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2015 Vol. 36, No. 5

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Review of Underground Nuclear Power Plant
Niu Xinqiang, Luo Qi, Zhao Xin, Zhang Wenqi, Liu Haibo, Li Xiang
2015, 36(5): 1-5. doi: 10.13832/j.jnpe.2015.05.0001
Abstract(11) PDF(0)
Abstract:
One of the new directions of safe development for the future nuclear energy after Fukushima accident is to investigate the innovations for a better design that is safer and more acceptable by the public, which may imply the return of underground nuclear power plant. In this review both foreign and domestic R&D achievements are introduced with dedicated preamble on recent studies of underground nuclear power plant in China, which demonstrate the convenience of surrounding rocks as barrier of preventing radioactive materials to the outside environment for the implementation of "practical elimination of large radioactive release" by design.
Research on Key Technologies of Large-Scale Underground NPPs
Niu Xinqiang, Luo Qi, Zhao Xin, Zhang Wenqi, Liu Haibo, Li Xiang
2015, 36(5): 6-11. doi: 10.13832/j.jnpe.2015.05.0006
Abstract(12) PDF(0)
Abstract:
Based on the mature technology of self-reliance 600 MW nuclear power plant design, and the construction feasibility study of large-scale underground nuclear power plants, the underground plant CUP600 with new models and technical scheme of independent intellectual property is developed and proposed for the first time in China, forming a complete set of engineering key technology. The evaluation demonstrates that the safety of this underground nuclear power plant is high, the construction technology is feasible, and the economy is reasonable with specific features of the fourth generation nuclear power technology, which fully meet the requirements of the latest nuclear safety standards of the state.
Mechanical Calculation and Evaluation of Supports for Nuclear Power Plant Steam Generator
Xu Yu, Chu Qibao, Wang Qing, Liu Le, Wen Jing, Li Hailong
2015, 36(5): 12-14. doi: 10.13832/j.jnpe.2015.05.0012
Abstract:
Standardized design is used for the steam generator lateral supports of a nuclear power plant. The types of the supports are various, including plate-type and shell-type supports, and linear-type support, and bolting and welded joints are also used, which makes the mechanical calculation and evaluation more complicate. Finite element analysis software is used to calculate the steam generator upper and lower lateral supports under level-D load case, then ASME ⅢNF and appendix F is used to evaluate all the types of supports. The key points of the mechanical calculation and evaluation of the SG supports are proposed.
Application of Two Response Spectrum Analysis Methods from RG1.92 for Steel Frame
Xu Shenghong, Liu Shenghua, Peng Xiaoping, Yang Shifeng
2015, 36(5): 15-17. doi: 10.13832/j.jnpe.2015.05.0015
Abstract(12) PDF(0)
Abstract:
In order to consider mass lose of frequency part of reaction response analysis, RG1.92(2006)separate nuclear power plant structure, systems, and components(SSCs) into two parts, which are referred to as the "periodic" response and the "rigid" response. Modes of steel frames in the containment vessel distribute in the periodic response region and rigid response region. A response spectrum analysis on a steel frame inside the containment is performed based on Gupta +Missing Mass and Lindley-Yow +ZPA to study the difference of the two methods. The software GTStrudl is used to perform the spectrum analysis based on the two methods above and to find a better way to complete the calculation of Steel frames in containment vessel.
FEM Analysis for Dropping Test of Fresh Fuel Cask
Li Nan, Han Zhi, Wang Jun, Wang Kunpeng, Zhang Chunming, Yi Dayong
2015, 36(5): 18-21. doi: 10.13832/j.jnpe.2015.05.0018
Abstract(14) PDF(0)
Abstract:
Several kinds of non-metallic material such as cork compound, nano-ceramic and coir mattress are used in the fresh fuel cask design for vibration control and fireproofing. FEM analysis for dropping test of the cask is performed using the ANSYS/LS-DYNA program in this paper. Firstly, the material model and material parameters are determined on the basis of experimental data with the cylindrical specimens. Secondly, the FEM model, contact condition and hourglass control of foam material model are defined. Finally, the experiment scheme of 1.2 meter dropping test is confirmed though FEM analysis of dropping tests with different dropping angles. The result presents that the stress intensity of cask is within the limit in ASME BPVC-III.
Evaluation Technology for Defects of Reactor Pressure Vessel Bolt Hole and Bolt
Zheng Liangang, Zou Mingzhong, Xie Hai
2015, 36(5): 22-24. doi: 10.13832/j.jnpe.2015.05.0022
Abstract(10) PDF(0)
Abstract:
During the manufacture and installation of reactor pressure vessel bolt hole and bolt, the existence of defects is inescapable. The analysis results show that most of the defects have no effect on the function and safety of the structure. This paper summarizes some kinds of defects of bolt hole and bolt. The received conditions and evaluation technology are provided here.
Optimal Analysis of Seismic Isolation of PCCS Water Storage Tank
Hou Gangling, Teng Fei, Jia Xiaofei, Song Tianshu
2015, 36(5): 25-29. doi: 10.13832/j.jnpe.2015.05.0025
Abstract:
PCCS in AP1000 includes two parts, namely a reinforced concrete containment, which is modeled as structural dynamic, and a gravity drain water tank, which is modeled as lumped masses known as sloshing mass, impulsive mass and rigid mass. Based on the seismic response characteristics of the water storage tank and the reinforced concrete containment, the conventional model and isolated model are built respectively. The optimal analysis of isolated layer parameters, namely the natural period and damping, are provided. Using the nuclear power plant dynamic characteristics and structural design codes, three models, namely conventional structure, isolated model on design code, and isolated model on the optimal parameter, are studied in MATLAB software. The seismic responses of all models are compared under the real earthquake ground motions. Results show that the isolated model with the optimal parameter provided by this paper has the best seismic function in all models.
