核动力工程

Study on Burnup Characteristics of Actinide Burnable Poisons in Plate-Type Fuel Assembly

Yu Tao, Liu Jinju, Xie Jinsen, Xie Qin, Chen Zhenping, Zhao Pengcheng, Liu Zijing, Zeng Wenjie, Xu Shikun

2020, 41(3): 1-7. doi: 10.13832/j.jnpe.2020.03.0001

Abstract(325) PDF(174)

Abstract:
In order to study the burnup characteristics of actinide burnable poisons in plate-type fuel assembly and the feasibility of actinide burnable poisons in extending core lifetime, the burnable poison loading, burnup depth and utilization of 235U in plate-type fuel assemblies with different enrichments but same initial kinf were calculated and analyzed in this paper. The results show that, in the case of low enrichment (4%~7%), 240Pu burnable poison exhibits a good conversion effect during the lifetime, and the utilization of 235U is high, which plays a role in extending lifetime. In the case of medium enrichment (25%~40%), the conversion effect of the 240Pu burnable poison is weakened. The 231Pa burnable poison exhibits a better conversion effect during the lifetime. In the case of high enrichment (70%~97%), the assemblies containing 231Pa can achieve the deepest burnup and its utilization of 235U is also the highest, although the conversion effect of the 231Pa burnable poison is weakened. It can be concluded that, 240Pu can be used as the burnable poisons for the low enrichment fuel for long lifetime purpose. However, 231Pa is preferred for the circumstances of medium and high enrichment fuels.

Development and Validation of Probability Table Module for Unresolved Resonance Region in NECP-Atlas

Bi Huchao, Zu Tiejun, Xu Jialong, Cao Liangzhi, Wu Hongchun

2020, 41(3): 8-13. doi: 10.13832/j.jnpe.2020.03.0008

Abstract(348) PDF(113)

Abstract:
Nuclear data processing is an important interface connecting nuclear data and neutronics transport codes. The calculation of the unresolved resonance region is one of the key technical points of nuclear data processing. The resonance self-shielding effect in the unresolved resonance region with dense resonance distribution shall be considered in neutronics calculation. The probability table method is a common method to obtain the self-shielding cross section of the unresolved resonance region. The Ladder Sampling method is the most widely used probability table generation method at present. It constructs pseudo resonance sequences to be equivalent to the actual resonance structure to calculate the probability tables. The probability table calculation module based on Ladder Sampling method is developed in NECP-Atlas. The sensitivity analysis of complex error function calculation, chi-square random number generation, probability table partition, resonance formulas selection, sort algorithm and ladder number is carried out, and the optimal calculation conditions are determined to obtain the accurate probability tables efficiently.

Initial Criticality Achieving Manner and Characteristic Analysis of Reactors in Advanced Nuclear Power Units of Third Generation

Wei Wenbin, Xing Chao

2020, 41(3): 14-18. doi: 10.13832/j.jnpe.2020.03.0014

Abstract(311)

Abstract:
This paper mainly introduces how the advanced nuclear power unit of third generation monitors the reactor to achieve the initial criticality through the source range detector of the external nuclear instrumentation system under the condition of low neutron flux, and compares and analyzes the change of the detector's count rate in the process of achieving the initial criticality. Through the analysis, it is found that under the condition of low neutron flux, the achieving of initial criticality of the reactor can be judged by the change of reactor start-up rate (or reactor period). At the same time, using the change rule of the detector count rate at different positions relative to the neutron source can monitor the approaching of initial criticality. This initial criticality mode can be used in the case of low neutron flux, such as the reactor start-up without neutron source.

Development and Preliminary Verifiation of Multi-Group Cross Section Libraries for PWR Shielding Design Using Discrete Ordinate Method

Peng Chao, Ding Qianxue, Mei Qiliang, Fu Yaru

2020, 41(3): 19-23. doi: 10.13832/j.jnpe.2020.03.0019

Abstract(290)

Abstract:
A broad-group coupled neutron/photon (47n+20γ) cross section library, formatted in ANISN format, dedicated to PWR shielding design, were generated based on the latest evaluation library ENDF/B-VIII.0. The broad-group library was produced by the NJOY, PUFF-IV and SCALE6.1 program. Firstly, a fine-group library was produced using the NJOY99 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format, and then the SMILER module of PUFF-IV was used for the conversion of GENDF format into the AMPX format. Secondly, the neutron resonance self-shielding and temperature effects of the fine-group library are treated using the Bondarenko method by the BONAMI module of SCALE6.1. Finally, the broad-group library was generated by collapsing the fine-group library through the MALOCS and ALPO module of SCALE6.1. The OECD/NEA PWR benchmark was used to validate the new library processed in this paper. Calculation results show that the processing method for the multi-group library proposed in this paper and the multi-group library developed in this paper are correct, which meets the engineering requirement of shield design.

Calculation and Analysis of C5G7-TD Benchmark Based on NECP-X Whole Core Transport Transient Solutions

Wang Bo, Liu Zhouyu, Chen Jun, Zhao Chen, Cao Liangzhi, Wu Hongchun

2020, 41(3): 24-30. doi: 10.13832/j.jnpe.2020.03.0024

Abstract(500) PDF(219)

Abstract:
The C5G7-TD benchmark problems are established by Organization for economic development and cooperation Nuclear Energy Agency (OECD/NEA), to verifiy the calculation capability and precise of 3D non-uniform transient transportation calculation code. NECP-X is a numerical reactor physics calculation code developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University. To better verify the space-time neutron kinetics module of high-fidelity neutronics NECP-X, the C5G7-TD benchmark problems are solved by the NECP-X employing faithful models of the core configuration and transient control parameters. For all cases, the NECP-X results are compared with the nTRACER transient simulation results. The results consisting of transient power behaviors and dynamic reactivity changes are presented. The results of the time consumption for three-dimensional cases and the detailed power distribution are also presented. Numerical results show that the results of NECP-X are with high precision and high resolution, the computational time is at the international leading level, and NECP-X can satisfy the requirements of high-fidelity calculation.

