核动力工程

Simulation Study on Fatigue Growth of Semi-Circular Axial Crack on Inner Surface of Core Barrel under Transient Loads

Shi Kaikai, Zheng Bin, Zheng Liangang

2021, 42(S2): 1-4. doi: 10.13832/j.jnpe.2021.S2.0001

Abstract(281) HTML (68) PDF(34)

Abstract:
In order to study the fatigue growth of inner surface cracks in the core barrel under transient loads, the change law of crack morphology of semi-circular axial crack (crack depth d is 10 mm and crack surface length L is 20 mm) on the inner surface of reactor pressure vessel core barrel under transient load is simulated by using the crack block analysis method in the Zencrack software, wherein, the fatigue crack growth rate model of core cylinder material is realized by subprogram. Through the simulation study, it is known that the subprogram realization of material fatigue crack growth rate model is an efficient way to analyze the simulation of crack fatigue growth; the stress intensity factor calculated by the crack block analysis method is similar to that calculated by the standard engineering analysis method;. For the axial cracks with a semi-circular initial state on the inner surface of the core barrel, the initial crack ovalization will occur during the fatigue growth of the crack under transient loads.

Analysis of Dynamic Characteristics of Water Level of Natural Circulation Steam Generator

Qiu Leilei, Zhang Xianshan, Wei Xinyu, Sun Peiwei

2021, 42(S2): 5-9. doi: 10.13832/j.jnpe.2021.S2.0005

Abstract(287) HTML (68) PDF(45)

Abstract:
Accurately obtaining the dynamic characteristics of natural circulation steam generator (SG) water level has important guiding significance for the design of SG water level control system. In this paper, based on the equations of mass conservation, energy conservation and momentum conservation of fluid, the sixteenth-order nonlinear dynamic model (N-M) of SG is established; based on the perturbation theory, the transfer function model (T-M) of SG is obtained by linearizing N-M; the same disturbance is introduced to the two models on the MATLAB/Simulink simulation platform, and the simulation results show that the transient results of the two models are consistent and can correctly reflect the water level dynamic characteristics of SG. The study of this paper is helpful to understand the mechanism of SG water level change and provide an analysis tool for the design of SG water level control system.

Development of Special Device for Cutting Irradiation Test Tube

Li Chengye, Wu Rui, Wang Wanjin, Wang Yajun, Liu Hao, Zhang Xianmeng

2021, 42(S2): 10-14. doi: 10.13832/j.jnpe.2021.S2.0010

Abstract(216) HTML (102) PDF(42)

Abstract:
In order to eliminate the harm of metal flying debris and the influence of abnormal deformation of cutting position in the process of underwater radial cutting of rod-shaped irradiation test pipe, we developed a rotary extrusion pipe cutting method; In this method, we invented a special cutting device for rotating rod bundle fuel test tube by using servo motor-worm gear drive structure, Linear Variable Differential Transformer (LVDT) displacement sensor control system and considering safety measures. The three-dimensional modeling and simulation analysis show that the device can ensure the radial cutting of the rod bundle test tube, the material at the notch deforms towards the tube, and there is no chip in the whole cutting process, which can ensure the safe underwater cutting of the test tube after irradiation.

A Study of Nitrogen-16 Isotope Transport Time Calculation in Steam Generator

Wu Ge, Li Donghui, Li Lei, Tian Yajing, Tang Chenhang, Su Tong, Hu Yu

2021, 42(S2): 15-19. doi: 10.13832/j.jnpe.2021.S2.0015

Abstract(126) HTML (70) PDF(21)

Abstract:
To calculate the transport time of ¹⁶N isotope from the damaged part of the heat transfer tube to the detection point in the main steam line in the steam generator (SG), this paper presents a calculation method of steady-state thermal hydraulic parameters of SG based on one-dimensional steady-state model and conservation equations of mass, momentum and energy. In this method, the finite difference method is used for discretization, and the drift flow model is used to process the two-phase flow in SG; On this basis, the ¹⁶N isotope transport time of SG with different power levels and different heat transfer tube leakage positions is calculated. By comparison, the results obtained by the calculation method proposed in this paper meet the engineering requirements and have been applied to HPR1000 nuclear power project.

