核动力工程

Experimental Investigation of Phase Radial Distribution in Medium Pressure Boiling Flow

SUN Qi, ZHAO Hua, YANG Rui-chang, HE Jun-shan

2005, 26(1): 1-5.

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This paper investigated the radial distributions of void fraction and bubble frequency and identified the flow patterns based on the lateral measurement in medium pressure boiling flow using high-temperature high-pressure single-sensor optical probe. The relations between void fraction radial distributions and flow patterns in medium pressure boiling flow were analyzed based on the experimental results. Experimental results showed that the lateral profiles of void fraction changed from near U shape towards saddle and arc as the equilibrium quality increased; local bubble frequency mainly distributed in U shape; the lateral profiles of void fraction took near U or saddle shape （local void fraction close to the wall was significantly higher than that in the core region） in bubbly flow; while in slug flow, local void fraction in the core region was only slightly lower than that in the region near the wall, and the lateral profiles took near arc shape.

Experiments on Heat Transfer Enhancement in 3D Inner Finned Helical Pipe

LI Long-jian, CUI Wen-zhi, XIN Ming-dao

2005, 26(1): 6-10.

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Experiments were performed to investigate the enhancement performance of single-phase flow and boiling heat transfer in the 3D inner finned helical pipes. The experimental results of two finned helical pipes with different inner fin geometry showed that the heat transfer and flow resistance in the 3D inner finned helical pipe gained greater augmentation than those in the smooth one. Within the measured range of Reynolds number, the average augmentation ratio of heat transfer for the two finned pipes are 71% and 103%, respectively, compared with the smooth helical pipe, and flow resistance are 90% and 140%. The tests for boiling heat transfer were carried out in the 3D inner finned helical pipes with the curvature of 0.0605 and the length of 0.668m.Comparied with that in the smooth helical pipe, the boiling heat transfer coefficient in the 3D finned helical pipe had been increased by 40% to 120% under various mass flow rates and wall heat flux, and the flow resistance coefficient increased by 18% to 119%.

Experimental Research on Density Wave Oscillation of Steam-Water Two-Phase Flow in Parallel Inclined Internally Ribbed Pipes

GAO Feng, CHEN Ting-kuan, LUO Yu-shan, YIN Fei, LIU Wei-min

2005, 26(1): 11-14.

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At p=3-10MPa, G=300-600kg/（m²·s）, Δt_sub=30-90℃, and q=0-190kW/m², the experiments on steam-water two-phase flow instabilities have been performed. The test sections are parallel inclined internally ribbed pipes with an outer diameter ofφ38.1mm, a wall thinkness of 7.5mm, a obliquity of 19.5o and a length more than 15m length. Based on the experimental results, the effects of pressure, mass velocity, inlet subcooling and asymmetrical heat flux on steam-water two-phase flow density wave oscillation were analyzed. The experimental results showed that the flow system were more stable as pressure increased. As an increase in mass velocity, critical heat flux increased but critical steam quality decreased. Inlet subcooling had a monotone effect on density wave oscillation, when inlet subcooling decreased, critical heat flux decreased. Under a certain working condition, critical heat flux on asymmetrically heating parallel pipes is higher than that on symmetrically heating parallel pipes, that means the system withl symmetrically heating parallel pips was more stable.

An Investigation on Heat Transfer Characteristics of Sub-Critical and Supercritical Pressure Steam-Liquid in an Inclined Internally Ribbed Tube

YIN Fei, LI Hui-xiong, CHEN Ting-kuan, LUO Yu-shan, WANG Wei-shu, ZHU Xiao-jing

2005, 26(1): 15-18.

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Within the range of pressures from 9 to 28MPa, mass velocity from 600 to 1200kg/（m²·s）, and internal heat flux from 200 to 600kW/ m², experiments have been performed to investigate the heat transfer characteristics of sub-critical and supercritical steam-liquid in an inclined upward internally ribbed tube with the inclined angle of 19.5 degree from the horizon, the outer diameter of 38.1mm, and the thickness of 7.5mm. Based on results, it was found that heat transfer enhancement of the internally ribbed tube can restrain the departure from nucleate boiling at sub-critical pressure. Near the critical pressure, heat transfer enhancement of the internally ribbed tube decreased. At supercritical pressure, the temperature differences between the wall and the fluid were small. Enough mass velocity needs to be supplied in the practical design. Correlation of critical mass velocity has been provided.

