核动力工程

Analysis and Design of Rod Position Detector Power Supply Control Method Based on Hysteresis Control in Nuclear Power Plants

Xu Mingzhou, He Jiaji, Zheng Gao, Li Guoyong

2018, 39(S1): 1-4. doi: 10.13832/j.jnpe.2018.S1.0001

Abstract(19) PDF(0)

Abstract:
The main function of the rod position detector power supply in the nuclear power plant is to provide the sine wave current for the detector primary windings. The output accuracy and load variation response have effects on the rod position measurement accuracy. In order to improve the above performance of the detector power supply, this paper introduces the current hysteresis control for its characteristics of fast load tracking and response, and optimizes the traditional hysteresis control to make the switch frequency keep constant. However, not only does the control method have relationship with the switch frequency, but also the device parameters. This paper also analyzes the inductive effects of parameters on the switch frequency spectrum. Lastly, simulation is carried out to verify the theory analysis. The result of simulation verification shows that the improved control method overcomes the defect of the traditional hysteresis control, and improves the electrical performance of the detector power supply.

Automatic Contrast Verification of DRAGON and Core Neutronics Software Developed on Self-Reliance

Ming Pingzhou, Li Zhigang, An Ping, Xia Bangyang, Lu Wei, Liu Dong, Yu Hongxing

2018, 39(S1): 5-9. doi: 10.13832/j.jnpe.2018.S1.0005

Abstract(15) PDF(0)

Abstract:
The development of the software CORCA-3 D and CORCA-K developed on self-reliance involves the correctness verification of core diffusion numerical calculation. In order to improve the efficiency of R&D, based on the core module named TRIVAC in DRAGON, the automatic contrast verification scheme of DRAGON and the core neutronics software developed on self-reliance is established for benchmark scenes of the square fuel assembly, which assists the development of the software. This scheme could be customized according to the content of calculation, and multiple benchmark diffusion problems are checked. The results of numerical experiments ensure the correctness of the core modules and serve as a basic test procedure to reduce the burden of software developers.

Analysis and Research of Neutron Spectrum in Reactor Shield Calculation

Tian Chao, Ying Dongchuan, Zhang Hongyue, Tang Songqian, Tan Yi

2018, 39(S1): 10-14. doi: 10.13832/j.jnpe.2018.S1.0010

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Abstract:
The monte-carlo neutron-photon coupled transport program（MCNP） is used in reactor shielding design to calculate the neutron flux rate of reactor pressure vessel and reactor components to evaluate the neutron irradiation damage to structural materials.In the calculation of such fixed source problems, the energy distribution of the source is mostly based on the Maxwell fission neutron energy spectrum or the Watt fission neutron energy spectrum of MCNP, which are the vector fission energy spectrum corresponding to the incident neutron of typical energy.However,the real fission neutron energy spectrum is the matrix fission neutron energy spectrum related to the incident neutron energy. Therefore, the influence of different neutron energy spectrum on the reactor shielding design is analyzed. The results show that the influence of incident neutrons of different energies on the fission neutron energy spectrum should be considered in reactor shielding design.

Analysis on Fluid Induced Vibration in Stop Valve of Piping System

Liu Lizhi, Chen Jiu, Cai Longqi, Tan Shuyang, Zhao Xuecen

2018, 39(S1): 15-19. doi: 10.13832/j.jnpe.2018.S1.0015

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Abstract:
Fluid induced vibration of the angle stop valve was obvious which brings some difficulties for controlling the vibration and the noise in the cooling water system of some nuclear devices. For reducing the vibration, the fluid induced vibration characteristics of both the angle stop valve and oblique stop valve are calculated and compared. Results show that for both valves, the pressure fluctuation after the valves are larger than that before the valves. However, the intensity of vortex, the force of valve core and the amplitude of pressure fluctuation for the oblique stop valve are less than that of the angel stop valve, so the fluid induced vibration characteristics of the oblique stop valve is better than that of the angle stop valve.

