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Determining the nuclear temperature dependence on source neutron-proton asymmetry in heavy-ion reactions at intermediate energy

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Chinese Physics C 2023年第5期47卷 152-159页

作者：曲国峰黄宇郑华刘星泉 R.Wada 林炜平黄美容韩纪锋任培培杨振雷张鑫冷强钟 Key Laboratory of Radiation Physics and Technology of the Ministry of Education Sichuan UniversityChengdu 610064China Institute of Nuclear Science and Technology Sichuan UniversityChengdu 610064China School of Physics and Information Technology Shaanxi Normal UniversityXi'an 710119China Cyclotron Institute Texas A&M UniversityCollege StationTexas 77843USA College of Physics and Electronics information Inner Mongolia University for NationalitiesTongliao 028000China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

In this article,we investigate the dependence of nuclear temperature on emitting source neutron-proton(N/Z)asymmetry with light charged particles(LCPs)and intermediate mass fragments(IMFs)generated from intermediate-velocity sources in thirteen reaction systems with different N/Z asymmetries,^(64)Zn on^(112)Sn,and^(70)Zn,^(64)Ni on^(112,124)Sn,^(58,64)Ni,^(197)Au,and^(232)Th at 40 MeV/*** apparent temperature values of LCPs and IMFs from different systems are deduced from the measured yields using two helium-related and eight carbon-related double isotope ratio thermometers,***,the sequential decay effect on the experimental apparent temperature deduction with the double isotope ratio thermometers is quantitatively corrected explicitly with the aid of the quantum statistical *** present treatment is an improvement compared to our previous studies in which an indirect method was adopted to qualitatively consider the sequential decay effect.A negligible N/Z asymmetry dependence of the real temperature after the correction is quantitatively addressed in heavy-ion reactions at the present intermediate energy,where a change of o.1 units in source N/Z asymmetry corresponds to an absolute change in temperature of an order of 0.03 to 0.29 MeV on average for LCPs and *** conclusion is in close agreement with that inferred qualitatively via the indirect method in our previous studies.

关键词： nuclear temperature N/Z asymmetry dependence Albergo thermometer sequential decay correction heavy-ion reactions at intermediate energy

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Modelling Analysis of Non-uniform Flow and Heat Transfer in Parallel Rectangular Channels During Flow Blockage Condition 23rd

Modelling Analysis of Non-uniform Flow and Heat Transfer in ...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Chen, Jiayue Chen, Huandong Wang, Xiaoyu Wang, Zefeng Sino-French Institute of Nuclear Engineering and Technology Sun Yat-Sen University Zhuhai610213 China Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811987793

In order to establish the non-uniform flow and heat transfer modelling method in parallel rectangular channels, as an additional technique and tool for safety analysis of plate-type fuel, one dimensional two-fluid model coupled with two-dimensional heat conduction of fuel are used to develop a special code for non-uniform flow and heat transfer simulation during flow blockage condition. The flow distribution and fuel temperature field are obtained under different blockage conditions. The two-dimensional heat conduction model are testes under four different power distributions. The simulation results show that mass flow and heat transfer between parallel channels would redistribute after blockage. And the two-dimensional heat conduction model predicted more uniform temperature profile of fuel plate cross-section. Thus, the modelling method used in this paper can be used in the parallel rectangular channels prediction in non-uniformly flow and heat transfer phenomenon. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Two phase flow

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reactor Containment Passive Safety Analysis: Steam Condensation in Presence of Non-condensable Gas Scaled Experiment and Modeling

Reactor Containment Passive Safety Analysis: Steam Condensat...

