核动力工程

Calculation of Higher Eigenmodes of Neutron Transport/Diffusion Forward and Adjoint Equations Using IRAM Algorithm Based on Domain Decomposition

Wu Wenbin, Luo Qi, Yu Yingrui, Li Qing, YAo Dong

2017, 38(S2): 1-6. doi: 10.13832/j.jnpe.2017.S2.0001

Abstract(19) PDF(0)

Abstract:
For the higher eigenmodes of neutron transport/diffusion forward and adjoint equations, there are various application perspectives for engineering purposes, such as core power distribution reconstruction, stability analysis, and uncertainty and sensitivity analysis. In this paper, the higher eigenmodes of both forward and adjoint equations are calculated using implicitly restarted Arnoldi method（IRAM） based on domain decomposition parallel computation and multi-group multi-domain coupled PGMRES accelerating algorithm. The IAEA3 D diffusion benchmark, a 1 D 2-group 2-domain transport problem and 2 D C5 G7 benchmark are used for verification. Numerical results demonstrate that the forward equations and the adjoint equations keep the same eigenvalue spectrum, and the obtained higher eigenfunctions are F-orthogonal or biorthogonal. These numerical results agree well with the expectations of theory analysis.

Preliminary Numerical Simulation of Fuel Assembly Deformation under Non-Homogeneous Irradiation Environment

Su Min, LI Yuanming, Chen Ping, Wang Haoyu, Pu Cengping, QI Min, Zhu Fawen

2017, 38(S2): 7-10. doi: 10.13832/j.jnpe.2017.S2.0007

Abstract(16) PDF(0)

Abstract:
Considering non-homogeneous fast neutron environment and material irradiation effects, numerical simulation of typical 17×17 fuel assembly overall deformation is preliminary realized based on assumptions and simplification. Under the inputs of large fast neutron fluence and fast neutron fluence gradient given in this paper, simulation results indicate that the fuel assembly has low stress level, but the control rod drop and refueling are affected by fuel assembly irradiation deformation.

Improvement in Neutron Transport of KYCORE Code

Tang Xiao, LI Qing, Chai Xiaoming, Tu Xiaolan, Wang Kan

2017, 38(S2): 11-15. doi: 10.13832/j.jnpe.2017.S2.0011

Abstract(22) PDF(0)

Abstract:
KYCORE is a 3 D nuclear reactor numerical calculation software and it is developed from KYLIN-2, a 2 D assembly calculation software developed in Nuclear Power Institute of China. The neutron transport part uses 2 D MOC and 1 D SN coupling to realize high precision through direct angular flux coupling and is accelerated with coarse mesh finite difference（CMFD）. Due to the simplification in the calculation procedure, the KYCORE code may detergency in calculation, thus it is improved in calculation method and mesh dividing. The salability and accuracy of the KYCORE 3 D neutron transport calculation are verified with the comparison to Monte Carlo code and C5 G7 extension case.

Numerical Research on Irradiation-Thermal-Mechanical Coupling Behavior of FCM Fuel

Tang Changbing, LI Wenjie, Chen Ping, LI Yuanming, ZhOu Yi, LI Wei

2017, 38(S2): 16-19. doi: 10.13832/j.jnpe.2017.S2.0016

Abstract(23) PDF(3)

Abstract:
User defined subroutines were coded under the calculation frame of ABAQUS software, and then, the irradiation-thermal-mechanical coupling behavior simulation method of fully ceramic micro-capsuled（FCM） fuel was preliminary established. With this simulation method, the in-pile behaviors of TRISO particle were modeled. Compared with the BISON calculation results, the validity of this simulation method was tested. Subsequently, the irradiation-thermal-mechanical coupling behavior of FCM fuel was modeled. The simulation results showed that the temperature distribution and stress distribution of FCM fuel exhibited a strong non-uniformity. The fission gas release had a strong effect on the mechanical performance of FCM fuel. The hoop stresses and temperature distribution of FCM fuel were not sensitive to transient power increasing.

Undirected Graph Partitioning Method of Matrix Algorithm in Reactor Sub-Channel Analysis

Ming Pingzhou, Pan Junjie, an Ping, Lu Wei, LIu Dong, Yu Hongxing, Sun Yufa

2017, 38(S2): 20-24. doi: 10.13832/j.jnpe.2017.S2.0020

Abstract(20) PDF(0)

Abstract:
In scale of fuel element level in the reactor core, the enthalpy conservation equation and momentum conservation equation are solved by using the structured mesh manner of according matrix algorithm in the sub-channel analysis code named CORTH. At present, parallel computing is introduced to enhance the calculation efficiency of CORTH. The coefficient matrix is with sparse symmetry which can be studied by converting to the undirected graph and using partitioning algorithms to achieve the goal of minimizing the communication volume, computation and load balance. Usually the problem is related to the non-linear polynomial（NP） algorithm. Several typical undirected graph partitioning methods and their effects are compared and discussed under sub-channel analysis scenario. Numerical experiments show that the sub-optimal solution of according matrices could be obtained by multi-level k-way undirected graph partitioning method in the application scenarios based on these algorithms. Furthermore, the computing performance of cluster is carried out for the test of pressure water reactor（PWR） full core sub-channel problem, and the parallel capability is well.