Application of Nuclear Piping Calculation Program
Ning Qingkun, Chen Li, Tang Yujian, Tian Jinmei
2015, 36(5): 30-32. doi: 10.13832/j.jnpe.2015.05.0030
Abstract:
A nuclear piping calculation program is developed, that can use a variety of codes for nuclear piping stress analysis and evaluation. This paper introduces the principle of the program. Taking the piping of a nuclear power station as an example, the stress analysis and evaluation is done using RCC-M and ASME codes, and the results are compared with SYSPIPE and PIPESTRESS respectively. The results show that the program calculation result is correct, and the precision meet the requirements.
Seismic Analysis of Collection Tank for Middle Level Radioactive Waste Liquid
Jing Dan, Wang Jun, Dai Shoutong, Luan Lin
2015, 36(5): 33-36. doi: 10.13832/j.jnpe.2015.05.0033
Abstract:
Collection tank for middle level radioactive waste liquid is a large standing liquid tank. Considering the fluctuant effect of the liquid, the 3D mass-spring finite element model is applied. The response under the earthquake were calculated by means of response spectrum analysis. The results show that the mass-spring model referenced to 3-D model is effective and the maximum stress concentration of the structure appears at the junction of the tank and the leg.
Experimental Study on Thick Rubber Bearings of Three-Dimensional Isolation of Nuclear Power Plants
Wang Tao, Li Jichao, Wang Fei
2015, 36(5): 37-40. doi: 10.13832/j.jnpe.2015.05.0037
Abstract:
The application of vertical isolation technique to nuclear power plants could shift the main frequency of floor response spectra(FRS) to 2~4 Hz, much lower than the operation frequency of the equipment and pipeline, thus reduce the damage to the inner equipment and pipeline due to the vertical earthquake. This paper provides an experimental study on the thick rubber bearings. Mechanical behaviors of two types of rubber-based bearings, normal bearings and thick rubber bearings were tested according to the code, including horizontal shear and vertical compress tests. The tests reveal that the thick rubber bearings possess a similar shear stiffness and smaller compressive stiffness nearly 1/8 compared with the normal rubber bearings. Therefore, the thick rubber bearings are suitable for loading element of new type three-dimensional isolation equipment.
Test and Research for Integrity Evaluation of Nuclear Pipe System under Seismic Effect
Zhang Shiwei, Shen Shuangquan, Liu Linlin, Xu Yugen, Li Xihua
2015, 36(5): 41-44. doi: 10.13832/j.jnpe.2015.05.0041
Abstract(13) PDF(0)
Abstract:
For the conservative treatment of sporadic seismic load, a lot of dampers and drop frames were used in the nuclear power plant piping system, which lead to the high cost for the manufactory, installation, in-service inspection and repair. According to the research scheme for integrity evaluation of pipe system under seismic effect, this paper completed the ultimate anti-seismic capacity test of nuclear pipes. Through the comparative study of the test and calculation results and the current specification, the safety margin of current pipeline design standard is clarified. A recommend value of damping ratio and stress evaluation of nuclear piping system is given.
Simple Method for Considering Soil-Structure Interaction in Three-Dimension Infinite Element Model of Nuclear Power Plant Building
Zhu Xiuyun, Qin Fan, Pan Rong, Li Jianbo
2015, 36(5): 45-49. doi: 10.13832/j.jnpe.2015.05.0045
Abstract(14) PDF(0)
Abstract:
In order to equivalently disperse the lumped soil dynamic stiffness of half space for the raft foundation of 3-D nuclear power plant building, the formula of spring-damping model of considering soil-structure interaction(SSI) for 3-D nuclear power plant building is deduced based on the lumped mass model considering SSI. This method of equivalently disperse is validated by modal analysis and dynamic time history analysis. This method solves the soil dynamic stiffness firstly, and then exerts the spring-damping element at the raft foundation of 3-D nuclear power plant building, which is easier and simpler to realize, compared to other artificial boundary methods.
Contact Nonlinear Computation Method for Multi-Subassemblies Interaction in Fast Reactor Core
Gao Fuhai, Li Nan
2015, 36(5): 50-53. doi: 10.13832/j.jnpe.2015.05.0050
Abstract:
The study focuses on the contact nonlinear computation method for multi-subassemblies interaction, and tries to verify the extendibility of the method proposed by Likhachev and the viability of resolving CEFR subassembly contact problem. Based on the computation method developed by Likhachev, the consistent deformation conditions and the system potential energy equations for multi-subassemblies are established at three contact levels instead of two levels. The minimum potential energy principle is utilized to acquire the contact forces. For simplicity, orthogonal transform and Lagrangian-double-problem transformation are used. Finally, the nonlinear contact analysis turns into the optimization of a quadratic function with conditional inequality constraints. Conclusions are as follows: Likhachev method can be extended for three contact levels from two, and is generally practicable; the final derived expression is a common problem in optimization theory, which could be easily solved numerically.
Application of Multi-Point Response Spectrum Method in Seismic Analysis of Nuclear Power Equipment
Wang Yanping, Peng Xingming, Tang Yujian
2015, 36(5): 54-56. doi: 10.13832/j.jnpe.2015.05.0054
Abstract:
This paper provides a brief theory description of the single-point response spectrum method and the multi-point response spectrum method, respectively. Taking the new nuclear fuel elevator as an example, the seismic analysis is carried out by the two methods with the software of ANSYS. The results of the calculation are contrasted. It shows that the result obtained by the single point response spectrum method is too large and that obtained by the multi-point response spectrum method is more reasonable. And some reference can be provided for the seismic analysis of this type of nuclear power equipment.
