doi: 10.13832/j.jnpe.2015.01.0018
Abstract(27) PDF(0)
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Numerical simulation was carried out by using fluid Volume function(VOF) method to study the equal-sized head-on binary droplet collision in steam separators. The typical coalescence and reflexive separation with low, middle, high We, were investigated. Droplet shapes during the collision process were observed. And transformation between kinetic energy and surface energy, energy dissipation were analyzed to achieve better knowledge on the collision process. It was found that in high We reflexive separation, the external liquid departures out of the droplet and forms a circular ring. The liquid ring is forced by the surface tension to contract and impact on the inner liquid, which leads to the surface’s fierce and complicated oscillation. This phenomenon affects the separation process and satellite droplet formation. And Kim’s satellite droplet model cannot make precise prediction. By simulating the equal-sized head-on collisions with various We, critical We between coalescence and reflexive separation was found as 15.6. And comparison with three models shows that, Qian model matches well.
Specificity Analysis of Intermittent Flow in a Vertical Heating Channel under Low-Pressure Conditions
ZHU Li, CHEN Jinbo, GONG Haiguang, TONG Lili, CAO Xuewu
2015, 36(1): 23-27. doi: 10.13832/j.jnpe.2015.01.0023
Abstract(25) PDF(0)
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Intermittent flow behavior in a vertical channel under low-pressure conditions is experimentally studied using a small scale experiment facility in the present study. The effects of different factors on the flow characteristics are investigated in detail. The results reveal that the steam eruption occurs periodically in the vertical channel. This flow pattern shows typical features, such as a time-consuming boiling delay period, and remarkable time-variation of periodicity parameters. The temperature and pressure in the flow channel fluctuates greatly accompanied by steam eruption and refill of super-cooled liquid. The flow period decreases by increasing the heat load and channel aspect ratio.
Study on Natural Circulation Flow under Reactor Cavity Flooding Condition in Advanced PWRs
TAO Jun, YANG Jiang, GUO Dingqing, CAO Jianhua, LU Xianghui
2015, 36(1): 28-32. doi: 10.13832/j.jnpe.2015.01.0028
Abstract(31) PDF(0)
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Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR 1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force.
Research of Bubble Growing Characteristics Based on Interface Mass Transport Equation of Molecular Dynamics and Microlayer Evaporation under Subcooled Conditions
XUE Longchang, PAN Liangming, YUAN Dewen, HUANG Haojie
2015, 36(1): 33-37. doi: 10.13832/j.jnpe.2015.01.0033
Abstract(28) PDF(0)
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A single bubble growing model is established under subcooled flow boiling conditions in vertical rectangular narrow channel to numerically investigate the bubble growth. The model combines interface evaporation/condensation based on molecular dynamics and microlayer evaporation. The phase change model shows the multiple influences of micolayer evaporation under the bubble, superheated liquid near the heating wall, and subcooled mainstream. Numerical simulations results coincide with the experimental results well. The results reveal that the microlayer evaporation accelerates the bubble growth at the beginning, and the mainstream condensation keeps its size after the growing period. Details of bubble shape evolution and changes of microlayer thickness during the growth of the bubble are gained. It is found that the wall effect of narrow channel has an obvious effect on the bubble growing.
Study on Effect of Fuel Oxidation on Fission Product Diffusion Release
JING Futing, YANG Hongrun, LU Huanwen, YU Hong
2015, 36(1): 38-40. doi: 10.13832/j.jnpe.2015.01.0038
Abstract(25) PDF(0)
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Effect of fuel oxidation on fission product diffusional release is analyzed, and it is concluded that the vacancy increase due to the fuel oxidation has greater effect on the enhanced fission product release than the temperature increase due to the degradation of fuel thermal conductivity. The fission product release fractions increase with the O/U ratio, and this phenomenon is more obvious as the linear heat rate increases.