Improvement of Intermediate Range Set Points Calibration in Nuclear Power Plants

Luo Liangwei, Zhang Haizhou

2020, 41(3): 31-34. doi: 10.13832/j.jnpe.2020.03.0031

Abstract(166)

Abstract:
Based on the fact that some units of CPR1000 series cannot complete the calibration of the set point on the Intermediate Range Channel at 48%Pn due to the current saturation, and considering the outage economic, a new method for the set points calibration of the intermediate range is studied and analyzed by MCNP program and the nuclear design software package SCEINCE. The analysis results show that the set points calibration of the intermediate range can be adjusted to 30%Pn platform, and the set points calibrated by the improved method is closer to the design value.

Experimental Investigation on Impact Behavior and Leidenfrost Phenomenon of Droplets on ATF Cladding Surfaces

Wang Zefeng, Ma Yunfei, Zhong Mingjun, Xiong Jinbiao, Yang Yanhua

2020, 41(3): 35-40. doi: 10.13832/j.jnpe.2020.03.0035

Abstract(274)

Abstract:
In order to investigate the Leidenfrost phenomenon of droplet impact on ATF (accident tolerance fuel) claddings, the impact behavior of droplet on sintered SiC and FeCrAl is recorded by high-speed camera, and Zr-4 is chosen as a reference. Based on the visualization results, the droplet impact behavior can be categorized into five regimes, i.e. deposition, rebound with secondary atomization, breakup with secondary atomization, rebound and breakup. Deposition corresponds to nucleate boiling. Rebound and breakup correspond to film boiling, and the rest two impact regimes are corresponding to transition boiling. It is found that the CHF temperature is relatively insensitive to Weber number, while the Leidenfrost temperature increases with Weber number and solid thermal effusivity. During film boiling, the droplet spreading dynamics seems to be independent of the surface temperature. As the Weber number increases, the droplet spreads more rapidly, and reaches a larger spreading diameter.

CFD Simulation of Quasi-Periodic Large Scale Vortices between Tight Lattice Subchannels

Wen Yan, Liu Maolong

2020, 41(3): 41-44. doi: 10.13832/j.jnpe.2020.03.0041

Abstract(278)

Abstract:
In order to accurately evaluate the mixing phenomenon between tight lattice subchannels, the open-source computational fluid dynamics (CFD) software OpenFOAM2.0 and an explicit geometric Reynolds stress turbulence model based on k-ω were used to simulate periodic large scale vortices. The variation of the periodic large-scale vortex wavelength and the peak frequency among the closely spaced channels are studied. The calculation results showed that the averaged peak frequency of the coherent fluctuations is increased linearly with the increasing of Re number. However, the wavelength can approximately be considered as a constant at different Re, and is only dependent on the geometry. The quasic-periodic large scale vortices cause a significant flow pulsation between tight lattice subchannels, which significantly enhances the turbulent mixing between adjacent subchannels.

Instantaneous Flow and Structural Characteristics During Opening of a Nuclear Check Valve

Liu Mengyao, Kang Can, Ding Kejin, Li Bing

2020, 41(3): 45-51. doi: 10.13832/j.jnpe.2020.03.0045

Abstract(246)

Abstract:
For a nuclear check valve, the flow characteristics and mechanical properties of the primary components during the opening process were investigated. The commercial computational fluid dynamics code of ANSYS Fluent and user defined function (UDF) were used in combination to simulate the flows inside the valve. Furthermore, the fluid-solid interaction method was used at various valve openings to predict the mechanical behaviors of the valve components exposed to the fluid flow. The stress distributions over the valve body and the valve plug were obtained. The deformation of the valve body and the valve plug was quantified. The results show that with the increasing of valve opening, high velocity arises at the throat and the valve outlet. High pressure gradients emerge at the valve inlet. The largest deformation of the valve body is observed at the throat and deformation magnitudes fluctuate with the valve opening. The maximum equivalent stress is produced at the elbow area; it varies inversely with the valve opening. For the valve plug, the largest deformation is produced at the side close to the valve inlet. The maximum equivalent stress arises at the contact region between the valve plug and the spring. Different from its counterpart associated with the valve body, the maximum equivalent stress in the valve core increases with the increasing of the valve opening.

Numerical Analysis of Flow and Heat Transfer Characteristics in Condenser Tube Bundle Module of Nuclear Power Plant

Li Guodong, Wang Tao, Zhou Jian, Guo Xinggang, Liu Guodong

2020, 41(3): 52-56. doi: 10.13832/j.jnpe.2020.03.0052

Abstract(260)

Abstract:
Aiming at the deformation problem of the condenser titanium tube of a nuclear power plant in China, the multi-phase flow CFD method was used to analyze the flow and heat transfer characteristics of the internal tube bundle module in different working conditions, and the stress of local titanium tube by finite element method. The research results show that when the unit stops temporarily in winter, the air cooling area of the condenser will be frozen. When the unit starts, Titanium pipes were damaged by ice moving due to the gravity and force of the flow field inside the condenser, caused large-scale deformation of the titanium tube around the air-cooling zone of the condenser.