Research on Vibration Characteristics Boundary Condition of Core Barrel of Nuclear Reactors
Tan Tiancai, Gao Lixia, Yu Danping, Ma Jianzhong
2015, 36(5): 57-60. doi: 10.13832/j.jnpe.2015.05.0057
Abstract(10) PDF(0)
Abstract:
Boundary condition of vibration characteristics both in air and still water about the empty core barrel of Qinshan Nuclear Plant 600 MW reactor and CAP 1400 type reactor were investigated. The structure of core barrel was modeled by solid element, and the fluid structure was simulated by fluid element. The close effect between the flange of core barrel and support step of reactor pressure vessel was assumed with three types of boundary conditions: weak close, medium close and strong close. At the same time, the friction between the flange of core barrel and support step of reactor pressure vessel was assumed with three types of boundary conditions: weak friction, medium friction and strong friction. The research shows that: The boundary condition of core barrel in still water is much stronger than that in air for Qinshan Nuclear Plant 600 MW reactor, but the boundary condition of core barrel both in still water and in air is the same for CAP 1400 type reactor. The preload of hold down spring and friction of flange have a certain effect on the core barrel frequency.
Implementation Evaluation of Small Flow Pipeline Vibration Transformation Plan for RIS System LHSI Pump
He Chao, Xi Zhide, Zhao Yue, Yuan Shaobo, Ma Jianzhong
2015, 36(5): 61-64. doi: 10.13832/j.jnpe.2015.05.0061
Abstract:
The implementation evaluation of small flow pipeline vibration damping transformation plan has been done for RIS system low pressure safety injection pump(referred to as LHSI pump) in a nuclear power plant. Combined CFD numerical calculation with simulation test circuit verification, the evaluation method of the damping effect has been adopted. Vibration limit value calculations are performed according to ASME 0M3. Pipeline stress check, hanger strength and the stiffness check of support have been completed using specialized pipeline software named syspipe. All the numerical results demonstrated that the proposed scheme is effective and feasible, and the scheme can be carried out by specific implementation.
Study on Dynamic Characteristics of Heat Transfer Tube in SG
Gao Lixia, Yu Danping, Tan Tiancai, Ma Jianzhong, Liu Litao
2015, 36(5): 65-67. doi: 10.13832/j.jnpe.2015.05.0065
Abstract:
The secondary cross flow induced vibration of tube bundles of steam generator is the main failure reason of tubes. The flow induced vibration analysis and evaluations are based on the natural frequencies and modes of the heat transfer tube. A study on the dynamic characteristics of the heat transfer tube of a certain type steam generator have been carried out by calculation and experiment, and the natural frequencies and modes were obtained, and the relative deviations between measured and calculated values are within 8%.
Vibration Control and Optimum Design of Piping System Based on Power Flow
Zhang Xiaoling, Liu Tianyan, Sun Lei, Zhang Kun, Qiao Hongwei, Lin Song
2015, 36(5): 68-71. doi: 10.13832/j.jnpe.2015.05.0068
Abstract(16) PDF(0)
Abstract:
Based on an engineering project, finite element analysis on vibration characteristics of piping system under the excitation of the pump is carried out. Then, by using the genetic algorithm, vibration attenuation is achieved by optimizing the positions of the supports of the piping system with the goal of minimizing the vibration power flow. And antishock performance of piping system must be ensured. The results show that at the aspect of vibration control of piping system, the adjustment of the position of the supports is another effective method for reducing the piping vibration.
Validity Analysis of Rolled Plugs of Steam Generators Tube
Shi Shaobo, Shen Ping, Tian Xianglu, Gao Luyang
2015, 36(5): 72-74. doi: 10.13832/j.jnpe.2015.05.0072
Abstract:
The steam generators tube is a vulnerable part of the primary pressure boundary in nuclear power plants,and its integrity directly affects the safety of NPPs.The tube needs to be plugged when the steam generator(SG) primary side water may leak out to the secondary side because of crack,wear or corrosion of the tube.This paper analyzes a SG tube rolling plug with FEM,simulates and calculates the contact stress between the plug and tube in the process of the rolling plug and after the rolling,to confirm the validity of the plug.The result shows the plug meets the strength requirements.
Numerical Simulation of Separation Efficiency and Pressure Drop of AP1000 Swirl-Vane Moisture Separator
Zhang Huang, Bo Hanliang, Chen Feng
2015, 36(5): 75-79. doi: 10.13832/j.jnpe.2015.05.0075
Abstract(20) PDF(0)
Abstract:
A mathematical model of gas-liquid two phase flow is built based on Eulerian-Lagrangian method, and this model is solved by the commercial CFD software CFX to simulate the flow characteristics in an AP1000 swirl-vane moisture separator in a special compression rate under cold condition. In this model, gas is supposed to be the continuous phase and water droplets moving in the separator are thought to be the dispersed phase. According to the motion characteristics of droplets in the flow field, we consider the droplet is controlled by the drag force, virtual mass force, buoyance force and gravity force, and in consequence build the momentum two-way coupling dynamic model of gas and droplets. Aiming at 9 working conditions, we used CFX software to solve this two phase flow model to get the trajectories of droplets with different diameters, in addition to obtain the separation efficiency and total pressure drop between inlet and outlet of the separator. The results show that the computed separation efficiency agrees well with the cold condition experimental data, and the changing trend of the calculated pressure drop is the same as the experimental value, which reveals the correction of our model.
