核动力工程

Digital Reactor: Development and Challenges

Yu Hongxing, Li Wenjie, Chai Xiaoming, Li Songwei

2020, 41(4): 1-7.

Abstract(1347) PDF(1381)

Abstract:
The digital reactor is an integral numerical simulation platform for the performance of nuclear reactor systems. In the first part of this paper, the development history of the nuclear reactor simulation technology is reviewed. The three technical elements constituting the digital reactor are elaborated, including the target scenario, advanced models and multi-physics coupling technology, and the integrating environment. Although there are several challenges for the development of digital reactors, such as the difficulties in multi-physics and multi-scale computation, the complexity in design optimization, and the insufficient database, the digital reactor can help better analyze key problems that limiting the reactor performances and safety, and better understand the mechanism of the phenomena that cannot be observed or measured experimentally.

Research of Neutron Transport Kinetics Solver ntkFoam Based on OpenFOAM

Ma Yu, Wang Yahui, Lu Wei, Yang Junhe

2020, 41(4): 8-11.

Abstract(441)

Abstract:
Due to the complexity of neutron transport simulation and the difficulty of its coupling calculation with other physical processes, the full-core detailed neutron transport-thermal-hydraulics multi-physical calculation is a difficulty of reactor engineering. Based on the finite volume C++ open source software OpenFOAM, this work establishes the numerical solution models of steady-state and transient neutron transport kinetic equations by using the finite volume method, and develops the neutron transport kinetics solver ntkFoam. The simulations of some benchmark problems show that the proposed ntkFoam solver can simulate the neutron transport kinetics problems accurately, and can adapt well to different dimensions and complicated geometries. The detailed coupling calculation of neutron transport and heat-mass transfer processes can be realized, which can provide some detailed coupling thoughts and techniques for neutron transport based full core multi-physical simulation.

Research on Multi-Physics Coupling Based on FLUENT

Wang Kun, Dong Xiuchen, Zhang Xin

2020, 41(4): 12-16.

Abstract(370)

Abstract:
Multi-physics coupling analysis based on FLUENT is a hot issue in current nuclear safety analysis. The reactor nuclear power calculation program (PKM) was prepared by using the point reactor neutron kinetic equations of 6 groups of delayed neutrons. The FLUENT/RELAP coupling analysis model and the FLUENT/PKM coupling analysis model were established by the external coupling method and secondary development coupling method respectively. The correctness and effectiveness of the coupled model were verified by the discharge problem of the horizontal branch pipe and the super-power transient problem of the linear reactivity introduced in a single-phase range. The coupled analysis method of the study can provide support for fluent multi-physics nuclear safety analysis.

Study on Heterogeneous Computing for MOC Neutron Transport Calculation with CPU-GPU Concurrent Calculation

Song Peitao, Zhang Zhijian, Zhang Qian, Liang Liang, Zhao Qiang

2020, 41(4): 17-21.

Abstract(318)

Abstract:
The Method of Characteristics (MOC) is capable to accurately solve the neutron transport equation with arbitrary geometry. However, the MOC suffers from some drawbacks: slow convergence and time consuming. Based on the spatial domain decomposition and the ray parallelization, the parallel 2D MOC algorithm was implemented with MPI+OepnMP/CUDA programming model to leverage the computing power of Central Processing Unit-Graphics Processing Unit (CPU-GPU) heterogeneous high-performance computing systems. In addition, a dynamic workload partitioning scheme was proposed to efficiently take advantage of all the CPU and GPU resources. The workload is appropriately assigned to the CPU and GPU according to their computational capabilities, and all CPUs and GPUs perform the calculation concurrently. The numerical results demonstrate that the parallel algorithm maintains the desired accuracy. Meanwhile, the dynamic workload portioning scheme can provide the optimal workload partition based on the runtime performance. As a result, about 14% improvement is observed in the overall performance compared with the MPI+CUDA parallelization when the CPU-GPU heterogeneous computation is performed on 5 heterogeneous nodes (including 20 GPUs).

Feasibility Study on Substitution of Primary Neutron Source with Secondary Neutron Source in PWRs

Su Genghua, Shi Xiaqing

2020, 41(4): 22-25.

Abstract(230)

Abstract:
This paper analyzes the feasibility of substituting the primary neutron source with the secondary neutron source (SNS) in pressurized water reactor under the background that the supply of the primary neutron source has risks. The neutron source intensity of SNS which experienced one fuel cycle irradiation in a working PWR was calculated. Based on the core parameters and loading sequence of the first cycle of a new CPR1000 reactor, the counting rates of neutron detectors inside and outside the reactor were calculated at various loading steps. The results show that the substitution of the primary neutron source with the irradiated SNS within 4 half-lives (about 240 d) in the CPR1000 reactor satisfies the requirements of neutron detector counting rates in related technical specifications, thus proving the feasibility of substituting the primary neutron source with SNS.