Experimental Investigation on Effect of Subcooling Degree on Behaviour of Single Vapor Bubbles in a Narrow Channel

Zhang Liqin, Huang Yanping, Zan Yuanfeng, Wang Junfeng

2020, 41(3): 57-61. doi: 10.13832/j.jnpe.2020.03.0057

Abstract(143)

Abstract:
Experimental study was carried out on the behaviour of single vapor bubbles under subcooling condition in a narrow channel. The effect of subcooling degree was analyzed. Results showed that the single vapor bubbles rose with the decreasing diameter and the changing shape under subcooling conditions. Condensation happened on the interface of single vapor bubbles. Increasing degree of subcooling, the diameter of single vapor bubbles decreased faster. The aspect ratio of single vapor bubbles with the same diameter fluctuated in a scope. Under subcooling conditions, the velocity in z and x direction both first increased then decreased with the increasing diameters. The two velocities both had the maximum value for the bubbles of 10 mm diameter. Larger degree of subcooling induced larger bubble velocity in z direction. The velocity in x direction swung around zero. Increasing the degree of subcooling, the bubble velocity in x direciton increased slightly. The degree of subcooling will affect the behavior of single vapor bubbles in narrow channels, and influence the formation and the evolution of flow regimes in further.

Predictive Control of Average Coolant Temperature for Small Modular Reactors

Pan Jinyi, Yang Ting, Qian Hong

2020, 41(3): 62-67. doi: 10.13832/j.jnpe.2020.03.0062

Abstract(351)

Abstract:
The application of small reactors is more flexible than that of large-scale PWRs, which needs to consider the demand of load variation under grid connection and island operation. However, the controlled system of coolant average temperature under rod speed control is non self balancing system, which has strong rigidity, open-loop instability and complex nonlinearity. Improved dynamic matrix controller (DMC) is designed in this paper. The controller widens the scope of the application of the traditional predictive control and overcomes the limitations of the algorithm. By comparing with the program unit control and PI control, it is verified that the steady-state error of the system is smaller, the response speed is faster and the tracking performance is better.

Suitability Evaluation of Containment Thermal-Hydraulics Codes in Marine NPP Reactor Compartment Safety Analysis

Yang Lei, Wang Ying, Zeng Daidan, Xia Zhimin

2020, 41(3): 68-73. doi: 10.13832/j.jnpe.2020.02.0068

Abstract(264)

Abstract:
At present, there lacks the code for the thermal-hydraulics safety analysis of the marine nuclear power plant reactor compartment. Based on the containment phenomenon identification and ranking table (PIRT) method and a reactor compartment pressure response during typical LOCA accident, the reactor compartment PIRT is established and the matching capability between the GOTHIC Codes validation matrix and PIRT is studied. Finally, the code is proved to be applicable for the safety analysis of the typical marine NPP reactor compartment. The analytical methodology is also a useful reference for the cross-domain application evaluation of the safety analysis codes for nuclear electrical field.

Feasibility Study of Two-Phase Flow Wetness Measurement by Ultrasound Velocity Technology

Leng Jie, Hu Xueyin, Tian Ruifeng

2020, 41(3): 74-80. doi: 10.13832/j.jnpe.2020.03.0074

Abstract(246)

Abstract:
Saturated steam is an important working medium for steam turbines. The wetness of the steam will directly affect the safe and economic operation of steam turbines. Therefore, it is of great significance to realize the reliable on-line measurement of the steam wetness. In order to verify the feasibility of two-phase flow wetness measurement technology based on ultrasonic sound velocity method, this paper firstly introduces some necessary assumptions, and proposes a theoretical model of wetness measurement by sound velocity method. Theoretical analysis shows that the medium wetness is the function of sound velocity, temperature and pressure. The function of direct variables has fewer direct measurement variables, and the overall structure of the model is relatively simple. On this basis, the cold-state experimental study is carried out with air-droplet as the object to further verify the feasibility of the model. The experimental results show that with the increase of wetness, the sound velocity of the medium gradually decreases, and the two have a strong linear relationship. In the wetness range of 0~20%, the speed of sound change is about 22 m/s. The two-phase flow wetness measurement scheme based on the ultrasonic sound velocity method is with certain feasibility.

Expansion Research of SG Overfill Analysis for SGTR

Liu Lixin, Liu Zhan

2020, 41(3): 81-85. doi: 10.13832/j.jnpe.2020.03.0081

Abstract(219)

Abstract:
Steam generator (SG) tube rupture overfill in a nuclear power plant have been analyzed with thermal-hydraulic system code, which validates that SG do not overfill under this accident. Expansion research of the overfill analysis for SGTR also has been conducted in this paper, considering multiple cases for tube break, including single tube double ended, multiple tubes double ended and tube small break. The analysis results of three cases are compared and the most conservative situation is provided in this paper. The results show that the case of single tube double ended is the most limit case, under which the SG overfill do not occur, and the overfill margin is less than that in the other two cases.