Research on Methods of Solving Microscopic Depletion Equation and Numerical Verification of Program

Guo Fengchen, Chai Xiaoming, Lu Wei, Ma Yongqiang, Tu Xiaolan

2018, 39(S1): 20-23. doi: 10.13832/j.jnpe.2018.S1.0020

Abstract(9) PDF(0)

Abstract:
The sovlving methods of depletion equation is researched, including Taylor method,Pade method, Scaling and squaring method（Scale）, Eigenvectors method（Eig）, Chebyshev rational approximation method（Cram）, Lagrange interpolation method, Newton interpolation method,Vandermonde method and Krylov subspace method. Based on the comparison and analysis of accuracy and efficiency, Cram is chosen as the candidate for the final depletion method. The program of solving depletion equation is developed and validated with burnup benchmark. Results show that the program developed can provide accurate solution.

Application of Diagonally Implicit Runge Kutta with Exponential Transformation on Point Kinetic Equations

Cai Yun, Zhang Zhizhu, Li Qing, Wang Shuai

2018, 39(S1): 24-27. doi: 10.13832/j.jnpe.2018.S1.0024

Abstract(9) PDF(0)

Abstract:
The point kinetics is very important to the safety of the reactor operation. However,these equations are stiff, with very small time step for solution. The diagonally implicit Runge Kutta（DIRK）with exponential transformation is studied for point kinetics. It keeps the advantage of diagonally implicit Runge Kutta method which is suitable for stiff equations and owns the better performance than the diagonally implicit Runge Kutta method when the positive reactivity is inserted. Several cases, such as step, ramp, and sinusoidal reactivity insertion are used to show that this method owns high accuracy.

Method Study on Static Buckling Load Analysis of Spacer Grid

Qin Mian, Pu Cengping, Chen Ping, Li Yuanming, Ru Jun, Lei Tao

2018, 39(S1): 28-33. doi: 10.13832/j.jnpe.2018.S1.0028

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Abstract:
Static buckling load is an important concept for assessing the structural performance of the spacer grid. The finite element analysis（FEA） of static critical buckling load for 3×3 spacer grid has been investigated with analyzing the influence of grid structural characteristics. Results show that the fuel rod retaining state can be accurately simulated by the suitable setting of welding spot and contact. The fuel rod loading state, initial retaining force and friction have great effects on the results of this method.

Finite Element Analysis of Interference Connection for ACP100 Reactor Vessel Internals

Liu Xiao, Wang Liubing, Zhang Hongliang, Luo Ying, Rao Qiqi, Wu Bingjie

2018, 39(S1): 34-36. doi: 10.13832/j.jnpe.2018.S1.0034

Abstract(13) PDF(0)

Abstract:
The ANSYS was employed in the interference connection design for the reactor vessel internals of ACP100. Compared with theoretical formula and the results of finite element method（FEM）, the FEM for the design of interference connection was acquired, and the FEM was used to study the stress condition of the lower reactor vessel internals. With the FEM, the reliability could be ensured in the interference connection design for reactor vessel internals.

Study on Logic of Reactor Coolant Pump Interlock in HPR1000 Nuclear Power Unit

Xu Tao, Zhu Jialiang, He Peng, Chen Xuekun, Du Mao, Chen Jing, Xu Sijie, Wang Xuemei

2018, 39(S1): 37-40. doi: 10.13832/j.jnpe.2018.S1.0037

Abstract(12) PDF(0)

Abstract:
Reactor coolant pump（RCP） is one of the key devices in the reactor coolant system in nuclear power Unit. The safety and stability of the RCP concerns the safety of the nuclear power plant. HPR1000 is the third generation nuclear power plant developed by China on self-reliance,with severe requirements for the safety. In this paper, we described the interlock logic of HPR1000（FUQING NPP UNITS 5&6 PROJECT）. Combining with the experiences from the previous nuclear power plants, we analyze and demonstrate whether the interlock logic satisfies the safety requirements in HPR1000. It is of great importance in the assimilation of the advanced design theories in other countries, thus to accumulate the experiences and achieve the domestic production of the reactor coolant pump as soon as possible.