Management Study on Alternate Long and Short Cycle of CNP650 Long Cycle Management

Liao Hongkuan, Yu Yingrui, Wang Yongming, Liu Tongxian, Huang Can, Chen Zhang, Liu Mingquan, Chen Liang, Hu Yuying

2020, 41(4): 26-29.

Abstract(210)

Abstract:
Unit 1 and unit 2 of Hainan Nuclear Power Plant adopt CNP650 reactor developed on self-reliance. Because of the obvious peak and valley period of the power consumption in Hainan power grid, the key to transition from annual refueling to long cycle is the great differences of the cycle length between the long and the short cycles. For these purposes, this paper completed the design of new fuel assembly types and enrichment, new fuel assembly number and layout, and the independent transition cycle, and then obtained the fuel management strategy of CNP650 rector. This strategy is composed of the alternative long and short equilibrium cycles, starting from the 5th cycle, through 4 transition cycles, into the equilibrium cycle. The equilibrium cycle are composed of 2 cycles which length are 517.3EFPD and 464.0EFPD. All the parameters meet the design requirements of the long fuel cycle management. This fuel management strategy effectively solves the specific requirements of Hainan Nuclear Power Plant, and can be directly used for the operation of Hainan Nuclear Power Plant.

Dynamic Rod Worth Measurement in Fuqing Nuclear Power Plant Based on Intermediate Range of Ex-Core Neutron Flux Measurement System

Meng Fanfeng, Geng Fei, Cai Guangming

2020, 41(4): 30-33.

Abstract(281)

Abstract:
By analyzing the performance of power range and intermediate range of the ex-core neutron flux measurement system (RPN), the data verification of the Dynamic Rod Worth Measurement (DRWM) was carried out using the measuring data of the intermediate range. The results show that it can be used as an alternative means to the traditional DRWM. Its measurement error is within the allowable range, and it is unnecessary to measure the background current with the use of DRWM based on the intermediate range of the ex-core RPN, which can reduce the dynamic carving time.

Research on Automatic Control Method of Power Regulation System of Xi’an Pulsed Reactor

Zhang Liang, Yuan Jianxin, Zhao Wei, Wang Baosheng, Zhang Qiang, Zhu Guangning, Yang Ning, Chen Lixin, Jiang Xinbiao

2020, 41(4): 34-40.

Abstract(200)

Abstract:
There are some differences between the software and hardware of the digital power regulation system and the previous analog systems, thus a new method for the automatic power regulation needs to be studied. The control model of digital power regulation system is built and the basic idea of the automatic control method is proposed. The key technical issues including avoiding short period problem and enhancing the stability of the motion of the control rod are studied. The automatic control method of the digital power regulation system is designed through iterative calculations. The simulation code of Xi'an Pulsed Reactor(XAPRSC) is applied to verify the design method, and the results show that the new method meets the design index and is beneficial to the operation safety of Xi'an Pulsed Reactor.

Study on Passive IVR Strategy Based on Steam Turbine

Ma Rubing, Sheng Tianyou, Yuan Yidan, Ma Weimin

2020, 41(4): 41-44.

Abstract(287)

Abstract:
In some NPPs with reactor pressure vessel (RPV) arranged at upper elevation, the water overflowing the cavity and/or condensed in the containment cannot flow back to the cavity only relying on the gravity during the IVR process, and so these NPPs cannot realize passive IVR in long term. To solve this problem, a new passive IVR strategy is put forward. In this strategy, the steam produced in the IVR is used to drive a steam turbine, and a pump is driven by the steam turbine to pump the water in the bottom of containment to the cavity again, to establish the cycle of steam and water and to realize the long term IVR. Detailed thermal hydraulic computation is carried out and the feasibility of the system is proved in this paper.

Research on Simplified Assembly and Core Structure Design of SCWR

Yao Lei, Xia Bangyang, Lu Di, Wang Lianjie, Li Xiang, Wang Shiqian, Li Qing

2020, 41(4): 45-49.

Abstract(284)

Abstract:
To solve the problem of insufficient neutron moderation of the super critical water-cooled reactor (SCWR), the water rod or the adding of the solid moderator is commonly used in assemblies, and a multi-pass flow scheme is adopted, which resulting in the complication of the fuel assembly and the core structure, and the strong neutron absorbing materials are introduced into the reactors. Based on the result of CSR1000 study, the simplification of assembly and core structure was carried out. The results significantly simplified the assembly and core structure of SCWR.

Study on Flow Patterns of Open Natural Circulation with Low Height Difference under Low Power Conditions

Li Yi, Sun Jianchuang, Peng Hang, Cheng Xiang, Quan Biao, Cao Xiaxin, Zhou Jian, Ding Ming

2020, 41(4): 50-54.

Abstract(228)

Abstract:
Open natural circulation systems are widely applied in the field of energy and chemical industries. Under some special conditions, open natural circulation systems with low height difference have to be applied, but little research focused on this type of systems. In this paper, the flow patterns of such a natural circulation system were experimentally studied at low power level by adopting the steam heating. In addition, the flow patterns and physical phenomena of the natural circulation system were analyzed in detail under different heating power conditions. It is found that there are four flow patterns under different power levels for the natural circulation system, because of the influence of non-condensable gas and subcooled boiling. The inlet and outlet temperatures of the heating section are analyzed in different flow patterns. The result shows that the outlet temperature of the heating section can be used as a judgment criterion of different flow patterns.

