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2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Yanli, Yuan Xianhui, Ye Lijuan, Li Feng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system. According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis. © 2020 American Society of Mechanical Engineers (ASME). All rig

关键词： Sensitivity analysis

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Numerical Study on Benchmark Analysis of NACIE-UP Facility with Uniform and Non-Uniform Power Distribution

SSRN

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SSRN 2024年

作者： Li, Haoyang Zhang, Dalin Zhou, Lei Deng, Jian Tian, Wenxi Su, Guanghui Qiu, Suizheng State Key Laboratory of Multiphase Flow in Power Engineering Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi’an Jiaotong University Xi’an710049 China Nuclear Power Institute of China 328 Changshun Avenue Chengdu610213 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

The NACIE-UP facility is located at the ENEA Brasimone Research Centre and is designed to investigate the thermo-hydraulic properties of LBE within a wire-spaced assembly under uniform and non-uniform power distribution conditions. In this study, the thermo-hydraulic characteristics of FPS are analyzed. ANSYS Fluent solver is applied with the COUPLE algorithm of RANS [[EQUATION]] model. Cheng&Tak correlation and computational conditions are implanted into Fluent in the form of UDFs. The results show that the RMSEs of ADP10SS1 and ADP10SS2 are 2.52 K and 3.69 K, 7.05 K and 8.48 K for ADP06SS1 and ADP06SS2, respectively, which indicate that the adopted models have good prediction ability for the temperature of the experimental conditions. Meanwhile, the heat transfer characteristics are investigated and compared with the correlations of Ushakov, Mikityuk, Kazimi and Carelli. Above all, this study provides quantitative and qualitative analysis for verification and validation of CFD/STH code and stand-alone CFD simulation. © 2024, The Authors. All rights reserved.

关键词： Heat transfer

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Modeling and Program Development of Supercritical CO2 Sub-Channel Based on Modelica Language

SSRN

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SSRN 2023年

作者： Guo, Jia Ming, Yang Zhao, Fulong Fang, Huawei Yin, Jingwei Tian, Ruifeng College of Nuclear Science and Technology Harbin Engineering University Harbin150001 China Heilongjiang Provincial Key Laboratory of Nuclear Power System & Equipment Harbin Engineering University Harbin150001 China Key Laboratory of Nuclear Reactor System Design Technology of China Nuclear Power Research and Design Institute Chengdu610000 China

Small nuclear reactor (SMR) system has many advantages, including low construction cost, short construction period and fast economic benefits. SMR has broad prospects in power supply in remote areas, small power grids, low-temperature heating, nuclear hydrogen production, seawater desalination, nuclear powered ships, etc. In order to study the hydraulics and thermal parameters of supercritical carbon dioxide as coolant in SMR, this paper developed a supercritical carbon dioxide reactor sub-channel analysis program based on Modelica language. Based on the sub-channel analysis method, the reactor coolant flow channel is divided in several control bodies, then the program calculates hydraulics and thermal parameters of each control body. In order to verify the correctness of the model and program, sub-channel program will calculate the axial temperature distribution, radial temperature distribution, coolant flow rate and power level under steady-state conditions. Then compare these calculations with KAIST MMR in Korea. In the transient characteristic analysis, this paper will analyze the reactor power increase, reactivity insertion and coolant flow loss accident, then evaluate the safety performance of the reactor. This paper relies on theoretical modeling and program development method, studying the sub-channel thermal hydraulic characteristics of steady-state and transient conditions, which provides theoretical reference and program support for the study of transient response characteristics and safety characteristics of supercritical carbon dioxide reactor. © 2023, The Authors. All rights reserved.

关键词： Temperature distribution

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Analysis of Transient Characteristics of Droplets in the Dffb Regime During Single-Rod Channel Reflood Experiment

SSRN

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SSRN 2023年

作者： Chen, Deqi Chen, Mingjing Liu, Hanzhou Liu, Haidong Zhong, Lei Deng, Jian Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing400044 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