Effect of Different Initial Temperatures of IRWST on Heat Exchange Performance of Passive Residual Heat Removal Heat Exchanger

Yan Linfeng, Wu Xinci

2021, 42(S2): 20-24. doi: 10.13832/j.jnpe.2021.S2.0020

Abstract(289) HTML (77) PDF(29)

Abstract:
CFD computations were conducted to investigate the effect of different initial temperatures of the AP1000 In-containment Refueling Water Storage Tank (IRWST) on the heat exchange performance of Passive Residual Heat Removal heat exchanger (PRHR HX), and to discuss possible benefits of adjusting the water temperature of IRWST. The standard $k - \varepsilon $ model is selected for the turbulence model and the SIMPLEC algorithm is selected for the correction of pressure and speed. The results show that reducing the initial temperature of IRWST can improve the heat exchange performance of PRHR HX. In addition, the initial temperature of IRWST is directly proportional to the average temperature at the tube bundle outlet. Furthermore, the relative temperature drop at the inlet and outlet of the tube bundle increases linearly with the decrease of the initial temperature of IRWST. On this basis, the following research and development directions are prospected, and the following design thought is put forward: after an accident, first increase the IRWST water temperature to reduce the thermal stress damage and fluid vibration of the core components, and then slowly reduce the IRWST water temperature to maintain the heat exchange capacity, so that it not only keeps the integrity of the reactor core, does not damage the reactor core components, but also can maintain the long-term cooling of the reactor core.

Dynamic Characteristic Analysis of Steam Bypass Exhaust System of Steam Turbine

Zhang Xianshan, Sun Peiwei, Wei Xinyu

2021, 42(S2): 25-28. doi: 10.13832/j.jnpe.2021.S2.0025

Abstract(309) HTML (135) PDF(37)

Abstract:
Based on the basic conservation equation of fluid, uses the critical flow model to calculate the steam flow at each outlet, and deduces and establishes the dynamic model of steam bypass exhaust system. In order to obtain the dynamic characteristics of the steam bypass exhaust system of steam turbine, a transient model is established on the MATLAB & Simulink simulation platform by using the mechanism method. By introducing a variety of boundary condition disturbances, the dynamic characteristics of the steam bypass exhaust system are determined. The transient response of steam flow and exhaust condenser valve disturbance is analyzed. The transient simulation results accord with the physical law. These results are of great significance to the subsequent design of steam turbine steam bypass exhaust control system.

Safety Analysis of China Fusion Engineering Test Reactor Helium Cooled Blanket

Zhou Bing, Wang Xiaoyu, Wang Yanling, Hu bo

2021, 42(S2): 29-32. doi: 10.13832/j.jnpe.2021.S2.0029

Abstract(389) HTML (101) PDF(34)

Abstract:
In order to evaluate the safety of helium cooled blanket system of China Fusion Engineering Test Reactor (CFETR) and optimize the blanket design, this study presents the initial events of accidents, and determines the accident list and design basis accidents based on the design characteristics of CFETR helium cooled blanket. The thermal hydraulic steady-state analysis of CFETR helium cooled blanket module is carried out by using the best transient estimation program RELAP5. The preliminary safety analysis of accident transient is carried out according to the requirements of design basis accident. The results show that the LOCA accidents in different positions should be focused on in the accident safety analysis of helium cooled blanket system; The in-vessel LOCA large break accident is analyzed and calculated, and the maximum temperature of the first wall and the maximum pressure in the vacuum vessel are given, which meet the requirements of the acceptance criteria.

Development of Automatic Inspection Device for Appearance of Spacer Grid

Wang Dan, Sheng Feng, Li Feng, Li Jiehua

2021, 42(S2): 33-36. doi: 10.13832/j.jnpe.2021.S2.0033

Abstract(273) HTML (163) PDF(33)

Abstract:
At present, the appearance inspection of spacer grid in China has always been manual visual inspection, which is low in inspection efficiency, easily leads to visual fatigue of inspectors, and even has the risk of false inspection and missed inspection. A set of automatic inspection device for the appearance of spacer grid is developed by using photoelectric, image acquisition and software processing technology, which realizes the automatic inspection of the appearance inspection items of spacer grid; improves the inspection level and efficiency of spacer grid, avoids the manual missed inspection as much as possible, and further reduces the risk of the operation of the reactor core fuel assembly. The inspection accuracy has reached the design goal. It has been successfully applied to the appearance inspection of spacer grid, running in good condition.

Implementation of Monte Carlo Load Balancing Based on Nearest Neighbor Algorithm

Cui Xiantao, Qiang Shenglong, Kuang Denghui, Yin Qiang, Zhang Wenxin, Liu Yuan, Wu Bin

2021, 42(S2): 37-40. doi: 10.13832/j.jnpe.2021.S2.0037

Abstract(263) HTML (96) PDF(38)

Abstract:
In order to solve the problem of low parallelism caused by particle load imbalance of Monte Carlo program at the end of iteration, this paper analyzes the advantages and disadvantages of different algorithms, and the nearest neighbor algorithm is used on MOI Monte Carlo program to solve load balancing. Finally, we adapt an example of ten million grid to test this algorithm, it reduces the computation time by at least 10%, and as the burnup deepens, the calculation time will decrease even further. The results show that the algorithm is effective for the problem of particle load imbalance.