CFD Simulation of Subcooled Boiling in a Full-Size 5×5 Rod Bundle with Spacer Grids

LI Quan, Avramova M, LIu Yanghua, Zheng Meiyin, Zhao Yanli, Chen Jie, Jiao Yongjun, Yu Junchong

2017, 38(S2): 25-28. doi: 10.13832/j.jnpe.2017.S2.0025

Abstract(22) PDF(0)

Abstract:
In this paper, two-phase CFD approach is used to simulate the subcooled boiling flow in PSBT full-size rod bundle with spacer grids. The calculated surface averaged void fraction in the central four subchannels shows good agreement with the experimental data at high void fraction conditions, while higher than the measured data at low void fraction conditions. Through CFD analysis, the effect of the spacer grid on the void distribution can be obtained. The void fraction locates majorly around the central 9 hot rods upstream the mixing vane spacer grid, while more void locates in the central region of subchannels downstream the spacer grid. The numerical simulation of boiling two-phase flow can be used for the prediction of CHF and provides a new approach for the optimization of spacer grid.

Numerical Simulation Study on Resistance Characteristics of Top Nozzle in Fuel Assembly Based on CFD

Wei Zonglan, Du Sijia, Wang Xiaoyu, Wu Guanghao, Liu Songtao, Zhang Yu

2017, 38(S2): 29-33. doi: 10.13832/j.jnpe.2017.S2.0029

Abstract(22) PDF(0)

Abstract:
In this paper, numerical simulation study is conducted for the flow of coolant in the top nozzle of the fuel assembly based on CFD methodology and the general simulation settings have been proposed. Comparison between the results of measured coefficient in the experiment and CFD simulation validated the feasibility of this calculation method. Based on CFD simulation, the resistant characteristics of different components of the top nozzle had been evaluated. The analysis and evaluation showed thatt the main reason for the difference of the experiment data of the top nozzle with the reference data is the resistance coefficient value of the fuel assembly top nozzle. Recommended value is proposed.

Research in U₃Si₂-Al Fuel Swelling Characteristics

Guo Zixuan, Li Yuanming, Lu Liangliang, Su Min, Xin Yong

2017, 38(S2): 34-37. doi: 10.13832/j.jnpe.2017.S2.0034

Abstract(27) PDF(1)

Abstract:
Based on diffusion dynamics, U₃Si₂-Al interdiffusion layer growth model and fuel swelling model under irradiation condition were established. With fuel irradiation test data for China ENgineering TEst Reactor（CENTER）, the validity and applicability of the models all above were demonstrated. The results showed that the interdiffusion layer growth model agrees well with the post-irradiation measured data now available, and could be used in the calculation of U₃Si₂-Al fuel swelling characteristics; the fuel swelling model is conservative for the swelling prediction for CENTER fuel element.

Study on Flow Mixing inside Rod Bundle Based on Image Processing Algorithm

Wang Xiaoyu, Tan Sichao, Du Sijia, Zhang Yu, Wei Zonglan, LIu Yu, Deng Jian

2017, 38(S2): 38-41. doi: 10.13832/j.jnpe.2017.S2.0038

Abstract(21) PDF(0)

Abstract:
Flow characteristics of the coolant inside the nuclear fuel rod bundle is an important research topic for reactor thermal-hydraulic safety analysis. Visualization experiment results of the rod bundle with spacer grid were studied in this paper, and fluid mixing inside the rod bundle was analyzed qualitatively and quantitatively based on image processing algorithm, to provide the information for the reactor thermal-hydraulic safety analysis. Image processing algorithm was introduced at three levels, i.e., data verification, regularity summarizing and mechanism analysis, and the image processing algorithm was applied in analyzing of visualization experiment results of rod bundle. The analysis results indicate that the flow mixing mitigate 5 cm downstream the spacer grid, and Canny edge detector is applicable of spacer grid turbulence image.

Sensitivity Analysis of Key Parameters for Dynamical Analysis of Reactor Coolant System

Xiong Furui, YE Xianhui

2017, 38(S2): 42-45. doi: 10.13832/j.jnpe.2017.S2.0042

Abstract(21) PDF(1)

Abstract:
Dynamical analysis of reactor coolant system（RCS） under extreme accidents is a key technical approach for nuclear power plant safety assessment. Quantitative examination of the sensitivity of key RCS structural parameters against system dynamical responses is a crucial aspect for the reliable evaluation of RCS responses. This paper presents a sensitivity analysis of steam generator（SG） support stiffness against the RCS load distribution under seismic load by means of global sensitivity analysis and correlation analysis. It is shown that SG support stiffness is more influential to load distributions at local scale, namely, close to SG, and less influential to the load distribution of distant reactor pressure vessel（RPV）. Moreover, the input-output relationship that characterizes the mapping from key parameters to RCS load distributions is constructed and a regression model via artificial neural network（ANN） is built. The ANN model enables the fast and accurate estimation of RCS load distribution upon structural design modifications of SG supports.