Study on Separation and PurificationTechnology of 63Ni from Neutron Irradiated Nickel Target

Su Dongping, , Liang Banghong, Zhang Jingsong, Chen Yunming, Li Bing, Li Shuntao, Zhou Chunlin

2020, 41(3): 86-90. doi: 10.13832/j.jnpe.2020.03.0086

Abstract(215)

Abstract:
⁶³Ni is produced by the neutron irradiation of ⁶²Ni in a reactor. After separation and purification, the highly purified ⁶³NiCl₂ solution was obtained. In this paper, using the nickel target which was irradiated in High Flux Engineering Test Reactor (HFETR) for more than 30 years as the research subject, the ZGANR170 nuclear anion resin was used to purify ⁶³Ni from the solution of irradiated nickel target. In the end, the high purity ⁶³NiCl₂ solution products were obtained. In addition, the separation principle, procedures and the optimal separation condition were both introduced in detail. Moreover, the effect of acidity on separation was studied experimentally. The result demonstrated that only when the concentration of HCl is greater than or equal to 9 mol/L, a good separation effect could be obtained. Besides, the decontamination factor of γ radionuclides in the whole process of nickel purification is 2.33×10³. And the nuclear purity of the 63NiCl2 solution is more than 99.9%. After dissolution, system conversion and separation, the recovery rate in the whole process is 86.5%.

Effect of Sintering Temperature on Microstructure and Phase Composition of U-Hf Burnable Absorber Fuel

Zhang Jia, Wang Xinjie, Peng Xiaoming, Liu Jinhong, Sun Chao, Kang Wu, Zeng Cheng, Li Jia

2020, 41(3): 91-96. doi: 10.13832/j.jnpe.2020.03.0091

Abstract(196) PDF(105)

Abstract:
The effect of sintering temperature on physical properties, microstructure and phase composition of U-Hf (containing 25wt%HfO2) burnable absorber fuels was studied by SEM, EDS and XRD, etc. The results suggest that:(1) The dried and calcined U-Hf fuel possessed good sintering activity, the densifying of U-Hf fuel was effectively promoted, and the density of fuel increased slightly when the sintering temperature was higher than 1550℃. (2)The sintered U-Hf crystal appeared in the form of UO2 solid solution matrix grains and the small second phase grains which dispersed in matrix grains. The small second phase grains were solid solution which Hf content was higher than stoichiometric, and as the sintering temperature increased, the small grains grew bigger and fewer. (3)The HfO2 crystal underwent a phase transition when the sintering temperature was higher than 1550℃. At about 1750℃, the monoclinic phase HfO2 transformed to the tetragonal phase incompletely, and dissolved into matrix UO2 incompletely.

Computational Simulation Research of Unstable Mechanics of Fission Gas Release under Condition of Fuel Cladding Breach

Li Chenyue, Dong Bing, Yin Junlian, Li Leihao, Wang Dezhong

2020, 41(3): 97-103. doi: 10.13832/j.jnpe.2020.03.0097

Abstract(266)

Abstract:
When the claddings of the PWR fuels breach, the fission gas accumulated in the gap of pellets and claddings releases into the coolant. Naturally, the process of fission gas release is a process of two-phase flow, while the microscopic mechanics is still undiscovered. To reveal the law of the interaction between the coolant and the gas during the process of fission gas release, a CFD method is built to simulate the transient process of fission gas release, in which VOF and k-ε model is used. The result shows that the coolant will flow into pellet-cladding gap, then it will be evaporated, which will lead to the increasing of the pressure in the gap, and the fission gas will release into the subchannel; the processing that the fission gas releases from the pellet-cladding gap can be divided into 2 steps. In step 1, the differential pressure between the pellet-cladding gap and the subchannel is relatively larger, which makes the gas jet into the subchannel, and the duration of this step is short, and the ratio of the release of the fission gas is higher, meanwhile, it changes violently. In step 2, the difference of the pressure between the pellet-cladding gap and the subchannel is relatively smaller and stable, and the fission gas enters the subchannel by the convective mass transfer through the vortex in the breach. The duration of this step is short, and the ratio of the release of the fission gas is lower and stable.

Transient Dynamic Analysis of Subcritical Energy Blanket for Uranium-Based Fusion-Fission Hybrid Reactor

Liu Zhiyong, Qu Ming, Huang Hongwen, Zeng Herong, Wang Shaohua, Guo Haibing, Ma Jimin

2020, 41(3): 104-109. doi: 10.13832/j.jnpe.2020.03.0104

Abstract(154)

Abstract:
Based on the transient dynamic analysis of the subcritical energy blanket and the support fixed strcture, this paper gives the stress distribution contours and deformation contours of all parts. The analysis results show the maximum stress value of the subcritical energy blanket of various parts is below the allowance stress and satisfy the strength requirement. The deformation for various parts is reasonable, and no significant displacement and instability phenomena occurs .

Simulation Analysis of Fuel Assembly Entering Floating Nuclear Reactor Core

Guo Yiding, Guo Jian, Tan Mei

2020, 41(3): 110-114. doi: 10.13832/j.jnpe.2020.03.0110

Abstract(268)

Abstract:
Unlike that in the onshore nuclear power plant, the refueling operation in the floating nuclear reactor will be affected by the wave environment, so new requirements for refueling operation technology and equipment are put forward. In this paper, the refueling scheme of Marine Nuclear Power Platform is selected to simulate the process of fuel assembly entering the reactor core under rocking condition. The analysis results show that the fuel assembly meets the strength design requirements in the process of introducing universal joint fuel assemblies entering the reactor core.