Research of ALSTM-GPC in Coordinated Control System of Nuclear Power Plant

Deng Zhiguang, Qing Xianguo, Wu Qian, Zheng Xiao, Zhu Biwei, Zhu Jialiang, Lyu Xin

2021, 42(S2): 41-47. doi: 10.13832/j.jnpe.2021.S2.0041

Abstract(337) HTML (107) PDF(34)

Abstract:
Aiming at solving the problem of poor control effect of conventional proportion integration differentiation (PID) controller when dealing with complex system, combining the advantages of deep learning in feature extraction, regression prediction and predictive control in dealing with multivariable and strong coupling, a predictive model controller is built by ALSTM deep networks, the model takes one-dimensional time series signal of object as its input, completes feature extraction through LSTM. Then, by introducing the attention mechanism, the weight parameters of LSTM extraction features are optimized, the target features are screened and retained, and the redundant features are filtered to achieve accurate extraction of effective time series features. And then a generalized predictive control (GPC) is used as a rolling optimization controller, thus building attention long short-term memory-generalized predictive controller (ALSTM-GPC). Then, the simulation experiment was carried out in the nuclear power plant's double-input and double-output multivariable coordinated control system. Through a series of simulations such as set-point disturbance, internal disturbance and external disturbance, ALSTM-GPC controller is proven having better control effect than conventional PID control.

Study on a Data-Enabled Physics-Informed Reactor Physics Operational Digital Twin

Gong Helin, Chen Zhang, Li Qing, Cheng Sibo

2021, 42(S2): 48-53. doi: 10.13832/j.jnpe.2021.S2.0048

Abstract(342) HTML (478) PDF(100)

Abstract:
To realize the fast and accurate online calculation and to predict the operation behavior of nuclear reactors, a physics-informed data-enabled reactor physics operational digital twin is proposed, to achieve rapid and accurate calculation of physical fields such as fast and thermal neutron flux and power distribution in the core. The physics-informed property is achieved through a fast calculation model of neutronics based on model order reduction technology and machine learning; the data enabled property is realized through an inverse model based on the fast calculation model. The test of the design and operation data of HPR1000 reactor shows that the digital twin meets the engineering requirements in terms of time and accuracy, and has the potential for online monitoring applications in real engineering.

Study on Analysis and Optimization Design Method of Sealing Characteristics of Spring Metal C-ring

Jiang Lu, Li Hui, Zhang Ying, Shao Xuejiao, Zhang Liping, Fu Xiaolong

2021, 42(S2): 54-59. doi: 10.13832/j.jnpe.2021.S2.0054

Abstract(319) HTML (144) PDF(39)

Abstract:
To improve the analysis and design method of spring metal C-ring sealing characteristics, by establishing a fine analysis model, the compression and rebound characteristic curve is accurately simulated, and the effectiveness and correctness of the numerical method are verified by the test data; In addition, the sensitivity of structural parameters is systematically analyzed by test design method, and the interaction law of design parameters on sealing characteristics is deeply studied; the optimization model is established based on Multi-Island Genetic Algorithm (MIGA), and the optimization design of key parameters is studied. The results show that the simulation results of compression and rebound characteristic curve are in good agreement with the test data, and the numerical method has very good stability; The diameter of spring wire, the thickness of sealing layer and cladding layer have an important influence on the contact pressure characteristics of sealing surface, but the diameter of spring wire should not be too large or too small; Using the optimization model and method established in this paper, the optimal combination of key parameters can be obtained quickly and the sealing performance can be improved effectively.

Design of Fast Neutron Detector Irradiation Sample and Study of Its Temperature Characteristics

Jiang Tianzhi, Wei Zheng, Bao Chao, Liao Longtao, Sun Congjian, Yang Zhenlei, Li Jin, Lin Chao, Yu Heng, Lu Jiawei

2021, 42(S2): 60-64. doi: 10.13832/j.jnpe.2021.S2.0060

Abstract(302) HTML (82) PDF(45)

Abstract:
To achieve the purpose of monitoring the fast neutron fluence rate outside the nuclear power plant, the development of a miniaturized fast neutron detector based on zinc sulfide (ZnS) scintillator was carried out. To verify the neutron radiation resistance performance of the fast neutron detector, the radiation sample design of the fast neutron detector was carried out in combination with the characteristics of the irradiation device and the requirements of the irradiation components, and the Monte Carlo software and COMSOL Multiphysics 5.5 simulate and analyze the particle transport and overall temperature field distribution in a fast neutron detector based on zinc sulfide (ZnS) scintillator. The irradiation sample of the detector is designed iteratively, and the irradiation sample of the fast neutron detector of ZnS scintillator is designed. At the same time, based on the in-depth analysis of the simulation data, the temperature characteristics of various materials under the irradiation field are extracted, which provides theoretical guidance for the subsequent processing and manufacturing of fast neutron detector irradiation samples and the irradiation verification test of a single material.