Corrosion Fatigue Cracking of Inconel 690(TT) with Crack-tip Area under Small Scale Yielding
XIAO Jun, CHEN Luyao, FU Zhenghong, QIU Shaoyu, CHEN Yong, LIN Zhenxia
2015, 36(1): 41-45. doi: 10.13832/j.jnpe.2015.01.0041
Abstract(24) PDF(0)
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Corrosion fatigue crack-tip plastic zone of Inconel 690(TT) was investigated, and the effects of crack-tip plastic zone and load ratio on the corrosion fatigue cracking in the simulated pressurized water reactor environment were studied. The accelerating effects of the simulated pressurized water reactor environment on fatigue crack growth rates under small scale yielding was as large as 3.
Iodine Induced Stress Corrosion Cracking of N36 Zirconium Tube in Hoop Tensile Condition
YAN Meng, WANG Pengfei, LIANG Bo, HONG Xiaofeng
2015, 36(1): 46-49. doi: 10.13832/j.jnpe.2015.01.0046
Abstract(27) PDF(0)
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Hoop tensile test is performed on N36 zirconium tube ring specimen at 350℃ to study the iodine induced stress corrosion cracking(I-SCC) behavior of zirconium tube in iodine gas environment, and the stress-elongation curve in partial pressure of iodine gas as 102 Pa,103 Pa,and 104 Pa separately are drawn. Stress-elongation curve can evaluate the I-SCC susceptibility of zirconium tube. There is no obvious necking stage in stress-elongation curve when I-SCC occurs, and the stages corresponding to crack begins and propagates can be found in the curves. Elongation and fracture energy decrease because of I-SCC, and with higher iodine contents, it decreases even more. Groups of cracks propagate parallel to the fracture surface, and the intergranular crack is located at the edge of crack.
Effects of Cl- and Cu2+ on Stress Corrosion Cracking of Alloy 690
LIN Zhenxia, QIU Shaoyu, XIAO Jun, FU Zhenghong, CHEN Yong
2015, 36(1): 50-54. doi: 10.13832/j.jnpe.2015.01.0050
Abstract(28) PDF(0)
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The effects of Cl- and Cu2+ on the stress corrosion cracking(SCC) of Alloy 690 made in China were investigated using the slow strain rate tensile(SSRT) test. The surface and cross section morphologies of tested specimens, as well as the fractographs, were observed by 3D measuring laser microscope and SEM, respectively. It is found that the SCC resistance of Alloy 690 decreases evidently in the solutions containing 100mg/L Cl- and 1000mg/L Cu2+. Many corrosion pits appear in the surface of the specimens and there are SCC initiations on the bottom of the pits. Moreover, SEM fractographs show that it is typical intergranular fracture. The observations of texted specimens are consistent with SSRT results, indicating that, it is the combined effect of both Cl- and Cu2+ that brings about SCC of Alloy 690. The mechanism of combined effect is discussed briefly.
Reliability of Test and Measurement for Hydride Reorientation
CHEN Le, XIE Meng, PEI Qilin, DAI Xun, XU Chunrong
2015, 36(1): 55-59. doi: 10.13832/j.jnpe.2015.01.0055
Abstract(30) PDF(0)
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The effect of circumferential stress and thermal cycles on the hydride orientation factor(Fn) of N36 zirconium alloy tubes after hydriding has been studied, and several deviations during the test and the uncertainty of measurement for Fn has been mainly analyzed and computed. The results show that, the uncertainty among different layers in the same sample is larger than that among different samples in the same condition. A type uncertainty which results from the heat treatment, cutting, hydriding and reorientation test plays the key role, and the measurement software and personal error is relative smaller.
Modification Design and Implementation of Core Cooling Monitoring System in Daya Bay NPP
WANG Yuan, XIONG Guohua
2015, 36(1): 60-63. doi: 10.13832/j.jnpe.2015.01.0060
Abstract(22) PDF(0)
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In order to solve the problems of aging of core cooling monitoring system, and the unstable work and the flash failure alarm of CCMS due to unavailability of some old parts in Daya Bay Nuclear Power Station, the modification of the whole upgrading program is proposed. The design of new CCMS using localization instrument control platform(Firmsys) is described. The research and analysis for the key technology of new CCMS function and interface is done. The installation and commissioning stage of handling technical problems and the solution is described. The new system has been functionally verified. Finally the independent design and renovation of CCMS is realized.
