检索结果-内蒙古大学图书馆

Development and Verification of Safety Analysis Code to Model and Simulate Steam power Conversion System in Sodium-Cooled Fast Reactor

学校读者我要写书评

暂无评论

SSRN 2023年

作者： Liu, Shuo Jin, Zhao Wang, Chenglong Zhang, Dalin Tian, Wenxi Qiu, Suizheng Su, Guanghui Yang, Jun School of Nuclear Science and Technology State Key Lab. of Multiphase Flow in Power Engineering Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China China Institute of Atomic Energy Beijing102413 China

Sodium-cooled fast reactor (SFR) is one of the most promising generation IV reactor type. SFR system is composed of three heat transfer circuits. SFR will generate electricity, which requires accuracy prediction of steam power conversion system. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR has been developed. A modular modeling method was applied. Each equipment is packaged as a separate module. The system model can be obtained by connecting modules to achieve flexible modeling. Compared with RELAP5, the code uses a general algorithm to solve ODE, which enhances the universality of code. The model is verified with design results of Daya Bay Nuclear power Plant and BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. It is proven that the code developed can be used in the analysis and design the steam power conversion system. The steady-state analysis of SFR steam power conversion circuit is carried out. The design could provide a reference for the future development of steam power conversion system in SFR. © 2023, The Authors. All rights reserved.

关键词： Sodium-cooled fast reactors

Dynamic Modeling of Steam power Conversion System in Sodium-Cooled Fast Reactor

学校读者我要写书评

暂无评论

SSRN 2023年

作者： Liu, Shuo Jin, Zhao Wang, Chenglong Zhang, Dalin Tian, Wenxi Qiu, Suizheng Su, G.H. Yang, Jun School of Nuclear Science and Technology State Key Lab. of Multiphase Flow in Power Engineering Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China China Institute of Atomic Energy Beijing102413 China

Sodium-cooled fast reactor (SFR) is the most promising generation IV reactor type. SFR systems are composed of three heat transfer circuits. Currently, research on steam power conversion system is scarce because of complex structure. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR has been developed. Models of main components have been established, including once-through steam generator (OTSG), turbine, deaerator, feedwater heater, condenser, pump and moisture separator reheater (MSR). The model is verified with design results of Daya Bay Nuclear power Plant and BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. It is proven that the code developed can analysis and design the steam power conversion system. The steady-state analysis of SFR steam power conversion circuit is carried out. The design could provide a reference for the future development of steam power conversion system in SFR. © 2023, The Authors. All rights reserved.

关键词： Sodium-cooled fast reactors

Experimental research on the characteristics of steam-water counter-current flow in the Pressurizer Surge Line assembly

学校读者我要写书评

暂无评论

Experimental Thermal and Fluid Science 2018年 96卷 180-191页

作者： Jiangtao Yu Dalin Zhang Leitai Shi Zhiwei Wang Wenxi Tian G.H. Su S.Z. Qiu State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University 28 Xianning West Road Xi’an 710049 China Shaanxi Key Lab. of Advanced Nuclear Energy and Technology School of Nuclear Science and Technology Xi’an Jiaotong University 28 Xianning West Road Xi’an 710049 China

To investigate Counter-Current flow (CCF) characteristics in the Pressurizer Surge Line (PZR SL) assembly of the large advanced passive nuclear reactors, steam-water CCF experiments are carried out on a stainless steel test section, which is a 1:4 scaled-down model of the AP1000 PZR SL assembly. The inner diameter of the test PZR SL pipe is 90 mm, which also belongs to the large-diameter pipes, as same as the prototype PZR SL pipe, in the field of Counter-Current flow Limitation (CCFL) researches. The present steam-water CCFL experiments are conducted under normal pressure and saturated temperature, with the PZR simulator collapse water level ranging from 350 to 900 mm. CCFL becomes severer at higher steam flow rate, and the most limiting CCFL effect locates between the PZR simulator bottom head and the vertical part of PZR SL pipe. The onset of CCFL and zero liquid penetration (ZP) are two critical conditions in the CCF development process, dividing the process into three stages: Before-CCFL, Partial-CCFL, and CCFL-ZP. According to the local CCFL conditions, the development of the CCFL process is also divided into four-regions since onset of CCFL. The present steam-water CCF data are well normalized in terms of dimensionless Kutateladze (Ku) numbers, and a Ku-type empirical partial CCFL correlation is developed. The comparisons of the present CCF data and partial CCFL correlation with the CCF data and correlations of former experiment researches validate that the present empirical partial CCFL correlation is conservative to predict steam-water partial CCFL in the prototype PZR SL assembly of the large advanced passive nuclear reactors.