An Investigation of Genetic Algorithm Based Mechanical Property Optimization Method for Nuclear Piping

Bai Xiaoming, Zheng Liangang, Wang Xinjun, Lu Xifeng, Zhang Rui, Yangjie

2017, 38(S2): 46-49. doi: 10.13832/j.jnpe.2017.S2.0046

Abstract(22) PDF(0)

Abstract:
In the nuclear steam supply system, the magnitude of nuclear piping is large and their layout are complicated. In order to keep the nuclear piping satisfing the design requirements, the optimization of the location and function of supports on the piping is significant during the design. The traditional method for optimization is manually trial calculation. The limitations of traditional method are high labor costs and experience dependency. Moreover, it is hard to obtain layout scheme with best mechanical property in the traditional method. In present work, a genetic algorithm based intelligent optimization method for piping layout is proposed. The mechanical analysis and genetic algorithm is combined in the present method, and the optimization process becomes automatic. The results show that this method can optimize the location and function of supports effectively. Comparing with traditional method, the present method is more efficiency and more practical.

Analysis of Dynamic Characteristics of Thrust Bearing for Nuclear Power Plant RCP during Loss of Power Coastdown

Deng Xiao, Wang Yan, Deng Liping, Mao Yuanfan, Zhou Ning, LIu Jia

2017, 38(S2): 50-54. doi: 10.13832/j.jnpe.2017.S2.0050

Abstract(24) PDF(1)

Abstract:
The dynamic characteristics of the thrust bearing for one nuclear power plant reactor coolant pump（RCP） during loss of power coastdown has been simulated and analyzed. Firstly the axial force of the RCP rotor has been derived. Then a mathematical model for the thrust bearing has been established based on Reynolds lubrication theory, afterwards the model is numerically solved by finite difference method and a calculation code is programed. The code is then validated through the calculation and comparison with some design operation conditions of the RCP thrust bearing. The dynamic characteristics of the thrust bearing during loss of power coastdown under hot and cold condition is simulated and calculated separately, and the calculation results suggest that when the RCP being coastdown to low speed under hot condition, the minimum oil film thickness of the thrust bearing will be less than the empirically safe margin and the maximum temperature increases sharply, which is adverse to the safe operation of the RCP. Meanwhile coastdown in cold condition is predicted to be safe.

Numerical Simulation of Thermal Striping in Pressurizer Surge Line

Cao Simin, Huang Wei, Yang Ming

2017, 38(S2): 55-59. doi: 10.13832/j.jnpe.2017.S2.0055

Abstract(18) PDF(0)

Abstract:
This paper studies the thermal striping in the pressurizer surge line to find the reason causing the phenomenon and acquire the temperature oscillation frequency and amplitude during reactor heat up processes. Numerical studies on the thermal striping were carried out by using dynamic Smagorinsky eddy viscosity model. The simulation results show that the vortex caused by contrary flew between hot and cold fluid around the interface is the reason for the thermal striping. Thermal striping in surgeline makes the temperature of the pipe inner wall oscillate from 4℃ to 20℃and the frequency is lower than 5 Hz.

Study on Dynamic Simulation of Rod Dropping Behavior under Ocean Conditions

Zhu Zihao, Peng Hang, Luo Ying, LIu Jia, Du Hua

2017, 38(S2): 60-63. doi: 10.13832/j.jnpe.2017.S2.0060

Abstract(17) PDF(0)

Abstract:
The reactor control rod drive line is an important part for the reactor safety control function. It consists of the control rod drive mechanism, the control rod assembly and its guiding system, which is the only structural equipment with relative movement in the heap. Under the ocean conditions, the working environment of the drive line of the floating nuclear power plant is more complicated for the influence of wind and sea water flow. Dropping time is an important indicator of the drive line design assessment, and the main factors that affect the time of rod dropping include: gravity, buoyancy, fluid resistance, mechanical friction and so on. In this paper, the method of theoretical simulation and software simulation is used to obtain the simulation method of control rod drive line under ocean conditions. The comparison between the simulation results and the experimental data verifies the rationality of the obtained method.