Vibration Analysis of Coupled Concentric Cylindrical Shells in Contact with an Annular Fluid Region
Dong Yu, Yang Yiren, Lu Li
2015, 36(5): 80-82. doi: 10.13832/j.jnpe.2015.05.0080
Abstract:
The fluid-structure interaction analysis of the vibratory modal characteristics of a vertical thin-walled cylindrical shell containing water, in an adjacent coaxial region is presented by ANSYS software. The cylindrical shells were modeled by 3D solid shell elements(Solsh190) and the fluid was modeled by 3D fluid elements(Fluid30). The influence of the fluid viscosity and compressibility are negligible. The natural frequency of cylindrical shell and inner rigid shell separated by the fluid annulus, is obtained by finite element method. The results show that with the decreasing of fluid annulus gap, the natural frequency decreases. When simulation with ANSYS Fluid 30, the fluid annulus gap g≥1.5 mm and the Reynolds number Rk≥1478, the calculation error is less than 20%
Analysis of Relevant Problems of Shape Flow Resistance in Nuclear Power Plants
Yao Yangui, Shi Yang, Zhang Kai, Lu Mingchao
2015, 36(5): 83-86. doi: 10.13832/j.jnpe.2015.05.0083
Abstract:
In the nuclear power plant, the flow resistance is an important factor for the design of the primary system. This paper clarified the cause for the shape flow resistance by analyzing the sudden expansion flow. The effect of the downstream resistance object on the total resistance was also examed by analyzing the flow resistance of the sudden expansion configuration close to a uniform pore plate or a sudden contraction configuration. The designers were advised to analyze the coupling effect of the upstream and downstream resistance objects in detail. To a specific problem, special analysis and demonstration were necessary to make sure whether the total resistance was equal to the sum of the resistance of each object or not.
Random-Turbulence Excitation Research of Steam-Generator Inlet Tubes
Gao Lixia, Tan Tiancai, Yang Jie, Yu Danping, Ma Jianzhong
2015, 36(5): 87-90. doi: 10.13832/j.jnpe.2015.05.0087
Abstract:
Random-turbulence excitation research of steam-generator inlet tubes simulated sector model under different pitch velocities has been done by using equivalent power spectral density(EPSD) method given by Axisa and Taylor et al. Compared the measured EPSD to Taylor and Pettigrew’s suggested guidelines, a peak of EPSD at fR≈1 has been observed, which was closed to the guideline boundary. According to test results, the guidelines were been modified and new guidelines were been obtained which were more closed to actual vibration of tubes.
Research on Sloshing Characteristics in Passive Cooling Storage Tank of AP1000 under Long-Period Earthquake
Zeng Xiaojia, Lu Daogang, Dang Junjie, Liu Yu
2015, 36(5): 91-95. doi: 10.13832/j.jnpe.2015.05.0091
Abstract(11) PDF(0)
Abstract:
The structure with long natural vibration period, such as the large-scaled gravity drain tank on the top of AP1000 containment, is vulnerable to long period ground motion. Particularly, the sloshing phenomena in a low frequency range are of great concern. For this kind of nonlinear problem, numerical simulations are complex and possess certain limitations, therefore, we established the experimental models to measure the impact force of the top cover in different positions with various fill depths. The experiments were carried out on the shaking table and the three cycle sine which has the same frequency with the sloshing water was selected as the input excitation. In addition, by comparing with the calculation method proposed by Lu Daogang, the feasibility and accuracy of this calculation method was further verified. At the same time, we found that the calculation results of small amplitude sloshing were relatively small and the results of large amplitude sloshing was relatively large. Hence, the calculation method is considered to be certain conservative.
Application of Normalization Method to J-R Curve Testing of SA335 P11 Alloy Steel at Elevated Temperature
Zhang Xu, Pan Keqi, Liang Bingbing, Dou Yikang
2015, 36(5): 96-100. doi: 10.13832/j.jnpe.2015.05.0096
Abstract(10) PDF(0)
Abstract:
Elastic-plastic fracture toughness parameters of nuclear piping materials at operation temperature are of significant importance to the integrity assessment and fracture analysis based on EPFM for in-situ nuclear piping structures. Conventional elastic unloading method meets limitations at elevated temperature, while the NDRT(Normalization Data Reduction Technique) performs well without the need for simultaneous crack growth monitoring. J-R curve testing using C(T) specimens are carried out at elevated temperature in this paper. To validate the NDRT, the J-R curve at elevated temperature determined using the elastic compliance method is compared with that obtained by NDRT. The result matches each other very well, which provides a reference for researchers in the investigation of the material fracture property at elevated temperatures.
FAT Iterative Method Based on Dual Conical Indentation to Obtain Properties of Materials
Chen Hui, Cai Lixun, Bao Chen
2015, 36(5): 101-104. doi: 10.13832/j.jnpe.2015.05.0101
Abstract(12) PDF(0)
Abstract:
Load-displacement curves are obtained based on the monotonous conical indentation test under half cone angle of 70.3° and 60°. The relationship between the stress-strain curves and the load-displacement curves are found according to Kick’s law of load-displacement curves. After reasonable iteration, eventually uniaxial elastoplastic constitutive relationships of materials are obtained by FE calculation. Through the reverse application of two kinds of titanium alloy, the constitutive relationships obtained by iterative calculation are consistent with the classical uniaxial tensile test results.
Research on Seismic Test of ACP1000 Control Rod Drive Line
Du Jianyong, Li Pengzhou, Li Qi, Liu Linlin, Xu Yugen, Li Tianyong, Ma Jianzhong
2015, 36(5): 105-107. doi: 10.13832/j.jnpe.2015.05.0105
Abstract:
To verify the structural integrity and operability under the seismic condition required by the third generation of nuclear power plants, the seismic test of ACP1000 control rod drive line(CRDL) was conducted on the multi-exciters test unit. One horizontal direction and one vertical direction of the CRDL were excited with multi-frequency earthquake time history. Drop-time of CRDL under earthquake was obtained. The acceleration and strain response values of CRDL under OBE and SSE level were measured. Operation functions under OBE level and safety function under SSE level of CRDL were validated. The seismic test shows that ACP1000 CRDL can keep the structural integrity and operability under the seismic condition.