Study on a CFD Simplified Approach for Steam Generator U-Tube

Xin Sufang, Li Songyu, Ren Chunming, Wang Wei, Xiao Peng

2018, 39(S1): 41-44. doi: 10.13832/j.jnpe.2018.S1.0041

Abstract(17) PDF(0)

Abstract:
It is necessary to use the computed fluid dynamic（CFD） method to study the problems considering the U-tube distribution, such as the reverse flow in the steam generator. But it is difficult to model all the U-tubes in a steam generator because of the huge number of the U-tubes and the limited capability of computers. This paper discussed a simplified modeling approach for single-phase fluid flow in U-tubes, using a square tube instead of a round tube. The result of a reverse flow calculation for a steam generator using the simplified modeling approach is basically the same as the result using a non-simplified model, which indicates that the approach can be used in CFD analysis of this kind of problems.

Structure Design and Analysis of Cable Shearing Tool for Temperature Measuring Element in Pulsed Reactors

Huang Xindong, Chen Shuhua, Wang Bingyan, Huang Hui, Ren He, An Yanbo

2018, 39(S1): 45-48. doi: 10.13832/j.jnpe.2018.S1.0045

Abstract(17) PDF(0)

Abstract:
This paper introduces the design requirement and characteristics of the cable shearing tool for the temperature-measuring element in pulsed reactors. Accrording to the characteristics of the cable shearing and the man-machine engineering theory, a shearing structure scheme is proposed, in which the shear motion is combined with up/down motion and the rotation,and the scheme of the force-amplification structure is the centric slider-crank connecting rod mechanism combining with lever mechanism. Then mathematics model is built and the computation of using interpolation for typical working conditions is finished based on the model. The computation shows that the scheme is reasonable and feasible, with the advantages such as small space, big mu-factor, and easy and safe operation.

Discussion of Time of Ultrasonic Examination for Low Alloy Steel Forgings of Reactor Pressure Vessels

Yin Qiwei, Luo Ying, Qiu Tian, Wang Xiaobin, Yang Zhihai

2018, 39(S1): 49-52. doi: 10.13832/j.jnpe.2018.S1.0049

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Abstract:
To determine the most appropriate stage of ultrasonic test for reactor pressure vessels（RPV） steel forgings, this paper confirmed that RCC-M requires the ultrasonic test be done after final finishing, and analyzed the rationality of the requirement based on the ultrasonic examination mechanism, referring to other norms. It is proposed that a more appropriate stage of ultrasonic examination would be after the heat treatment for mechanical properties and before final finishing.

Numerical Simulation Analysis of Main Steam Relief Valve Opening Characteristics by Dynamic Mesh

Yu Deyong, He Xun, Yu Xiaoquan, Li Cong, Zhang Zhenhua

2018, 39(S1): 53-57. doi: 10.13832/j.jnpe.2018.S1.0053

Abstract(19) PDF(0)

Abstract:
The study was performed based on numerical CFD simulation by means of dynamic meshing. A CFD model was established to reproduce the valve opening behavior, which means that a dynamic valve disc model has been built using the trimmed volume mesh and the deformable mesh coupled with 1 D freedom model of the valve disc. The simulation shows that the total opening time of the pilot-valve is principally affected by the steam discharging time, the disc motion time of about 0.1 s. The disc movement characteristics was similar at different pressure conditions, and the higher the pressure, the longer the disc opening time, but the shorter the disc motion time.

Simulation and Analysis of Reactor Coolant System for Two-Loop Nuclear Power Plant

Zeng Chang, Zhao Yu, Ye Zhu, Ren Yun

2018, 39(S1): 58-61. doi: 10.13832/j.jnpe.2018.S1.0058

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Abstract:
The reactor coolant system simulation model for the two-loop nuclear power plant is established using the thermal hydraulic analysis software Flowmaster. The power operation steady condition, the bias-loop operation steady condition during startup/shutdown reactor, the pump coastdown operation transient condition after loss of off-site power are simulated. The results show that the simulation data matches well with the design values and actual operation values, and the deviations are less than 4%. The simulation model provides a reference for the design optimization and operation of the subsequent reactors of the same type.