Impact-Buffer Process Analysis Method for Control Rod Drive Line Buffer Structure

Yan Dapeng, Du Hua, Liu Jia, Duan Chunhui, Zhu Zihao

2017, 38(S2): 64-69. doi: 10.13832/j.jnpe.2017.S2.0064

Abstract(24) PDF(1)

Abstract:
The paper discusses the application of continuous contact force model in the process of collision between a typical buffer structure and bearing components. The differences between compressible fluid hydraulic buffer model and incompressible fluid hydraulic buffer model are compared. Combining the continuous contact force model and two kinds of hydraulic buffer model respectively, a comprehensive calculation and analysis method was studied, and the calculation results were compared with the experimental data. The combination of continuous contact force model with compressible fluid model is more suitable for the piston hydraulic buffer structure.

Study on Sealing Behavior of Laterally One Side Restrained C-Ring under Internal Pressure by Numerical Simulation

Dong Yuanyuan, Luo Ying, YIN Qiwei, Yang Zhihai, Wang Xuxin

2017, 38(S2): 70-74. doi: 10.13832/j.jnpe.2017.S2.0070

Abstract(14) PDF(0)

Abstract:
This paper established a 3 D numerical simulation method to study the laterally one side restrained C-ring under internal pressure. In this paper, it found that with the internal pressure forcing and sealing groove squeezing, C-ring would generate additional plastic deformation, which would decrease the sealing performance of C-ring. When p＞20 MPa, the ratio of decreasing becomes bigger. In order to guarantee the sealing reliability of C-ring, the internal pressure cannot exceed 100 MPa.

Dynamic Simulation Study on Pulse-Campbell Join Section of Fission Chamber

Luo Tingfang, Zhu Hongliang, GAo Zhiyu, BAo Chao, LIu Lixin

2017, 38(S2): 75-78. doi: 10.13832/j.jnpe.2017.S2.0075

Abstract(16) PDF(0)

Abstract:
Fission chamber is with the advantages of wide measuring range and high immunity on Gama ray. But there is a difficult problem of the convergence when the pulse counting method and Campbell method are introduced in the measurement. Tests conducted based on the reactor are with high cost and limited time. This paper tries to use the computer simulation technology, and taking the pulse-campbell join section of the fission chamber as the research object, establishes the simulation model and calculates the dynamic change process about the pulse-campbell join section of the fission chamber. Through the analysis and processing of simulation data, this paper discusses the influence of the filter parameters and the correlation between the period and the measurement error.

Analysis of Control Rod Dropping and Buffering Behavior of Supercritical Water-Cooled Reactor Based on Dynamic Grid Technology

Xiao Cong, Luo Ying, Zhang Hongliang, LIu Xiao, Du Hua, HuanG Kedong, MO Chao

2017, 38(S2): 79-83. doi: 10.13832/j.jnpe.2017.S2.0079

Abstract(26) PDF(0)

Abstract:
Dynamic simulation analysis of the control rod dropping behavior of the supercritical water-cooled reactor（SCWR） is carried out based on the dynamic mesh technology of computational fluid dynamics（CFD） in this paper. And the characteristic parameters and characteristic curves of the control rod drop are obtained by calculation. The results show that the control rod dropping time can meet the requirements of safety analysis of the reactor, but impact of control rod is too large to damage the structure of control rod drive line. Based on the above results, the optimization design of the control rod drive line structure is further studied, and the hydraulic buffer structure is added, which reduces the control rod dropping terminal velocity effectively.

Device and Method for Nuclear Auxiliary Pump New Cavitation Test

Wan Yi, Chen Zhihui, LI Yi, Xiao Kunjian, Lu Xin, He Ronghui, LI Chunmei

2017, 38(S2): 84-86. doi: 10.13832/j.jnpe.2017.S2.0084

Abstract(21) PDF(0)

Abstract:
The cavitation test device and method for the nuclear auxiliary pump is designed, which is a closed type of using water ring vacuum device and valve micro degree for the cavitation test. In the process of test, there is no need to switch the water ring vacuum device and valve, only the startup and the stop of the water ring vacuum can completed the cavitation test. At the same time, the test device can also realize the flow, pump head and other comprehensive performance test, It has the characters of high degree of automation, simplification, strong versatility, and low cost.

Study on Effect of Transition Section on Safety Injection and Corresponding Mitigation Measures in Floating Nuclear Power Plants

HAo Chengming, Wan Yi, SuN Guanyu, ZhAo Jing, SuN Yan, Wang Yu, LI Chunmei, LIang Tiebo, Yan Siwei

2017, 38(S2): 87-92. doi: 10.13832/j.jnpe.2017.S2.0087

Abstract(30) PDF(0)

Abstract:
In this paper, floating nuclear power plant is studied and RELAP5 is used to analyze the effect of water-seal-structure on the safety injection when occurring the lose of coolant accident with a particular broken size, and investigate the influence pattern of water-seal-structure on the safety injection. The result shows that: the water-seal-structure is harmful to the safe injection of water into the reactor effectively which may lead the reactor to a danger state. This harmful influence can be controlled by increasing the pressure and flow rate of safety injection, which give rise to the increasing demand of system capacity. According to the design features of floating nuclear power plants, the primary passive residual heat removal system can bypass the water-seal-structure after LOCA,which can guarantee the effective injection of water and auxiliary removing the decay heat of the reactor. The operation mode is conducive to the optimization of the capacity design of safety injection.