Assessment of 3# Primary Pump Vibration of Unit 1 in Qinshan NPP3
Yuan Shaobo, Chen Zhigao, Guo Longzhang, Huang Yongbo, Yu Danping, Cong Bin, He Chao, Liu Linlin
2015, 36(5): 108-110. doi: 10.13832/j.jnpe.2015.05.0108
Abstract(11) PDF(0)
Abstract:
The vibration of 3# primary pump of unit 1 in Qinshan NPP3 is serious during the start-up and operation. In order to find out the reasons for the vibration, the main parameters in four fields were measured rotating machinery vibration test, mode test, the vibration test and the heat displacement test of the primary pump. Based on the analysis of the measurement data, the root cause for the vibration of the primary pump is diagnosed, and the modification scheme is put forward. After the implementation of the modification scheme, the vibration is reduced greatly, and the alarm defect of the primary pump vibration has been eliminated.
Improvement of Vibration Measurement for Nuclear-Class Pipes in Nuclear Power Plants during Decommissioning
Zhao Yue, He Chao, Xu Weizu, Wei Chao, Bao Yu
2015, 36(5): 111-113. doi: 10.13832/j.jnpe.2015.05.0111
Abstract:
ASME(American Society of Mechanical Engineers) OM-S / G-2000 Part 3 of insufficient operational guidelines exists defects, while the vibration measurements for nuclear-class pipes in domestic nuclear power plants refer this guideline, with similar test method, which can not fully and accurately predict the pipeline vibration maximum points. In this paper, taking a nuclear power plant during the commissioning of nuclear grade pipe vibration measurements as an example, the screening of test objects, selection of the critical equipment or components, selection of measurement point, field test and vibration evaluation are discussed, and the recommendation for improvement is proposed. Finally, taking some pipe vibration of someone nuclear power plant EAS system(containment spray system) as the research object, we analyze the method of measurement, calculation and evaluation of vibration limits.
Fuel Assembly Model Test Based on SIMO
Xu Yugen, Liu Linlin, Wang Xu, Sun Lei
2015, 36(5): 114-116. doi: 10.13832/j.jnpe.2015.05.0114
Abstract:
Model frequencies and model shapes reflect the inherent characteristics of one kind of fuel assembly. The breakdown or damage of fuel assembly have bad influence on its operation, which is one of the most important reasons for the engineering accident. Based on the multiple curves fitting of transfer function using LMS model analysis software, a hammering test of Single-Input Multi-Output(SIMO) analysis method have been performed on a kind of fuel assembly. The orders of mode frequencies and model shapes were obtained by the test as force to be single-input, and multiple accelerometers to be output. Comparing to the calculation modal, the test results are proved to be validity and reliability.
Study on End-Loading Qualification Test on Nuclear Class 1 Gate Valve and Notice
Xu Yugen, Wang Xu, Liu Linlin, Jiang Shenghan
2015, 36(5): 117-119. doi: 10.13832/j.jnpe.2015.05.0117
Abstract(12) PDF(0)
Abstract:
A type of nuclear class 1 gate valve was tested in the end-loading qualification test. The test intends, test content, test device, test method and results were described in this paper. By performing the force loading on the pipe flange of the valve, the force and stress of the valve was tested. The rigidity, intensity and the functionality of the valve was examined during and after this end-loading qualification test. The test results show that the integrity and operability of this gate valve satisfy the requirement of related documents.
Fracture Mechanics Analysis for Inlet Nozzle of RPV with Pipe Load Considered
Wang Dasheng, Liu Pan, Xiong Guangming
2015, 36(5): 120-123. doi: 10.13832/j.jnpe.2015.05.0120
Abstract:
The structure and load is complex at the nozzle of pressure vessel, a crack is supposed in the corner angular of the RPV inlet nozzle and it is a mixed model crack while considering the pipe load. A three-dimensional finite element model with the supposed crack is established and the SIF(Stress Intensity Factor) distribution and variation of the crack tip with pipe load and internal pressure is analyzed. The results show that the SIF is approximately in symmetrical distribution in the crack tip and the SIF KI, KII, KIII are small with pipe load, and internal pressure do not affect the SIF KII and KIII of the crack tip.
Fatigue and Strength Analysis of CRF Pump with Flaw
Tan Xiaohui, Wang Wei, Luo Ting, Liu Zhiling, Chen Guangyi
2015, 36(5): 124-127. doi: 10.13832/j.jnpe.2015.05.0124
Abstract:
Fatigue and strength analysis are carried out for the CRF pump shaft with correction flaws which are found during outage of Ling’ao Nuclear Power Plant. Moreover, the corrosion flaws are considered as initial cracks and stability of these cracks are evaluated. Life of the pump shaft is further confirmed. Through the aforementioned analysis procedure, structural safety of the pump shaft with flaws can be evaluated, which provides the technical support for on-site decision.
Theory and Simulation Investigation on Crack Tip Blunting for Ductile Material
Pan Keqi, Zhang Xu
2015, 36(5): 128-131. doi: 10.13832/j.jnpe.2015.05.0128
Abstract(12) PDF(0)
Abstract:
High ductile steel are usually adopted for the piping material. The stress concentration and the plastic deformation can be found at the crack tip for the ductile material under external load which make stress and strain of crack tip domain redistributed and the blunting deformation appeared. Based on the relationship of the elastic-plastic material, 1/4 singular elements are used for crack front discretization, under the displacements control, the blunting process are simulated. The hardening zone and the fracture parameter are studied at different loading rates.