Study on Spectral Accelerations for Seismic Design of Nuclear Power Plants on Non-Bedrock Sites Based on Statistical Data

Wang Yushi, Li Xiaojun, Liu Aiwen, Lu Tao, Zhao Jiaxiang, Wang Ning, Li Yiqiong

2020, 41(3): 115-120. doi: 10.13832/j.jnpe.2020.03.0115

Abstract(214)

Abstract:
In order to obtain more accurate spectral accelerations for seismic design, 2661 strong-motion acceleration records in the Next Generation Attenuation (NGA) West2 database were statistically analyzed, the effects of earthquake parameters on the middle to long periods of the ground motion spectral accelerations were studied, and normalized horizontal spectral accelerations on bedrock (site classification I in the Chinese standard) and non-bedrock (site classification II and III in the Chinese standard) for seismic design were established. The results indicated that the spectral accelerations of the ground motion were significantly influenced by site conditions and earthquake moment magnitudes, meanwhile they were slightly influenced by distance parameters of the earthquake source. Comparing to the spectral accelerations for seismic design suggested in RG1.60 of USNRC and GB50267-97 in China, the spectral accelerations of the ground motion obtained in this paper could evaluate the effects of site surface geology characteristics and earthquake moment magnitudes on the middle to long periods of the ground motion more reliably. Finally, the spectral accelerations for seismic design considering the influences of site conditions and seismotectonic environments were established, which could be used as the ground motion inputs in the site selection and seismic design of nuclear power plants on non-bedrock sites.

Development of NPP Steam Generator Multifunctional Simulator

Tang Chenhang, Wu Ge, Li Donghui, Li Huanming, Huang Yan

2020, 41(3): 121-124. doi: 10.13832/j.jnpe.2020.03.0121

Abstract(182)

Abstract:
During the in-service inspection and the maintenance of the nuclear steam generator, a steam generator simulator is necessary for the debugging of maintenance tools and personal training. Based on the functional requirements and design requirements, a multifunctional steam generator simulator is developed. The overall structure of the simulator was designed to be 1/2 of the prototype, the composite material was used, and it is equipped with observation windows. The developed simulator is with the characteristics of compact structure, high economic efficiency and observable. Function tests verify that the simulator can be used for nuclear steam generator in-service inspection and training.

Analysis of Maximum Span of Pipe Based on Timoshenko Beam Theory

Li Xinghua, Wu Gaofeng, Qin Manqing

2020, 41(3): 125-128. doi: 10.13832/j.jnpe.2020.03.0125

Abstract(204)

Abstract:
The traditional calculation method of the maximum span is based on Euler beam theory. This method only considers the bending deformation of the cross section and ignores the effect of shear deformation. Based on the simple support beam model of Timoshenko beam theory, the theoretical formula for the calculation of span considering shear deformation is given in this paper. By this formula, for the tubes of DN200 and below, the difference of maximum span based on Euler beam theory and Timoshenko beam theory is less than 1%. For that above DN200, maximum span based on Timoshenko beam theory is smaller than that based on Euler beam theory, and the difference is larger with the increasing of the outer diameter of the pipe. It is suggested to use the formula based on the Timoshenko beam theory to calculate the maximum span of pipes.

Investigation of Vibration Reason for Vertical Pumps in Nuclear Power Plant Safety System

Xiang Xianbao, Li Zhen

2020, 41(3): 129-132. doi: 10.13832/j.jnpe.2020.03.0129

Abstract(210)

Abstract:
Vibration problem have been existed on multiple safety injection and spray pumps in the safety systems in many nuclear power plants. Firstly, the vibration condition has been introduced, and the vibration characteristics have been summarized. Spectrum analysis has been implemented. The possible factors related to the fault have been investigated. Secondly, based on the vibration characteristics and frequency spectrum characteristics, a vibration mechanics model is set up. The root cause of excessive vibration is resonance in the system that is consisted of equipment and basement, not the soft feet. It has been validated in the test. Finally, the treatment suggestion of stiffness optimization has been proposed. It was advised that the modal analysis of integral equipment and basement system should be carried out in design phase.

Evaluation of Bypass Leakage Design Idea for Floating Nuclear Power Plants Based on Safety Enclosure

Chen Yanxia, Tan Mei, Chen Qiang, Guo Jian, Zhang Jincai, Li Pengfan

2020, 41(3): 133-136. doi: 10.13832/j.jnpe.2020.03.0133

Abstract(248)

Abstract:
Based on the concept of secondary containment of onshore nuclear power plants, the concept of safety enclosure of floating nuclear power plants is introduced in this paper. The radioactive containment “containment + safety enclosure +reactor compartment” is proposed. This paper studies the evaluation of the design idea of bypass leakage for safety enclosure and proposes a method for identifying the bypass leakage path and determining the bypass leakage rate. The design basis of negative pressure for safety enclosure is given in the end, which can be referred in the design of the ventilation system for the floating nuclear reactors in later stage.

Analysis of Human Errors in Severe Accident of Nuclear Power Plant Based on Cognitive Model and Fault Tree

Zhang Li, Chen Shuai, Qing Tao, Sun Jing, Liu Zhaopeng

2020, 41(3): 137-142. doi: 10.13832/j.jnpe.2020.03.0137

Abstract(501)

Abstract:
In order to analyze the human errors of the emergency crews in nuclear power plants while dealing with severe accidents, this paper establishes a cognitive model of the emergency crew, identifies the corresponding performance influencing factors, finds thirteen human error modes, known as insufficient information source, poor information reliability, premature end of parameter acquisition, incorrect processing of important data, error in negative impact assessment of mitigation measures, selection of strategies unsuitable for the current situation, delayed decision-making, omission of important information / alarm, delayed detection, soft control operation error, information feedback failure, equipment installation, connection or operation error, and delayed implementation. Then the paper analyzes the root causes of human errors based on fault tree, including communication failure, time pressure, uncertainty of accident progress, information delayed reception, monitoring error, poor human-computer interface and environmental factors. The results can be used to predict the human errors in the process of severe accident mitigation, as well as helping the implementation of severe accident management and the technical improvement, thus providing guidance for improving the safety of nuclear power plants in severe accidents.