Experimental Study of Pool Boiling Heat Transfer of Water-Based Magnetic Fluid on a Vertical Heater

LIU Jun-hong, GU Jian-ming, LIAN Zhi-wei, LIU Hui

2005, 26(1): 19-22.

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The effect of magnetite concentration and applied magnetic field on pool boiling heat transfer of a water-based magnetic fluid was investigated on a vertical heater. The experiments showed that there was an optimum magnetite concentration in which the enhancement of heat transfer was best for the dilute magnetite particles; and this conclusion was the same with the applied magnetic field. The boiling heat transfer of magnetic fluid was enhanced by the applied magnetic field.

Solution of Multi-Dimensional Inverse Heat Conduction Problem Using Genetic Algorithm

WANG Xiu-chun, ZHI Hui-qiang, MAO Yi-zhi, YANG Zeng-jun, HAN Peng

2005, 26(1): 23-27.

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Genetic algorithm was applied to the research of multi-dimensional inverse heat conduction problem, and a valid mathematic model was presented. By the numerical simulation, the practicability and the availability of the method were verified, and the influence of the number of observation points on the solution precision was discussed, and the selection method of the number of observation points was presented. The results showed that the method was practicable with high precision. The only precondition using the method is that the corresponding direct problem shall be solved.

Application of Monte Carlo Method for Verification Calculation in Fuel Loading Prediction for First Criticality of HTR-10

JING Ying-qing, YANG Yong-wei, XU Yun-lin

2005, 26(1): 28-34.

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Before the 10MW pebble-bed High Temperature Gas-cooled Reactor-Test Module （HTR-10） reached the first criticality, the prediction calculations of the fuel loading of the core was a main physics design concern. In order to ensure the reliability of the physics design, verification calculations with Monte Carlo code MCNP has been carried out. Firstly, the sophisticated structure of the fuel element and reactor were descripted approximately and rationally, based on the HTR-10 characteristics. Then, the parameters of the eigenvalue calculation were chosen rationally, including number of cycles to be skipped before beginning, nominal source size per cycle, and number of cycles for calculation of eigenvalue; and the effect of different descriptions for HTR-10 fuel element structure was studied. Finally, the better scheme was determined.
The results of the scheme showed that under the atmosphere of air with a temperature of 27℃, the results of prediction calculation for the number of spheres including the fuel elements and the graphite elements, using MCNP and VSOP, were 16864 and 16821, respectively,with a relative error of 0.25%. According to the results of the first criticality of HTR-10, the experimental and predicted results were agreed well with low relative error less than 1 %.

The Simulation Research for The Dynamic Performance of Integrated PWR

YUAN Jian-dong, XIA Guo-qing, FU Ming-yu

2005, 26(1): 35-39,50.

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The mathematical model of the reactor core of integrated PWR has been studied and simplified properly in this paper. With the lumped parameter method, we have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, we have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct.

High Efficiency Algorithm for 3D Transient Thermo- Elasto-Plastic Contact Problem in Reactor Pressure Vessel Sealing System

XU Ming-yu, LIN Teng-jiao, LI Run-fang, DU Xue-song, LI Shui-an, YANG Yu

2005, 26(1): 40-43,69.

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There are some complex operating cases such as high temperature and high pressure during the operating process of nuclear reactor pressure vessel. It is necessary to carry out mechanical analysis and experimental investigation for its sealing ability. On the basis of the self-developed program for 3-D transient sealing analysis for nuclear reactor pressure vessel, some specific measures are presented in this paper, to enhance the calculation efficiency in several aspects such as the non-linear solution of elasto-plastic problem, the mixed solution algorithm for contact problem as well as contact heat transfer problem and linear equation set solver. The 3-D transient sealing analysis program is amended and complemented, with which the sealing analysis result of the pressure vessel model can be obtained. The calculation results have good regularity and the calculation efficiency is twice more than before.

Sensitivity Analysis for Floor Response Spectra of Nuclear Reactor Buildings

LI Zhong-xian, LI Zhong-cheng, SHEN Wang-xia

2005, 26(1): 44-50.

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In this paper, against the reactor building of a 1000MW Pressurized Water Reactor （PWR） nuclear station, the influence of the change of dynamic shear modulus of foundation soil on the floor response spectra of nuclear reactor building on a site of hard soil is investigated by considering soil-structure interaction （SSI）, and the sensitivity of the floor acceleration response spectra of the reactor building to the change of dynamic parameters of foundation soil is quantitatively analyzed, which may provide a reliable method and an important reference to evaluate the anti seismic safety of nuclear reactor buildings on similar sites of hard soil.