This study aims to explore the dynamic distribution and hydrodynamic characteristics of droplets in the process of strong transient reflood. However, the quasi-steady state analysis is generally adopted in the current study, ignoring the dependence of droplet formation and breaking mechanism in the Dispersed flow film boiling (DFFB) mechanism on the time-related historical behaviors such as the advance of the quench front. A series of reflood cooling visualization experiments of zirconium alloy high-temperature rod-shaped heating surface were carried out to achieve this research goal. The wall surface and fluid temperature history curves were recorded, and the parameters, such as the size and velocity of DFFB regime droplets, were quantitatively extracted by visual image processing technology. Based on the formation and breakup mechanism of droplets in the dispersed flow, the reasons for droplet size distribution and velocity change with input heat flux, initial wall temperature, inlet subcooling, and inlet flow rate are well analyzed and explained. At the same time, the experimental results also show a strong correlation between droplet behavior and the reflood process in time. This study can provide more detailed physical support for the subsequent development of droplet behavior models more suitable for the strong transient reflood process and the evaluation of existing models. © 2023, The Authors. All rights reserved.

关键词： Heat flux

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Analysis of Transient Characteristics of Droplets in the Dffb Regime During Single-Rod Channel Reflood Experiment

SSRN

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SSRN 2023年

作者： Chen, Mingjing Chen, Deqi Liu, Haiong Liu, Hanzhou Zhong, Lei Deng, Jian Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing400044 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

This study aims to explore the dynamic distribution and hydrodynamic characteristics of droplets during the highly transient process of reflood. However, the quasi-steady state analysis is generally adopted in the current study, ignoring the dependence of droplet formation and breaking mechanism in the Dispersed flow film boiling (DFFB) mechanism on the time-related historical behaviors such as the advance of the quench front. A series of reflood cooling visualization experiments of zirconium alloy high-temperature rod-shaped heating surface were carried out to achieve this research goal. The wall surface and fluid temperature history curves were recorded, and the parameters, such as the size and velocity of DFFB regime droplets, were quantitatively extracted by visual image processing technology. Based on the formation and breakup mechanism of droplets in the dispersed flow, the reasons for droplet size distribution and velocity change with input heat flux, initial wall temperature, inlet subcooling, and inlet flow rate are well analyzed and explained. At the same time, the experimental results also show a strong correlation between droplet behavior and the reflood process in time. This study can provide more detailed physical support for the subsequent development of droplet behavior models more suitable for the strong transient reflood process and the evaluation of existing models. © 2023, The Authors. All rights reserved.

关键词： Heat flux

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Investigation of turbulent prt model and the segmentation rule for lbe turbulent heat transfer

Investigation of turbulent prt model and the segmentation ru...

引用

2020 International Conference on Nuclear Engineering, ICONE 2020, collocated with the ASME 2020 Power Conference

作者： Peiying, Li Jian, Deng Lei, Zhong Libo, Qian Rong, Cai Yugao, Ma Hongxing, Yu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791883785

With liquid metal like lead-bismuth alloy(LBE) acting as a coolant for nuclear reactors, it is necessary to use a more accurate heat transfer relationship and a more reliable Prt model for the low Pr fluid. Because of the low Pr of liquid metal, the thermal conductivity is more dominant than the momentum transfer, which is quite different from ordinary fluids. In this case, the turbulent Prt can better reflect the heat transfer process. In this study, the Prt=A1+A2/Pr form is selected, and the corresponding coefficients are obtained by the renormalization group analysis method, then corrected by Pr. Furthermore, the applicable range and segmentation rule of the turbulent Prt model are discussed, and the obtained Prt segmentation theoretical model is written into CFD. The result shows that, compared with the previously unmodified model, the radial temperature distribution and Nusselt number (Nu) of the annular and bundle channel obtained by RANS method with the improved Prt model is in good agreement with experimental results, and the deviations are within 5%. It is proved that the turbulent Prt segmentation theoretical model proposed in this study is effective and can represent the heat transfer characteristics of liquid metal from the mechanism. © 2020 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Liquid metals

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Representation Learning based and Interpretable reactor system Diagnosis Using Denoising Padded Autoencoder

arXiv

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arXiv 2022年

作者： Li, Chengyuan Qiu, Zhifang Yan, Zhangrui Li, Meifu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology Nuclear Power Institute of China Chengdu610213 China