Research of Automatic Start-up and Shutdown Technology of Pool Reactors

Liu Chun, Zhang Caike, Xie Chenglong, Nie Wen, Zhang Yadong

2020, 41(3): 143-146. doi: 10.13832/j.jnpe.2020.03.0143

Abstract(173)

Abstract:
The automatic start-up and shutdown technology used in the reactors can effectively reduce the working intensity of the operators, reduce the misoperation, and improve the safety and reliability of the reactors. Based on the analysis of the process characteristics and the start-up and shutdown operation of a typical pool reactor, this paper researches the control range, hierarchical structure, breakpoints and typical control logics of automatic start-up and shutdown (APS) system in the pool reactor, and completes the simulation and verification system of the pool reactor APS technology. The APS system can achieve the automatic start-up and shutdown, and there is no manual operation in the process, thus reduces the possibility of personnel misoperation.

Research on Dependence Assessment in Human Reliability Analysis of Nuclear Power Plants

Li Lusu, Su Xiaoyan, Qian Hong, Zhou Jie

2020, 41(3): 147-152. doi: 10.13832/j.jnpe.2020.03.0147

Abstract(241)

Abstract:
Human reliability analysis (HRA) is an important part of nuclear power plant risk analysis. The dependence analysis of human errors is an essential part of HRA. Ignoring the dependence between human errors will lead to the underestimation of the risk level of nuclear power plants. This paper presents a model of dependence assessment based on D numbers and AHP-DEMATEL method. Firstly, the factors affecting the dependence between events are determined and their structural relationships are established, and the membership function and anchor point of the dependence degree are established for each factor. Secondly, the comprehensive weight of each factor is determined by using AHP-DEMATEL method. Finally, the dependence degree of each factor is evaluated according to the actual situation and the D numbers is constructed. And, according to the D numbers and the comprehensive weight, the total dependence degree and the reliability of the two human failure events are calculated. The validity of the model and method is verified by an example.

Design of High Pressure Water Jet Decontamination Device with Pressure and Flow Synchronous Control Function

Teng Lei, Wang Shuai, Wang Xiaobing

2020, 41(3): 153-157. doi: 10.13832/j.jnpe.2020.03.0153

Abstract(405)

Abstract:
In the process of maintenance or decommissioning of nuclear power plants, the high-pressure water jets are often used to remove the radioactive contamination from the site. Based on the conventional high-pressure water jet decontamination device, a proportional-integral-derivative(PID)-based electric adjustment control was proposed. The application of separate pressure and flow control in high-pressure water jet decontamination was studied. The results of theoretical analysis combined with decontamination laboratory verification show that: using the improved flow and pressure control methods, in the high-pressure water jet decontamination process, with the same other influencing factors, when a larger water flow rate is used to decontaminate, the decontamination factor is rather small, but this trend is tending to be gentle. Therefore, by adopting the improved high-pressure water jet decontamination device to achieve the same decontamination effect under the same pressure, the amount of secondary radioactive waste liquid generated can be significantly reduced, and this device is with high market application value.

Simulation Research on Feed Water Control System of Demonstration Fast Reactor

Bi Derui, Duan Tianying, Zhang Houming, Jia Yuwen, Liu Yong

2020, 41(3): 158-163. doi: 10.13832/j.jnpe.2020.03.0158

Abstract(214)

Abstract:
Demonstration Fast Reactor sets up eight once-through steam generators in each circuit, and based on that, two designs of feed water control system are proposed. The first is to control the total water in each circuit, and the second is to control the water of each module. The simulation models of the two feed water systems are built, and the process of sodium temperature and steam superheat at evaporator outlet is analyzed. The results show that the first design scheme helps to stabilize the steam superheat at evaporator outlet, and the second scheme facilitates the control of the sodium temperature at evaporator outlet.

Automatic Ultrasonic and Signal Analysis for Closure Stud of Nuclear Power Station

Chen Zhicong, Ren Jianbo, Zhu Jiazhen

2020, 41(3): 164-169. doi: 10.13832/j.jnpe.2020.03.0164

Abstract(321)

Abstract:
According to RSE-M standard, the main bolts of reactor pressure vessel (RPV) in nuclear power plants need to be inspected by ultrasonic regularly. In order to ensure the sensitivity of ultrasonic inspection of notches in different depths of stud thread area and rod area, this paper carries out the sound field simulation calculation for the detection process, analyzes the relevant and non relevant signals in data acquisition, and focuses on the characteristics of crack signals to verify the reliability of ultrasonic technology. Combined with the field case, the abnormal signal can be determined effectively by the high reflectivity of 45° S-wave and other detection methods such as eddy current and penetration testing.

Condition Monitoring Model for Sensors of Reactor Coolant Pump Based on PCA

Zhu Shaomin, Xia Hong, Peng Binsen, Wang Yan, Wang Zhichao, Zhang Jiyu, Jiang Yingying

2020, 41(3): 170-176. doi: 10.13832/j.jnpe.2020.03.0170

Abstract(295)

Abstract:
In order to meet the requirements of operation control and parameter monitoring, a large number of sensors are arranged in the main and auxiliary systems of the reactor coolant pump. As the reactor coolant pump operates, the sensors may age or malfunction. In order to improve the weakness of the periodic tests and calibration schemes of sensors in nuclear power plants, the principal component analysis (PCA) was utilized for the condition monitoring of the reactor coolant pump sensors. The PCA monitoring model was established based on the operating data of the reactor coolant pump of a nuclear power plant, and the small drift fault and common mode fault of the sensors were identified by the model. The simulation results show that the monitoring model has a good effect on the condition monitoring of the reactor coolant pump sensors.