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2024 Pacific Basin Nuclear Conference, PBNC 2024

作者： Bhowmik, Palash K. Schlegel, Joshua P. Sabharwall, Piyush Division of Reactor System Design and Analysis Idaho National Laboratory Idaho FallsID United States Department of Nuclear Engineering and Radiation Science Missouri University of Science and Technology RollaMO United States

ISBN: (纸本)9798331307653

This study presents steam condensation scaled experiments and semi-empirical models in presence of nitrogen (N)—a noncondensable gas (NCG), simulating air in the reactor containment—to support water-cooled small modular reactors (SMRs) passive containment cooling system (PCCS) design and analysis. Previous experimental studies on PCCS are focused on fixed and smaller tube (mostly 2-in.) geometries and specific test condition variations, bringing challenges with geometric scaling and mismatching with SMR prototypic design. To address these challenges, this study presents steam condensation test dataset obtained from three scaled test sections of 1-, 2-, and 4-in.-diameter steam condensers with an annular/ jacket cooling of 2-, 3-, and 6 in.-diameter tubes, respectively. Test data were collected for steam ranges from 58 to 63 kg/hr., and NCG flow of to 13.3 kg/hr. Annular cooling water flow was varied to obtain required testing conditions of saturated steam 4.4 inlet and fully condensed outlet. Axial temperature test data of bulk cooling water, steam and condensate were collected by thermocouples for three test sections and various steam-NCG mixing/testing conditions. A standard data reduction method was adopted—utilizing iterative and nodalized mass and heat transfer calculation—to estimate axial local heat fluxes, heat transfer coefficients (HTCs), condensation rates, film thickness, and Nusselt number. Based on the obtained dataset semi-empirical model results—a ratio of experimental and Nusselt’s theoretical HTC are presented. Such results and findings are supportive of developing scaled-up testing facility, to model validations and accelerate next generation of reactors development and deployment. © 2024 Pacific Basin Nuclear Conference, PBNC 2024. All rights reserved.

关键词： Thermocouples

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Direct Contact Condensation Characteristics of Steam Injection into Cold-Water Pipe Under Rolling Condition 23rd

Direct Contact Condensation Characteristics of Steam Injecti...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Wang, Zhiwei He, Yanping Duan, Zhongdi Huang, Chao Liu, Shiwen State Key Laboratory of Ocean Engineering Shanghai Jiao Tong University Shanghai China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789819910229

Direct contact condensation (DCC) is widely occurred in nuclear power systems and leads to undesired phenomena such as condensation-induced water hammer. For ocean nuclear power ships, DCC is inevitable in the passive heat removal system and influenced by sea conditions. In this paper, the characteristics of DCC under rolling conditions are analyzed. The numerical model of DCC is established based on computational fluid dynamics approach. The VOF model, SST k–ω turbulence model and the additional inertia force model are incorporated to describe the liquid-gas two-phase flow under the rolling motion. The condensation model based on surface renewal theory (SRT) is used to simulate steam-water DCC phenomenon. The simulation results are compared with the experimental data and show reasonable agreement. The effects of rolling motion on DCC for steam injection into a horizontal pipe filled with cold water are numerically investigated. The results show that the additional inertial forces and the average condensation rate increase with the increase of the rolling angle and frequency. The reverse flow of the seawater induced by rolling motion leads to the accumulation of the steam at the lower part of the pipe, resulting in a large pressure pulse. With the increase of rolling angle and frequency, the pressure pulse induced by DCC increases. © 2023, The Author(s).

关键词： Condensation

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Research and design on the Control Rod Drive Mechanism Power Supply and Control Strategy for the Nuclear Power Plants Based on Silicon-Controlled Rectifier

Research and Design on the Control Rod Drive Mechanism Power...

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International Conference on Power Electronics and Motion Control (IPEMC)

作者： A. Mingzhou Xu B. Gao Zheng C. Zerun Zhao D. Ping Yang E. Yusheng Peng Science and Technology on Reactor System Design Technology Laboratory China Southwest Jiaotong University China

ISBN: (数字)9798350351330

ISBN: (纸本)9798350351347

Due to the characteristics of large output current ripple and long current rise and fall times in the current rod control power supply basedsilicon-controlled rectifier, a decoupling nonlinear adaptive control is *** proposed control strategy can improve the rod control power supply of the speed of respond and the steady-state current accuracy. Practical application has shown that this method solves the output current ripple and the rise and fall time of the current, meeting the working requirements of rod control power supply in nuclear power plants.