Improvement of Design Method for CPR1000 Steam Generator Blowdown Heat Recycling
XIAO Sanping, WU Hao, GAN Quan, LI Shu, QIAN Hui, WANG Liangliang, CHEN Shushan
2015, 36(1): 64-67. doi: 10.13832/j.jnpe.2015.01.0064
Abstract(23) PDF(1)
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The paper conducts the secondary side thermal economy analysis of China improved 3rd Ring Rd Pressurized Water Reactor(CPR1000) nuclear power plant according to the principal method of thermal system calculation. The stream generator of Ling’ao nuclear power plant(phase II), as an example, is calculated by analyzing the blowdown heat recycling. It is concluded that the thermal economy of the whole unite could be increased by 0.088% if the blowdown heat was recycled by the low pressure heater No. 4.
Technology Improvement for Reactor Pressure Vessel of Ling’ao Phase II Nuclear Power Station, Shenzhen, Guangdong, 518124, China
CHEN Zhenwei, WU Chaorong, GUAN Jianwei, XING Rujun, YANG Chunle, GUAN Chunxiang
2015, 36(1): 68-71. doi: 10.13832/j.jnpe.2015.01.0068
Abstract(24) PDF(0)
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The technology improvement for the reactor pressure vessel on Ling’ao Phase II Nuclear Power Station is given. This paper presents the description of the integrated core shell, integral reactor pressure vessel top cover, the control of irradiation sensitivity element content, co resident content reduction, the improvement of radial support blocks material and the increase the quantity of irradiation surveillance program capsules. The advantage of these improvement and impacts on equipment performance, construction, operation and maintenance of Ling’ao Phase II Nuclear Power Station is analyzed.
Design and Research of Seal Structure for Thermocouple Column Assembly
RAO Qiqi, LI Na, ZHAO Wei, MA Zhigang
2015, 36(1): 72-76. doi: 10.13832/j.jnpe.2015.01.0072
Abstract(30) PDF(0)
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The new seal structure was designed to satisfy the function of thermocouple column assembly and the reactor structure. This seal structure uses the packing graphite ring and adopts the self-sealing principle. Cone angle is brought to the seal face of seal structure which is conveniently to assembly and disassembly. After the sealing principle analysis and stress calculation of graphite ring which adopt the cone angle, the cone angle increases the radial force of seal structure and improves the seal effect. The stress analysis result shows the seal structure strength satisfies the regulation requirement. The cold and hot function test result shows the sealing effect is good, and the design requirement is satisfied.
Design of Tritium Trap for 3He Power Transient Tests of LWR Fuel Element
LI Binglin, SUN Sheng, WANG Hai, TONG Mingyan, DAI Yubing
2015, 36(1): 77-80. doi: 10.13832/j.jnpe.2015.01.0077
Abstract(24) PDF(0)
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The purpose of this study is to develop the tritium trap for LWR fuel element irradiation test with power transient, to purify the 3He gas from tritium produced by neutron capture. Titanium sponge particles were selected as the storage materials. The design of prototype vessel was determined by factors such as the amount of tritium, the titanium-hydrogen reaction, the temperature by the electric heat, the maximum possible pressure of 3He, and helium release condition by the decay tritium. The titanium bed was heated internally to reduce the heat loss, and the primary and secondary containments of stainless steel were applied to prevent the permeation of tritium. Analysis results for the structure, heat transfer, shielding, and confinement of this trap are introduced.
