核动力工程

Comparison of Elastic-Plastic FE Method and Engineering Method for RPV Fracture Mechanics Analysis

SUN Ying-xue, ZHENG Bin, ZANG Feng-gang

2009, 30(4): 1-3,8.

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This paper described the FE analysis of elastic-plastic fracture mechanics for a crack in RPV belt line using ABAQUS code. It calculated and evaluated the stress intensity factor and J integral of crack under PTS transients. The result is also compared with that by engineering analysis method. It shows that the results using engineering analysis method is a little larger than the results using FE analysis of 3D elastic-plastic fracture mechanics, thus the engineering analysis method is conservative than the elastic-plastic fracture mechanics method.

Study on Criterion for Leak Before Break Assessment of Pressure Pipes

YANG Lin-juan, SHEN Shi-ming

2009, 30(4): 4-8.

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Based on the elasto-plastio fracture mechanics, this paper established the expression formulas of limit buckling pressure P_u on the ligament of axial and circumferential surface cracks and the initial pressure for the through cracks P_c. A new Leak Before Break （LBB） assessment criterion was put forward to predict the failure mode of pressure pipes, i.e., when P_u is less than P_c, the pipe will leak; when P_u is equal to or larger than P_c, the pipe will break, which is verified by the test data reported in literatures.

Creep Based Stress Relaxation Damage Model for High Temperature Components

GUO Jin-quan, XUAN Fu-zhen, WANG Zheng-dong, TU Shan-dong

2009, 30(4): 9-12.

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Based on Schlottner-Seele average creep rupture rate theory and relaxation function, this paper constructs a relaxation damage model, with which to predict the relaxation damage life of high temperature bolting material 1Cr10NiMoW2VNbN. Validation results of relaxation tests indicate that the developed model has led to better consistent results with the experimental data.

Study on Acoustic Emission Signals of Active Defect in Pressure Piping under Hydraulic Pressure

AI Qiong, LIU Cai-xue, WANG Yao, HE Pan, SONG Jian

2009, 30(4): 13-16,58.

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Experimental investigations of acoustic emission （AE） of active defect in pressure piping with a prefabricated crack under hydraulic pressure tester were conducted. AE signals of fatigue-crack-growth in pressure piping were monitored incessantly in all processes, and all signals recorded were analyzed and processed. The result of signal processing show that the amplitude and energy of acoustic emission signals from defect in pressure pipeline increase gradually with the load time, and thus the active defects in pipeline can be identified; the amplitude, energy and count of acoustic emission signals increase sharply before the defect runs through, and we can forecast the penetrated leakage of pipeline .

Reliability Analysis on Eddy Tube-Defects Quantitative Detection in Nuclear Power Plants

YAO Yun-ping, HAN Jie, LIAO Shu-sheng

2009, 30(4): 17-20.

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Finite element method is used in the reliability analysis on the quantitative detection of ECT （eddy current test）, and based on the calculation and analysis of of the tube signals in nuclear power plants, the quantitative excursion of tube defects and qualitative relations under different depth and width conditions are obtained, which proved that reliability analysis is more practical when the width of the tube defects is very small, and the efficiency and precision of ECT can be improved.

Fracture Mechanics Analysis of Flaw in Outlet Nozzle of RPV

SUN Ying-xue, ZHENG Bin, ZANG Feng-gang

2009, 30(4): 21-23.

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This paper describes the fatigue crack propagation analysis and fast fracture mechanics analysis for the flaw in the outlet nozzle of Reactor Pressure Vessel（RPV） in Daya-bay Nuclear Power Station, using fracture mechanics analysis method. The calculation result is evaluated by the code, and it indicates there is no risk in outlet nozzle with the flaw.

Hybrid Low-Order Harmonics and Linear Perturbation Expansion Method for Fast Loading Pattern Evaluation—Theory

ZHANG Shao-hong, WANG Tao, LU Dong, FU Xue-feng, CHAO Yung-an

2009, 30(4): 24-27,36.