Research on Radiation Source Term Characteristics with Large Fuel Rod Break of PWR Nuclear Power Plant

Jing Futing, Lyu Huanwen, Zhu Jianping, Gao Xilong, Huang Qianming

2021, 42(S2): 65-69. doi: 10.13832/j.jnpe.2021.S2.0065

Abstract(173) HTML (76) PDF(39)

Abstract:
During the operation of the reactor, the large break of the fuel rod will cause the release of fuel core materials and the obvious rise of the source term in primary coolant circuit, which will affect the safe operation of the reactor. In this paper, the fuel damage status is evaluated based on the measured value of the source term of a nuclear power plant. The analysis shows that there is a large break in the core fuel rod and the release of fuel, which is basically consistent with the damage inspection results after shutdown. Research shows that in the case of a large break in the fuel rod, the source term of the primary coolant circuit has the following characteristics: there is the continuous rise of ¹³⁴I source term in coolant; the activity spectra of typical fission products are similar to those of contaminated uranium; and there is no obvious iodine peak during power transient; and ²³⁹Np source term in coolant could be detected. These laws can be used to analyze the fuel damage of the reactor and help to identify the large fuel rod break in the reactor core.

Experimental Study on Vortex Shedding in Water Medium of Closed Three-Way Side Branch Pipe

Liu Shuai, Jiang Xiaozhou, Feng Zhipeng, Huang Xuan, Zhang Rui, Zhang Yixiong

2021, 42(S2): 70-76. doi: 10.13832/j.jnpe.2021.S2.0070

Abstract(130) HTML (71) PDF(18)

Abstract:
Through the study on the vortex shedding in the water medium of the closed three-way side branch pipe, and according to the vortex shedding mechanism of the three-way pipe, a test device for the closed three-way side branch pipe is established to verify the mechanism. The Particle Image Velocimetry (PIV) test method is used to obtain the cross-sectional streamlines and velocities at the three-way positions of the closed three-way side branch pipe under different flow conditions, and analyze the location and change characteristics of the vortex shedding of the three-way pipe. Tests have shown that due to the unique flow field structure of the three-way pipe, a pressure wave will be formed at the location of the three-way, and a vortex will be formed in the side branch pipe; a large velocity gradient occurs in the area where the fluid flows through the main pipe and the side branch pipe, resulting in the formation of a small vortex that is falling off at the front edge of the side branch pipe; due to the viscosity of the fluid, the fluid at the front edge of the side branch pipe will be driven by the fluid in the main flow area to flow downstream, creating a vacuum area at the front edge; the velocity fluctuation at the interface between the main pipe and the side branch pipe, the size of the vortex in the side branch pipe and the vortex shedding velocity at the front edge of the side branch pipe will increase with the increase of velocity.

Development and Verification of a Comprehensive Rod Bundle CHF Mechanism Model

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu

2021, 42(S2): 77-81. doi: 10.13832/j.jnpe.2021.S2.0077

Abstract(309) HTML (116) PDF(37)

Abstract:
In order to continuously and accurately predict departure from nucleate boiling (DNB) and dry-out (DO) type critical heat flux (CHF) under wide parameter range for rod bundle channel, a comprehensive rod bundle CHF mechanism model covering different types of CHF based on the CHF classification criterion in the bundle channel and the mechanism model of DNB type CHF for evaporation of superheated liquid layer and combined with the mature DO type CHF mechanism model that has been studied. The 5 × 5 full-length rod bundle CHF experimental data of Nuclear Power Institute of China (NPIC) is used to verify the CHF mechanism model of the comprehensive rod bundle. The results show that all predicted value/measured value (P/ M ) data of the comprehensive rod bundle CHF mechanism model are evenly distributed around 1, and the maximum relative deviation is within ± 22%, indicating that the developed comprehensive rod bundle CHF mechanism model can continuously and accurately predict the DNB and DO type CHF in rod bundle channel.

Research of Double-Heterogeneity Physical Boundary on Dispersed Particle-type Systems

Lou Lei, Chai Xiaoming, Yao Dong, Li Mancang, Chen Liang, Liu Xiaoli, Zhang Hongbo, Li Sinan, Tang Xiao, Zhou Nan

2021, 42(S2): 82-88. doi: 10.13832/j.jnpe.2021.S2.0082

Abstract(119) HTML (35) PDF(17)

Abstract:
Dispersed particle-type systems cannot be described with traditional neutronics calculation programs because of the double-heterogeneity (DH), and the direct use of Volumetric Homogenization Method (VHM) will bring reactivity calculation deviation. In this paper, by analyzing the calculation deviation of volumetric homogenization reactivity of dispersed particle fuel and different types of dispersed particle burnable poison and its relationship with optical length, it is proposed that the influence factors of calculation deviation shall be integrated into corrected optical length. The physical boundary of double-heterogeneity of dispersed particle-type system is proposed. When the corrected optical length is greater than 10⁻⁴, the reactivity calculation deviation of the volumetric homogenization method will be more than 100pcm, so the double-heterogeneity of the dispersed particle-type system needs to be taken into account.