Song Lei, He Xiangyan, Cheng Yanhua, Cui Dawei

2020, 41(4): 55-59.

Abstract(250)

Abstract:
A large-scale and refined three-dimensional numerical simulation of the upper plenum flow field of the pressure vessel of the three-loop pressurized water reactor is carried out by using the commercial CFD software STAR-CCM+ code. The coolant flow at the outlet of 157 fuel assemblies is calculated by the component tracking method. A mixing matrix of the upper plenum with 3×157 elements is constructed, which can be used to quantitatively and accurately describe the complex flow process of the coolant flowing out from the reactor core and mixing in the upper plenum and redistributing to the hot legs. It is found that the mixing of the coolant flowing from the core is not thorough and complete in the upper plenum of pressure vessel. The coolant flowing from fuel assembly at different positions in the radial direction is with obvious corresponding relationship in the interface area between the upper plenum and the hot legs, and the difference of the radial power distribution of fuel assembly will inevitably lead to the formation of the thermal stratification of the coolant in hot legs.

Neural Network and Chaotic Characteristics Analysis of Water Film on Corrugated Plate Wall

Wang Bo, Chen Bowen, Tian Ruifeng, Ke Bingzheng, Li Ru, Lu Chuan

2020, 41(4): 60-63.

Abstract(287)

Abstract:
The corrugated plate dryer is an important steam-water separation device in marine nuclear power systems. The flow characteristics of the free falling liquid film on the corrugated board wall have a great effect on the steam-water separation efficiency of the dryer and the economic and safety indicators of the marine nuclear power plant. The liquid film thickness of different Reynolds number is measured by the Plane Laser Induced Fluorescence (PLIF) technique. Based on the small data volume method, the maximum Lyapunov exponents of liquid film thickness time series under different experimental conditions are calculated. The chaotic characteristics of the liquid film on the corrugated plate wall and phase space reconstruction is analyzed and performed, respectively. The BP (back propagation) neural network which has the advantage of solving nonlinear problems is applied to predict the liquid film thickness, the prediction model of single hidden layer BP neural network is established, and the nonlinear characteristics of the thickness of the free liquid film is analyzed. Results reveal that the maximum Lyapunov exponent is positively correlated with the liquid film Reynolds number. The liquid film volatility which is increased by the isolated peaks generated in the large Reynolds number region are coupled with gravity and the superposition effect of different liquid films, which makes the chaotic characteristics of the liquid film more obvious.

Experimental Study on Spread of Droplet Impacting on Dry Inclined Wall

Chen Bowen, Wang Bo, Tian Ruifeng, Li Ru

2020, 41(4): 64-69.

Abstract(302) PDF(155)

Abstract:
High-speed photography and pixel analysis methods are used to study the droplets impact on the inclined dried wall. The effects of different inclined angles and Weber number (We) on droplet splashing and spreading are analyzed. It is concluded that the increasing of the inclined angle is beneficial to suppress the droplet splashing; when the We is constant, the front spreading factor of the droplet increases with the increasing of the inclined angle, while the back spreading factor decreases with the increasing of the inclined angle; the front and back initial spreading velocity of the droplet is larger than the impact velocity of the droplet, and decreases with the increasing of the inclined angle.

Effect of Drainage Hole on Separation Efficiency of Swirl-Vane Separator

Xu Han, Lu Mingchao, Xiong Zhenqin, Zu Hongbiao, Gu Hanyang, Xie Yongcheng

2020, 41(4): 70-75.

Abstract(138)

Abstract:
In order to improve the separation efficiency of the swirl-vane separator, the swirl-vane steam separator with normal triangular drainage holes was built and studied by experimental and numerical methods using air-water mixture. The proportion of the separated water in each row of drainage holes and the effect of the drainage hole location on the separation performance were analyzed. The results show that when the superficial velocity of air is low (≤14.4 m/s), the proportion of the separated water in each row of drainage holes decreases at first and then increases with the increasing of height. When the superficial velocity of air is high(>14.4 m/s), the proportion of the separated water in each row of drainage holes decreases with the increasing of height. The study compares the separation efficiency of the swirl-vane separator when the drainage hole is located at 0.8D (the distance of the drainage hole from the swirl vane is 1 times of the inner diameter), 1.0D and 1.2D. When the superficial velocity of air is low (<12.9 m/s), the separation efficiency of the separator with the drainage hole at 1.2D is the highest. While the superficial velocity of air is high(≥12.9 m/s), the separation efficiency is the highest when the drainage hole is located at 0.8D.

Experimental Study on Wake of Single Vapor Bubble under Subcooling Condition in a Narrow Channel

Zhang Liqin, Huang Yanping, Zan Yuanfeng, Wang Junfeng

2020, 41(4): 76-78.

