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< Page ,Total 89 >
Preliminary conceptual design and analysis of a 100 kW(e) level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) EI SCIE Scopus
期刊论文 | 2022 , 46 (14) , 19653-19666 | INTERNATIONAL JOURNAL OF ENERGY RESEARCH
SCOPUS Cited Count: 2
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Abstract :

To meet the requirements of the unmanned underwater vehicles (UUVs) for the power source, a conceptual design of a 100-kW(e) level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) is proposed. The NUSTER-100 is a 1-MWt reactor that has been designed to couple with a 100-kW(e) thermoelectric generator system. The high temperature sodium heat pipes are adopted to cool the reactor core. The cascaded wide temperature range thermoelectric generation (TEG) system is used to achieve thermoelectric conversion, which ensures high power generation efficiency and low operating noise. The neutronic and thermal-mechanical analysis results are provided for the design. The core power distribution and energy spectrum are obtained. The neutronic analysis results are imported into ANSYS as input for thermal-mechanical analysis. In this paper, the fuel-HPs channel and the 1/8 symmetrical core are selected as the analysis objects, and the temperature distribution and statics parameters under the normal operation condition are obtained. The analysis results show that under normal operation conditions, the core parameters meet the design requirements. These design parameters are all lower than the safety limit. This work provides a certain reference for the design and application of the heat pipe reactor power system.

Keyword :

conceptual design heat pipe cooled reactor safety analysis thermal-electrical reactor unmanned underwater vehicle

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GB/T 7714 Huang, Jinlu , Wang, Chenglong , Tian, Zhixing et al. Preliminary conceptual design and analysis of a 100 kW(e) level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) [J]. | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2022 , 46 (14) : 19653-19666 .
MLA Huang, Jinlu et al. "Preliminary conceptual design and analysis of a 100 kW(e) level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100)" . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH 46 . 14 (2022) : 19653-19666 .
APA Huang, Jinlu , Wang, Chenglong , Tian, Zhixing , Guo, Kailun , Su, G. H. , Tian, Wenxi et al. Preliminary conceptual design and analysis of a 100 kW(e) level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) . | INTERNATIONAL JOURNAL OF ENERGY RESEARCH , 2022 , 46 (14) , 19653-19666 .
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LES study on the turbulent thermal stratification and thermo-mechanical fatigue analysis for NPP surge line EI SCIE Scopus
期刊论文 | 2022 , 178 | INTERNATIONAL JOURNAL OF THERMAL SCIENCES
SCOPUS Cited Count: 7
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Abstract :

Thermal stratification caused the significant temperature differentials from top to bottom inside the pipe that imposed fluctuating high stresses on the wall of the pipe and eventually could lead to severe structural integrity concerns and degrade fatigue strength. This study aimed to evaluate the transient thermal distribution and fatigue damage caused by thermal stratification in the pressurizer surge line (SL) of Nuclear Power Plants (NPP). In this paper, the effects of nonuniform temperature distributions caused by thermal stratification were investigated by performing a detailed Computational Fluid Dynamic (CFD) study using the Large Eddy Simulation (LES) approach. The fluid flow was simulated using a full buoyancy model and conjugate heat transfer at the pipe inner wall and the fluid interface. Numerical results of the investigation in the SL pipe were validated with similar experimental setup data obtained from a prototype experimental facility of a pressurized water reactor. Fatigue damage due to induced thermal stresses in the SL model was investigated and fatigue hotspots were also identified. This work provides a comprehensive case study of thermal stratification for estimating pressurizer surge pipeline structural integrity.