Study of Non-proportionally Multiaxial Cyclic Deformation Behavior of Domestic 508-3 Steel at Different Temperatures

Tian Jun, Tang Yanjie, Zhang Liping, Fu Xiaolong, Kuang Linyuan, Zhang Ying, Jiang Lu, Li Hui, Liu Zhenyu

2021, 42(S2): 89-92. doi: 10.13832/j.jnpe.2021.S2.0089

Abstract(226) HTML (76) PDF(15)

Abstract:
Domestic 508-3 steel is the key material for the pressure vessel of pressurized water reactor. To study the non-proportionally multiaxial cyclic deformation behavior of domestic 508-3 steel at different temperatures, whole-life fatigue tests under multiaxial non-proportionally path with strain and stress controlled were performed on domestic 508-3 steel. Hourglass type and butterfly type non-proportionally paths were chosen in the tests, which revealed cyclic deformation characteristics and ratcheting behavior under different loading paths and different loading conditions. Results show that notable temperature dependent non-proportionally multiaxial cyclic deformation behavior of domestic 508-3 steel can be observed, and due to the effect of dynamic strain aging, the material exhibits cyclic hardening or stable characteristic at 350℃.

Numerical Simulation Study of Hold-down Force of Leaf Spring Hold-down System of Fuel Assembly in Reactor Environment

Wang Haoyu, Qin Mian, Pu Zengping, Zhu Fawen, Ran Renjie, Miao Yifei, Yuan Pan, Liu Menglong

2021, 42(S2): 93-98. doi: 10.13832/j.jnpe.2021.S2.0093

Abstract(153) HTML (114) PDF(29)

Abstract:
In order to accurately predict the hold-down force during the design lifetime of the leaf spring hold-down system of nuclear fuel component, considering the plastic model and creep model of the leaf spring material under fast neutron irradiation, numerical simulation method for the hold-down force of the leaf spring hold-down system in the reactor environment was established based on the fine structure model. By comparing the calculation results with the calculation results of the fuel assembly leaf spring hold-down system analysis software HOFA, the rationality of the numerical simulation method is verified (the maximum error of hold-down force is 8.83%). The influence comparison results show that the maximum hold-down force is reduced by 4.34% when radiation creep is considered; When the cycle length increases, the hold-down force at the end of each cycle increases slightly.

Application of HALT Test Method in Development of Digital Instrument and Control Equipment

Zhang Weichuan, Zhao Hui, Qin Fan, Zhao Yang, Qing Xianguo, Chen Jie, He Xiaopeng

2021, 42(S2): 99-103. doi: 10.13832/j.jnpe.2021.S2.0099

Abstract(140) HTML (103) PDF(13)

Abstract:
In order to solve the following questions that digital instrument and control equipment cannot avoid its potential failure mode in the design stage, traditional development and test methods cannot simulate the environment profiles and task profiles of the equipment in the whole life cycle, some failure modes of the equipment cannot be exposed in the development stage, this paper proposes to carry out a HALT test on a certain type of digital instrument and control equipment after adaptive simplification. Based on the statistical results of a large number of similar tests abroad and the requirements of equipment environmental adaptability, the criterion of stress level range to measure the necessity of fault improvement is formulated; At the same time, combined with the experience feedback of similar instrument and control equipment, the test scheme is adjusted; Finally, combined with the existing BIT testing methods of the target equipment, a real-time condition monitoring and fault analysis tool is developed to complete the real-time monitoring and auxiliary failure analysis in the test process. The experimental results show that the customized test method can effectively expose and reproduce the potential faults of this typical digital instrument and control equipment, and the method can be applied to other similar instrument and control equipment with strong operability.

Reactor Startup Characteristics of Heat Pipe Cooled Reactor with Multiple Feedback Mechanism

Zhong Ruicheng, Ma Yugao, Deng Jang, Liu Yu, Chai Xiaoming, Wang Jiageng

2021, 42(S2): 104-108. doi: 10.13832/j.jnpe.2021.S2.0104

Abstract(431) HTML (539) PDF(85)

Abstract:
The heat pipe-cooled reactor (HPR) has a high temperature solid core, so during the startup, it will experience large temperature rise and thermal expansion. Besides, the state of working fluid in the heat pipe will change, which would bring complex reactivity feedback. In order to study the startup characteristics of HPR under feedback mechanism, taking the megawatt HPR MegaPower designed by Los Alamos National Laboratory (LANL) as the research object, using the HPR transient analysis code HPRTRAN, the startup conditions of MegaPower are calculated with three startup modes, and the startup characteristics of each mode are studied. The results show that, relatively large power fluctuation occurs during the startup process of HPR, there may be a "temperature platform" that is not conducive to the startup, and that rotating control drum at intervals is a safer and more practical startup scheme than rotating control drum continuously.

