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15th IEEE Global Reliability and Prognostics and Health Management Conference, PHM-Beijing 2024

作者： Wang, Miao Wei, Wenbin Wan, Bo Huang, Qichang Sun, Shangqing Lu, Weiwei Wu, Chenlu Hong, Shixin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China CNNP Guodian Zhangzhou Energy Co. Ltd. China Nuclear Power Engineering Co. Ltd. Zhangzhou China

ISBN: (数字)9798350354010

ISBN: (纸本)9798350354010

Large reaction cross-sections, low operating voltage, extended lifespan, good stability, and suitability for high-temperature and intense gamma radiation settings are among the features of boron-coated proportional counter tubes. Consequently, it has been extensively employed in the nuclear-measuring system outside of the reactor, where the source range detector is a boron-coated proportional counter tube. This paper studies boron-coated proportional counter tubes using a comprehensive approach that combines Monte Carlo simulation with lifespan prediction in health management. The wavelet packet decomposition approach was used to extract feature information from the pulse signals of the boron-coated proportional counter tube. A counter tube defect prediction model was established based on a multi-kernel correlation vector machine, where the impact of different wavelet packets and correlation vector machine kernel functions on the model's prediction performance was examined. The study's findings demonstrated that the particle swarm optimization algorithm-optimized multi-kernel correlation vector machine model exhibited superior prediction accuracy for the counter tubes' future operating conditions. The primary goal is to comprehend how gas loss affects gas pressure and to determine the lifespan of detectors within a detector system. © 2024 IEEE.

关键词： Proportional counters

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MATLAB/Simulink-Based Simulation Study of Small Pressurized Water reactor Nuclear Steam Supply System

MATLAB/Simulink-Based Simulation Study of Small Pressurized ...

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Information Engineering (ACIE), IEEE Asia Conference on

作者： Jie Chen Kai Xiao Mengwei Mengwei Zhao Zhao Ke Huang Pengcheng Yang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

Accurate and reliable modeling and simulation is the key prerequisite for intelligent control technology research of nuclear power plant. For the nuclear steam supply system (NSSS) of small pressurized water reactor under bi-stable steady state operation scheme, this paper firstly establishes the models of the core, regulator, steam generator and other components. Then, the nuclear steam supply system platform (NSSSP) of small pressurized water reactor is built based on Matlab/Simulink. Simulations are carried out to analyze the normal operation transient and load rejection transient in the water supply flow control mode. The accuracy and reliability of NSSSP are verified to beconsistent with the commercial program RETRAN.

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Research and Verification of the One-Step Resonance and Transport Methods Based on the Openmoc Code

SSRN

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SSRN 2024年

作者： Zhao, Chen Wang, Lianjie Chen, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

With the continuous improvement of computer hardware capabilities, the one-step method in reactor physics has become one of the important research directions in recent two decades. OpenMOC is an open-source MOC code developed by MIT, with C++/Python hybrid programming and three-dimensional transport calculation capabilities. Based on the OpenMOC code, the following work was carried out. Firstly, the global-local resonance method with multi-group and continuous neutron libraries was researched and established. Next, based on the 2D and 3D MOC solver, the 2D/1D and the MOC/DD transport methods were realized in OpenMOC. Finally, verification of the transport and resonance methods was conducted using the C5G7 macro benchmark and the VERA micro benchmark. The numerical results demonstrated that the average eigenvalue deviation was 44pcm and average maximum pin power distribution deviation was 0.37% in the VERA-2 benchmark, which showed the good accuracy of the resonance method. As for the transport method, 3DMOC method exhibited better accuracy in strong anisotropic cases, but the computational time was 38 times that of the 2D/1D method. © 2024, The Authors. All rights reserved.

关键词： Computer hardware

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Research on Hybrid Modeling Method Based on Mechanism Model and BP Neural Network

Research on Hybrid Modeling Method Based on Mechanism Model ...

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Information Engineering (ACIE), IEEE Asia Conference on

作者： Jie Chen Yanqiu Zheng Qing Chu Guanfu Jiang Jing Li Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

reactor system modeling is the premise of reactor control system design. The common commercial thermal and hydraulic analysis programs like RELAP5, RETRAN have complex structure and slow operational speed, which are not conductive to the transplantation and improvement of reactor system. In order to realize the rapid development of reactor control system, this paper carries out the mechanism modeling of the reactor system in Matlab/Simulink, and then introduce BP neural network to establish the hybrid model, which improves the modeling accuracy to support the design of reactor system. For small pressurized water reactor, we firstly take analytical modeling method to build the core mechanism model and its state space model through three conservation principles and the operating mechanism of core. Based on analyses, we can conclude that the parameter error and structural error of the mechanism model is caused by the assumption and simplification during modeling, and choose the parameter with the most influence (the feedback coefficient of the coolant temperature) as the corrected parameter in the BP neural-based hybrid modeling process. After that, according to the dual constant operation scheme of small pressurized water reactors, and the gradient descent method of BP network, the hybrid model of core is built and the parameter error and structure error of the mechanism modeling are corrected online. The simulation results show that the proposed hybrid modeling method can accurately simulate the reactor operate state. Thus, the proposed hybrid model is accurate and effective.