With the mass construction of Gen III nuclear reactors, it is a popular trend to use deep learning (DL) techniques for fast and effective diagnosis of possible anomalies to ensure the safety of reactor systems under accidents. To improve the robustness of diagnostic model to crippled and noisy monitoring data, and to enhance the interpretability of the diagnostic outputs, this paper focuses on three aspects. First, a novel Denoising Padded Autoencoder (DPAE) model is proposed for representation extraction of monitoring data, and such representation extractor is effective on disturbed data with signal-to-noise ratios up to 25.0 and monitoring data missing up to 40.0%. Secondly, a diagnostic paradigm using DPAE encoder followed by shallow statistical learning algorithms is proposed, and such stepwise diagnostic approach is tested on disturbed datasets with 41.8% and 80.8% higher classification and regression task evaluation metrics, in comparison with the end-to-end diagnostic approaches. Finally, a hierarchical interpretation algorithm using SHAP and feature ablation is presented to analyze the importance of the input monitoring parameters and validate the effectiveness of the high importance parameters. The results of this study inform the development of robust, interpretable, intelligent reactor anomaly diagnosis systems in situations with high safety requirements. Copyright © 2022, The Authors. All rights reserved.

关键词： Monitoring

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Experimental Study of Coherent Structures Downstream Mixing Vaned Spacer Grid of Different Inclination Angles in a 5×5 Rod Bundle by Tr-Piv

SSRN

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SSRN 2022年

作者： Qu, Wenhai Xie, Hao Shen, Danhong Wang, Zefeng Xiong, Jinbiao School of Nuclear Science and Engineering Shanghai Jiao Tong University Shanghai China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

Experimental study of coherent structures downstream of a spacer grid with different inclination angles of mixing vanes in a 5×5 rod bundle was conducted by TR-PIV under four Reynolds *** coherent structures downstream of spacer grid change with several patterns. The developments of lateral time-averaged velocity, Reynolds stress, dominant frequency, integral spatial length scales and integral temporal length scales are determined by coherent structures. Smaller coherent structures are generated with Reynolds number increasing, and time-averaged velocity, Reynolds stress, St numbers, spatial and temporal length scales decrease. Larger inclination angle generates larger coherent structures, and spatial and temporal length scales increase with inclination angle increasing. The coherent structures indicate strong anisotropy of turbulent flow in rod *** experimental findings provide valuable information for understanding sub-channel mixing phenomena caused by mixing vaned spacer grids. The experimental data can be used for sub-channel mixing model developments and validations of turbulent models. © 2022, The Authors. All rights reserved.

关键词： Reynolds number

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Development, Validation and Application of Sub-Channel Thermal-Hydraulic Analysis Code for Liquid Metal Cooled reactor

SSRN

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SSRN 2024年

作者： Wang, Jinshun Chen, Ronghua Liang, Yu Gui, Minyang Zhu, Xinyang Tian, Wenxi Qiu, Suizheng Su, Guanghui School of Nuclear Science and Technology Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University Xi’an710049 China Nuclear Power Institute of China 328 Changshun Avenue Chengdu610213 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China College of Nuclear Science and Technology Harbin Engineering University Harbin150001 China

The liquid metal reactor, specifically the lead-based fast reactor and the sodium-cooled fast reactor, are generally recognized as the most significant representatives of the Gen IV reactors due to their sustainability, economy, security and reliability. In order to conduct thermal-hydraulic safety analysis for liquid metal cooled fast reactors, adaptive improvements are made to the model and solution algorithms based on the sub-channel analysis method, and a Sub-channel Analysis Code Of Safety for Liquid Metal cooled reactor (SACOS-LMR) is developed. The validation of SACOS-LMR is conducted using classic experimental data of liquid metal. The results show that the dimensionless temperature deviations of ORNL 19-pin test and Toshiba 37-pin test are within 10%, and the relative temperature deviations of KIT-KALLA 19-pin test are within 25%. These findings confirm the accuracy and reliability of SACOS-LMR in predicting the temperature distribution of liquid metal reactors. Ultimately, calculations conducted on the ALFRED lead-cooled fast reactor demonstrated the practical utility of the code and have certain guiding significance for its subsequent design and development. © 2024, The Authors. All rights reserved.

关键词： Liquid metals

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Experimental Study of Coherent Structures Downstream Mixing Vaned Spacer Grid of Different Inclination Angles in a 5×5 Rod Bundle by Tr-Piv

SSRN

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SSRN 2022年

关键词： Reynolds number

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