Analysis and Improvement of Phase Differential Protection Malfunction of Main Transformer in Nuclear Power Units

Chen Jiqing

2017, 38(S2): 93-96. doi: 10.13832/j.jnpe.2017.S2.0093

Abstract(19) PDF(0)

Abstract:
This paper elaborates the principle of the phase differential protection, analyzing the failure according to an event of the phase differential protection malfunction of the transformer. It proposes the improvement in the type of current transformer and the setting, to avoid the simple faults and provide experience for the related technologist.

Reasons and Solutions for Temperature Difference between Two Sides of ADG Deaerator during Start-up

Chen Jiaqing

2017, 38(S2): 97-99. doi: 10.13832/j.jnpe.2017.S2.0097

Abstract(24) PDF(0)

Abstract:
In the hot commissioning of Fangjiashan nuclear power project, a relatively large temperature difference was found between the two ends of the deaerator in the feedwater deaerator system（ADG）. Displays in the main control room show that the temperature difference was nearly 40 degrees. After a detailed analysis, this paper found the cause for the temperature difference and put forward a variety of solutions. Through argumentation comparison, a deaerator recirculation pump was added, which successfully solved the problem in the start-up phase.

Analysis and Treatment of Weld Defects in Negative Pressure Cavity Valve of Main Pipeline in Nuclear Reactor

Shao Zhen, Cheng Xiaowen, Le Yu

2017, 38(S2): 100-103. doi: 10.13832/j.jnpe.2017.S2.0100

Abstract(19) PDF(0)

Abstract:
Negative pressure cavity valve for coolant flow rate in the primary loop main pipeline of nuclear power plants connects with the main pipeline by the way of welding. In this paper, the treatment of weld defect before a negative pressure cavity valve for 30 MW unit in Qinshan Nnuclear Power Plant has been taken as an example, comprehensive description has been made from the preliminary analysis of the causes of weld crack defects, the propose and demonstration of online casing sealing welding repair scheme, as well as technical difficulties in its implementation. Based on further analysis of the morphology and microstructure of the weld, the primary cause of the weld crack is confirmed. On this basis, technological change have been taken to solve the effect of the insufficient structure on equipment operation, and to avoid the occurrence of weld crack defects, thus ensure the safe and stable operation of the unit.

Discussion of Nuclear Hydrogen Ignition Equipment Selection

Dong Yuling

2017, 38(S2): 104-109. doi: 10.13832/j.jnpe.2017.S2.0104

Abstract(24) PDF(0)

Abstract:
In order to eliminate concentration hydrogen in the containment under the severe accidents of nuclear power plants, it is necessary to add Hydrogen Ignition Equipment. The paper analyzes the main theory of the hydrogen igniter and passive autocatalytic recombiner which are used mostly, and how to select the nuclear hydrogen ignition equipment, based on the improvement of adding the nuclear hydrogen recombiner of a nuclear plant. The improvement of the nuclear plant shows that it is very important to select the most suitable hydrogen ignition equipment for nuclear plants and society.

Influence Analysis of SVDU Fault on Nuclear Power Plant Control Model and Corresponding Technical Solutions

Du Congbo, Yuan Shangcao

2017, 38(S2): 110-114. doi: 10.13832/j.jnpe.2017.S2.0110

Abstract(32) PDF(0)

Abstract:
This paper introduces the fault analysis of FJS NPP SVDU of the unit control mode, analyzes the reasons and gives corresponding technical solutions. After the implementation of the technical program, this paper analyzes from several aspects: the elimination of 5 groups of KIC/BUP switch logic interlocking switches will not affect the safe operation of the unit. It also avoid the switch of unit control mode to BUM or even the retreat mode of the unit resulting from SVDU failure when the KIC control mode is available.

Disposal of Low-Polluting Metal Wastes in Decommissioning of Nuclear Power Plants

Fan Kai, HuanG Wentao, Wang Yongchao, JIang Shenghan

2017, 38(S2): 115-118. doi: 10.13832/j.jnpe.2017.S2.0115

Abstract(22) PDF(0)

Abstract:
Nuclear power plant decommissioning will produce large amounts of metal wastes, and the disposal of the parts with radioactive contamination should follow the appropriate classification and processing guidelines. Based on the analysis of international and domestic standards, this paper puts forward the feasible classification and treatment plan, focusing on the restricted use of low-polluting metal waste. The practical application in a nuclear power plant decommissioning project verified that it can minimize the wastes and reduce the cost of disposal.