Probabilistic Analysis for Characteristic Values of Fracture Toughness
Li Yuebing, Gao Zengliang, Lei Yuebao
2015, 36(5): 132-135. doi: 10.13832/j.jnpe.2015.05.0132
Abstract(13) PDF(0)
Abstract:
The statistical significance for the method of minimum of three equivalent(MOTE) was revealed on the uncertainty of fracture toughness data, using the probability theory. Large numbers of fracture toughness data sets sampled from a parent distribution, which were fitted by 41 fracture toughness values, were used to simulate the test data. Then the characteristic values of fracture toughness for every data set were determined with the MOTE method. At last the failure probability of a typical cylinder containing surface crack was analyzed for every characteristic value, and compared with ASME low bound. The results show that the confidence level is relatively low to estimate a low quantile of parent distribution. The result of assessment will be more reliability with much more test data. However, the confidence level is only 81.5% to ensure that the characteristic value of fracture toughness from the MOTE method is lower than ASME low bound, even if the number of test data exceeds three.
Mechanic Problems in PWR Fuel Assembly Research and Development
Li Pengzhou, Li Qi
2015, 36(5): 136-139. doi: 10.13832/j.jnpe.2015.05.0136
Abstract:
Fuel assembly bowing may induce the reloading or unloading difficulties, and the incomplete insertion of control rod. The fretting between fuel assemblies could result in the component damage and the leakage of nuclear fuels. These phenomena directly affect the safe operation and economic benefits of nuclear power plants. This paper summarizes the typical mechanic problems, such as bowing, fretting, and integrality analysis in PWR fuel assembly research and development. The key factors, resolving methods and prospect are also proposed.
Analysis of A Large Commercial Aero Plane Crash on AP1000 Reinforced Concrete Shield Building
Cheng Shujian, Wang Xiaowen, Ge Honghui, Xia Zufeng, Huang Xiaolin
2015, 36(5): 140-143. doi: 10.13832/j.jnpe.2015.05.0140
Abstract(10) PDF(0)
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This paper mainly concerns about the behavior of the AP1000 RC shield building located at Sanmen and Haiyang when a large narrow-body commercial aero plane impacts on it. An aero plane suggested by HAD with the weight of 90 t is chosen as the impact aircraft. The middle of the cylindrical wall, the air inlet zone and the conic roof are selected as the targets. Their behaviors will be simulated by LS-DYNA. The result of these analysis shows that the RC shield building of AP1000 Plant can resist the impact of this aero plane.
Mechanical Analysis for Personnel Air Lock in Third-Generation Nuclear Power Plants
Shi Ji, Xu Xiaogang, Yang Linmin
2015, 36(5): 144-147. doi: 10.13832/j.jnpe.2015.05.0144
Abstract:
The personnel air lock is one part of the inner reactor containment, which is the third layer of the barrier. Persons and small equipment can pass the personnel air lock when the reactor is running or hot shutdown without damaging the containment sealing function. The personnel air lock should prevent the unacceptable release of radioactive material. Mechanical analysis for personnel air lock in third-generation nuclear power plants is carried out using ANSYS finite element analysis software. A three-dimensional model of the personnel air lock is built to give the various working parameters. The impact of civil works(the forced displacement) is considered, and the stress distribution and the deformation by simulation are obtained.
Effect of Tube Structural Style on Thermal Stress in CEFR Intermediate Heat Exchanger
Zhang Tianyi, Zhang Zhenxing, hu Lina, Xu Yu
2015, 36(5): 148-151. doi: 10.13832/j.jnpe.2015.05.0148
Abstract(11) PDF(0)
Abstract:
Between different layers of CEFR intermediate heat exchanger tubes and inside sleeve, temperature gradient is distributed in radius direction, and thus results in various axial loads. The design of CEFR intermediate heat exchanger tube thermal compensation is bend tube to reduce the thermal stress of equipment. The thermal stress of different tube designing(straight tube and bend tube scheme) is calculated. The result indicates that maximum stress is at the top tube sheet area and stress of bend tube model is greater than the straight tube models. After heat exchanger tubes stress analysis, axial load of different layers heat exchanger tubes is balanced well by compensatory bend part. Because the tangential displacements of compensatory bend part is more, it will bring more tangential load to the top tube sheet.
Stress Analysis and Structural Integrity Evaluation of Piping System in TMSR
Gong Wei, Zhang Xiaochun, Wang Xiao, Fu Yuan
2015, 36(5): 152-155. doi: 10.13832/j.jnpe.2015.05.0152
Abstract:
The design temperature of pipeline can up to 700℃ in the thorium-based molten salt reactor(TMSR).Thus, the high temperature structural design code ASME-NH is adopted for the structural integrity evaluation. The load-controlled stress, the total accumulated inelastic strain as well as the creep-fatigue damage need to be checked in evaluating the structural integrity for elevated temperature. The software ANSYS is used to do the finite element analysis of the overall loop system. The calculation results demonstrate that the demands of the pipe stress are satisfied by repeated optimizing.