Characteristics of Passive Technology in Nuclear Power Plants
LI Zhaojun, WANG Xin, WANG Yuan
2015, 36(1): 81-84. doi: 10.13832/j.jnpe.2015.01.0081
Abstract(20) PDF(0)
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This paper analyzed the present passive technology in nuclear power plants, and based on the special requirement on the application of the passive methods used in the nuclear power plants under the marine conditions, the theoretical research, experimental technology, design technology, equipment technology and simulation technology are studied, a set of passive technology system for the marine nuclear power plants is established and successful implemented in a marine nuclear power plant.
Fault DSm T Fusion Method of Core Barrel Based on Wavelet Energy Analysis
GUO Qing, XIA Hong, HAN Wenwei
2015, 36(1): 85-89. doi: 10.13832/j.jnpe.2015.01.0085
Abstract(24) PDF(0)
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Being aimed at the problems that it is difficult to recognize effectively the faults of core barrel rupture and fall off of core barrel fasteners parts, a method of the wavelet packet energy analysis and DSm T is proposed. By using the wavelet decomposition and reconstruction method, the fault signal of core barrel is decomposed to the basis function family formed from wavelet expanding and contracting, then sub-band energy distributed in different bands are obtained, and used as BPA evaluation signal of core barrel DSm T. By the simulation and analysis, the effectiveness of this method is verified.
Research of Preventive Measures for Heterogeneous Boron Dilution Accident in Daya Bay NPP
DONG Chaoqun, XIE Bo
2015, 36(1): 90-93. doi: 10.13832/j.jnpe.2015.01.0090
Abstract(15) PDF(0)
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The nuclear power plants are possible to suffer the core uncovering damage due to the heterogeneous boron dilution which is out of the design basis. This paper analyzes the mechanism of criticality accident caused by the heterogeneous boron dilution, and study the measures to prevent the erroneous dilution. The implementation of the improvement measures greatly improves the safety level of the Daya Bay NPP, especially effectively improves the safety level under the reactor shutdown conditions.
Analysis of LCA/LCB DC Power Supply Loss in Fuqing Nuclear Power Plant
SUN Mingchen, WU Shengguo
2015, 36(1): 94-97. doi: 10.13832/j.jnpe.2015.01.0094
Abstract(23) PDF(0)
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This paper comprehensively analyzes the impact on the nuclear power plant security when LCA / LCB loses the power. When LCA loses the power, the normal boronization of the reactor coolant is unavailable, only the direct boronization can be used; the normal discharge, excess discharge and low pressure discharge of the pressurizer are unavailable; the reactor trip breakers disconnect and P4 signal appears; the reactor trips or safety injection probably happens for the excessive cooling reactor coolant or too low pressure. When LCB loses the power, normal boronization is available, and normal discharge can be used by manual operation, and the reactor shutdown or safety injection probably also happen for the same reason.
Research on Alarm Triggered Fault-Diagnosis Expert System for U-Shaped Tube Breaking Accident of Steam Generators
QIAN Hong, LUO Jianbo, JIN Yuxiao, WANG Du, ZHOU Jinming
2015, 36(1): 98-103. doi: 10.13832/j.jnpe.2015.01.0098
Abstract(34) PDF(0)
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According to the U-shaped tube breaking accident of steam generator(SGTR), this paper designs a fault-diagnosis expert system based on the alarm triggering. By analyzing the fault mechanism of SGTR accidents, the fault symptom is obtained. The parameters of the belief rule are set up based on the simulation experiment. The information fusion is conducted on the fault-diagnosis results from multiple expert systems to obtain the final diagnose result. The test result shows that the expert system can diagnose the SGTR accident accurately and rapidly, and provide with the operation guidance.
Analysis and Solution of Spike Current of Intermediate Range for Nuclear Instrumentation System
LI Xingqiang, WANG Yinli, XIAO Yu, XUE Bin
2015, 36(1): 104-107. doi: 10.13832/j.jnpe.2015.01.0104
Abstract(27) PDF(0)
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During the initial start-up of HYH NPP unit 1, spike currents occurred in intermediate range channel of RPN system and reactor trip was triggered. After analyzing the operation principle of intermediate range channel and site inspection, and doing simulating test, the root cause was fixed on the bug of range switching of intermediate range channel. Then a solution based on parameters optimizing was made and executed on site.