Application of DMRM on LOCA Integrated Effect Experiment

Huang Tao, Li Zhongchun, Sun Wei, Deng Jian, Ding Shuhua, Liu Yu, Wu Dan, Qian Libo, Shen Yaou, Du peng, Wu Zenghui

2021, 42(S2): 109-112. doi: 10.13832/j.jnpe.2021.S2.0109

Abstract(109) HTML (38) PDF(25)

Abstract:
China Nuclear Power Research and Design Institute independently developed the loss of coolant accident (LOCA) evaluation model of Deterministic Model-Realistic Method (DMRM), and developed the advanced reactor system analysis code ARSAC based on the model. In this paper, this method is applied to large-break LOCA conditions to verify the correctness and rationality of the method. In the verification process, firstly, a conservative bench system model is established through sensitivity analysis; On this basis, the input conditions (system pressure, safety injection tank water temperature, etc.) that have a great impact on the peak cladding temperature (PCT) are randomly sampled to form 59 groups of working conditions for calculation and analysis; These working conditions are analyzed by ARSAC to obtain the final results, and the final results are compared with the experimental data. The results show that the pressure and other parameters calculated by the code are consistent with the experimental data, which proves the correctness of the ARSAC calculation; the PCT temperature obtained by DMRM is significantly higher than the experimental value, which proves the conservativeness of DMRM.

Verification and Validation of the Advanced Neutronics Component Program KYLIN V2.0 Based on B&W Critical Experiment Benchmark Task

Zhao Chen, Peng Xingjie, Zhang Bin, Chai Xiaoming, Li Qing

2021, 42(S2): 113-118. doi: 10.13832/j.jnpe.2021.S2.0113

Abstract(288) HTML (62) PDF(32)

Abstract:
KYLIN V2.0 is an advanced neutronics component program independently developed by the Nuclear Power Institute of China. The program uses subgroup resonance method, characteristic line method and Chebyshev rational approximation method to calculate resonance, transport and burnup, so as to provide multi-group cross-section library. This paper uses B&W critical experiment benchmark task to verify and validate the KYLIN V2.0 program, including 21 examples of B&W1484 critical benchmark task and 23 examples of B&W1810 critical benchmark task. The calculation results show that the calculation results of KYLIN V2.0 program are in good agreement with the measured data, which verifies that KYLIN V2.0 program has good calculation accuracy and non-uniform calculation ability.

Analysis of the Models in CISER2.0 Code

Liu Lili, Zhang Ming, Deng Jian, Yu Hongxing, Chen Liang, Xu Youyou, Luo Yuejian

2021, 42(S2): 119-123. doi: 10.13832/j.jnpe.2021.S2.0119

Abstract(143) HTML (65) PDF(20)

Abstract:
This paper first analyzes the models of CISER2.0, a code of In-vessel retention (IVR) strategy effectiveness analysis. The CISER2.0 code consists of four three-layer melting pool models: Esmaili & Khatib-Rahbar model, Seiler model, Salay & Fichot model and self-developed model. It is found that compared with Esmaili & Khatib-Rahbar model, Seiler model is more conservative; Although the Salay & Fichot model is based on thermodynamic theory in calculating the composition of oxide layer and heavy metal layer, the method of user hypothesis is adopted in determining the composition of light metal layer, and it is considered that the light metal layer is formed automatically at the top of the melting pool; The self-developed melting pool structure model calculates the structure of the melting pool based on the accident process. Compared with the Salay & Fichot model, it can automatically calculate the composition of the light metal layer. In this paper, taking the 1000MW advanced reactor as an object, the morphology of the melting pool formed in the lower chamber after the accident of small break in the cold section of the main pipe is calculated based on the different layering models of the melting pool in the code. However, the content of stainless steel in the melt of this research object is too small to form a three-layer structure that meets the Seiler model. In addition, the heat flux distribution on the outer wall of the pressure vessel is given according to the calculated three-layer melting pool structure. The results show that the difference of melt composition in the corresponding layer of each melting pool leads to the difference of heat flux distribution on the outside of the pressure vessel. Even if the corresponding layer thickness of Esmaili & Khatib-Rahbar model and Salay & Fichot model is set to be basically the same, the difference of heat flux distribution between them is large. At the same time, different from the previous three models, the self-developed model also gives the transient heat flux of the outer wall of the pressure vessel when the melt falls down the chamber.