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In order to address the issue of large computational workload in the process of optimum loading pattern search for commercial PWR NPP, a new method--Hybrid Harmonics and Linear Perturbation （HHLP）is proposed to perform the fast loading pattern evaluation. The method is characterized by separating the local perturbation effect of neutron flux distribution induced by fuel shuffling from the global flux tilting effect, and using low order harmonics and linear perturbation base function expansion to represent these two effects respectively. Since by adoption of the weighted-residual method, the solution of the original large-scale eigenvalue problem can be replaced by the solution of a small-scale eigenvalue problem on expansion coefficients, the proposed method can significantly reduce the computational time for a loading pattern evaluation. This paper gives an introduction of the theory of HHLP and another accompanying paper will provide examples for various kinds of numerical verification of using HHLP.

Exponential Function Method for Solving Point Reactor Neutron Kinetics Equations

LI Hao-feng, CHEN Wen-zhen, ZHU Qian, LUO Lei

2009, 30(4): 28-31,67.

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An exponential function method for solving the point reactor neutron kinetics equations is presented. The reactor kinetics equations are reduced to a differential equation in matrix form, which is con- venient to express explicitly by using exponential functions as basis function. The coefficients of the series have been obtained according to initial conditions. The method is applied to the step reactivity insertion, ramp reactivity insertion and oscillatory reactivity insertion respectively,and the results show that the exponential function method is efficient and accurate to several significant figures.

Study on Method of Characteristics Based on Cell Modular Ray Tracing

TANG Chun-tao, ZHANG Shao-hong

2009, 30(4): 32-36.

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To address the issue of accurately solving neutron transport problem in complex geometry, method of characteristics （MOC） is studied in this paper, and a quite effective and memory saving cell modular ray tracing （CMRT） method is developed and related angle discretization and boundary condition handling issues are discussed. A CMRT based MOC code-PEACH is developed and tested against C5G7 MOX benchmark problem. Numerical results demonstrate that PEACH can give excellent accuracy for both k_eff and pin power distribution for neutron transport problem.

Analysis of Characteristics of Different Working Fluids for Gas Turbine Cycle with High Temperature Gas-Cooled Reactor

ZHANG Jian-cheng, ZHANG Hui-sheng

2009, 30(4): 37-40.

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Gas turbine cycle with high temperature gas-cooled reactor is the main direction of nuclear energy generation, which is with the advantages in terms of the safety and economy. The thermal and physical properties of helium, nitrogen, carbon dioxide and the mixtures were compared and analyzed in this paper. Further more, the heat transfer coefficient, pressure loss and the stage number of turbo-machines have been also compared. Results indicate that taking the mixture of helium and carbon dioxide as the working fluid of gas turbine cycle with high temperature gas-cooled reactor can not only improve the heat transfer coefficient and decrease the stage number of turbo-machinery, but also can limit the pressure loss to a certain level.

Two Phase Heat Transfer Characteristics of Horizontal and Vertical Tube Heat Exchangers in Pool with Internal Heat Source

LI Long-jian, CUI Wen-zhi, WANG Xiao-jun

2009, 30(4): 41-46.

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Based on the geometrical structure and operation condition of test prototype of medical isotope reactor, the three-dimensional physical and mathematical models on the natural convection and bubbly driven two-phase convective heat transfer in the pool with internal heat source and with horizontal and vertical tube heat exchanger immerged were proposed, and the corresponding numerical simulation was performed. The computed heart transfer coefficients for different internal heat source and bubble flow rate were gained and compared with the detected ones by employing the analogy theory and dimension analysis, the empirical criteria correlations of Nusselt number of two phase heat transfer for the horizontal and vertical heat transfer exchanger was derived and compared with each other, which was provided with theoretical reference for the selection of the heat exchanger tube layout

Analysis of Two-Phase Flow Pressure Drop in Inclined Annular Channel

HONG Gang, SUN Qi, ZHAO Hua

2009, 30(4): 47-51,95.