Research on Double Stage Vibration Isolation System of Vibratory Machine for Reactor Coolant System Based on Four-Pole Parameter Method

JIang Shenghan, Zhang Kun, Fan Kai, Li Zhibin, Li Pengzhou, Wang Yongchao

2017, 38(S2): 119-123. doi: 10.13832/j.jnpe.2017.S2.0119

Abstract(29) PDF(0)

Abstract:
The four-pole parameter model of the isolation system was established to study the effect of the main parameters, such as middle mass and stiffness of the isolator, on the vibration isolation of the system by using the four-pole parameter method, and the principle which could be referenced in the design work was presented. The research shows that the middle mass should be chosen within a proper range based on the vibration isolation index requirement, and the isolators could be the same under the premise that the middle mass is lesser.

Analysis and Treatment of Shell Reduction of Moisture Separator and Reheater for Nuclear Power Stations

LI Hongjun, Wang Jian, Yang Biao, YIN Kaiju, Shi Jingfeng, Zhou Yu, Peng Yuejun

2017, 38(S2): 124-127. doi: 10.13832/j.jnpe.2017.S2.0124

Abstract(24) PDF(0)

Abstract:
After long-term operation, the thickness of the shell of the 1# Moisture Separator and Reheater in CNNO Qinshan PhaseⅠPower Plant has been reduced. The analysis reveals that the flow accelerated corrosion is the root cause. If the corrosion is allowed to continue, the equipment strength will be further reduced, resulting in great operational security risk to the nuclear power station. Analysis of the repair process analysis and repairement of the moisture separator and reheater equipment with suitable welding material, welding procedure and heat treatment process can effectively improve the operation reliability of the moisture separator and reheater, and ensure the operation safety of the nuclear power station.

Research on Intergranular Corrosion of Alloy 690 Used in SG Tubes

Lin Zhenxia, Dang Ying, Xu Qi, Chen Yong

2017, 38(S2): 128-130. doi: 10.13832/j.jnpe.2017.S2.0128

Abstract(16) PDF(0)

Abstract:
According to the standard（GB/T 15260-1994）, the experiments of intergranular corrosion of homemade and imported Alloy 690 used in SG tubes are conducted in boiling HNO₃ solutions and boiling H₂SO₄-Fe₂（SO₄）₃ solutions. The intergranular corrosion rate and microstructure of tested specimens are obtained from the experiments. It is found that the homemade Alloy 690 has excellent intergranular corrosion resistance, and the intergranular corrosion resistance of homemade Alloy 690 is generally better than that of imported Alloy 690.

Corrosion Impact of Chilled Water Leaking into RCW System and Countermeasures

Lu Yeting, YOu Zhaojin, Shen Yafang

2017, 38(S2): 131-135. doi: 10.13832/j.jnpe.2017.S2.0131

Abstract(21) PDF(1)

Abstract:
The circulating cooling water system（hereinafter referred to as RCW system） and chilled water system using inhibitor is different in CNNO NPP3. The chemical analysis showed that the corrosion inhibitor by nitrite is frozen water leaking into the hydrazine/lithium hydroxide as RCW inhibitors will make the iron in water content increased significantly. In this paper, by investigating the rising of the nitrite ion in RCW system, the impact of chilled water leaking into RCW system is evaluated, and the corrective measures to be taken are analyzed. Ccombined with the laboratory and field tests, it is concluded that although the corrosion rate of metal materials resulting from the nitrite ion in RCW system exceeded that in normal operation, but it is within the acceptable range. At this stage, the corrosion can be mitigated by natural water feedand drainage to reduce the impurity ions. In view of long term operation, it is recommended to replace the inhibitor of the chilled water at the appropriate time, thus to solve the impact of chilled water leaking into RCW system completely.

Application of NASPIC in Fangjiashan Nuclear Power Plant

Ma Bin, Wang Xi, Pan Donghui

2017, 38(S2): 136-139. doi: 10.13832/j.jnpe.2017.S2.0136

Abstract(19) PDF(0)

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Fangjiashan nuclear power plant environmental radiation and meteorological monitoring system（KRS） signals from Fangjiashan off-site emergency center building. There are security risks of network attacks, affectting the stable operation of the whole unit DCS. To solve potential risk of information security, the project design KRS signal is seperated from the current DCS network, and NASPIC is used as the preferred solution to this modification project. This paper introduces the principle of NASPIC, the main function of the equipment, network configuration, software configuration and the application in Fangjiashan nuclear power plant. The modification project has been completed and the system operates normally.

Study on Core Characteristics of Thorium-Fueled CANDU Reactor

Meng Zhiliang, Fan Shen, Wu Tianyuan, Chen Mingjun, Zhang Zhenhua

2017, 38(S2): 140-142. doi: 10.13832/j.jnpe.2017.S2.0140

Abstract(21) PDF(0)

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All the operating CANDU heavy water reactors are using natural uranium as the fuel, however, owing to the unique design features, the heavy water reactor actually is with high fuel flexibility, low-enriched uranium, anduranium and thorium to be reprocessed as well. The paper intends to study the potential impact of fuel changing from natural uranium to thorium fuel on core characteristics and core safety features. This paper uses DRAGON code to set up the infinite lattice model, and then calculate and compare the important core parameters for natural uranium and thorium-fueled heavy water reactor core. The study results show that the core features of thorium-fueled reactor are different from those of natural uranium-fueled reactor, but utilization of thorium will effectively help to improve the safety features of the heavy water reactor.