Abstract(195)

Abstract:
The wake of single vapor bubbles is a key factor affecting the flow regime formation and the evolution of vapor-water two phase flows in narrow channels. The wake of single vapor bubbles were examined under subcooling conditions in a narrow rectangular channel using Particle Image Velocimetry and post processed software Insight3G. Results showed that the wake structures were asymmetric with the vortex shed alternately in bubble rising process. The larger the bubble is, the stronger and the longer the vortex and the vortex strength are. As a result, the wake of large-diameter vapor bubble has a stronger effect on the surrounded bubbles become stronger. When the channel gap decreases, the strength and fluid velocity of the wake weaken, so as to the influence on the surrounded bubbles. Wake characteristics of single vapor bubbles can provide basic support for two phase flow model construction in narrow channels.

Experimental Study of Flow-Induced Acoustic Resonance in Side Branch on Large Diameter Pipeline

Huang Chao, Zhang Kai, Xiao Yao, Li Junlong, Zu Hongbiao, Gu Hanyang, Xie Yongcheng

2020, 41(4): 79-84.

Abstract(276)

Abstract:
The design of the safety valve branch pipe of steam pipeline in the nuclear power plant is prone to cause the acoustic resonance phenomenon of the coupling of the flow field and the sound field. Aiming at the acoustic resonance phenomenon caused by the side branch, this paper conducted an experimental study on the effect of the structure of the side branch on the acoustic resonance. The experimental section consists of the main pipe with the diameter of 305 mm and the side branches with the diameters of 64 mm and 44 mm, and the experimental medium is air. The effects of different lengths, diameters and side-by-side structures on the acoustic resonance intensity, characteristic frequency and flow characteristics were studied at the flow rate of 10~65 m/s experimentally. The experimental results show that the second-order hydraulic mode features are significant on the large diameter pipeline, and the St-number of its acoustic resonance peak is about twice the St-number of acoustic resonance peak for first-order hydraulic mode. The increase of the length of the side branch significantly reduces the intensity of the acoustic resonance. The decrease of the diameter of the side branch suppresses the occurrence of the second-order hydraulic mode, changes the flow characteristics and reduces the acoustic resonance intensity. Compared with the single structure, the tandem structure has a significant increase in the acoustic resonance intensity and the flow velocity region of the acoustic resonance with the first-order hydraulic mode and the first-order acoustic mode. Two branch pipes of tandem structure are similar in characteristics.

Fretting Friction Wear Behavior of Zircaloy and Ni-Based GH4169 Alloy

Gao Wen

2020, 41(4): 85-90.

Abstract(335)

Abstract:
The fuel rod tubes are disturbed under the condition of micro-vibration when the coolant flow pasts the surface of the fuel rod tubes from the bottom to the top during the operation of nuclear reactors, which can cause the fretting wear at the contact area of the frame spring and the tube cladding, and thus radio-active products might be leaked out under certain serious conditions, which will result in the shutdown of nuclear reactors, influencing its safety. The fretting wear behavior of Zirconium alloy Zr-4 and N36 mated with Ni-based GH4169 alloy was investigated in this work. The worn surfaces were examined with scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS) and 3D Microscopy. The obtained results showed that the friction efficient increases with the load increasing, and the Zr-4/Zr-4 has the maximum friction efficient, while the GH4169/N36 has the lowest friction efficient. Meanwhile, the preliminary oxidation has great influence on the friction behavior, and the friction coefficient of the pre-oxidized samples is higher than that of the non-pre-oxidized.

Measurement of Elastic Modulus of Zirconium Alloy Oxide Films

Zhang Junsong, Long Chongsheng, Xiao Hongxing, Liao Jingjing, Wei Tianguo

2020, 41(4): 91-95.

Abstract(227)

Abstract:
The internal stress of the oxide film has an important influence on the corrosion behavior of the zirconium alloy, and the elastic modulus is the key physical parameter for calculating and analyzing the internal stress of the oxide film. Due to the difficulty of testing, the elastic modulus of the zirconium alloy oxide film is usually estimated based on the data of bulk materials. In this paper, the elastic modulus and the hardness of the oxide film of the zirconium alloy under various conditions was analyzed by nano-indentation tests. It was found that the elastic modulus and hardness of the oxide surface and oxide cross section are different. Compared with the alloys corroded in pure water, the elastic modulus and hardness of the oxide film are smaller for the alloys corroded in lithium-containing water.

Nonlinear Analysis and Performance Evaluation of Prestressed Concrete Containment Structure under Severe Accident Condition

Jin Song, Li Zhongcheng, Lan Tianyun, Dong Zhanfa, Gong Jinxin

2020, 41(4): 96-100.