Helium Sampling and Analyzing System of 10MW High Temperature Gas-Cooled Reactor

ZHU Jiang, WANG Yu-cheng, HUANG Zhi-yong, WU Xiao-dong, YAO Wei, NIU Dao-qing

2005, 26(1): 51-53,83.

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The helium purification system of 10MW high temperature gas-cooled reactor consists of the purification equipments and their accessories. The purification equipments include a copper oxide bed, a molecular sieve absorber, a low temperature absorber, and etc. The gas sampling and analyzing system is consisted of the gas chromatograph, moisture probe, and infrared analyzing instrument. The moisture probe and infrared analyzing instrument both conform to the design requirement, and consecutive inspection of H₂O、CO、CO₂ can be carried out for the primary helium circuit. The gas chromatograph can also meet the design requirement, and so the intermittent sampling and analyzing of H₂O₂, N₂, CH₄, CO, and CO₂ can be carried out for the primary helium circuit.

Feasibility Study on Pre-stressed Concrete Pressure Vessel in Modular High Temperature Gas-Cooled Reactor

YIN Qing-liao, HE Shu-yan, WU Xin-xin

2005, 26(1): 54-58.

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Considering the disadvantages of the steel pressure vessels in modular high temperature gas-cooled reactor （MHTR）, such as high cost and difficulties in manufacture and transportation, the technical feasibility study is performed for the pre-stressed concrete pressure vessel （PCPV） in MHTR. The paper briefly presents the development status and technical characteristics of PCPV, analyzes and studies the technical crux and feasibility of PCPV as the primary system pressure boundary components of MHTR. The preliminary design of PCPV is proposed for MHTR. The analysis proves that PCPV utilized in MHTR is feasible, which could not only solve the technical crux, such as mechanics problems and decay heat removal in the multi-cavity of PCPV, but also improve the s economic competition and safety of MHTR.

Study Mechanism of Zigzag-Corrugated Baffle Separator

TIAN Rui-feng, ZHANG Zhi-jian, CHEN Jun-liang, LI Feng

2005, 26(1): 59-62,87.

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The flow field in the zigzag-corrugated baffle separator was calculated using vorticity-velocity algorithm. The tracks of water droplets in the flow field were calculated using SPD model. And the separating efficiency and pressure reduction were calculated too. By comparing calculation and experiment results, the mechanism of zigzag-corrugated baffle separator was gotten. The research result is that,it is suitable to the leading angle equal to 45degree when design zigzag-corrugated baffle; And if the steam with great moisture flow into Separator, the groove zigzag-corrugated baffle should be adopt firstly.

Improvement of Static Load Test Technology for Polar Crane in Ling’ao Nuclear Power Station

CHEN Zhen-wei, SANG Shu-hui

2005, 26(1): 63-64,87.

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The test instrument and method for polar crane are improved to facilitate the load test （especially double-hook combine static load test ）. The paper introduces the test requirement and conventional test method for polar crane static load test for 1000MWe PWR nuclear power station and the static load test technology improvement for polar crane in Ling’ao Nuclear Power Station .

Effect of Hot-Rolling Temperature on the Texture of N18 Zirconium Alloy Plate

PENG Qian, LIU Yan-zhang, ZHAO Wen-jin, SUN Zhang-long

2005, 26(1): 65-69.

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The texture evolution of plates after rolling at different temperatures was investigated. The result shows that after hot-rolling, cold-rolling and re-crystallized annealing, textures of the N18 plates varied, i.e., f_n and f_r increased,and f_t decreased. The effect of the temperature of hot rolling on the texture can be summarized as follows: when hot rolling temperature changed from 750℃ to 780℃, f_n decreased, and f_t and f_r increased slightly. But when hot rolling temperature was higher than 780℃, there was little difference in texture with the increasing of temperature. β phase can be formed in N18 when temperature was higher than 780℃, and the volume fraction of β phase increased with the increasing of temperature. The deformation of N18 at 750℃ followed the HCP crystal deformation mechanism. When rolling temperature was 780℃ ~820℃, the textures of N18 were mainly controlled by the deformation of α phase. But when temperature was high enough, non-crystal mechanism and crystal boundary slide mechanism were also activated, besides the slip mechanism. Crystal boundary slide mechanism can be promoted by β phase distributing along α grain boundaries.