Measurement and Analysis of Thermal Strain in Emergency Decay Heat Removal System in CEFR
Yu Huajin, Tang Long, Qi Min, Yu Danping
2015, 36(5): 156-160. doi: 10.13832/j.jnpe.2015.05.0156
Abstract:
The coolant of China experimental fast reactor(CEFR) is sodium, and the emergency decay heat removal system is the safety features in CEFR, which can ensure the removal of the reactor core decay heat passively by air heat exchanger when earthquake, stop of system power supply, loss of steam generator feedwater occur. The emergency decay heat removal system design temperature is 550℃, the maximum operating temperature is 516℃, all piping is double-wall structure, the medium is high temperature liquid metal sodium. Through thermal strain measurement and analysis of emergency decay heat removal system piping, we can grasp the stress and strain of this system, monitor the stress state when the system in operation.
Design of Alarm Software for NPP Virtual Control System
Dong Chenpeng, Leng Shan, Cheng Junjie
2015, 36(5): 161-164. doi: 10.13832/j.jnpe.2015.05.0161
Abstract(14) PDF(0)
Abstract:
Based on the analysis of Ovation alarm function of digital I&C system in nuclear power plants and the simulation requirements of virtual control system, this paper has designed an alarm software specifically for the virtual control system. This software takes the operation results of control function module in the virtual DPU server as the source of the alarm information, and it applies the socket communication model to realize the transmission of the alarm information, and Active X controls have been used for real-time interaction of the alarm display screen. The alarm software can simulate the main features of Ovation alarm system in nuclear power plants, therefore it can provide strong support for training the operators and improving the ability to deal with the fault process in nuclear power plants.
PWR Power Distribution On-Line Monitoring Based on Harmonics Expansion Method
Li Zhuo, Wu Hongchun, Cao Liangzhi, Li Yunzhao
2015, 36(5): 165-168. doi: 10.13832/j.jnpe.2015.05.0165
Abstract(19) PDF(0)
Abstract:
This paper studied the on-line monitoring of PWR 3D power distribution based on the harmonics expansion method. In this method, the power distribution is expanded by the neutron diffusion equation’s harmonics. And the expansion coefficients are calculated based on the in-core detector measurements. Non-linear iteration semi-analytic nodal method combined with Krylov sub-space method is used to obtain the harmonics of neutron diffusion equation. This nodal method is 100 times faster than the fine-mesh finite differencing method combined with Krylov sub-space method for harmonics calculation. A 3D power distribution on-line monitoring system named NECP-ONION(On-line monitoring system) has been developed based on the harmonics expansion method. Measurements from a typical PWR core in China are used to validate this system. Numerical results show that the root-mean-square errors of assembly averaged powers are less than 2%. NECP-ONION has high calculation accuracy.
Experimental Study on Thermal Mixing in Emergency Core Cooling System
Ren Wuyue, Bian Jiawei, Yu Guojun, Tian Wenxi, Zhang Dalin, Su Guanghui, Qiu Suizheng
2015, 36(5): 169-172. doi: 10.13832/j.jnpe.2015.05.0169
Abstract(10) PDF(0)
Abstract:
Through the postulated loss of coolant accident(LOCA) in pressure water reactors(PWRs), the cooling water from the emergency core cooling system(ECCS) start and then inject into the cold leg. The scaling-down experiment aims to study the characteristics of thermal mixing in T-junction, including two subprojects: single phase thermal mixing test and two phase flow direct contact condensation test. The conclusion indicated that: The temperature field in the cross-section of pipes was affected by the jet flow on single phase thermal mixing test; In two phase flow direct contact condensation test, the mass of condensation after cooling goes to linear distribution with the jet flux.
Study on Molten Fuel Fragmentation Behaviors Using Moving Particle Semi-Implicit Method
Zhang Rui, Tian Wenxi, Chen Ronghua, Su Guanghui, Qiu Suizheng
2015, 36(5): 173-177. doi: 10.13832/j.jnpe.2015.05.0173
Abstract(11) PDF(0)
Abstract:
In a core-melt nuclear reactor accident, the important phenomenon, the interaction between molten fuel and residual water, is difficult to simulate numerically. In the present study, MPS(Moving Particle Semi-implicit Method) is employed to simulate the molten fuel ball impacting liquid pool and its fragmentation behaviors during falling into the water pool and in the water tank. The simulation results indicate that during the impacting process, the molten fuel ball becomes flat and Rayleigh-Tayor instability occurred and the boundary layer stripped. In the initial phase, its velocity rapidly decreased about 15%. The results showed a good agreement with the picture observed in experiments, which verified the rationality of molten fuel behavior simulated using MPS.
Analysis of Characteristics for Liquid Entrainment through ADS-4 during a Small Break LOCA in AP1000
Wang Weiwei, Liu Lifang, Meng Zhaoming, Fu Xiaoliang, Tian Wenxi, Yang Yanhua, Su Guanghui, Qiu Suizheng
2015, 36(5): 178-183. doi: 10.13832/j.jnpe.2015.05.0178
Abstract(10) PDF(0)
Abstract:
In the ADS blowdown phase during SBLOCAs in AP1000, ADS-4 valves open and vent directly to the containment. If a critical vapor velocity is reached in the two hot legs, the liquid phase can be entrained to the containment through the ADS-4 in the form of liquid droplet. In the present study, the RELAP5 code is modified with the entrainment correlations obtained from the ATLATS test facility which was constructed at OSU in USA. Furthermore, the characteristics of liquid entrainment through ADS-4 during a 5.08cm(2 inch) cold leg SBLOCAs in AP1000 was investigated. Results showed that the current liquid entrainment model in RELAP5 underestimated the liquid entrainment quantity through ADS-4 and resulted in a non-conservative safety analysis results.