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Unveiling deformation behavior and damage mechanism of irradiated high entropy alloys

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Journal of Materials science & technology 2024年第29期196卷 71-87页

作者： Shuo Wang Yang Chen Jia Li Bin Liu Ruiqian Zhang Peter K Liaw Qihong Fang State Key Laboratory of Advanced Design and Manufacturing for Vehicle Body College of Mechanical and Vehicle EngineeringHunan UniversityChangsha 410082China State Key Laboratory of Powder Metallurgy Central South UniversityChangsha 410083China Science and Technology on Reactor Fuel and Materials Laboratory Nuclear Power Institute of ChinaChengdu 610213China Department of Materials Science and Engineering The University of TennesseeKnoxvilleTN 37996USA

The oxide dispersion strengthening(ODS)high entropy alloy(HEA)exhibits the high elevated temperature performance and radiation resistance due to severe atomic lattice distortion and oxide particles dispersed in matrix,which is expected to become the most promising structural material in the next generation of nuclear energy ***,microstructure and damage evolution of irradiated ODS HEA under loading remain elusive at submicron scale using the existing simulations owing to a lack of atomic-lattice-distortion information from a micromechanics ***,the random field theory informed discrete dislocation dynamics simulations based on the results of high-resolution transmission electron microscopy are developed to study the dislocation behavior and damage evolution in ODS HEA considering the influence of severe lattice distortion and nanoscale oxide ***,the damage behavior shows an unusual trend of the decreasing-to-increasing transition with the continuous loading *** are two main types of damage micromechanics generated in irradiated ODS HEA:the dislocation loop damage in which the damage is controlled by irradiation-induced dislocation loops and their evolution,the strain localization damage in which the damage comes from the dislocation multiplication in the local plastic *** oxide particle hinders the dislocation movement in the main slip plane,and the lattice distortion induces the dislocation sliding to the secondary slip plane,which promotes the dislocation cross-slip and dislocation loop annihilation,and thus reduces the material damage in the elastic damage *** findings can deeply understand atomic-scale damage mechanism and guide the design of ODS HEA with high radiation resistance.

关键词： Oxide dispersion strengthening High entropy alloy Radiation Dislocation Damage model

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Noninvasive Data-Driven Prediction of reactor Power Field: An MLP-Based Approach 9

Noninvasive Data-Driven Prediction of Reactor Power Field: A...

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9th IEEE Smart World Congress, SWC 2023

作者： Shi, Binhui Qiu, Qi Gong, Helin Li, Qing Luo, Yang University of South China School of Computer Hengyang China Shanghai Jiao Tong University ParisTech Elite Institute of Technology Shanghai China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9798350319804

The reactor Operation Digital Twin (RODT) is a numerical representation method designed for the operational phase of nuclear reactors, aiming to solve forward problems (such as state prediction) based on input parameters and inverse problems (such as parameter identification) utilizing additional measurement data, in order to enhance the economic benefits and safety of nuclear power. In this paper, we propose a solution for the forward problem in RODT based on a Multilayer Perceptron (MLP) approach, where the MLP is used to directly learn the mapping relationship from low-dimensional inputs to high-dimensional outputs. Compared to previous methods, the proposed approach shows improved prediction accuracy, especially in situations with limited training data, demonstrating superior generalization capabilities and accuracy. To validate our contributions, we conducted a comparative experiment based on a engineering case, namely the reactor core simulation in the life cycle of a real-world nuclear ractor, which confirmed the effectiveness of the proposed method. © 2023 IEEE.

关键词： limited training data Multilayer Perceptron (MLP) reactor core simulation reactor Operation Digital Twin (RODT) reactor power field

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Synthesis Optimization of SSZ-13 Zeolite Membranes by Dual Templates for N_(2)/NO_(2) Separation

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Chemical Research in Chinese Universities 2022年第1期38卷 250-256页

作者： LI Jun RONG Huazhen CHEN Cong LI Ziyi ZUO Jiayu WANG Wenjian LIU Xinjian GUAN Yixing YANG Xiong LIU Yingshu ZOU Xiaoqin ZHU Guangshan School of Energy and Environmental Engineering University of Science and Technology BeijingBeijing 100083P.R.China Faculty of Chemistry Northeast Normal UniversityChangchun 130024P.R.China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213P.R.China