Abstract(196)

Abstract:
As the important leak-tight barrier in nuclear power plants, the prestressed concrete containment structure is essential to maintain the safe operation of nuclear power plants and ensure the safety of personnel. Based on the sequentially coupled thermal-stress method, nonlinear finite element analysis of the containment structure under severe accident loading is carried out in this study. Thermal and internal pressure effect is considered, and the displacement and strain response at current zone and singularity zone of the containment structure are studied in detail. Results indicate that the displacement response in the singularity zones of the containment structure along the thickness direction has the most significant difference, while that in the current zones are relatively small. Leakage failure mode of the containment is controlled by the location of the equipment hatch, and the break failure mode is controlled by the location of the cylinder wall. 50% percentile and 95% percentile pressure capacity corresponding to the dominant leakage mode of the containment is 1.266 MPa and 1.072 MPa, respectively, and 50% percentile and 95% percentile pressure capacity corresponding to the dominant break failure mode is 2.224 MPa and 1.883 MPa, respectively. The prestressed concrete containment analyzed in this paper meets the requirement of minimum margin not less than 2.5.

Analysis of Inspection Methods for Power Spectral Density in Seismic Qualification Test of Equipment in Nuclear Power Plants

Sun Yugang, Chu Meng, Ding Zhenkun, Yuan Fang

2020, 41(4): 101-104.

Abstract(225)

Abstract:
In order to meet the power spectral density (PSD) requirements of the input motion in the seismic qualificaiton test of equipments for nuclear power plants, the inspection method of PSD is analyzed and evaluated in this paper, based on the study on the background of the standards and the target PSD algorithm and the comparative analysis of typical examples. The results show that the most intuitive method of PSD check is to compare the PSD for input motion with the target PSD. According to the accuracy and conservativeness of the results of different types of methods for developing target PSD, it is recommended to use the synthetic time-history method in Appendix B of U.S. NRC SRP 3.7.1(2014) to calculate the target PSD. Although this method is generally applicable to the site-specific response spectra of nuclear power plants, it is also applicable to the test response spectra (TRS) for seismic qualification by replacing the target response spectra with TRS in the process of the synthetic time history. When using the proposed method to calculate the target PSD, the PSD check for the seismic identification input motion should be consistent with SRP 3.7.1, which envelops 70% of the target PSD between 0.3Hz and the upper bound frequency of the target response spectra.

Fragility Research for Storage Batteries of High Temperature Gas-Cooled Reactor

Jiang Zhuoer, Zhao Jun, WangHaitao, Shi Li, Wang Xiaoxin, Sun Weidong

2020, 41(4): 105-110.

Abstract(198)

Abstract:
For the verification of the electrical safety of nuclear power plants when seismic external events occur, it is required to conduct the seismic qualification test for the storage batteries. In this paper, with the seismic qualification test data and engineering experience of the storage batteries, we focus on the Class 1E storage batteries of high temperature gas-cooled reactor (HTR). The identification and the quantification of the fragility variables are then made for fragility curves and the high confidence of low probability of failure (HCLPF) capacity through the methods of fragility computations based on dynamic testing. The results show that the HCLPF capacity of uninterrupted power system Class 1E storage batteries are much higher than the design basis earthquake level of HTR.

Effects of Spatial Coherency on Seismic Response of Nuclear Island Structures Considering Soil Structure Interaction Analysis

Jiang Wei, Qu Yunguang, Xu Zhengyu

2020, 41(4): 111-115.

Abstract(124)

Abstract:
In order to study the effects of spatial coherency on the seismic response of nuclear island structures considering soil structure interaction analyses, in this paper, ACS-SASSI software was used to analyze the soil structure interaction of a typical PWR reactor building. Considering different soil layers and elevations, the effects of spatial coherency on the seismic floor response spectrum of nuclear island structures were evaluated. The results show that at high frequency the floor response spectrum will be reduced by 10%~70% for firm rock and Upper Bound Soft to Medium site. For soft soil sites, it will be reduced by 10%~40%. Thus, the floor response spectrum is conservative at high frequency and is not safe at low frequency without considering the spatial coherency.

Study on HRA Method in Seismic PSA of Nuclear Power Plants

Wang H, ing

2020, 41(4): 116-121.

Abstract(284)

Abstract:
External event Human Reliability Analysis (HRA) is different from the HRA in the internal event level PSA. Therefore, based on the investigation of the development history and method defects of the HRA of external events, the technical standards of the reliability analysis of external events at home and abroad are studied, and the human events in the seismic PSA of a nuclear power plant are taken as the research object. The method study illustrates the difference between the seismic HRA and the internal events HRA, the process of the seismic human response, and the analysis method of the human reliability under the earthquake, so that the analysis results obtained are closer to the actual situation of the external event, and it can provide a technical basis for the development of external event PSA.

Reliability of Passive Residual Heat Removal System Based on PSA

Zhao Xinwen, Guo Haikuan, Cai Qi, Deng Chunrui

2020, 41(4): 122-127.