Mechanical Property Study of CF3 Fuel Assembly Bottom Nozzle

Su Min, Chen Ping, Kuang Linyuan, Li Qi, Lei Tao, Feng Linna, Zheng Meiyin

2018, 39(S1): 62-65. doi: 10.13832/j.jnpe.2018.S1.0062

Abstract(10) PDF(0)

Abstract:
Bottom nozzle, used to support and locate, is an important structural component of CF3 fuel assembly. FEM and load test method were adopted to study the mechanical properties of CF3 fuel assembly bottom nozzle. The result shows that the stress on every condition meets the requirement of the ASME code and the carrying capacity meets the design requirement of CF3 fuel assembly.

Research on Containment Pressure Suppression and Heat Transfer of Small PWR

Jiang Xiaoyu, Deng Jian, Yu Hongxing, Li Zhe, Shen Yaou

2018, 39(S1): 66-69. doi: 10.13832/j.jnpe.2018.S1.0066

Abstract(16) PDF(0)

Abstract:
Small PWR has adopted the suppression tank and the steel containment with outside pool to mitigate the containment pressure and temperature increasing after a relevant accident. The features of suppression and heat transfer are essentially different from those of a conventional PWR.This paper studied the features of the pressure suppression through the suppression tank and the heat transfer through the steel containment, respectively. After that, the requirement of the suppression tank and the heat transfer area was studied. Analysis results demonstrated that the pressure increasing can be suppressed effectively by the suppression tank, while the requirement of the heat removal can be fully satisfied if the heat transfer area increases to 350 m².

Design and Development of Multi-unit Simulator Management System in Tianwan Nuclear Power Station

Zhang Na, Luo Hongchun, Zhao Xin, Wang Jiachang, Xiao Anhong, Zeng Hui

2018, 39(S1): 70-74. doi: 10.13832/j.jnpe.2018.S1.0070

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Abstract:
The full-range simulator in Tianwan nuclear power plant is mainly used for the training and authorization of the operators in the master control room. Since it is an important tool and means for operator training, its consistency with the main unit and its availability shall be guaranteed. In order to that the simulator management system meets the requirement of the standards in the national energy industry, and the requirement of digitalized comprehensive maintenance and management, Nuclear Power Institute of China designed and integrated the multi-unit simulator management system. The system ensures that the simulator’s maintenance and management satisfies the industrial standards, and the system enhances the simulator management in nuclear power station. Currently, the system has been successfully applied in the simulators in units 1/2 of Tianwan Nuclear Power Station. It can be applied directly in the simulators in units 3/4 and units 5/6 in the future. At the same time, the system can applied to other NPPs in service and to be built by a small amount of customized development.

Application of Direct Temperature Detection Technology for Primary Pipe in HPR 1000

Zhu Jialiang, He Zhengxi, Xu Tao, Du Mao, Chen Jing, Li Xiaofen, Chen Xuekun

2018, 39(S1): 75-78. doi: 10.13832/j.jnpe.2018.S1.0075

Abstract(12) PDF(0)

Abstract:
Traditional nuclear power plant（M310） adopts the bypass measurement method to measure the key safety parameter, i.e., the reactor coolant temperature, but this method cannot satisfy the requirement of the 3 rd generation nuclear power for the complicated process loop, large amount of connection on the main pipe and difficulty in the maintenance. This paper analyzes this physics phenomenon at first, and extracts a direct temperature detection scheme for the primary pipe which is suitable for HPR 1000 nuclear power plant, then verifies this scheme from the aspect of safety analysis. The result of the safety analysis verification indicates that the direct temperature detection techonoly for the primary pipe can be applied on HPR 1000 nuclear power plant with some special configuration.