关键词： Steam-water counter-current flow Partial CCFL correlation Pressurizer surge line assembly Advanced passive nuclear reactors ADS Automatic Depressurization System ADS1-3 First to Third Stages of ADS ADS4 Fourth Stage of ADS CCF Counter-Current flow CCFL Counter-Current flow Limitation CCFL-1 The 1st grade of CCFL region CCFL-2 The 2nd grade of CCFL region CCFL-3 The 3rd grade of CCFL region CCFL-L Local CCFL at the T-junction of HL pipe and PZR SL pipe CCFL-S Local CCFL along the PZR SL pipe CCFL-U Local CCFL at the location between the PZR simulator bottom head and vertical part of PZR SL pipe CCFL-ZP Zero liquid penetration region or stage CMT Core Makeup Tank COCF Co-Current flow HL Hot Leg IRWST In-containment Refueling Water Storage Tank Ku Kutateladze PRHR Passive Containment Cooling System PWRs Pressurized Water Reactors PZR SL Pressurizer Surge Line RC Reactor Core RCS Reactor Coolant System RPV Reactor Pressure Vessel SBLOCAs Small-Break Loss-of-Coolant Accidents ZP zero liquid penetration

Non-Hermitian Delocalization in a Two-Dimensional Photonic Quasicrystal

学校读者我要写书评

暂无评论

Physical Review Letters 2024年第26期132卷 263801-263801页

作者： Zhaoyang Zhang Shun Liang Ismaël Septembre Jiawei Yu Yongping Huang Maochang Liu Yanpeng Zhang Min Xiao Guillaume Malpuech Dmitry Solnyshkov Key Laboratory for Physical Electronics and Devices of the Ministry of Education & Shaanxi Key Lab of Information Photonic Technique School of Electronic Science and Engineering Faculty of Electronics and Information PHOTON-N2 Université Clermont Auvergne CNRS Clermont INP F-63000 Clermont-Ferrand France International Research Center for Renewable Energy and State Key Laboratory of Multiphase Flow in Power Engineering Department of Physics National Laboratory of Solid State Microstructures and School of Physics 75231 Paris France

Theoretical and experimental studies suggest that both Hermitian and non-Hermitian quasicrystals show localization due to the fractal spectrum and to the transition to diffusive bands via exceptional points, respectively. Here, we present an experimental study of a dodecagonal photonic quasicrystal based on electromagnetically induced transparency in a Rb vapor cell. First, we observe the suppression of the wave packet expansion in the Hermitian case. We then discover a new regime, where increasing the non-Hermiticity leads to delocalization, demonstrating that the behavior in non-Hermitian quasicrystals is richer than previously thought.

关键词： Electromagnetically induced transparency Localization Photonics Non-Hermitian systems Quasicrystals

Nuclear Thermal Coupling Optimization of Baffled Ntp Reactor

学校读者我要写书评

暂无评论

SSRN 2022年

作者： Duan, Zimian Zhang, Jing Han, Zichao Wang, Mingjun Wu, Yingwei He, Yanan Tian, Wenxi Qiu, Siuzheng Su, Guanghui Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University Xi’an710049 China Department of Nuclear Science and Technology State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University Xi’an710049 China

Nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. In particular, CERMET fuel reactor has received a lot of research due to its simple form and structure. A general CERMET fuel reactor model is established by using three-dimensional Monte Carlo neutron dynamics program and commercial CFD program Fluent. After nuclear thermal coupling optimization, three partition schemes of baffled reactor were proposed to reduce the temperature difference between the wall and the coolant and lower the relative spike factor in the inner region. A feasible design of redundant control system was then proposed, with a cold clean excess reactivity greater than $3.44 and a minimum dry shutdown reactivity lower than -$5.95, where there was a high control accuracy of 6.7 pcm/mm. The reactor realized safety of the core temperature lower than 3000K under the conditions of total exit temperature of 2650K and thrust of 25000lb. © 2022, The Authors. All rights reserved.

关键词： Cermets

Sacos-Plate: A New Thermal-Hydraulic Subchannel Analysis Code for Plate Type Fuel Assemblies

学校读者我要写书评

暂无评论

SSRN 2023年

作者： Sun, Ruiyu Wang, Jianjun Wang, Jinshun Chen, Ronghua Jiang, Xing Tian, Wenxi Qiu, Suizheng Su, Guanghui Gui, Minyang Harbin Engineering University China School of Nuclear Science and Technology Xi’an Jiaotong University Shaanxi Xi’an710049 China Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University China State Key Lab. of Multiphase Flow in Power Engineering Xi’an Jiaotong University China Shanghai Nuclear Engineering Research Design Institute Shanghai China

Over the past few decades, plate-type fuel elements have been widely used in advanced research reactors, small modular pressurized water reactors, production reactors, etc., with the advantages of compact structure arrangement, good heat transfer characteristics and high fuel consumption. Currently, a thermal-hydraulic subchannel analysis code, namely SACOS-PLATE code, was newly designed to satisfy the calculation of different types of plate fuel elements under various steady-state operating conditions. Based on the original SACOS code, this updated version modifies the flow distribution model, the plate heat conduction model, the heat distribution model as well as the flow heat transfer model, and then realizes the simulation of the special thermal-hydraulic phenomena of the fuel and the rectangular flow channel. In this present study, the RA-6 experimental data was used to validate the code, and the simulation calculation was performed on the JRR-3 research reactor, which proved the good computing capability of SACOS-PLATE code. © 2023, The Authors. All rights reserved.

关键词： Plates (structural components)

Experimental Study on Leakage of High-Energy Fluids Through Narrow Elliptical Slits

学校读者我要写书评

暂无评论

SSRN 2024年

作者： Bi, Qincheng Zhang, Tao Zhao, Jinle Wang, Teng Yang, Zhendong Gui, Miao State Key Lab Multiphase Flow Power Engn Xi'an Jiaotong University Shaanxi Xian710049 China Department of Power Engineering North China Electric Power University Baoding071003 China State Key Laboratory of Eco-hydraulics in Northwest Arid Region of China Xi'an University of Technology Shaanxi Xian710048 China School of Nuclear Science and Technology Xi'an Jiaotong University Shaanxi Xian710049 China

This study is focused on the leakage flow rate of high-energy fluid passing through narrow elliptical slits with large aspect ratios, which is an issue in the application of the leak-before-break concept. Experiments were conducted using a high-temperature and high-pressure experimental platform at Xi’an Jiaotong University, where the temperature reached up to 320 ℃ and the pressure reached up to 10 MPa. These temperatures and pressures encompass a wide range of operating conditions (single-phase cold water, liquid-vapor (two-phase) mixture, and superheated steam). A series of narrow elliptical slits with different aspect ratios (528-2083) and hydraulic diameters (226-412 um) were used in the experiments. The experimental results were compared with previous model. Based on the experimental data and the the effects of the aspect ratio, the mathematical model was corrected, and the results obtained from the modified model conform well with the experimental results. © 2024, The Authors. All rights reserved.