Abstract(268)

Abstract:
Currently the equipment reliability has been studied when the probabilistic safety assessment (PSA) of advanced nuclear reactor is used to analyze the reliability of the passive system. However, the physical reliability has not yet been considered in PSA. On the other hand, the study of equipment reliability has been focused on all the start-up equipment rather than on the equipment in operation when the equipment reliability and physical reliability are discussed in a holistic fashion. In view of the aforementioned issues, this paper analyzed the equipment reliability sensitivity of the system based on the AP1000 passive residual heat removal system（PRHRS）subjected to the loss of normal feedwater supply. The system reliability is integrated into PSA model by self-developed synthesis method, which takes into account the demand failure of active equipment and the operation failure of passive equipment, and analyzes the sensitivity of system equipment reliability. The results show that the spectrum of the accident sequence obtained by the comprehensive method is more real and comprehensive when it is used to analyze the reliability of PRHRS, which is superior to the traditional method.

Assessment Model of Operator’s Implementation Reliability for Portable Equipment in Nuclear Power Plants

Chen Shuai, Zhang Li, Qing Tao, Lin Yaozu, Li Linfeng, Chen Chao

2020, 41(4): 128-134.

Abstract(259)

Abstract:
Portable equipment acts as necessary facilities for mitigating severe accidents, of which the accessing process are mostly complex in nuclear power plants. In order to analyze the reliability of nuclear power plant operators when operating portable equipment, this paper studies the characteristics of the implementation of improvement items added after Fukushima Nuclear Accident, proposes an assessment model based on Human Error Mode and Effect Analysis by defining human error probability, human error effects, human error recovery probability as risk factors, and combines experts evaluation with fuzzy linguistic theory. A case study of mobile power supply task under Station Black Out accident is given to get the importance ranking of human error modes as well as reasonable risk views, through which the feasibility of the model is verified.

Study on Radionuclide Migration Path and Concentration Distribution in Coastal Waters of Haiyang Nuclear Power Plant in Winter and Summer

Li Zichao, Zhou Tao, Si Guangcheng, Zhang Boya

2020, 41(4): 135-140.

Abstract(158)

Abstract:
In order to formulate the emergency response plan for the nuclear leakage accident in Haiyang Nuclear Power Plant, it is necessary to quickly predict the migration path and concentration distribution of the radio-nuclides in coastal waters. Firstly according to the real-time meteorological data, the radionuclide diffusion model of the nuclear power plant was established based on the methods of Lagrange and Euler respectively, and then the reliability of the model was verified. Secondly the radionuclide migration path and concentration distribution in summer and winter were analyzed. The results show that, due to the influence of the southeast wind and the ocean current in summer, the radio-nuclides of Haiyang Nuclear Power Plant migrate mainly to the northwest along the coastline. Due to the northwest wind and the ocean current in winter, the radio-nuclides migrate mainly to the east along the coastline, and then migrate to the southeast rapidly. In winter and summer, the radionuclide concentration in the coastal waters of the nuclear power plant decreases by about 9 orders of magnitude than the release amount after 5 days, and the radionuclide concentration decreases by about 10 orders of magnitude than the release amount after 10 days.

Experimental Study of Debris Bed Relocation in Sodium-Cooled Fast Reactor by Bottom Gas-Injection Method

Teng Chunming, Zhang Bin, Shan Jianqiang, Zhang Xisi, Cao Yonggang

2020, 41(4): 141-147.

Abstract(217)

Abstract:
In order to simulate the transient process of the debris bed relocation behavior of sodium-cooled fast reactor (SFR), a large number of debris bed relocation experiments were carried out by bottom gas-injection method to study the effects of the experimental factors, such as particle properties, gas flow rate, gas injection location and horizontal flow rate, on the debris bed relocation behavior. In general, the larger gas-injection flow rate, the larger horizontal flow rate and the concentrated gas-injection near the center can promote the relocation of the debris bed. While the larger particle size, the irregular particle shape, the larger particle density and the smooth particle surface can hinder the relocation of the debris bed. The relocation characteristics of the mixed size particles are between the uniform size particles we used for mixing.

Research on Spatial Effect Correction Method of Xi’an Pulsed Reactor Reactivity Measurement

Zhao Wei, Zhang Liang, Wang Baosheng, Yuan Jianxin, Zhang Qiang, Zhu Guangning

2020, 41(4): 148-152.

Abstract(285)

Abstract:
Reactivity is an important physical parameter of the reactor. Adding the reactivity real-time monitoring function in Xi’an Pulsed Reactor can provide operators with the real-time responsiveness and trends, which is conducive to its safe operation. The inverse dynamic method is widely used in the reactivity measurement because of its good real-time performance, which is capable to measure any reactivity introduction. However, due to the movement of the control rod, the spatial distribution of the neutron injection rate is inconsistent, and the deviation occurs. In this paper, the inverse dynamic method and the static spatial effect factor are analyzed theoretically, and the corresponding calculation method is given. Taking Xi’an Pulsed Reactor as the research object, the three-dimensional response function of the detector is calculated by using Monte Carlo (MCNP) program, and the normalized nodal power density is calculated, thus the static spatial effect factor is obtained. Finally, the experiment of interpolating the control rod at different rod speeds is carried out on the pulsed reactor and the integral value curve of the control rod is obtained by processing the experimental data. The results show that it is necessary to correct the spatial effect. The value curve of the control rod is more stable and the error between the calculation result and the real value is smaller.