Keyword :

LES Temperature fluctuations Thermal stratification Thermo-mechanical fatigue

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GB/T 7714 Muhammad, Naveed , Wang, Mingjun , Tian, Wenxi et al. LES study on the turbulent thermal stratification and thermo-mechanical fatigue analysis for NPP surge line [J]. | INTERNATIONAL JOURNAL OF THERMAL SCIENCES , 2022 , 178 .
MLA Muhammad, Naveed et al. "LES study on the turbulent thermal stratification and thermo-mechanical fatigue analysis for NPP surge line" . | INTERNATIONAL JOURNAL OF THERMAL SCIENCES 178 (2022) .
APA Muhammad, Naveed , Wang, Mingjun , Tian, Wenxi , Su, Guanghui , Qiu, Suizheng . LES study on the turbulent thermal stratification and thermo-mechanical fatigue analysis for NPP surge line . | INTERNATIONAL JOURNAL OF THERMAL SCIENCES , 2022 , 178 .
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Isothermal experiments on steam oxidation of magnetron-sputtered chromium-coated zirconium alloy cladding at 1200 degrees C EI SCIE Scopus
期刊论文 | 2022 , 206 | CORROSION SCIENCE
SCOPUS Cited Count: 30
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Abstract :

The isothermal steam oxidation behavior of chromium-coated zirconium alloy cladding at 1200 degrees C is studied. The cladding substrate is Zr-1 Nb alloy with outer diameter of 9.5 mm and thickness of 0.57 mm. The Cr coating is prepared by magnetron sputtering method with thickness of about 11 mu m. The oxidation durations are 300 s similar to 14,400 s. After oxidation, blisters are formed on the surface due to the compressive stress in coating, and the coating surfaces are porous or exhibit a morphology with ravines and granular crystals. A four-layer structure (Cr2O3, Cr, Zr(Cr,Fe)(2) and Zr from outside to inside) is formed in the early stage of oxidation. However, Zr4+ can migrate into Cr coating through grain boundaries and reacts with Cr2O3. The redox reaction between Zr and Cr2O3 leads to the coating failure, resulting in the oxidation of Zr alloy substrate. Due to selective oxidation, a certain thickness of Cr coating is retained during the thickening of outer ZrO2. The emergence of outer ZrO2 layer is significantly delayed by Cr coating, and Cr coating is able to reduce the growth rate of outer ZrO2 layer. Therefore, Cr coating has a protective effect in 1,200 degrees C steam environment and can improve the oxidation resistance of Zr alloy cladding.

Keyword :

Accident tolerant fuel Chromium coating High -temperature steam oxidation Zirconium alloy

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GB/T 7714 Wang, Dong , Zhong, Ruhao , Zhang, Yapei et al. Isothermal experiments on steam oxidation of magnetron-sputtered chromium-coated zirconium alloy cladding at 1200 degrees C [J]. | CORROSION SCIENCE , 2022 , 206 .
MLA Wang, Dong et al. "Isothermal experiments on steam oxidation of magnetron-sputtered chromium-coated zirconium alloy cladding at 1200 degrees C" . | CORROSION SCIENCE 206 (2022) .
APA Wang, Dong , Zhong, Ruhao , Zhang, Yapei , Chen, Peng , Lan, Yicong , Yu, Jian et al. Isothermal experiments on steam oxidation of magnetron-sputtered chromium-coated zirconium alloy cladding at 1200 degrees C . | CORROSION SCIENCE , 2022 , 206 .
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Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor EI SCIE Scopus
期刊论文 | 2022 , 392 | NUCLEAR ENGINEERING AND DESIGN
SCOPUS Cited Count: 6
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Abstract :

Nuclear thermal propulsion utilizes the nuclear reactor rather than the combustion chamber to yield thermal energy. Propellant hydrogen could dissociate in the high-temperature reactor, which has an important effect on thermal hydraulic performance of the reactor. In this study, a one-dimensional steady-state analysis code has been developed for studying the behavior of hydrogen flowing through the high-temperature coolant channel. Thermal dissociation and real gas thermophysical property models of hydrogen were proposed and considered in the calculation models. It was found that the model validation deviations of thermophysical properties were within +/- 5% in the range of 200 - 3000 K and 0.01 - 10.0 MPa. Developed models were reliable and accurate with validation. Thermal-hydraulic behaviors of hydrogen in NRX-A6 reactor channel were analyzed. When dissociation occurred, the variation of properties was larger than those without dissociation, which enhanced heat transfer. The degree of dissociation was small and xH,out was just 0.456% under the design condition. The power density was the most significant influence factor, especially under the high power density. xH,out was 3.2 times than that of design condition as the power density grew 20%. This study can provide an approach to study hydrogen dissociation phenomena in nuclear thermal propulsion reactors.