Numerical Simulation of Debris Bed Formation in Severe Accident
Zhang Bin, Wu Jian, Shamsuzzaman M, Shou Tianxinglu, Dan Jianqiang, Chen Yaodong
2015, 36(5): 184-186. doi: 10.13832/j.jnpe.2015.05.0184
Abstract:
This paper presents 2-D numerical simulations using the modified discrete element method(DEM) on debris bed formation. The modified DEM facilitates the application by reorganizing the calculation parameters. A series of experiments performed by gravity driven discharge of solid particles into a quiescent water pool was used to validate the present simulation method. We made comparison of the particle dispersion angle and particle fall time in the pool, and the shape of particle bed between the experimental and simulation results. In this comparison, the general trend of simulation results indicates a reasonable agreement with the experimental observations.
Visual Experimental Investigation on Oscillations of Pressure Drop and Flow Rate of Two-Phase Flow in Mini-Channel
Chen Deqi, Wang Qinghua, Lu Qi, Pan Liangming
2015, 36(5): 187-193. doi: 10.13832/j.jnpe.2015.05.0187
Abstract(12) PDF(0)
Abstract:
The pressure drop in the narrow space channel is different from that in the regular size channel, and the bubble behavior in confined narrow space has great influence on the flow stability. A visual investigation on the gas-liquid two-phase flow was carried out in a narrow space channel with 2 mm in diameter under atmospheric pressure(0.101 MPa). Experiments were carried out with different working conditions and gas-phase working fluids, including air, carbon dioxide and argon. It is found that the pressure drop increases along with the increasing of liquid mass flux and constant gas flow rate, and the corresponding interface morphology is elongated slug flow. However, the pressure drop decreases at first and then increases along with the gas flow rate with constant liquid mass flux, and the corresponding interface morphology is still elongated slug flow.
Experiment Research on Corrosion Damage of Rubber O-Ring in High Temperature and High Pressure High CO2 Environment
Zeng Dezhi, Li Tan, Zhou Zhiru, Zhang Zhi, Shi Taihe, Cui Qingwu
2015, 36(5): 194-198. doi: 10.13832/j.jnpe.2015.05.0194
Abstract:
Taking the hydrogenated nitrile rubber O-ring as the research object, through the simulation of the working conditions of the rubber seal, corrosion tests on O-rings were conducted at pressure 25 MPa, temperature 120 ℃, test period 168 h, and gas phase composition 5 vol%CO2, 95 vol%N2. Through comparing the material structure, composition, mechanical properties and fracture morphology of O-rings before and after corrosion, the corrosion damage behavior of rubber seal in high temperature and high pressure high CO2 environment was studied. The result shows that after corrosion, the mechanical properties of O-rings decreased, and the corrosion degree in the compressive state was weaker than that in the free state, therefore, the analysis results in the compressive state should be the leading reference in a practical application. The molecular structure and fillers were damaged after corrosion, causing the performance of the rubber material decreased. The tensile fracture of the initial rubber O-ring mostly was ductile fracture, while after corrosion, it mostly was brittle fracture.
Evaluation of Thermal Fatigue Life for SSRF High-Heat-Load Components
Xiao Weiling, Chen Haibo, Yin Yan
2015, 36(5): 199-203. doi: 10.13832/j.jnpe.2015.05.0199
Abstract(11) PDF(0)
Abstract:
This paper focuses on a typical high heat load component at the front end of Shanghai Synchrotron Radiation Facility(SSRF). The temperature and elastoplastic stress-strain are simulated with the finite element method. The modified Von Mises equivalent strain model, combining with Miner linear cumulative damage theory, is used to predict the fatigue lives. Meanwhile, taking into account the effects of surface roughness and hold time, we ultimately provide a thermal fatigue life evaluation method for the high heat load components. The finite life design method is accordingly proposed, which aims to improve the current conservative design method and to promote the overall performance of the facility.
Numerical Simulation of Hydraulically Expanded Tube-to-Tube Sheet Joints in Steam Generator
Ni Peng, Hui Hu, Wang Xiaodong, Lin Shaoxuan, Zhang Liyan, Chen Qingqi
2015, 36(5): 204-207. doi: 10.13832/j.jnpe.2015.05.0204
Abstract:
For the tube sheet material and the tube material of the steam generator, according to the room temperature mechanical performance test,, the true stress-strain curve was obtained. On this basis, the numerical simulation of the hydraulically expanded tube-to-tube sheet joints in the steam generator is conducted, and the effect of the expansion pressure and length on the residual contact pressure and residual stress were studied in this paper. The results show that, the residual contact pressure increases with the increasing of the expansion pressure and has nothing to do with the expansion length. Stress corrosion is likely to occur on the inner wall of the tube transition zone.
Beryllium Localization and Verification Test of PHWR Fuel
Wang Wenli
2015, 36(5): 208-210. doi: 10.13832/j.jnpe.2015.05.0208
Abstract:
In order to manufacture the CANDU fuel bundles with domestic Beryllium completely, this paper analyzed the operation requirements for the fuel bundles coating with domestic Beryllium, and developed a special production project on the basis of the process qualification, and the bundles had been loaded in the unit 2 of the third Qinshan Nuclear Power Plant for verification test. It shows that the quality and performance of the fuel bundles coating with domestic Beryllium meets the requirements of the Qinshan CANDU units. It will be the best quality to optimize the coating process control.
Assessment of Mechanical Properties of Domestic A508-3 Steel
Yu Meifang, Luo Zhen, Zhao Yujin
2015, 36(5): 211-214. doi: 10.13832/j.jnpe.2015.05.0211
Abstract(11) PDF(0)
Abstract:
Data of domestic A508-3 steel are relatively scattered. The authors collected tensile property, Charpy V-notch impact energy, and fracture toughness data of domestic A508-3 steel from open literatures, and made comparisons with the US reactor pressure vessel(RPV) steel A533 B and Gernam RPV steel 20 Mn Mo Ni55.