Application of System Engineering Methodology in Digital Experiment of Nuclear Reactor Engineering

Zeng Xiaokang, Huang Yanping, Zhang Liqin, Lang Xuemei, Zan Yuanfeng, Yuan Dewen

2020, 41(3): 177-182. doi: 10.13832/j.jnpe.2020.03.0177

Abstract(191)

Abstract:
Complexity of nuclear reactor experiment systems always is one of important factors which limit the experimental technology development and innovation. Digital experiment of nuclear reactor engineering is introduced to improve the ability to deal with the complexity of experiment systems. The objective of digital experiment is to establish a unified and efficient work environment throughout the whole life cycle of the experiment programs of nuclear reactors. The high-level system architecture of digital experiment is constructed based on the method of system engineering in this paper. It consists of business architecture which is constituted by V model and business scenario diagram, experimental infrastructure architecture which is constituted by logic diagram of intellectualization, and layered architecture of digital experiment platform which is constituted by business administration system in the business layer, experimental design simulation environment system in the application layer, experiment knowledge system in the knowledge layer and basic data administration system in the resource layer. The application example for PRS experiment validates the effectiveness of the above system architecture of digital experiment as the whole logic framework of the digital experiment platform.

Application of Significance Determination Process at Power Condition in Nuclear Safety Regulations

Ma Guoqiang, Li Juan, Ding Shanshan, Zhang Yanyun, Liu Chengyun

2020, 41(3): 183-187. doi: 10.13832/j.jnpe.2020.03.0183

Abstract(163)

Abstract:
This paper describes the basic principles and methods of the significance determination process (SDP) at power condition, which was developed by the National Nuclear Safety Administration, and was used to analyze the significance and sensitivity of the failure of the pneumatic auxiliary feed pump (ASG004PO) in nuclear power plants. The result has indicated that the period of the recirculation flow test for the steam-driven auxiliary feed pump was too long, and the optimization proposal for this issue is to optimize the cycle of ASG004PO recirculation flow test to less than 34 days.

Effect of Rod Swelling and Rupture on LOCA Process

Wu Dan, Deng Jian, Ding Shuhua, Xin Sufang, Xian Lin, Bi Shumao, Mao Huihui

2020, 41(3): 188-192. doi: 10.13832/j.jnpe.2020.03.0188

Abstract(337)

Abstract:
Fuel rods experience the heat up process several times during the loss of coolant accident (LOCA). Swelling and rupture will happen if the temperature rises to a certain degree which will have great effect on the LOCA process. Both the inside and outside of the fuel cladding will be oxidized, and this will exacerbate the Zr-H2O reaction. In this study, code ARSAC-K is used to analyze the LOCA process in the generation-III nuclear power plants designed and constructed by China. In order to study the effect of rod swelling and rupture on the LOCA process, several power shapes are chosen. It is found that at the moment of rod rupture, the temperature of the rod cladding will decrease in a few tens of seconds. After few tens of seconds, the cladding temperature continues to rise till the peak cladding temperature is reached.

Study on Nuclear Class Piping Hanger and Support Root Intelligent Selection Based on SOM Clustering Algorithm

Tang Yongtao, Duan Yongqiang, Huang Jie, Su Rongfu, Yu Hongxing, Liu Yuchen, Wen Jian

2020, 41(3): 193-196. doi: 10.13832/j.jnpe.2020.03.0193

Abstract(183)

Abstract:
Based on the data mining technology, the pre-processing method of intelligent root selection data for nuclear-level pipeline hanger and support is studied. The classification method of root selection pre-processing is studied. The data pre-processing flow is designed and the priority order of root selection of hanger and support is determined; Based on SOM clustering algorithm, the calculation process of intelligent selection data of the root of hanger and support is studied, the experimental platform is designed. Based on the actual engineering data, the feasibility and the effectiveness of the algorithm are verified, the data pre-processing is proved and the clustering effect is obvious.

Preliminary Study on Influencing Factors of Fuel Economy for Heat Pipe Reactor

Wang Jinyu, Yu Hongxing, Chai Xiaoming, Zhang Zhuohua, Li Wenjie, Su Dongchuan, Zeng Chang, He Xiaoqiang, Li Songwei

2020, 41(3): 197-201. doi: 10.13832/j.jnpe.2020.03.0197

Abstract(313)

Abstract:
The heat pipe reactor directly extracts heat from the core through the high temperature heat pipe, the system design itself is extremely simplified, and it is more suitable as a technical selection of a small nuclear power source. Fuel economy is an important basis for the selection of reactor technology routes. In order to study the impact of heat pipe reactor design on its fuel cycle economy in detail, a preliminary analysis model of the fuel economy influencing factors of the heat pipe reactor is established in this paper. The eVinci reactor is taken as an example to develop the fuel cycle economy. Exploring and researching on the influencing factors of fuel economy, we obtained the influence trend of the fuel economy such as the overall scheme power scale and core operating temperature. The results show that due to the comprehensive impact of fuel prices, uranium loading, and enrichment, the preferred power scale range for relatively good fuel economy of heat pipe reactors is between about 1MWt and 5MWt. Increasing the core operating temperature can greatly improve fuel economy, and economic best power range is extended to high power scale.