Keyword :

Code validation Dissociation Hydrogen Nuclear thermal propulsion Thermal hydraulics Thermophysical properties

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GB/T 7714 Fang, Yuliang , Wang, Chenglong , Tian, Wenxi et al. Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor [J]. | NUCLEAR ENGINEERING AND DESIGN , 2022 , 392 .
MLA Fang, Yuliang et al. "Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor" . | NUCLEAR ENGINEERING AND DESIGN 392 (2022) .
APA Fang, Yuliang , Wang, Chenglong , Tian, Wenxi , Zhang, Dalin , Su, Guanghui , Qiu, Suizheng . Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor . | NUCLEAR ENGINEERING AND DESIGN , 2022 , 392 .
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Heat Pipe Design and Heat Transfer Performance in Residual Heat Removal System EI
期刊论文 | 2021 , 55 (6) , 1000-1006 | Atomic Energy Science and Technology
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Abstract :

In this paper, the optimal design process of a single heat pipe and its heat and mass transfer mathematical and physical model were established. Considering the working environment of the heat pipe, the heat pipe used for heat exchanger in the passive residual heat removal system of the nuclear reactor was completely designed, and its heat transfer characteristics were analyzed. The analysis shows that the composite wick heat pipe meets the heat transfer requirements of the residual heat removal system, and its heat transfer power is mainly affected by the capillary limit, boiling limit and total thermal resistance of the heat pipe. With the same wick thickness, the capillary limit of the composite wick heat pipe is 100% to 700% higher than that of the single wire mesh wick heat pipe. Changing the outer diameter of the heat pipe or thickness of the wick, that is, reducing the diameter of the steam cavity, the boiling limit significantly reduces. When the heat transfer power of a single heat pipe is greater than 1 kW, the length of each section of the heat pipe is 0.4, 0.2 and 0.4 m, the outer diameter is 30 mm, and the wick with a thickness of 2 mm is a composite wire structure of 400 mesh+50 mesh. This paper provides theoretical support for the design of high performance heat pipe heat exchangers and the analysis of heat transfer characteristics. © 2021, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Capillary flow Composite structures Design Heat exchangers Heat pipes Heat resistance Heat transfer performance Mass transfer Mesh generation Nuclear reactors

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GB/T 7714 Duan, Qianni , Wang, Chenglong , Zhang, Dalin et al. Heat Pipe Design and Heat Transfer Performance in Residual Heat Removal System [J]. | Atomic Energy Science and Technology , 2021 , 55 (6) : 1000-1006 .
MLA Duan, Qianni et al. "Heat Pipe Design and Heat Transfer Performance in Residual Heat Removal System" . | Atomic Energy Science and Technology 55 . 6 (2021) : 1000-1006 .
APA Duan, Qianni , Wang, Chenglong , Zhang, Dalin , Qiu, Suizheng , Su, Guanghui , Tian, Wenxi et al. Heat Pipe Design and Heat Transfer Performance in Residual Heat Removal System . | Atomic Energy Science and Technology , 2021 , 55 (6) , 1000-1006 .
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数值堆热工流体程序CVR-PACA验证及典型应用 EI CSCD
期刊论文 | 2021 , 55 (9) , 1569-1580 | 原子能科学技术
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Abstract :

谱元方法是一种高精度的数值计算方法,采用该方法开发了数值堆高精度热工水力并行C FD计算程序CVR-PACA.应用CVR-PACA对单棒光棒通道湍流流场、3×3光棒棒束湍流流场、Matis-H压水堆棒束通道基准题、19棒带绕丝组件通道湍流流场进行了仿真计算.通过与实验测量值对比,研究定量验证了大涡模拟(LES)模型及非稳态雷诺时均(URANS)模型对各类棒束通道流场预测的准确性.算例在建模过程中采用网格分裂技术实现了复杂几何的纯六面体网格划分,用于支撑谱元方法计算.研究较为全面地积累了高精度谱元方法模拟流场流动及换热的建模经验,获取了各类棒束通道内丰富的流动和换热细节,获得的建模经验能更加精准有力地指导相关设计的优化改进.