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The experiment is investigated under the condition of: P=1~3 MPa, G=190~1 050 kg/（m²·s）, ΔT_sat=20~70℃, q=304~1873 kW/m² in an annular channel which is heated by a tube at its center with the co-current stream-water system. By analyzing the change of forces, a criterion for working out how pressure drop is influenced by inclination and void fraction has been developed. Based on the criterion, a modified two-phase flow pressure drop correlation for angles from vertical upward to horizontal has been developed. The result indicates that the modified correlation is fit for calculating the pressure drop of two-phase flow in the inclined channel.

Study on Fault Diagnosis in Nuclear Power Plant Based on Rough Sets and Support Vector Machine

XU Jin-liang, CHEN Wu-xing, TANG Yao-yang

2009, 30(4): 52-54,85.

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The faults of Nuclear Power Plant （NPP） are featured with complication and uncertainty. A NPP fault diagnosis method based on Rough Sets （RS） and Support Vector Machine （SVM） is proposed. Firstly, the uncertain data is reduced based on RS theory. According to the chosen reduction a SVM multi-classifier is designed for fault diagnosis. Finally this method is used to diagnose four typical failures, i.e., steam generator tube rupture accident, cold leg rupture accident, vapour phase rupture accident and loss of heat sink accident. The result shows that this method can diagnose the faults of the NPP rapidly and accurately.

Application of GO-FLOW Methodology to RHS Reliability Analysis in PWR

HUANG Tao, CAI Qi, ZHAO Xin-wen, SHANG Yan-long

2009, 30(4): 55-58.

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The Go-FLOW chart of residual heat-removal system （RHS） has been drawn, and the success probability of RHS in different time points has been calculated with GO-FLOW algorithm based on the principle of GO-FLOW methodology. The results show that GO-FLOW can describe the functions of a system with a complex operating sequence and multiple states. GO-FLOW methodology is an effective, intuitionistic, and accurate method for analyzing the system reliability.

Classification Analysis of Organization Factors Related to System Safety

LIU Hui-zhen, ZHANG Li, ZHANG Yu-ling, GUAN Shi-hua

2009, 30(4): 59-63.

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This paper analyzes the different types of organization factors which influence the system safety. The organization factor can be divided into the interior organization factor and exterior organization factor. The latter includes the factors of political, economical, technical, law, social culture and geographical, and the relationships among different interest groups. The former includes organization culture, communication, decision, training, process, supervision and management and organization structure. This paper focuses on the description of the organization factors. The classification analysis of the organization factors is the early work of quantitative analysis.

Preliminary Study on Outage Cycle Extension of Qinshan CANDU NPP

JIANG Fu-ming, WANG De-zhong

2009, 30(4): 64-67.

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Currently, most NPP operators have attached great importance to the optimization of outage duration. By taking various optimization measures, the operators have managed to optimize the outage duration and improve the operational performance. CANDU NPP, with its unique feature of on-power refueling and less constraint from fuel burn-up, offers the opportunity of longer outage cycle. With the extension of outage cycle, it is possible to reduce the number of outages in the life time of NPP, minimize the number of shutdown and start-ups, increase the outage preparation lead-time, and improve the manpower preparation and training for outage, which could effectively lead to improved capacity factor, outage and operational performance in the life time of the NPP. This paper provided the preliminary study results and process for outage cycle extension taking into account the situation of CANDU NPP in China and abroad, aiming to further improve the operational performance and competitiveness of CANDU NPP in China.

Analysis and Solutions of Existing Acceleration Protection Problems in 900 MW Unit of Daya Bay Nuclear Power Station

TIAN Feng, LUO Xiang-dong

2009, 30(4): 68-70,80.