Friction and Wear Characteristics of Plasma Spray Metal Coatings on 1Cr17Ni4 Stainless Steel

ZHANG Peng-cheng, BAI Bin, TAN Xing-long, QIU Shao-yu

2005, 26(1): 70-74,95.

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FeCrAlY,NiCrAlY₂O₃,NiCrAlY coatings on 1Cr17Ni4 stainless steel were prepared by plasma spray. Friction and wear characteristics of these coatings were studied under dry sliding of small distance. The strength of these coatings was measured by tensile method. The friction coefficients of these coatings are almost the same, and much smaller than that of 1Cr17Ni4 matrix. The wear resistance of Ni-based coatings had been improved, and wear resistance of NiCrAlY coating is the best. Wear resistance of FeCrAlY coating is worse than that of 1Cr17Ni4 matrix. The predominant wear mechanism of the coatings appeared to the plastic deformation of abrasive wear and the fatigue delamination.

Study of Factors with Effects on the Grain Size in Gd₂O₃-UO₂ Pellets

PAN Ying, CHEN Yi, YI Wei, ZUO Guo-ping, JIANG Hong-man

2005, 26(1): 75-79.

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The factors with effects on the grain size in Gd₂O₃-UO₂ pellets have been investigated by metalloscope and image analyzer. We have investigated the effects on the grain size in Gd₂O₃-UO₂ pellets due to the sintering atmosphere, blending methods, pellets position in sintering furnace, and additive U₃O₈, ammonium oxalate and oxides （Al, Ti, V） in Gd₂O₃-UO₂ pellets. The results have shown that grain growth and grain size distribution homogeneity of Gd₂O₃-UO₂ pellets were enhanced by mixing methods using ball process, additive of oxides （Al,Ti,V） and sintering of low oxidizing atmospheres. The effect of adding U₃O₈ and ammonium oxalate on the grain growth of pellets is not obvious.

Application of Extended Cut Sets Matrix in Fault Tree Rapid Analysis Strategy

YANG Yu, LIU Xiao-ping, LIU Ping, WU Yi-can

2005, 26(1): 80-83.

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One of the most concerned issues in probabilistic security assessment （PSA） of nuclear power plants is living PSA（LPSA）, and the key issue in LPSA is speed. Fault tree rapid analysis strategy is very important LPSA. The extended cut sets matrix method has been presented, and applied in fault tree rapid analysis strategy to generate module and fault tree minimal cut sets （MCSs）. Some optimal policies based on extended matrix also haven been presented.

Research of Electrochemistry Behavior of Slower for Nuclear Steam Generator

XU Qi, HUANG Xin-quan, SU Xing-wan, MA Wei-gang

2005, 26(1): 84-87.

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Corrosion of construction material in high temperature water is one of the important problems of nuclear reactor. Lately research shows that TiO₂ inhabiting species can alleviate local corrosion, such as stress corrosion cracking （SCC） and intergranular attack （IGA） of steam generator tubes, and thus prolong the life of SG tube. This paper has studied the corrosion potentials and polarization carve of steel stainless in caustic solution with TiO₂ inhabiting species, and it shows that TiO₂ have positive effect.

Development of Operation Management Database for Research Reactors

ZHANG Xin-jun, CHEN Wei, YANG Jun

2005, 26(1): 88-90,95.

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An Operation Database for Pulsed Reactor has been developed on the platform for Microsoft visual C++ 6.0. This database includes four function modules, fuel elements management, incident management, experiment management and file management. It is essential for reactor security and information management.

Study of Advanced Evaluation System Based on a Fuzzy AHP Method for Nuclear Reactors

ZHANG Qi, ZHU Shu-tang, ZHOU Zhi-wei, GAO Shan-shan, HUA Jun

2005, 26(1): 91-95.

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Based on current status and developing trend of nuclear power technology, an advanced evaluation system is developed to assess nuclear reactors based on the study of a set of pre-established evaluation indices, including the selection and classification of the indices, and the determination of their weighting factors in applying AHP method. In this system the fuzzy method is used to build a fuzzy evaluation model with respect to the relativity and fuzziness of NPPs’ specifications.This evaluation system is computer network based, with high flexibility for modification, and it is easy to use and to display evaluation results. Some preliminary evaluation results are provided and discussed.

2005 Vol. 26, No. 1