Modification and Optimization of Narrow Range Temperature Probe in CPR1000 Unit
CHEN Yongwei, QIU Hewen, ZHANG Liguo, YOU Dailun, YANG Xing
2015, 36(1): 108-112. doi: 10.13832/j.jnpe.2015.01.0108
Abstract(24) PDF(0)
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With the extension of the operation period, the characteristics of narrow range temperature probe will be drifting and aging in CPR1000 units. At the same time, the change of thermal hydraulic characteristics factors will also lead to the unreasonable setting of the parameters of average temperature summary and temperature deviation summary. Therefore, it requires to modify the characteristics of narrow range temperature probe and the parameters of summary according to the cross comparison data. Aiming at the possible deviation, the modification method is put forward: the fitting modification method for three point curve of narrow range temperature probe and the fitting modification of two point curve of the average temperature summary and temperature deviation summary. And the optimization and improvement scheme is put forward, including excluding natural temperature deviation factor, weight coefficient in different operation conditions and horizontal / vertical trend comparison. These methods had been applied in the field, and were proved to be feasible and effective.
Study on Extension of Calibration Cycle for Radiation Monitoring Instrument in Tianwan Nuclear Power Plant
GUO Peibin
2015, 36(1): 113-115. doi: 10.13832/j.jnpe.2015.01.0113
Abstract(21) PDF(0)
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Based on the collection and statistical analysis of the radiation monitoring instrument calibration data, and taking the typical GIM204 gamma dose rate monitoring device as an example, the feasibility is demonstrated by the experience feedback method, to adapt to the implementation of the refueling scheme for a long period of time in Tianwan nuclear power station. Studies have shown that when the calibration cycle of GIM204 is extended from1 year to 1.5 year, the relative calibration error does not exceed the allowable values, and satisfies the requirement on refueling cycle extension.
Analysis of Process Parameters of Gaseous Radwaste Treatment by Activated Charcoal Delay Beds in Nuclear Power Plants
YU Shikun, LIU Yu, CHEN Shaowei, BAI Ying
2015, 36(1): 116-119. doi: 10.13832/j.jnpe.2015.01.0116
Abstract(28) PDF(0)
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Several factors can influence the activated charcoal adsorption coefficients of gaseous radwaste treatment by activated charcoal delay beds in nuclear power plants. This paper analyzes the effects of temperature, relative humidity, gases concentration and gases flow rates on the activated charcoal adsorption coefficients(mainly to krypton and xenon) based on the treatment process by activated charcoal delay beds. Especially, the rationality of gaseous radwaste system process parameters in AP1000 nuclear plants has been discussed. There are several factors impacting the selection of activated charcoal such as adsorption capacity, adsorption selectivity and service lifetime. In order to improve the treatment effects of gaseous radwaste process, the gas superficial velocity should be within the range of 0.1 to 0.7 centimeter per second, the temperature of chilled water should be decreased appropriately, the relative humidity of gaseous radwaste should be lowered than 25% and the operational pressure should be increased correctly.
AP1000 Service Water System Availability Analysis in Sanmen NPP
KONG Weijie
2015, 36(1): 120-123. doi: 10.13832/j.jnpe.2015.01.0120
Abstract(29) PDF(0)
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Service water system operation requirements and circulate water system maintenance needs are analyzed, and it is determined that the circulate water system maintenance will result in the availability of service water pumps. Solutions are proposed to resolve the conflicts between the circulate water system maintenance and the service water system availability, that will improve the safety and economy of the plant.