Research on Conceptual Design of Sodium Cooled Standing Wave Reactor

Zheng Meiyin, Chen Ping, Zhang Dalin, Tian Wenxi, Su Guanghui, Qiu Suizheng

2018, 39(S1): 79-83. doi: 10.13832/j.jnpe.2018.S1.0079

Abstract(13) PDF(0)

Abstract:
The code developed on self-reliance was used to perform the conceptual design of the sodium cooled standing wave reactor. To reduce the power peak of the inner core, the radial partition design was used. To control the radial power distribution, the outward fuel shuffling strategy was adopted. MCNP-ORIGEN coupled code MCORE was applied to perform the core physical analysis. The results show that the core and assembly in different positions reach equilibrium state after 30 fuel shuffling steps; the reactivity and multiplication coefficient only fluctuate with the fuel shuffling, and the power and the neutron fluence, shaped like M, mainly concentrate in the intermediate region of the inner core, with low power peak; the nuclide density of ²³⁹Pu gradually increases along the radial direction of the inner core from inside to outside, and the nuclide density distribution of fission products is opposite with that of ²³⁸U; the equilibrium cycle discharged burn-up and maximum discharged burn-up of the core are 27.6% and 29.3%,respectively.

Review of Critical Heat Flux Prediction Methods in Pressurized Water Reactors

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu, Dan Jianqiang

2018, 39(S1): 84-87. doi: 10.13832/j.jnpe.2018.S1.0084

Abstract(16) PDF(2)

Abstract:
In order to accurately predict the critical heat flux（CHF） in pressurized water reactor core fuel assembly, the effects of spacer grids, cold walls and non-uniform heat flux in rod bundles on CHF are analyzed. Simultaneously, six CHF mechanism models based on different physical hypothesis are comparatively researched, and the development methods for CHF correlations are put forward. Advantages and disadvantages of CHF mechanism models and empirical correlations in the prediction of bundle CHF are obtained. It is suggested that the expanding of existing CHF experimental database, the optimizing of development methods of CHF correlations, and the developing of rod bundle CHF mechanism models shall be carried out.

Study on Mechanical Calculation and Analysis of One Ship-Type FNPP

Li Song, Tang Huapeng, Xu Yu, Chen Zhi

2018, 39(S1): 88-93. doi: 10.13832/j.jnpe.2018.S1.0088

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Abstract:
In this paper, one ship-type Floating Nuclear Power Plant（FNPP）, which is fixed by multi-point anchor chain, was studied. The loads applied on FNPP was calculated and analyzed under various kinds of external forces, such as wind, wave, current and ice, and under the sea conditions encountered in one hundred year and two hundred year. The results show that the chain does not break under the worst sea condition. The maximum tension and the maximum deviation of the chains meet the requirement, and the safety factor is within the maximum limit. Analysis results show that the multi-point anchoring can satisfy the operational requirements of of a floating nuclear power plant.

Analysis and Improvement of Small Branch Pipe Leakage of Reactor Coolant in Nuclear Power Plant

Yao Chong

2018, 39(S1): 94-97. doi: 10.13832/j.jnpe.2018.S1.0094

Abstract(21) PDF(0)

Abstract:
The leakage of the pressure boundary of the nuclear power plant is an important operation event. The response and treatment after the occurrence of abnormal events is focus on the safety management of the nuclear power plant. Based on the treatment of the leak point on the extension branch pipe of the pressure boundary of 300 MW Qinshan Nuclear Power Plant,combined with the evolution of the process parameters of the unit, the ideas and principles to find the leak point are analyzed, and the response and treatment of the similar leakage in nuclear power plants are analyzed. At the same time, according to the method of root cause analysis, combined with the requirements of the American Society of Mechanical Engineers（ASME） specifications,this paper analyzes the scheme, basis and main points of solving this problem, and puts forward the subsequent improvement suggestions for the design and the daily in-service inspection. Especially the method to effectively manage the small branch pipe is proposed to ensure the safety of the nuclear power plant. It can provide a good reference for the management of pressure boundary for small branch pipes in domestic nuclear power plants.

New Requirements of Defense in Depth and Extension of Its Application

Wang Chengcheng, Wu Yuxiang

2018, 39(S1): 98-101. doi: 10.13832/j.jnpe.2018.S1.0098

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Abstract:
The applications of the defense in depth in the design of the irradiated fuel water pool storage, I&C system, electrical system, and the protection systems for internal and external hazards of the nuclear power plant are reviewed in this paper. Base on the study of the newest standard and codes, the design requirements of each level of defense in depth in the above systems are confirmed. The analysis of the design features for the defense in depth in an advanced nuclear power plant in China shows the application value of defense in depth in various fields.