Study on CUDA-based Heterogeneous Parallel for Advanced Assembly Neutronics Program

Zheng Yong, Lu Wei, Ma Yongqiang, Cui Xiantao, Guo Fengcheng, Ma Dangwei, Tu Xiaolan

2021, 42(S2): 124-129. doi: 10.13832/j.jnpe.2021.S2.0124

Abstract(241) HTML (86) PDF(25)

Abstract:
To improve the calculation performance of the advanced assembly neutronics program KYLIN-II when handling the complicated boundary condition, the current paper studied the heterogeneous parallel in the KYLIN-II program based on heterogeneous parallel technology of programmable graphics card, implemented massive thread parallel computing of resonance and transport modules, and reduced the number of atomic operations of heterogeneous parallel program by optimizing iterative strategies. In order to verify the calculation accuracy and acceleration effect of heterogeneous parallel programs, calculations were carried out for test examples such as AFA3G super assembly, hexagonal plate fuel assembly and multilayer sleeve fuel cell. The results indicate that heterogeneous parallel programs will not affect the accuracy of calculation results. The KYLIN-II program after heterogeneous parallel of a single graphics card can achieve an acceleration ratio of more than 10 times. Optimizing the iterative process can effectively reduce the calculation time. Compared with the traditional multi-core parallel mechanism based on central processing unit (CPU), heterogeneous parallel of graphics card significantly reduces the economic cost of large-scale parallel of KYLIN-II program, which can be used as the direction of further parallel optimization of KYLIN-II program.

Performance Analysis and Irradiation Test of HPR1000 CF3 Fuel Rod

Zhang Kun, Jiao Yongjun, Chen Ping, Xing Shuo, Li Guoyun, Pu Zengping, He Liang, Fan Hang, Wang Yanpei, Qiu Bowen, Hui Yongbo

2021, 42(S2): 130-135. doi: 10.13832/j.jnpe.2021.S2.0130

Abstract(316) HTML (160) PDF(86)

Abstract:
CF3 fuel assembly is an independently developed large PWR fuel assembly in China, which would be loaded in HPR1000, and support HPR1000 outlet. In order to meet the requirements of HPR1000 reactor, the irradiation test of the CF3 pilot assembly was carried out on the commercial reactor. The fuel element in-reactor data was obtained through the pool-side inspection, and the model was developed based on the in-reactor data. The performance analysis program suitable for the CF3 fuel element was established. Based on this, the performance of CF3 fuel element is predicted according to the demand of HPR1000 reactor. The analysis results show that the CF3 fuel element meets the design requirements of HPR1000 reactor.

Study of Online Operation Hazard Analysis Method of Nuclear Power Plant

Li Wei, Qing Xianguo, Zhao Yang, Zhang Zhongyue, Dai Kailei

2021, 42(S2): 136-139. doi: 10.13832/j.jnpe.2021.S2.0136

Abstract(204) HTML (94) PDF(28)

Abstract:
In order to online evaluate whether the operator’s operational actions is harmful, so as to avoid human error and ensure the safe operation of nuclear power plant, an operation hazard analysis method based on system goal-function model is presented, which determines the goal and function state according to the operation parameters of nuclear power plant. According to the cause-effect dependency between goal and function, qualitative reasoning is used to analyze the effects of operational actions on system goals. The results show that the proposed method can prompt the potential harmful consequences of the operator's operational actions on the specific system goals and functions. The method has good robustness and high real-time performance, and is helpful to reduce and avoid human errors.

Research on Equipment Grading Based on System Reliability Allocation

Chen Baowen, Zhang Dan, Zhu Dahuan, Jiang Xiaowei, Fang Hongyu, Mi Zhengpeng, Cheng Ruiqi, Zhong Mingjun

2021, 42(S2): 140-145. doi: 10.13832/j.jnpe.2021.S2.0140

Abstract(253) HTML (140) PDF(35)

Abstract:
Traditional equipment grading method is based on the realization of safety function, which can not take into account the economy of nuclear power plant (NPP). A new method of equipment grading based on the system reliability allocation and system importance and nuclear safety importance factor is proposed for solving this problem and achieving high safety and high economy of NPP. Taking the condensate water supply system of floating nuclear power plant as an example, the equipment reliability grading research is carried out. The quantitative results of reliability considering system importance and nuclear safety importance are given. The results can be used as the basis for equipment grading, and can be used for maintenance, fault diagnosis and health management of power plant equipment.