Effect of Island Underlying Surface on Radionuclide Atmospheric Dispersion under Nuclear Accidents

Xu Fan, Ma Yuanwei, Wang Dezhong, Wang Dingzhi, Wu Siyuan

2020, 41(4): 153-160.

Abstract(125)

Abstract:
In order to study the diffusion of airborne nuclides in marine environment after a nuclear accident, this paper took the island nuclear power plant as the research object, and simulated the underlying surface elements, wind speed changes and nozzle velocity changes of the island nuclear power plant by using computational fluid dynamics (CFD) simulation method, and verified the accuracy of CFD simulation by wind tunnel experiments. The results show that the mountain around the nuclear power plant has a significant effect on the maximum ground diffusion factor. When the height of the platform increases, the maximum ground diffusion factor decreases, which is closer to the nozzle. The increasing of the wind speed will cause the position of the maximum ground diffusion factor to shift back. As the nozzle velocity increases, the maximum diffusion factor on the ground becomes smaller and its position shifts back. Under the condition that the velocity of the nozzle changes, the CFD simulation value of the ground maximum diffusion factor has a multiple relationship with the calculated value of the Gaussian plume model.

Open Failure Analysis of Parallel Slide Gate Valves in Steam Entrance of Steam Pump for PWR Auxiliary Water Supply System

Chen Fei, Chen Shiyi, Meng Duiqiang

2020, 41(4): 161-165.

Abstract(154)

Abstract:
In view of the failure of ASG steam pump steam inlet isolation valve which cannot be opened as required in many domestic PWR nuclear power plants, the fault tree analysis method is used to analyze the root cause of the failure. It is determined that the root cause of the failure is too much friction of the packing or too little spring force. The failure is completely solved by replacing the graphite packing of the valve with Teflon packing.

Domestic Development of Nuclear Grade Continuous Valve Position Indicator

Luo Shihong, Wang Yuxiang, Guo Song, Yuan Shengyi, Tian Xiaoshuai, Zhang Donglin, Li Xiaozhong

2020, 41(4): 166-169.

Abstract(200)

Abstract:
Based on the measurement principle and the functional requirements of nuclear grade continuous valve position indicator, a continuous valve position indicator is developed by means of modular design, simulation analysis and verification test. The test results show that the nuclear grade continuous valve position indicator described in this paper can be applied to the valve position measurement of important valves in nuclear power plants.

Simulation Verification of SINBAD Shielding Benchmark Experiment

Huang Hao, Xie Qin, Yu Tao, Xie Jinsen, Chen Zhenping, Hou Chen, Su Shi

2020, 41(4): 170-173.

Abstract(386)

Abstract:
In order to verify the accuracy of the shielded benchmark questions and further improve the standard benchmark question bank in China, this paper adopts MCNP fine modeling and calculation method to evaluate and verify the accuracy of 3 representative benchmark questions in the SINBAD shielded benchmark question bank. Combined with the experimental results of the literature, it is shown that the simulation results can provide the data verification support to other independent software within the allowable error range, and the verification results can be included in China shielded benchmark database.

Research on Radiation Shielding Design Optimization for Ship Reactors Based SuperMC

Sun Yuanli, Song Zhihao

2020, 41(4): 174-177.

Abstract(212)

Abstract:
The weight of the radiation shielding seriously is with great effects on the maneuverability of the ship. It is important to improve the maneuverability of the ship by optimization of radiation shielding in a fixed shielding space and reducing the weight of the shielding. In this paper, the optimization methods of radiation shielding in a fixed space is studied based on SuperMC intelligent nuclear design module, and the shielding optimization design is carried out by using Savannah reactor to demonstrate the effectiveness of the method. The results show that the shielding weight of the optimized scheme is greatly reduced compared with the original design scheme, which proves that this approach provides a new technical means for guiding the material selection and shielding layout of ship reactor radiation shielding.

Optimization of Pre-Service Inspection Scheme for CEPR CRDM

Xiao Kaihua, Shao Chunbing

2020, 41(4): 178-180.

Abstract(242)

Abstract:
After the installation of the pressure housing of the control rod drive mechanism (CRDM) in CEPR units, it was found that there were unqualified samples in the certificate test of CRDM welding. In order to shorten the period for the replacement of CRDM and reduce the impact on the overall progress of the project, a new pre-service inspection strategy is put forward by studying the pre-service inspection specifications in the process of the replacement of CRDM pressure housing. The practice shows that the off-line pre-service inspection of all CRDMs is completed within 10 days by using the optimized pre-service inspection scheme, which is about 20 days ahead of the original plan; through ultrasonic and eddy current inspection of some installed CRDMs, it is found that the results of off-line and on-line inspection are consistent, and there is no accessibility problem in on-line inspections.