Numerical Study on Pressure Pulsation Characteristics of Liquid LBE Medium Axial-Flow Pump

Wang Yan, Yu Hongxing, Guo Yanlei, Yan Mingyu, Sui Haiming, Zhang Yulong, Ren Yun

2020, 41(3): 202-207. doi: 10.13832/j.jnpe.2020.03.0202

Abstract(211)

Abstract:
Base on the Reynolds time averaged N-S equation and Renormalization group (RNG) k-ε two equation turbulence model, this paper studies and analyzes the hydraulic performance and the pressure fluctuation distribution characteristics of the axial-flow LBE pump in room temperature, clear water and liquid LBE. The results show that: for the hydraulic design of the axial-flow pump completed according to hydraulic design method of clear water medium and relevant empirical coefficient, under the condition of LBE medium, the head and the efficiency of the pump are improved, the head increases obviously with the increasing of the flow, and the phenomenon of flow separation in the inner boundary layer of the pump is obviously weakened under the LBE medium. The main frequency of the pressure coefficient fluctuation at the inlet monitoring point of the pump impeller is 81hz, which is equal to the blade rotation frequency, and the second frequency harmonic occurs at two and three times of the blade frequency; the medium viscosity does not affect the pressure fluctuation coefficient amplitude at the inlet of the impeller, and the medium density, that is, the inertial force, is linearly positively related to the pressure fluctuation amplitude at the inlet of the pump impeller.

Research on Fault Diagnostic Technology of Primary Loop of Nuclear Power Plant Based on iForest-Adaboost

Ai Xin, Liu Yongkuo, Jiang Liping, Xia Hong, Zhou Xinqiu

2020, 41(3): 208-213. doi: 10.13832/j.jnpe.2020.03.0208

Abstract(210)

Abstract:
The traditional fault diagnostic methods such as principal component analysis and BP neural network are with poor generalization capability and low fault identification accuracy in complex nonlinear systems. The isolation forest (iForest) algorithm uses the idea of isolation tree partition to identify the abnormal data, which is applicable to the state monitoring of nonlinear systems. The Adaboost algorithm is a boosting algorithm based on the idea of combined classification, and the overall algorithm has better generalization capability through the superposition of multiple weak classifiers. Therefore, the isolation forest algorithm and Adaboost algorithm are used in this paper to establish the iForest-Adaboost primary loop fault diagnostic system for nuclear power plants, and the GSE real-time simulation platform and the simulation data in unit 1 of Fuqing Nuclear Power plant are used for the test. The test results show that the isolation forest algorithm can identify the system anomalies faster than the principal component analysis and QTA threshold algorithm. The Adaboost algorithm has a higher fault identification accuracy than BP neural network and support vector machine algorithm.

Research on Methods of Failure Analysis for Control Rod Drive Mechanism Based on Structure Noise Detecting Technique

Peng Cuiyun, He Pan, Peng Xiaowei, Liu Caixue, Wang Yao, Xu Hui

2020, 41(3): 214-216. doi: 10.13832/j.jnpe.2020.03.0214

Abstract(220)

Abstract:
The structure noise detecting technique for the control rod driving mechanism (CRDM) has been researched for CRDM of screw and roller driven. The sensitive characteristics and failure criteria in the rise and fall of CRDM have been obtained by the time and frequency analysis of the structure noise signals of CRDM. The test results indicated that the time and frequency analysis of the structure noise signals could reflect the failure state of CRDM and could discriminate the failure types of CRDM effectively.

Vibration Fault Diagnosis and Treatment of Vertical Long Shaft Pump Motor in Nuclear Power Plants

Fu Jiangyong, Wei Wenbin, Liu Mingli, Wang Yuehui

2020, 41(3): 217-220. doi: 10.13832/j.jnpe.2020.03.0217

Abstract(255)

Abstract:
In a nuclear power plant, the vertical pump motor has a large vibration difference between the two radial directions, and the vibration phenomenon slowly rises after the start. To solve the vibration problem, this paper uses methods such as spectrum analysis, natural frequency analysis, and phase analysis to diagnose the fault. It is found that the matching motor is with the coupling problem of structural resonance, rotor thermal bending and dynamic unbalance, and the vibration problem is solved by adjusting the bolt and the on-site dynamic balance method.

Design of Intelligent Wireless Vibration Sensor for Wireless Monitoring of Rotating Device Operating Condition

Yu Ren, Xie Xuyang, Qing Fatao, Peng Qiao, Wang Tianshu

2020, 41(3): 221-226. doi: 10.13832/j.jnpe.2020.03.0221

Abstract(341)

Abstract:
In order to improve the data acquisition and condition monitoring capability of the rotating device in the nuclear power plant that is already in operation, it is necessary to solve the problems of sensor installation and cable laying. In this paper, an intelligent wireless vibration sensor based on Zigbee IoT communication technology is designed. The field programmable gate array (FPGA) is adopted as the main control unit. The circuit composition and the work principle of the sensor are given, as well as the working flow of the embedded control software. The performance of the sensor is tested, and the results show that the power consumption of the sensor is low, and the vibration signal can be collected, analyzed and uploaded continuously. The sensor is simple to install, and there is no need to lay power supply and signal cables, and can be used to constitute the condition monitoring system for the rotating device of nuclear power plants.

2020 Vol. 41, No. 3