Research and Development of Active Vibration Controller Based on Shared Memory Technology

He Hongyang, Yang Heng, Zhou Yu, Zhang Kun, Li Pengzhou, He Ziang

2021, 42(S2): 146-149. doi: 10.13832/j.jnpe.2021.S2.0146

Abstract(194) HTML (57) PDF(21)

Abstract:
The vibration of the pump and pipeline of marine power plant is transmitted to the main body through the base or the support, which results in serious troubles. The controller is the core of the active vibration control (AVC) system. Aiming at the shortcomings of the traditional controller, an active vibration controller based on shared memory technology is proposed. Both analog-to-digital (A/D) conversion acquisition module and digital-to-analog (D/A) output module contain micro-control units (MCU). Field programmable gate array (FPGA), digital signal processing (DSP) and MCU in real time read data from or write data to the shared memory which is structured in superblocks and data blocks. At the same time, the maximum sampling frequency and the system delay error of the controller are verified, and the error channel identification test is carried out on the simulation bench. The results show that the controller can meet the requirements of low frequency line spectrum control, improve the data transmission efficiency, and ensure the real-time performance of the system.

Application Study of Cloud Computing in Stress Analysis and Optimization Design of Nuclear Pipe

Jiang Shenghan, Fan Kai, Wang Yongchao, Wang Shuai, Zhang Kun

2021, 42(S2): 150-153. doi: 10.13832/j.jnpe.2021.S2.0150

Abstract(149) HTML (51) PDF(33)

Abstract:
In order to solve the problem that it is difficult to obtain the optimal solution in the stress calculation of nuclear pipe, this research combines cloud computing with pipe stress analysis progress, and develops the preprocessing method of nuclear pipe by setting the limiting conditions, the pipe stress analysis flow of cloud computing pre-processing was put forward. Meanwhile, the stress qualification model was established by support vector machine to save computing time based on feature analysis and data modeling. A calculating example of minimizing the number of spring supports and hangers and reasonable positioning of the supports and hangers for nuclear pipe shows that this method can efficiently and rapidly solve the optimal solution of stress analysis for nuclear pipe under given conditions.

Research on Automatic Test Technology for Response Time of Multi-Controller Based on Synchronous Triggering Mechanism

Tian Haowen, He Xianjian, Wen Jing, Chen Zhao, Yuan Shengjun

2021, 42(S2): 154-158. doi: 10.13832/j.jnpe.2021.S2.0154

Abstract(250) HTML (74) PDF(25)

Abstract:
Existing test methods for response time based on oscilloscope generally have many defects, including limited test sample base, large test data error, and low test efficiency, which do not meet the requirements of test preciseness and current industrialization development. To solve these problems, this research, based on the distributed test platform of multi-controller, designs an automatic test technology for response time based on synchronous triggering mechanism, through analyzing the existing test technology for response time of Distributed Control System (DCS) in nuclear power plant. Practice has proved that this automatic test technology for response time has technical advantages and practicability in the field of industrial control, and has high popularization value.

Research on Start-up and Shutdown and Key Design and Maintenance Strategy of Research Reactor Based on System Engineering

Fu Jingqian, Miao Zhuang

2021, 42(S2): 159-164. doi: 10.13832/j.jnpe.2021.S2.0159

Abstract(452) HTML (77) PDF(34)

Abstract:
In order to solve the problems of the traditional text-based research reactor start-up/shutdown plan design method due to the numerous research reactor systems and the complexity of mutual matching interfaces, there are problems such as poor matching of strategies and systems, inter-system interfaces, and difficulty in tracing changes. Model-based systems engineering (MBSE) method, by decomposing demand use cases layer by layer and matching system functions, completes the start-up and shutdown plan design of the research reactor and the design analysis of key equipment maintenance strategy, and uses the simulator to verify the typical operation strategy. The results show that: The start-up and shutdown plan and maintenance strategy can achieve the design goals. The design method is scientific, efficient, and traceable, and can improve the efficiency of nuclear engineering design.

Simulation Analysis of FeCrAl-UN Fuel Rod Performance

Tu Teng, Gao Shixin, Zhou Yi, Chen Ping, Zhang Ruiqian, Yang Qingfeng, Liao Nan

2021, 42(S2): 165-170. doi: 10.13832/j.jnpe.2021.S2.0165

Abstract(348) HTML (77) PDF(43)

Abstract:
FeCrAl cladding and UN fuel pellet are important accident tolerant fuel (ATF) candidates, it’s necessary to analyze the performance of FeCrAl cladding and UN fuel pellet under PWR operation condition. Based on the latest FeCrAl cladding and UN fuel physical property data and behavior model at home and abroad, the fuel performance analysis program FUPAC is redeveloped to analyze the in-reactor performance of FeCrAl/UN, FeCrAl/UO₂, Zr-4/UN and Zr-4/UO₂ fuel rods under different linear power densities. Through comparison, the results showed that FeCrAl/UN fuel rod has good performance in aspects like pellet temperature, fission gas release, fuel rod internal pressure, etc. But due to low creep rate of FeCrAl cladding, once the cladding-pellet gap closes, the cladding stress increases rapidly. This phenomenon needs to be paid attention during the following research.

2021 Vol. 42, No. S2