Optimization Study of Improving Cavity Injection and Cooling System Performance

Zhao Jiaming, Wang Guangfei, Zhu Dahuan, Zhao Bin

2018, 39(S1): 102-105. doi: 10.13832/j.jnpe.2018.S1.0102

Abstract(10) PDF(0)

Abstract:
The optimization study is carried out for the disadvantage of cavity injection and cooling system（CIS） in HUALONG 1 plant, when dealing with serious accidents, and the new system configuration and operation of CIS is presented, including adopting control valves to control flow of different conditions, designing new cavity injection pumps, using Flowmaster software to model and calculate the system, to fix the main equipment parameters and verify the new system CIS satisfying the requirement of flow of serious accidents. Finally the problem of adopting low voltage SBO（station black-out） power for cavity injection pumps is solved, the rate of utilization of passive injection is increased, the CIS ability of dealing with various serious accidents is enhanced,to lay a solid foundation for improving cavity injection and cooling system in subsequent generationⅢ pressurized water reactor, which has great significance for the reliable operation of nuclear power plants.

Assumption of Severe Accident Management for Chinese Third Generation Nuclear Power Units

Sha Pingchuan, Kuang Huiwen, Yang Yun

2018, 39(S1): 106-108. doi: 10.13832/j.jnpe.2018.S1.0106

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Abstract:
After the accident in Fukushima nuclear power plant, it is more often to adopt a combination of active and passive accident mitigation systems in Chinese 3 rd generation nuclear power units, compared with the 2 nd generation plus nuclear power unit（M310）. Chinese 3 rd generation nuclear power units adopt a number of optimization design and improvement are adopted.The core damage frequency（CDF） and large-scale radioactivity release frequency（LRF） of the Chinese 3 rd generation nuclear power unit is lower than that of the M310. What’s more, the accident mitigation capacity of Chinese 3 rd generation unit is greater in dealing with severe accidents. For the architecture of severe accident management guideline（SAMG） in the 3 rd generation units, we can learn from the WOG（Westinghouse Owner Group）’s SAMG systems, which are used in 2nd-plus improved units（M310）. After SAMG’s structures have been optimized, the severe accident management of the 3 rd generation units will be improved greatly both in hardware and software.

Research on Standard System of Third Generation Nuclear Power Technology HPR1000

Zhang He, Wu Yuxiang

2018, 39(S1): 109-111. doi: 10.13832/j.jnpe.2018.S1.0109

Abstract(10) PDF(0)

Abstract:
This paper mainly studies the standard system of the third generation nuclear power technology HPR1000, and emphasizes on the top level design of the standard system. Firstly the paper summarizes the current standard situation adopted in HPR1000, specifies the standard system study area, and primarily establishes the framework construction of standard system, and then puts forward a systematic analysis method of HPR1000, and explains the analysis process with the general design as an example, so as to provide necessary study bases on the top level design of the standard system and provides suggestion to further find out the relationship among HPR1000 standards.

Study on Control and Command Recovery in Support of Nuclear Power Plant during Extreme External Hazards

Yu Yun, Yu Xinli

2018, 39(S1): 112-114. doi: 10.13832/j.jnpe.2018.S1.0112

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Abstract:
An effective emergency response system is the basis of accident management in case accident. The accident of Fukushima Daiichi Nuclear Power Plant suggests that extreme external hazards may affect the normal emergency system of the nuclear power plant, which results in the difficulty of accident mitigation. In hence, based on the research of the emergency system in the nuclear power plant, a recovery method of command and control is proposed aiming at extreme external hazards. In addition, multi-sites issue is concerned in the recovery analysis. The product of this paper can be a beneficial reference for NPPs to deal with the extreme external hazards.