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In this paper, an incident in Daya Bay 900 MW nuclear unit, in which the turbine-generator acceleration protection action was initiated by three-phase short circuit fault of power grid and then manually tripped by the operator, was introduced, and the related issues were analyzed.An improvement measure to cancel the response of the high pressure steam valve to the acceleration protection have been put forward. The safety stability of the grid and nuclear generator unit has been improved by improving the related automatism and protection and their control strategy.

Step-Orifice Applied in Nuclear Piping System

ZHANG Yi-xiong, MAO Qing, XIANG Wen-yuan, BI Qin-cheng, WANG Wei

2009, 30(4): 71-74.

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Taking the EAS test pipeline in Daya Bay Nuclear Power Station as an example, this paper studies the method for the analysis of piping vibration caused by the cavitations and the design method by using the step-orifice to reduce the cavitations. Using the CFD method, the flow characteristics and pressure distribution around the orifice are studied to determine if the cavitations occurs in the downstream of the orifice. Finally, the cavitation number approximate criteria for step-orifice are adopted to redesign a new throttle orifice. It has been proved that the redesigned step-orifice can effectively reduce the piping vibration and noise.

Analysis and Experimental Study on Hydraulic Balance Characteristics in Density Lock

GU Hai-feng, YAN Chang-qi, SUN Fu-rong

2009, 30(4): 75-80.

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Through the simplified theoretical model, the hydraulic balance condition which should be met in the density lock is obtained, when reactor operates normally and density lock is closed. The main parameters influencing this condition are analyzed, and the results show that the hydraulic balance in the density lock is characterized with self-stability in a certain range. Meantime, a simulating experimental loop is built and experimental verification on the self-stability characteristic is done. Moreover, experimental study is done on the conditions of flow change of work fluids in the primary circuit in the process of stable operations. The experimental results show that the hydraulic balance in the density lock can recovered quickly, depending on the self-stability characteristic without influences on the sealing performance of density lock and normal operation of reactor, after the change of operation parameters breaks the hydraulic balance.

Numerical Simulation of Interior Flow Field of Nuclear Model Pump

WANG Chun-lin, PENG Na, KANG Can, ZHAO Bai-tong, ZHANG Hao

2009, 30(4): 81-85.

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Reynolds time-averaged N-S equations and the standard k-ε turbulent model were adopted, and three-dimensional non-structural of tetrahedral mesh division was used for modeling. Multiple reference frame model of rotating fluid mechanical model was used, under the design condition, the three-dimensional incompressible turbulent flow of nuclear model pump was simulated, and the results preferably post the characteristics of the interior flow field. This paper first analyzes the total pressure and velocity distribution in the flow field, and then describes the interior flow field characteristics of each part such as the impeller, diffuser and spherical shell, and also discusses the reasons that cause these characteristics. The study results can be used to estimate the performance of nuclear model pump, and will provide some useful references for its hydraulic optimized design.

Numerical Simulation of Thermal-Dynamic Characteristics through a Helical Coiled Tube with Annular Cross Section for Laminar Flow

WU Shuang-ying, CHEN Su-jun, LI You-rong, LI Long-jian

2009, 30(4): 86-90.

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A numerical method for simulating three-dimensional laminar forced convective heat transfer in a helical coiled passage with annular cross section under uniform wall temperature condition is presented. The helical coiled passage is fabricated by bending a 0.03m inner diameter and 0.05m outer diameter straight tube into a helical-coil of two turns. The results presented in this paper cover a Reynolds number range of 200~1000, a pitch range of 0.1~0.2 and a curvature ratio range of 0.1~0.3. The numerical computations reveal the development and distribution of heat transfer and flow fields in the helical coiled passage when the inner annular wall is heated and the outer annular wall is insulated. In addition, the effects of Reynolds number, curvature ratio, and coil pitch on the average friction factor, average Nusselt number at different axial cross-section have been discussed. The results show that the secondary flow is weak and can be neglected at the entrance region, but the effect of the secondary flow is enhanced, the maximum velocity perpendicular to axial cross section shifts toward the outer side of helical coiled passage. Furthermore, the average Nusselt number and friction factor at every different axial location present different characteristics when the Reynolds number, curvature ratio and pitch change. Compared with the curvature ratio, the pitch has relatively little influence on the heat transfer and flow performance.