关键词： Aspect ratio

Experimental and Kinetic Modeling Study on the Oxidation and Laminar Combustion Characteristics of Natural Gas

学校读者我要写书评

暂无评论

SSRN 2023年

作者： Chen, Xiaoxiao Zeng, Wen Zheng, Weilin Hu, Erjiang Ma, Hongyu Liaoning Key Lab of Advanced Test Technology for Aerospace Propulsion System Shenyang Aerospace University Liaoning Shenyang110136 China State Key Laboratory of Multiphase Flow in Power Engineering Xi’an Jiaotong University Shanxi Xi'An710049 China AECC Shenyang Engine Research Institute Shenyang110015 China

The oxidation process in the flow tube reactor at the conditions of P=0.1 MPa, T=500-1850 K, f=0.5, 1.0, 3.5, and the laminar combustion process in the constant volume combustion bomb at the conditions of P=0.1, 0.2, 0.3 MPa, T=300, 350, 400 K, f= 0.7-1.4, the O2 contents of 15%, 18% and 21%, the CO2 and H2O contents of 0%, 10%, and 20% of natural gas (0.9 methane/0.07 ethane/0.03 propane, mole fraction) were experimental tested. In the oxidation process of natural gas, with the equivalence ratio increasing, the starting and ending temperatures for oxidation reaction of natural gas are increased, the reaction temperatures corresponding to CO generation and consumption are higher, and the amount of NO production is decreased. In the the laminar combustion process of natural gas, at the conditions of higher pressure, equivalence ratio, CO2 and H2O contents, the flame front is not smooth as a large number of cracks and cell structures appear, and the flame is unstable. Moreover, with the initial temperature and oxygen content increasing or the initial pressure, CO2 and H2O contents decreasing, the laminar flame speed is increased. Furthermore, combined the path sensitivity analysis, rate of production analysis, reaction path analysis and the decoupling method, a reduced reaction mechanism (including 40 species and 189 reactions) of natural gas was developed. Through comparing with the calculated results by the Aramco 2.0 mechanism and the corresponding test data, the reduced reaction mechanism can well predict the oxidation, ignition delay and laminar combustion characteristics of natural gas under various conditions. © 2023, The Authors. All rights reserved.

关键词： Sensitivity analysis

Numerical Study on the Impact Characteristics of a Single Molten Lead Bismuth Droplet on Water

学校读者我要写书评

暂无评论

SSRN 2024年

作者： Lin, Yue Zhang, Dalin Chen, Yutong Zhang, Xisi Deng, Jian Du, Peng Tian, Wenxi Qiu, Suizheng Su, Guanghui Department of Nuclear Science and Technology State Key Lab. of Multiphase Flow in Power Engineering Shaanxi Key Lab. of Advanced Nuclear Energy and Technology Xi’an Jiaotong University No.28 Xianningxilu Beilinqu Shaanxi Xi’an710049 China Institute of Reactor Engineering Technology China Institute of Atomic Energy Beijing102413 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu610213 China

As an important link in the possible occurrence of steam explosions, the interaction between molten materials and coolant is of great significance for safety analysis in severe accidents of nuclear reactors. This study is based on the two-dimensional multiphase and multi-component fluid dynamics code ACENA to model the process of single molten lead bismuth droplet impacting a water tank and obtain the velocity changes of the molten droplet as it moves through the water tank. Furthermore, a sensitivity analysis of relevant factors such as the heat transfer area fraction and drag coefficient of the droplet is carried out. To verify models in ACENA, the computation results are compared to experimental data. It is demonstrated that the overall trend of the velocity curve acquired by the ACENA code during the interaction between a molten lead bismuth droplet and water is consistent with the experimental results. The droplet falling velocity increases with the increase in droplet temperature when the temperature of water is the same, and the droplet falling velocity increases with the increase of water temperature at the same droplet temperature. The simulation results by ACENA code for MFCI are in good agreement with experiments which verifies the accuracy of relevant models in ACENA. It can provide an effective means for safety analysis of nuclear reactors in China. © 2024, The Authors. All rights reserved.

关键词： Drops