High-silica SSZ-13 zeolite membranes are promising in industrial separations of light gases and continuous membranes are highly demanded for better separation performances. Herein, pure-phase, continuous and thin SSZ-13 zeolite membranes were synthesized using dual templates of N,N,N-trimethtyl-1-adamantammonium hydroxide(TMAdaOH) and tetraethylammonium hydroxide(TEAOH). Systematical investigations of TMAdaOH/TEAOH ratios and their concentrations show that TMAdaOH acts as the main structure-directing agent in the formation of the SSZ-13 zeolite. TMAdaOH cooperatively plays with TEAOH in promoting the SSZ-13 crystal intergrowth to form a continuous polycrystalline membrane. Additionally, appropriate introduction of TEAOH is able to adjust the membrane thickness to the crystal-comparable size of ca. 2.0 μm. The SSZ-13 membranes are further applied for N2/NO2separation, which is firstly reported on zeolite membranes. The gas permeation results show that the SSZ-13 membrane synthesized by the dual-template approach exhibits selective separation of N2 over NO2 with N2/NO2 separation factor of 7.6 and N2 permeance of 1.66×10-8 mol·m-2·s-1·Pa-1.

关键词： Zeolite membrane Dual template SSZ-13 Gas separation Nitrogen dioxide

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A novel approach for radionuclide diffusion in the enclosed environment of a marine nuclear reactor during a severe accident

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Nuclear science and Techniques 2022年第2期33卷 53-65页

作者： Fang Zhao Shu-Liang Zou Shou-Long Xu Xuan Wang Jun-Long Wang De-Wen Tang School of Resource Environment and Safety Engineering University of South ChinaHengyang 421001China Hunan Provincial Key Laboratory of Emergency Safety Technology and Equipment for Nuclear Facilities Hengyang 421001China Shanghai Nuclear Engineering Research and Design Institute Co.LTD Shanghai 200000China Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental *** is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear ***,this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe *** radionuclide release fraction and temperature calculated by MELCOR were combined with the scSTREAM calculations to study the radionuclide diffusion behavior and the phenomenon of radionuclide diffusion in different space environments of the reactor under the conditions of varying wind velocities of the ventilation system and diffusion *** results show that the wind velocity of the ventilation system is very small or zero,and the turbulent diffusion of radionuclides is not obvious and diffuses slowly in the form of condensation sedimentation and gravity *** the wind speed of the ventilation system increases,the flow of radionuclides meets the wall and forms eddy currents,affecting the time variation of radionuclides diffusing into chamber *** wind velocity of the ventilation system and the diffusion speed has opposite effects on the time variation trend of radionuclide diffusion into the four chambers.

关键词： Radionuclide diffusion MELCOR coupled with scSTREAM Severe accident Marine nuclear reactor

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Multi-physics Coupling Analyses of Nuclear Thermal Propulsion reactor 23rd

Multi-physics Coupling Analyses of Nuclear Thermal Propulsio...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Han, Wenbin Guan, Zechuan Huang, Shanfang Deng, Jian Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China China Institute of Atomic Energy Beijing China

ISBN: (纸本)9789819910229

Nuclear thermal propulsion (NTP) reactors have high-temperature solid-state characteristics and significant thermal expansion, which therefore require multi-physics coupling analyses. In this paper, the framework of Neutronics, Thermal-Hydraulics and Mechanics coupling (N/T-H/M) of nuclear thermal propulsion reactor is developed, and the typical reactor XE-2 is analyzed with this method. The results show that the N/T-H/M coupling will bring -1049 pcm negative reactivity, of which the thermal expansion effect accounts for 22%, indicating that the nuclear thermal propulsion reactor has a certain capacity for self-regulation. However, thermal expansion will lead to 0.88 mm peak deformation and 233 MPa peak stress, which will severely threaten the mechanical tolerance of the materials. Therefore, there is a trade-off between the advantages and disadvantages of the high-temperature solid-state core while designing NTP reactors. © 2023, The Author(s).

关键词： Thermal expansion

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Study and Analysis of the Effect of Dynamic Load on the Application of Lbb technology to the Primary Pipe of reactor Coolant System

SSRN

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SSRN 2024年

作者： He, Feng Xiong, Furui Bai, Xiaoming Yang, Kang Lu, Xifeng Li, Bingjin Wang, Xinjun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

reactors may be subjected to dynamic loads, and in order to ensure that Leak-Before-Break (LBB) technology is safely and reliably applied to the primary piping of reactor coolant system, dynamic load impact analysis of LBB technology applied to the primary piping was performed. The dynamic mechanical properties test was carried out on the primary piping, and the crack leakage rate test under dynamic load and the impact load loading test under different loading speeds were carried out on the primary piping test piece with circumferential penetrating crack. Based on experimental data, the crack leakage rate analysis method and crack stability analysis method in quasi-static LBB technology were used to analyze the impact of applying LBB technology to the primary piping under dynamic load. The analysis results indicate that, when LBB technology is applied to the primary piping of reactor coolant system, the analysis results under dynamic load are more conservative than those under quasi-static load, and it is safer and more reliable under dynamic loads in practical engineering applications. © 2024, The Authors. All rights reserved.

关键词： Dynamic loads

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