Numerical Simulation for Thermal Stratification of PWR NPP Pressurizer Surge Line

ZHANG Li, LAI Jian-yong, HUANG Wei, LI Hai-ying

2009, 30(4): 91-95.

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In order to evaluate the structural integrity of pressurizer surge line affected by thermal stratification, this paper presents the numerical simulation for thermal stratification of PWR NPP pressurizer surge line by using Computational Fliud Dynamics（CFD） method. The flow in the surge line is studied, and the heat transfer characteristics, flow field and temperature distribution are obtained and the relationship between the phenomenon of thermal stratification and the surge velocity is analyzed. The result shows that the maximal temperature difference increases along with the increasing of surge velocity, as the surge velocity changes within certain range. The maximum temperature difference of cross section and its locations with different surge velocity are obtained, and vulnerability of the surge line when thermal stratification occurrs is pointed out.

Modeling of Reactor Power Control System and Closed Loop Verification

LIN Hua, LIN Meng, HOU Dong, YANG Yan-hua

2009, 30(4): 96-99,112.

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The model of nuclear power control system of Ling’ao NPP based on MATLAB/SIMULINK was established. Closed loop test was realized by coupling with the thermal-hydraulic model of primary loop which is based on RELAP5. The comparison of simulated result with test result from NPP shows a good coordination, and the correctness of the model was proved qualitatively and quantitatively.

Construction of Experience Feedback System for Equipment Supervision in Nuclear Engineering

ZOU Ping-guo, ZHANG Li-ying, ZHANG Wen-zhong

2009, 30(4): 100-104.

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Based on the analysis of the experience sources on equipment supervision in nuclear engineering, the details of the organization principle, working flow, and report requirement for the experience feedback system are introduced. The function range and its roll in the experience feedback system of the nuclear authority, nuclear power plant owners and equipment supervision organizations are illustrated. The standardization working requirements in the information gathering, analyzing, feedback and tracking process, and the characteristics and form of the incident report and feedback report are proposed. It emphasizes that the method for combined analysis of one significant incident and the whole incidents shall be adopted in the information analysis, and the experience feedback shall be considered in the development of equipment supervision technique and the equipment manufacturing, thus to maximize the use of experience feedback information to improve the pertinency and effectiveness of the experience feedback system.

Data Mining in Nuclear Engineering

JIANG Bo-tao, ZHAO Fu-yu

2009, 30(4): 105-107,112.

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Data mining （DM） is a process to find the useful and interesting information in huge data. Support Vector Machine （SVM） is a new technique in data mining, but Support Vector Regression （SVR） is the applying of SM in regression . Compared with the traditional regression methods, SVR has not been specified beforehand, and is fitted directly from the inner relationship of data, thus the simulation results are more accurate. This paper introduces the mathematical theory of SVR and uses SVR to process the data of the moving characteristics of molten metal droplets in serious nuclear engineering accidents.

Discussion of REA Boric Acid Tank Volume

SHENG Guo-long

2009, 30(4): 108-112.

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This paper discussed the design rules of the boric acid tank volume of reactor boron and water makeup system（REA） in PWR nuclear plant and the limit requirements of the tank volume in the operation technical specification. Based on the comparison and analysis of the arguments of Daya Bay nuclear plant 18 months refueling modification and Ling’ao nuclear plant 1/4 refueling modification, the paper pre-assessed and calculated the lack of the REA boric acid tank volume as the fuel enrichment upgraded to 4.95% in the future, and proposed the suggestion of increasing the volume or boric acid concentration of the REA system in new nuclear plant design in order to meet the operation technical specification requirements.

2009 Vol. 30, No. 4