TRACE Thermal Hydraulic Model Development of Qinshan Nuclear Power Plant Phase Ⅱ PWR
FENG Jinjun, ZHOU Kefeng, HU Wei, ZHAN Jiashuo, SHI Junying, CHAI Guohan
2015, 36(1): 124-126. doi: 10.13832/j.jnpe.2015.01.0124
Abstract(27) PDF(1)
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In order to improve nuclear power plant safety review work, NNSA imported reactor thermal hydraulic best estimate code TRACE from NRC. In this paper, the NPQJVC two-loop PWR thermal hydraulic model is setup and LBLOCA is simulated using TRACE code and engineering modeling assistant code SNAP. The results of simulation are reasonable.
Simulation of Steam Condensation Inside Vertical Tube with Noncondensable Gases Using MELCOR
HUANG Zheng
2015, 36(1): 127-131. doi: 10.13832/j.jnpe.2015.01.0127
Abstract(19) PDF(0)
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Using MELCOR, the numerical simulation of steam condensation inside vertical tube in the presence of noncondensable gases is carried out. Calculated results are compared with Kuhn’s experimental results. Generally the agreement is satisfactory except that the heat transfer process and condensed film accumulation were overestimated, leading to the larger heat transfer coefficient. After modification of parameters of MELCOR model, the thickness of film and the overall heat transfer coefficient were decreased and agreed well with experiment data. Thus the MELCOR model is valid.
Effect of Parameter Variation of Reactor Coolant Pump on Loss of Coolant Accident Consequence
DANG Gaojian, HUANG Daishun, GAO Yingxian, HE Xiaoqiang
2015, 36(1): 132-136. doi: 10.13832/j.jnpe.2015.01.0132
Abstract(18) PDF(0)
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In this paper, the analyses were carried out on Ling’ao nuclear power station phaseⅡ to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D(employed on Ling’ao nuclear power station phaseⅡ) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-down phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump’s homologous characteristic curves change. On the other hand, the pump’s free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different.
Analysis of Mechanism of Vane-Type Bubble Separator Developed for Thorium Molten Salt Reactor
WANG Jianjun, SUN Licheng, CAI Baowei, ZHANG Nana, YAN Changqi
2015, 36(1): 137-140. doi: 10.13832/j.jnpe.2015.01.0137
Abstract(33) PDF(1)
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The mechanism of the vane-type bubble separator developed for Thorium Molten Salt Reactor was investigated experimentally and numerically. The experimental phenomena and flow field obtained by the CFD software of Fluent were compared to analyze the process of the bubble coalescence and merging into a steady air core. It is demonstrated that, a rotational water flow with high radial pressure gradient forms after flowing through the swirl vanes of the separator, and that the bubbles in the flow moving to the center of the separator is due to that the centripetal force provided by the radial pressure gradient is greater than the centrifugal force generated by rotation. The bubbles will coalesce and merge to form a steady air core due to the large radial pressure gradient and low relative velocity in the central area of the separator, leading to the continuous separation of bubbles from the water flow.
Three-Dimensional Analysis of Flow of Hot and Cold Pools of Sodium Cooled Fast Reactor Based on Added Source Method
ZHU Huanjun, XU Yijun, QIAN Xiaoming
2015, 36(1): 141-143. doi: 10.13832/j.jnpe.2015.01.0141
Abstract(24) PDF(0)
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An integrated CFD model was built in which the main components of the first loop are included.The intermediate heat exchanger,decayed heat exchangers,reactor core and main pump are simulated by the method of additional source term.The integrated model is closer to the real situation as there are not artificial assumed boundary between various components.The integrated model are justified after the results were compared with CEFR design values.It shows that sodium pool were divided into low temperature cold sodium pool and high temperature hot sodium pool.The temperature difference in hot sodium pool is large,which indicate that the hot and cold fluid mixing is obvious in the hot pool.Meanwhile,the average temperature changes little at different height of cold sodium pool and hot sodium pool,which indicate that the heat shield separating the hot and cold sodium pools effects well.