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GB/T 7714 王明军 , 鞠浩然 , 赵民富 et al. 数值堆热工流体程序CVR-PACA验证及典型应用 [J]. | 原子能科学技术 , 2021 , 55 (9) : 1569-1580 .
MLA 王明军 et al. "数值堆热工流体程序CVR-PACA验证及典型应用" . | 原子能科学技术 55 . 9 (2021) : 1569-1580 .
APA 王明军 , 鞠浩然 , 赵民富 , 李伟卿 , 刘天才 , 胡长军 et al. 数值堆热工流体程序CVR-PACA验证及典型应用 . | 原子能科学技术 , 2021 , 55 (9) , 1569-1580 .
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Experimental and numerical investigation on characteristics of MCCI with exothermic thermite EI SCIE
期刊论文 | 2021 , 384 | NUCLEAR ENGINEERING AND DESIGN
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Abstract :

Investigation was conducted experimentally and numerically for ex-vessel melt behavior and concrete interaction during Molten Corium-Concrete Interaction (MCCI) in current research. The CINA (Corium-Concrete Interaction Apparatus) experiment was carried out with a total of 94 kg melt generated by exothermic thermite chemical reaction to react in a two-dimensional (2-D) cylindrical siliceous crucible with an inner diameter of 300 mm and a height of 500 mm. Meanwhile, the variation of melt temperature was monitored during the experiment. Besides, the accurate ablation depth and thickness of stratified layers were measured after the experiment. Numerical simulation was used to study the mechanism of high-temperature melt heat transfer and concrete ablation during the MCCI process. The slag film model to calculate the melt/concrete interface heat transfer coefficient and the stratified melt model were both used to simulate the above experiment. The results showed that the numerical results were in good agreement with experiment measurements by the slag film model under separated layers condition.

Keyword :

Concrete ablation MCCI Melt stratification Severe accident

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GB/T 7714 Xu, Zhichun , Zhang, Yapei , Zhang, Kui et al. Experimental and numerical investigation on characteristics of MCCI with exothermic thermite [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 384 .
MLA Xu, Zhichun et al. "Experimental and numerical investigation on characteristics of MCCI with exothermic thermite" . | NUCLEAR ENGINEERING AND DESIGN 384 (2021) .
APA Xu, Zhichun , Zhang, Yapei , Zhang, Kui , Wu, Zijie , Zhan, Dekui , Su, G. H. et al. Experimental and numerical investigation on characteristics of MCCI with exothermic thermite . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 384 .
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Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles EI
期刊论文 | 2021 , 55 (11) , 1930-1938 | Atomic Energy Science and Technology
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At present, the prediction of the critical heat flux (CHF) in the rod bundles is mostly based on the experimental correlations, which is limited by the specific application range, and cannot be effectively extrapolated or the prediction accuracy is reduced. In order to meet the prediction requirements of CHF in different light water reactors, it is necessary to develop a wide range of CHF prediction methods for different geometries and thermal boundaries. Based on the sub-channel analysis method, the two types of critical phenomena were considered in the paper, which were departure from nucleate boiling (DNB) and dryout, and the CHF in the bundles through coupling sub-channel analysis code and the CHF mechanistic model was calculated. By comparing with the experimental data of CHF, it is proved that the coupling code has better prediction accuracy for the CHF in the rod bundles. © 2021, Editorial Board of Atomic Energy Science and Technology. All right reserved.