Accuracy and Responsiveness Test Study of Cu/Cu2O Oxygen Sensor in Oxygen Saturated Lead-Bismuth Eutectic

Xu Yuheng, Niu Fenglei, Zhang Yu, Zhao Yungan

2020, 41(4): 181-184.

Abstract(262)

Abstract:
Cu/Cu2O oxygen sensor was tested on the pool type experimental device for solid-phase oxygen control. Experimental device was cooled from 500℃ step-by-step to 300℃. 95%Ar+5%O2 was injected during cooling process. The test indicates that the Cu/Cu2O oxygen sensor shows good performance in terms of accuracy and responsiveness in the temperature range from 300℃ to 500℃. Oxygen sensors can respond to the changes of oxygen concentration caused by temperature change quickly. The relative deviation between the electromotive force (EMF) of the oxygen sensor and the theoretical EMF remains within ±3%, and the oxygen concentration relative deviation is maintained within ±10%. Oxygen sensor signal fluctuation in the test is within 1.7 mV.

Testing Method of Redundancy Function in Nuclear Safety DCS Based on Failure Mode Analysis

Qi Min, Wu Yao, Zhu Jian, Lyu Xiuhong

2020, 41(4): 185-190.

Abstract(270)

Abstract:
To solve the problems in current redundancy testing of nuclear safety Digital Control System (DCS), such as the lack of systemic technology, a testing method based on failure mode analysis was proposed for nuclear safety DCS redundancy function. The basic failure modes for redundancy testing were established through the analysis of the design mechanism, and were combined into different kinds of failure scenario. The overall testing method was then designed based on failure mode combinations and system state transition analysis under the failures, including the design of redundancy testing scenarios and testing environment. The validity of the testing method is validated through the successful application experience in a nuclear power station, which shows that the redundancy testing method proposed could effectively detect defects in product design and is valuable for the quality control of the redundancy function in nuclear safety DCS.

Neutronic/Thermal-Mechanical Coupling in Heat Pipe Cooled Reactor

Ma Yugao, Liu Minyun, Yu Hongxing, Huang Shanfang, Chai Xiaoming, Xie Biheng, Han Wenbin, Liu Yu, Du Zhengyu, He Xiaoqiang

2020, 41(4): 191-196.

Abstract(806)

Abstract:
The structural thermal expansion of the heat pipe cooled reactor will significantly affect the heat transfer and neutron physical transport process. A geometric updating strategy and a reactivity feedback method considering the significant expansion of the solid core is presented in this paper, and a neutron physics/thermal/mechanical coupling algorithm based on dynamic geometry is established. In the coupling, the change of the cross section caused by temperature, the change of the density of the materials and the change of the core size caused by thermal expansion are considered. Furthermore, the coupling analyses are applied to the MegaPower heat pipe cooled reactor. The convergence of the core power distribution and the radial power factor under different relaxation factors is analyzed. Results show that the neutron leakage of the side-channel increases due to thermal expansion, which generates negative reactivity feedback and yet aggravates the non-uniformity of the power distribution and the deterioration of the heat transfer. The radial power factor increases from 1.20 to 1.23 when considering the thermal-mechanical feedback. The peak fuel temperature increases by 11 K.

Development and Assessment of Improved Spacer Grid Model in Rod Bundle Subchannel

Liu Wei, Liu Yang, Li Jie, Peng Shinian, Jiang Guangming, Liu Yu, Du Sijia, Qiu Zhifang, Deng Jian

2020, 41(4): 197-202.

Abstract(317)

Abstract:
In order to improve the prediction accuracy of two-phase local parameters in the rod bundle subchannel, based on the distributed resistance method, an improved spacer grid model is established in this paper, with the appropriate expression of the friction forces and a detailed model of mixing vane. The Carlucci turbulent mixing model is used to calculate the turbulent mixing rate, and the blocking factor is introduced to consider the turbulent mixing effect caused by the spacer grid. Coupled with subchannel analysis code ATHAS, the improved spacer grid model is used to calculate and analyze the PWR subchannel and bundle tests(PSBT) data. The results indicate that the improved spacer grid model has higher accuracy in predicting the distribution of the void fraction and outlet temperature in rod bundle subchannel. It lays a foundation for the accurate calculation of local parameters and CHF prediction in the rod bundle subchannel.

Research on Overall Design of Operation Mode in PWR Nuclear Power Plants

Cui Huaiming, Zhou Jinman

2020, 41(4): 203-207.

Abstract(318)

Abstract:
Based on the grid demand and the construction cost, this paper chooses the proper control manner of the reactor power, and the functional requirement of the operation mode. The operation mode, the control system and load rejection are designed. Finally, the core power capability and relevant accidents are analyzed for the nuclear power plants adopting this operation mode. The results show that the nuclear power plants that adopt this operation mode is safe. The paper introduces the design process of ACP600 operation mode, and it proves that the design method of the operation mode introduced in this paper is correct and practical.

2020 Vol. 41, No. 4