Analysis of Assessment Requirement on COSINE System Analysis Code
FU Xiaoliang, LIU Lifang, YU Nan, DU Zheng, LIANG Guoxing, YANG Yanhua
2015, 36(1): 144-147. doi: 10.13832/j.jnpe.2015.01.0144
Abstract(24) PDF(0)
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COSINE is the first software package for the nuclear power plant design and analysis developed in China. The system analysis code of COSINE has two versions, one is the conservative model version and another is the best-estimate model version. According to the latest evaluation model development and assessment methodology-EMDAP method, the important phenomenon and process to be evaluated by the conservative model and the best estimation model for the large break LOCA of COSINE system code are identified and ranked, and a PIRT table of large break LOCA has been made. At the same time, according to the assessment requirement, a nuclear power software evaluation database is constructed.
Numerical Study on Effects of Shroud in Core Lower Plenum on Flow Field
ZHENG Jiantao, LI Huaqi
2015, 36(1): 148-151. doi: 10.13832/j.jnpe.2015.01.0148
Abstract(32) PDF(0)
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A numerical simulation is carried out to investigate the influence of shroud in lower plenum on flow field. The geometrical model and boundary conditions in the simulation are the same as the integral hydraulic experiment which based on domestic CNP1000 nuclear power plant. CFD method is adopted, the detailed velocity and pressure distributions in core under different shroud diameters have been obtained by numerical analysis and validation between experimental and computational results. Finally the influence of shroud on mass flow and pressure distributions in core inlet and the mixing characteristics in lower plenum have been confirmed, which could be used for the optimization of core construction.
Evaluation of RELAP5 Code by Experiments of Integral Test Facility
LU: Yufeng, DU Kaiwen
2015, 36(1): 152-156. doi: 10.13832/j.jnpe.2015.01.0152
Abstract(19) PDF(0)
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With reference to the parameters of integral natural circulation test facility OSU-MASLWR, the system analysis code RELAP5/MOD3.3 was applied to investigate the thermal-hydraulic characteristics. Transient safety studies were done for loss-of-coolant accidents within the containment. The results show that the reactor core can be provided with a stable cooling source adequate to remove decay heat without significant cladding heatup under all scenarios. Further, the heat rejected through the containment wall to the surrounding pool of water will be greater than the amount of the decay heat produced by the reactor core. The thermal stratification phenomenon occurs in the containment. Comparison of calculated results and experiment data shows that the phenomena of interest are predicted by the code.
Investigation of Effect of Supercritical Water Properties in Pseudocritical Region on Heat Transfer Characteristics
ZANG Jinguang, YAN Xiao, HUANG Shanfang, CENG Xiaokang, HUANG Yanping
2015, 36(1): 157-160. doi: 10.13832/j.jnpe.2015.01.0157
Abstract(28) PDF(0)
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The water property will have a big variation in the pseudo-critical region under supercritical conditions and largely impact the heat transfer and fluid flow characteristics of the supercritical water. The water property data will directly affect the credibility of the numerical prediction results. In this study, the FLUENT software was used as the simulation platform to analyze the sensitive factor of water properties to heat transfer behaviors based on two supercritical water heat transfer experiments and a new method of improving the capacity of water properties was established with user defined functions(UDF). The new water property dealing method improved the prediction capability in comparison with the experimental results.
Experimental Investigation on Motion Characteristics of Relative Large Bubble
LI Zhongchun, SONG Xiaoming, JIANG Shengyao, YU Jiyang, Mamoru Ishii
2015, 36(1): 161-164. doi: 10.13832/j.jnpe.2015.01.0161
Abstract(25) PDF(0)
Abstract:
The bubble hydrodynamics characteristics such as the bubble rising velocity, bubble shape and drag coefficient was studied. The bubble was generated using a hemisphere cup to collect specific volume and released to the still water manually. The bubble motion was captured by high speed camera. The parameter was obtained by digital image process routines. The rising behaviors of bubble with different size in still water were investigated. The comparison with existed drag coefficient correlation indicated that Tomiyama was better than Ishii and Chawa. There were similarities for the shape for relative large bubble. The experimental result was beneficial to the basic understanding of the flow dynamics and interfacial exchange where relative large bubble existed.