Keyword :

Codes (symbols) Forecasting Heat flux Light water reactors Nucleate boiling

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GB/T 7714 Gui, Minyang , Tian, Wenxi , Wu, Di et al. Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles [J]. | Atomic Energy Science and Technology , 2021 , 55 (11) : 1930-1938 .
MLA Gui, Minyang et al. "Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles" . | Atomic Energy Science and Technology 55 . 11 (2021) : 1930-1938 .
APA Gui, Minyang , Tian, Wenxi , Wu, Di , Chen, Ronghua , Zhang, Kui , Su, Guanghui et al. Development and Preliminary Validation of CHF Mechanistic Model for Rod Bundles . | Atomic Energy Science and Technology , 2021 , 55 (11) , 1930-1938 .
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From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction EI SCIE Scopus
期刊论文 | 2021 , 165 | Annals of Nuclear Energy
SCOPUS Cited Count: 10
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Abstract :

When a severe accident occurs in a nuclear reactor, fuel–coolant interaction (FCI) may occur and cause steam explosion. Energetic FCI or steam explosion will threaten the integrity of the containment and even cause radioactive leakage. Therefore, the research and evaluation of the morphology and development of the melt in the FCI process is very important. In recent decades, many scholars have conducted theoretical and experimental research on this phenomenon, and formed a large number of mathematical and physical models for the degradation, evolution and relocation of the melt. This study retrospects and summarizes the experimental and theoretical research on the FCI phenomenon, and describes the behavior of the melt in the entire evolution process from the melt jet to debris bed. In addition, on the basis of literature review, the models of each stage are classified according to the formation mechanisms and applicable conditions. Finally, in the present study, the future development direction is prospected on the ground of the previous research results. © 2021 Elsevier Ltd

Keyword :

Coolants Debris Fuels Nuclear reactors

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GB/T 7714 Sun, Ruiyu , Wu, Liangpeng , Ding, Wen et al. From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction [J]. | Annals of Nuclear Energy , 2021 , 165 .
MLA Sun, Ruiyu et al. "From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction" . | Annals of Nuclear Energy 165 (2021) .
APA Sun, Ruiyu , Wu, Liangpeng , Ding, Wen , Chen, Ronghua , Tian, Wenxi , Qiu, Suizheng et al. From melt jet break-up to debris bed formation: A review of melt evolution model during fuel-coolant interaction . | Annals of Nuclear Energy , 2021 , 165 .
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Operation performance analysis of a liquid metal droplet radiator for space nuclear reactor EI SCIE
期刊论文 | 2021 , 158 | ANNALS OF NUCLEAR ENERGY
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Liquid droplet radiator (LDR) is the frameless space radiator designed for heat dissipation of large-power space nuclear reactors. This paper analyzes the radiative heat transfer and evaporative characteristics of liquid metals adopted in LDR including lithium, aluminum and tin. Firstly, the cooling process of different working fluids is simulated and the temperature distribution at the droplet collector is obtained. The radiant energy of per unit mass fluid with different optical thicknesses and temperatures is calculated. Secondly, the evaporative characteristics of the working fluid are studied. Finally, the system operation performances are evaluated including the system life and radiant energy with different optical thicknesses and temperatures. Results show that liquid lithium is appropriate for heat dissipation system working under 600 K, while liquid aluminum and tin are more appropriate for working under 1200 K and 1500 K, respectively. This work provides the theoretical support for liquid droplet radiator thermal design. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

Evaporative loss Liquid droplet radiator Liquid metal Radiative heat transfer Space nuclear reactor

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GB/T 7714 Yang, Linyi , Wang, Chenglong , Qin, Hao et al. Operation performance analysis of a liquid metal droplet radiator for space nuclear reactor [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 158 .
MLA Yang, Linyi et al. "Operation performance analysis of a liquid metal droplet radiator for space nuclear reactor" . | ANNALS OF NUCLEAR ENERGY 158 (2021) .
APA Yang, Linyi , Wang, Chenglong , Qin, Hao , Zhang, Dalin , Tian, Wenxi , Su, G. H. et al. Operation performance analysis of a liquid metal droplet radiator for space nuclear reactor . | ANNALS OF NUCLEAR ENERGY , 2021 , 158 .
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