Physics Tests to Verify Performance of Vanadium Detectors of Qinshan Pressurized Heavy Water Reactor after First Replacement

Shi Xingjin

2017, 38(S2): 143-145. doi: 10.13832/j.jnpe.2017.S2.0143

Abstract(25) PDF(0)

Abstract:
With the increasing of neutron irradiation time, the sensitivity of the vanadium detectors will reduce and even lose efficacy. Therefore, the vanadium detectors need to be integral replaced after operating a certain period of time. This paper mainly discusses and analyzes the design and implementation of the physics tests to verify the performance of the vanadium detectors of the CANDU6 reactor after replacement for the first time. The results meet the requirements through the analysis and verification, lying an important foundation for the safe and stable operation of the plant in the future and providing an important reference for the similar work in the future.

Analysis and Confiring of Acceptance Criteria in Diverse Actuation System Signals Design for Nuclear Power Plants

Tian Haowen, Guan Zhonghua, Xiao Peng

2017, 38(S2): 146-148. doi: 10.13832/j.jnpe.2017.S2.0146

Abstract(25) PDF(0)

Abstract:
This paper introduces the design flow and method of diverse actuation system（DAS） signals, and discusses the process of analyzing and confirming the acceptance criteria for fixing diverse actuation system protecting signal set. The method is clarified, and the acceptance criteria is confirmed for different operation conditions. Besides, the paper illustrates a preconceived accident in Fujian Fuqing Nuclear Power Plant to verify the acceptance criteria suggested, ensuring its validation and applicability, and establishing foundations for the subsequent work.

Study on DCS Operation and Maintenance Strategy of Nuclear Power Stations

Tian Lu

2017, 38(S2): 149-151. doi: 10.13832/j.jnpe.2017.S2.0149

Abstract(17) PDF(0)

Abstract:
This paper analyzes the key impacts that DCS system brings to the nuclear power plant operations with real cases, and based on these analyses, it proposes the requirement that non-safety DCS systems could also apply Single-Failure criteria in nuclear power plants. In the meantime, this paper proposes the idea of dynamic management to the Single Point Vulnerability components respectively to different operation conditions and plant status of nuclear power plants. In addition, it discusses the importance of the DCS online maintenance capacity. The studies in this paper is helpful for the DCS operation, maintenance and management of a nuclear power plant.

Study on Forming Technology of Anti-Water Erosion Shields Made of Stellite Alloys

Wang Miaomiao, Zhang Xingtian, Wang Jun, Geng Jianqiao, Yu Weiwei, Chen Mingya

2017, 38(S2): 152-154. doi: 10.13832/j.jnpe.2017.S2.0152

Abstract(19) PDF(0)

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The molding process, microstructure and structural mechanics of the Stellite sheet are discussed, and the potential cause of cracking is shown in this paper. The surface stress of the Stellite alloys is changed after the mechanical polish, and the size of the grain in the surface and internal of the Stellite alloys is different, which will affect the mechanical properties of the material.

WCAP Methodology Related to Chemical Effect of Containment Sump Strainer in PWR

Tang Ming, Wang Tao, Xia Xiaojiao, Jia Xiaohong, Xing Dianchuan, LI Yong

2017, 38(S2): 155-159. doi: 10.13832/j.jnpe.2017.S2.0155

Abstract(33) PDF(0)

Abstract:
The block of containment sump strainer is one of the primary safety concerns. A regulatory guide of RG1.82 WATER SOURCES FOR LONG-TERM RECIRCULATION COOLING FOLLOWING A LOSS-OF-COOLANT ACCIDENT was published by US NRC. The chemical element in generated debris will be dissolved in the containment spray and sump fluid after LOCA, and those chemical elements will precipitate following the decreasing of temperature of the sump fluid. The precipitates will deposite on the debris bed on the sump strainer surface to worsen the performance of sump strainer’s pressure loss. These phenomena called chemical effect. An analysis methodology related to chemical effect called WCAP is introduced in this paper.