关键词： Accuracy Power supplies Conferences Asia Rectifiers Power electronics Steady-state

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Effects of Inlet Conditions on the Two-Phase Flow Water Hammer Transients in Elastic Tube 23rd

Effects of Inlet Conditions on the Two-Phase Flow Water Hamm...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhao, Zixiang Duan, Zhongdi Xue, Hongxiang Yuan, Yuchao Liu, Shiwen State Key Laboratory of Ocean Engineering Shanghai Jiao Tong University Shanghai China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789819910229

Two-phase flow water hammer events occur in the pipelines of the nuclear power systems and lead to transient and violent pressure shock to tube structures. For the sake of operation safety, the occurrence and severity of the two-phase water hammer should be carefully assessed. This paper presents a parameter analysis of the inlet conditions on the two-phase flow water hammer transients, with considering the elastic effect of the tube walls. A numerical model is established for the vapor-liquid two-phase flow based on the two-fluid six-equation modelling approach, with incorporating correlations and criterions for two-phase flow regime, interfacial interactions and heat transfer. The governing equations are transformed to matrix form expressed by characteristic variables, and solved using the splitting operator method and the total variation diminishing scheme. The accuracy of the model is verified against the experimental data in open literature. Then, the model is applied to investigate the effect of inlet velocity and inlet water temperature on the two-phase flow water hammer transients. The simulation results show that the increase of inlet velocity increases the pressure peak values and brings forward the onset of water hammer, and the increase of inlet temperature decreases the pressure shock. A comparison of the water hammer results between the elastic tube and rigid tube is further presented, and the effect of the elastic modulus on the water hammer is analyzed. The results also show that the pressure peak is largely affected by the tube diameter. © 2023, The Author(s).

关键词： Condensation

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Reliability Analysis of Nuclear Safety-Class DCS ESFAS Based on FTA 7th

Reliability Analysis of Nuclear Safety-Class DCS ESFAS Based...

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7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Zhang, Xu Peng, Hao Feng, Shi-Man Yang, Jing-Hua Chen, Shi-Yong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China PowerChina Sichuan Electric Power Engineering Co. Ltd Chengdu610041 China

ISBN: (纸本)9789819934546

Nuclear safety-class DCS realizes reactor safety protection function, which is directly related to the safe and stable operation of nuclear power plants, so its reliability should be analyzed and optimized. In this paper, the engineered safety features actuation system in nuclear safety-class DCS is analyzed through the control instruction logic flow involved in the structure, according to the components of the series and parallel or voting logic and other logical relations to bulid a fault tree model. Using fault tree model, the reliability degradation trend is obtained, and the minimum cut set, importance degree and other parameters are quantitatively calculated. According to overhaul cycle of nuclear power plant, a maintenance optimization method for engineered safety features actuation system is proposed, and some optimization suggestions are given. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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NUMERICAL INVESTIGATION ON DIRECT CONTACT CONDENSATION CHARACTERISTICS OF STEAM INJECTION INTO COLD-WATER PIPE UNDER HEAVING CONDITION 30