Study on Passive Cavity Injection and Cooling Strategy for Small Module Reactors
DENG Jian, ZHU Dahuan, WANG Xiaoji, XIANG Qingan
2015, 36(1): 165-167. doi: 10.13832/j.jnpe.2015.01.0165
Abstract(27) PDF(0)
Abstract:
Typical severe accident scenarios are calculated to obtain the characteristic parameters of the molten pool in the lower vessel plenum, based on the MELCOR model for small module reactors. The indigenous CISER code is then used to analyze and validate the effectiveness of the cavity injection and cooling strategy for the small module reactors, considering conservatively the issues of phenomenological uncertainties.
Simulation of Subcooled Boiling and Critical Heat Flux in Uniformly Heated Tubes
LI Quan, JIAO Yongjun, YU Junchong
2015, 36(1): 168-172. doi: 10.13832/j.jnpe.2015.01.0168
Abstract(27) PDF(0)
Abstract:
Based on Eulerian two-fluid framework and non-equilibrium subcooled boiling model, the numerical model is established to simulate the subcooled boiling. This model is used to simulate the subcooled flow and CHF of DNB type in uniformly heated tubes. The numerical model is proved correct by comparing to the subcooled flow experiment of Bartolomei. The boiling curves is calculated, and it is suggested to treat the first point in the critical area of boiling curves as the criterion of DNB. 12 CHF points drawn from the latest 2006 CHF look-up table are simulated. These 12 points are chosen with different mass flux, pressure or local equilibrium quality, but all are under high pressures and high mass flux. The CHF predicted confirm well to the experiment and the location that CHF occur is also well captured.
Study on Neutronics Performance of Flower Shape Advanced Supercritical Water Cooled Fast Reactor with Different Solid Moderators
YU Tao, LI Zhifeng, PENG Honghua, XIE Jinsen
2015, 36(1): 173-176. doi: 10.13832/j.jnpe.2015.01.0173
Abstract(21) PDF(0)
Abstract:
The supercritical water cooled fast reactors worked at such harsh condition with high temperature and high pressure, huge hydrogen balance pressure and thermal shock can result in a great loss of hydrogen. The released hydrogen would be out of control under accident situations. Keff, conversation ratio, moderator temperature effect, Doppler effect and void effect of different material such as Zr H1.7, Be, Be O, C and Si C are discussed. Be O and Si C hold better integrated performance among these materials. Besides, moderators have less effect on the Doppler effect of fuel.
Methodology Study on Safety-Critical Software Reliability Evaluation of Digital I&C Systems at Nuclear Power Plants
CHI Miao, YANG Ming, SHI Liping
2015, 36(1): 177-181. doi: 10.13832/j.jnpe.2015.01.0177
Abstract(22) PDF(0)
Abstract:
Based on the comparative analysis of the current DCS software evaluation methods, the authors adopted the BTP7-14 of NUREG-0800 as the reference standard for evaluating AP1000 nuclear power plants. An overall framework based on Bayesian Belief Network(BBN) for reactor protection system software reliability evaluation was therefore presented. The dependency between evaluation model criteria was decided. A quantitative evaluation was performed based on the software evaluation model. Furthermore, a sensitivity analysis method is proposed for identifying the key criteria and the refinement degree of each criterion which will be helpful to make clear the direction of the software quality improvement.
Discussion of Management Review in Nuclear Power Plants
DUAN Hongwei, LI Juan, WANG Jing, ZHANG Hui, WANG Yanqi, TIAN Feng
2015, 36(1): 182-184. doi: 10.13832/j.jnpe.2015.01.0182
Abstract(24) PDF(0)
Abstract:
According to the nuclear safety regulations and guidelines in China, the information of the IAEA and the NRC, this paper discusses the origin, purpose, method and report of the management review, the basic concepts and requirements. In order to identify the difference between the management review and the internal quality assurance audit, the paper also compares their purpose, nature, the pursuant documents, and etc.