Development of Detection Device Pressure Vessel for Control Rod Position Sensor in Nuclear Power Plants

Wang Yongchao, Fan Kai, JIang Shenghan, Zhang Bin, CAo Qi, LI Yun, Feng Weiwei

2017, 38(S2): 160-163. doi: 10.13832/j.jnpe.2017.S2.0160

Abstract(19) PDF(0)

Abstract:
The pressure vessel design scheme was completed based on the structure features and function requirements for the step-type control rod position sensor of Tianwan Nuclear Power Plant. The pressure vessel was mainly formed by the upper head, the cylinder body and the lower head. According to standards and working conditions, the strength calculation and the opening reinforcement of the vessel body device for cylinder body, the upper and lower head was completed. At the same time, the strength calculation of the non-standard flange supports was completed, and then the verification of the vessel was completed based on the results of strength calculation, to ensure that the vessel can form a pressure boundary and seal the working medium to achieve the purpose of maintaining the working pressure and temperature.

Analysis of Effect of Low Pressure Critical Heat Flux

Wu Xiaohong, Zhao Erlei, Zan Yuanfeng, LI Pengzhou, Zhuo Wenbin

2017, 38(S2): 164-166. doi: 10.13832/j.jnpe.2017.S2.0164

Abstract(22) PDF(0)

Abstract:
This paper mainly presents the test section, measure of parameters, technique research of low pressure CHF（critical heart flux） experiment and analysis of experimental result. In the experimental study, experimental methods and key experimental technique are explored by a uniformly heated vertical tube in low pressure. A number of different coordinate systems are used to present the experimental results, and coordinate the empirical correlation for predicting CHF.

Study on Fracture Failure Characteristics of Instrument Casing in Nuclear Power Plants

Yan Liang, LI Yang, Song Mingliang

2017, 38(S2): 167-171. doi: 10.13832/j.jnpe.2017.S2.0167

Abstract(19) PDF(0)

Abstract:
In this paper, the analysis on the casing and cladding metal were carried out by macro and micro methods, including the chemical composition analysis, microstructure analysis, fracture metallography and scanning electron microscopy（SEM）. The results showed that the microstructure and hardness of the casing and cladding metal were not uniform. Combined with theoretical analysis, it was concluded that the fracture property of retaining tube was fatigue fracture.The cause for fracture was welding residual stress, which results in the crack initiation under the vibration of fluid. It was suggested that the welding of the retaining sleeve should be avoided.

Effect of Back Pressure on Pressurizer Safety Valve

Zhang Donglin, Yuan Shengyi, Wang Yuxiang, Guo Song, Yang Yong, Luo Shihong

2017, 38(S2): 172-174. doi: 10.13832/j.jnpe.2017.S2.0172

Abstract(32) PDF(0)

Abstract:
The effect of back pressure on opening and closing pressure, capacity and tightness of safety valve is analyzed in this paper. According to the analysis and computing, the effect is slight when the back pressure fluctuates. The reasonable design of pilot valve disk and bellow’s sealing areas can reduce the effect of back pressure on the safety valve.

Research on Fatigue Monitoring System Development for Nuclear Power Plants

Zhao Chuanli, Chen Yinqiang, Xu Feng, He Ziang, JI Yuanyuan

2017, 38(S2): 175-178. doi: 10.13832/j.jnpe.2017.S2.0175

Abstract(21) PDF(0)

Abstract:
The difference of the international fatigue monitoring systems were compared. As the key technologies of fatigue monitoring system development, the processing of operational load data and the stress influencing function development at monitored locations were researched in the paper. Finally, the on-line fatigue monitoring system for pipelines was established.

Reducing Low Pressure Injection Pump Vibration with Orifice in VVER

Zhou Zhijun, Zhang Shuai, He Ziang

2017, 38(S2): 179-181. doi: 10.13832/j.jnpe.2017.S2.0179

Abstract(21) PDF(0)

Abstract:
This paper analyzes the free end vibration of low-pressure injection pump, during the commissioning test on unit 3 of Tianwan Nuclear Power Plant.Through the quantity of orifice plates and optimization of aperture calculation, the pump and pipeline vibration problems are solved. It provides important empirical feedback and reference to the follow-up mechanical calculation for nuclear power unit pipeline.

Application of Nonlinear Iteration Method in Reactor Real-Time Simulation

Duan Xinhui, Wang Bingshu, JIang Ping

2017, 38(S2): 182-187. doi: 10.13832/j.jnpe.2017.S2.0182

Abstract(25) PDF(0)

Abstract:
It is difficult to satisfy both high precision and speed of the 3 D distribution calculation of reactor power, during the development of full scope simulator for nuclear power plants. A 3 D real-time simulation code is developed for reactor core based on the Nonlinear Iteration Method（NIM）. The coupled correction factor is calculated with long step time using Nodal Expansion Method（NEM） by its high accuracy. The flux and power are calculated in real-time using Coarse Mesh Finite Difference Method（CMFD） by its rapid speed. In contrast with the calculation of typical benchmark examples, the NIM simulation results showed that the calculation of reactor core using corrected coupling factor revised CMFD can not only keep the precision, but also can improve the calculation speed. NIM is a good 3 D real-time iteration strategy satisfied with the full scope simulator in nuclear power plants.

2017 Vol. 38, No. S2