Performance of Plate Heat Exchanger for Component Cooling Water System in HPR1000

Yu Pei, Fu Haoran

2018, 39(S1): 115-118. doi: 10.13832/j.jnpe.2018.S1.0115

Abstract(8) PDF(0)

Abstract:
The four methods are used to simulate the plate heat exchanger in Units 1 and 2 in Fuqing Nuclear Power Plant, i.e., the criterion number heat exchange formula, HTRI heat transfer factor empirical formula, Matin heat transfer factor empirical formula and A. Muley heat transfer factor empirical formula. The minimum difference of heat exchange area from the design value is calculated by using the criterion number heat exchange formula. This method is used to validated the plate heat exchanger of the component cooling water system in HPR1000 nuclear power plant,and the result shows that the heat exchanger supplied by the factory comply with the engineering requirement.

Calculation of Mixed Nuclear Piping Model with Elbow

Ning Qingkun, Bai Xujuan

2018, 39(S1): 119-121. doi: 10.13832/j.jnpe.2018.S1.0119

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PIPESTRESS program is used to analyze and evaluate the nuclear safety class 1 and class 2 piping model with elbows, and the influence of two calculation methods of the separate calculation and the restart calculation is compared. The research shows that there is a deviation in the result of the separate calculation. In order to get the result that meets the standard requirements,the restart calculation should be adopted. This paper can provide a reference for the piping calculation of generation Ⅲ and Ⅲ nuclear power plants.

Study on Planeness Deviation of Bottom Steel Liner in Nuclear Power Plant Containment during Construction

Wang Di, Xiong Meng, Wu Shaowei

2018, 39(S1): 122-124. doi: 10.13832/j.jnpe.2018.S1.0122

Abstract(13) PDF(0)

Abstract:
In this paper, the method of planeness deviation analysis in the construction stage of the bottom steel liner is adopted to calculate the bottom steel liner of different sizes, the limit value of the planeness deviation of the bottom steel liner of different sizes is obtained. Calculation results are compared and the analysis shows that the planeness deviation in the bottom steel liner is related to the size of the block. Different planeness deviations can be used according to different block schemes.

Study on Design Method of Polar Crane Brackets with Their Embedded Parts under Complex Loads in Reactor Containment

Bai Yunxiu, Xiong Meng, Wang Di, Xiao Qiongguan

2018, 39(S1): 125-127. doi: 10.13832/j.jnpe.2018.S1.0125

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In this paper, the design of a polar crane bracket is optimized using EXCEL and ANSYS. Based on the so-called fiber model, ABAQUS and MATLAB were used to simulate the mechanical behavior of the embedded parts under the combined action of axial force and bending moment. This design method can quickly respond to engineering needs and save manpower and time costs, leading to a substantial increase in computing efficiency and accuracy.

Study on Integral Assembly Transportation Scheme of AP1000 Nuclear Island Dome

Wang Haining, Zhang Shang

2018, 39(S1): 128-132. doi: 10.13832/j.jnpe.2018.S1.0128

Abstract(13) PDF(0)

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At present, the domestic AP1000 nuclear island dome is assembled in the surrounding area of the nuclear island, and it is restricted by the site and traffic constraints, and the program is not universal. According to the needs of the whole assembly, based on the dome transport bracket, road transport vehicles and the analysis of the feasibility of the scheme, the optimization of the integration of the dome assembly site and module assembly site is proposed, to save space and cost and shorten the construction period while uphold the AP1000 modular construction concept at the same time, and to provide new ideas for the follow-up of AP1000 nuclear island dome assembly.

Suggestions on Optimization Method of Emergency Action Level of Advanced Passive Light Reactor

Zang Xiaochuan, Liu Tao, Tong Jiejuan

2018, 39(S1): 133-137. doi: 10.13832/j.jnpe.2018.S1.0133

Abstract(21) PDF(0)

Abstract:
Emergency Action Level（EAL） is one important tool to determine the emergency classification level in nuclear emergency response. Because of its design characteristics, the risk spectrum characteristics of a large advanced passive light water reactor are different with that of the nuclear power plants currently in operation, and its EALs have been improved. This paper proposes a risk-informed assessment method, using the probabilistic safety assessment（PSA） model, to establish an approximate relationship between the EAL and the conditional core damage probability（CCDP）. Based on the range of emergency classification by CCDP, the EAL mismatched with the emergency classification level can be screened to achieve the adjustment and optimization.

2018 Vol. 39, No. S1