NUMERICAL INVESTIGATION ON DIRECT CONTACT CONDENSATION CHARA...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Wang, Zhiwei He, Yanping Duan, Zhongdi Huang, Chao Liu, Shiwen State Key Laboratory of Ocean Engineering Shanghai Jiao Tong University Shanghai200240 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Direct contact condensation (DCC) is widely appeared in the nuclear power plants and will lead to serious temperature and pressure fluctuations. For ocean nuclear power plants, the DCC is inevitably affected by the sea conditions, such as rolling, heaving motion. In this study, the integration of volume of fluid model (VOF) and condensation model is adopted to simulate the DCC characteristics under heave conditions. The interfacial heat and mass transfer are calculated by user-defined functions (UDF). The additional inertial force method is used to describe the heave motion. First, the DCC numerical model under heave motion is validated by the experimental data, which coincides well with the experiment results. Then, the characteristics of DCC under different heave parameters are numerically studied. The results show that the average condensation rate increases with an increase in heave amplitude, decreases with an increase in heave period. This is mainly because the heave motion will cause the liquid phase to impact the upper pipe wall, and increasing the contact area of vapor and subcooled water. The pressure oscillations induced by DCC are more complicated under heaving conditions. With heave period continuing to change, the formation mechanisms of the pressure peak are different. When the heave period is relatively small, the pressure peak is mainly induced by the acceleration of water phase under high frequency heaving motion. With an increase in the heave period, the heaving motion accelerated the formation of isolated steam bubble, and caused the condensation induced water hammer (CIWH) events, which in turn induced the pressure peak. In addition, the pressure peak is proportional to the heave amplitude. In summary, the direct contact condensation characteristics are more sensitive to the heave period under heaving conditions. These qualitative conclusions may serve as a reference for direct contact condensation simulations under heaving conditions. © 2023 by

关键词： Nuclear power plants

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Experimental Study of Turbulent Bubbly Flow in 180-Degree Elbow by PIV-PS Technique

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International Journal of Advanced Nuclear reactor design and technology 2022年第3期4卷 109-120页

作者： Chen, Wuguang Li, Songwei Mu, Xiaocheng Yang, Fan Yan, Zhigang Sun, Jiajin Huang, Guangyuan Cai, Kangbei Hu, Yuanbiao Yin, Junlian Wang, Dezhong School of Nuclear Science and Engineering Shanghai Jiao Tong University Shanghai200240 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China CGN Inspection Technology Co. Ltd. Jiangsu Suzhou215100 China

The curvature effect and bubbles are the key factors that affect the gas-liquid two-phase flow and turbulent characteristics in the 180-degree elbow. Limited by the difficulty of experimental measurement and numerical investigation, little is known about the two-phase flow evolution and turbulent modulation in 180-degree elbow. In this paper, a gas-liquid two-phase measurement technique, simultaneous Particle Image Velocimetry and Pulsed Shadowgraphy (PIV-PS) has been developed by combining two cameras through a spectroscope and optical filters and was applied to explore the two-phase characteristics in 180-degree elbow. In addition, an image post-processing method, based on bubble area detection and remaining particles estimation, was proposed. The characteristics of velocity and turbulent fields in the 180-degree elbow are experimentally studied in different conditions, and the evolution of the flow field in the elbow was also explored in detail. The results show that the horizontal velocity on the centerline of the elbow increases linearly with the liquid flow rates. However, with the introduction of bubbles, the linear feature is eliminated. The turbulent fields show that the introduction of bubbles suppresses turbulence at the low void fraction condition, and enhances at the high void fraction in bubble accumulation areas. © 2022 Xi'an Jiaotong University

关键词： Turbulence

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Prediction of Electromagnetic Shielding Effectiveness of DCS Cabinet Based on Neural Network 7th

Prediction of Electromagnetic Shielding Effectiveness of DCS...

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7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Zhang, Xu Yang, You-wei Yang, Jing-hua Li, Jun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China PowerChina Sichuan Electric Power Engineering Co. Ltd Chengdu610041 China

ISBN: (纸本)9789819934546

The shielding effectiveness (SE) of structural components is one of the important indicators of electromagnetic compatibility (EMC) performance, and it is related to the size, material, frequency and test points position in the structure. The measurement of SE is governed by a variety of factors. SE of structural components is affected by the measuring antenna size which is adopted to different frequencies and the testability of certain observation points. To address this issue, this paper introduces an example of the SE of DCS cabinet of nuclear power plant control system, where the neural network algorithm was used to build the relationship between spatial location and SE in the cabinet at specified frequency, so that the SE of unknown points could be predicted by the SE of known points. Besids, the model parameters were optimized by training with a large amount of data, the HPR1000 safety-class DCS cabinet was adopted for the verification, and high prediction accuracy was obtained. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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