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Numerical investigations of thermal mixing performance of a hot gas mixing structure in high-temperature gas-cooled reactor 被引量:2
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作者 Yang-Ping Zhou Peng-Fei Hao +1 位作者 Xi-Wen Zhang Feng He 《Nuclear Science and Techniques》 SCIE CAS CSCD 2016年第1期149-155,共7页
A numerical simulation study was performed to clarify the thermal mixing characteristics of coolant in the core bottom structure of the high-temperature gas-cooled reactor(HTR). The flow field and temperature field in... A numerical simulation study was performed to clarify the thermal mixing characteristics of coolant in the core bottom structure of the high-temperature gas-cooled reactor(HTR). The flow field and temperature field in the hot gas chamber and the hot gas duct of the HTR were obtained based on the commercial computational fluid dynamics(CFD) program. The numerical simulation results showed that the helium flow with different temperatures in the hot gas mixing chamber and the hot gas duct mixed intensively, and the mixing rate of the temperature in the outlet of the hot gas duct reached 98 %. The results indicated many large-scale swirling flow structures and strong turbulence in the hot gas mixing chamber and the entrance of the hot gas duct, which were responsible for the excellent thermal mixing of the hot gas chamber and the hot gas duct. The calculated results showed that the temperature mixing rate of the hot gas chamber decreased only marginally with increasing Reynolds number. 展开更多
关键词 高温气冷堆 混合性能 混合结构 热气体 数值研究 计算流体动力学 数值模拟 热气导管
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Lead-Bismuth and Lead as Coolants for Fast Reactors 被引量:1
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作者 G. I. Toshinsky A. V. Dedul +2 位作者 O. G. Komlev A. V. Kondaurov V. V. Petrochenko 《World Journal of Nuclear Science and Technology》 2020年第2期65-75,共11页
Fast reactors used lead-bismuth eutectic (LBE) and lead as coolants possess very high level of inherent self-protection and passive safety against severe accident. So, population radiophobia can be overcome. That type... Fast reactors used lead-bismuth eutectic (LBE) and lead as coolants possess very high level of inherent self-protection and passive safety against severe accident. So, population radiophobia can be overcome. That type of reactors can be simultaneously more safely and more cheaply. As all other coolants, LBE and lead coolant (LC) possess the certain virtues and shortcomings. The presented report includes the comparative analysis of characteristic properties of those coolants, their impact on reactor safety, reliability and operating characteristics. The conclusion is made about promising usage of FRs with these coolants in future NP after the experience in operating of the prototypes of such reactors has been obtained. 展开更多
关键词 SVBR-100 Fast reactor lead-bismuth coolANT LEAD coolANT Nuclear Power Plant Inherent SELF-PROTECTION Melting Point 210Po BISMUTH Recourses
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Design of the material performance test apparatus for high temperature gas-cooled reactor 被引量:1
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作者 REN Cheng YANG Xing Tuan +2 位作者 LI Cong Xin LIU Zhi Yong JIANG Sheng Yao 《Nuclear Science and Techniques》 SCIE CAS CSCD 2013年第6期132-136,共5页
Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor(HTGR).To solve the problem,a material performance te... Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor(HTGR).To solve the problem,a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR.The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus.Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field.A typical experimental process was analyzed in the paper,which lasted 76 hours including seven stages.Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified.Test temperature in the apparatus could be elevated up to 1600oC,which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor. 展开更多
关键词 高温气冷反应堆 性能测试装置 屏蔽材料 设计 性能试验装置 高温气冷堆 还原气氛 实验过程
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Study on neutronics design of ordered-pebble-bed fluoride-salt- cooled high-temperature experimental reactor 被引量:3
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作者 Rui Yan Shi-He Yu +11 位作者 Yang Zou Qun Yang Bo Zhou Pu Yang Hong-Hua Peng Ya-Fen Liu Ye Dai Rui-Ming Ji Xu-Zhong Kang Xing-Wei Chen Ming-Hai Li Xiao-Han Yu 《Nuclear Science and Techniques》 SCIE CAS CSCD 2018年第6期36-44,共9页
This paper presents a neutronics design of a 10 MW ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor. Through delicate layout, a core with ordered arranged pebble bed can be formed,which ca... This paper presents a neutronics design of a 10 MW ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor. Through delicate layout, a core with ordered arranged pebble bed can be formed,which can keep core stability and meet the space requirements for thermal hydraulics and neutronics measurements.Overall, objectives of the core include inherent safety and sufficient excess reactivity providing 120 effective full power days for experiments. Considering the requirements above, the reactive control system is designed to consist of 16 control rods distributed in the graphite reflector. Combining the large control rods worth about 18000–20000 pcm, molten salt drain supplementary means(-6980 to -3651 pcm) and negative temperature coefficient(-6.32 to -3.80 pcm/K) feedback of the whole core, the reactor can realize sufficient shutdown margin and safety under steady state. Besides, some main physical properties, such as reactivity control, neutron spectrum and flux, power density distribution, and reactivity coefficient,have been calculated and analyzed in this study. In addition, some special problems in molten salt coolant are also considered, including ~6Li depletion and tritium production. 展开更多
关键词 中子物理学 反应堆 试验性 高温度 学习 设计 脉冲编码调制 控制系统
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Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor 被引量:1
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作者 ZHOUTao YANGRui-Chang JIADou-Nan 《Nuclear Science and Techniques》 SCIE CAS CSCD 2005年第1期46-52,共7页
The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calcu... The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calcu- lated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore- sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen’s formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell’s temperature. Its maximum value is 14%. 展开更多
关键词 高温气冷反应堆 压水堆 放射性微粒 热敏致电沉积 安全防护
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Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment 被引量:1
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作者 左嘉旭 靖剑平 +2 位作者 毕金生 宋维 陈堃 《Nuclear Science and Techniques》 SCIE CAS CSCD 2015年第5期112-117,共6页
Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized wat... Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done. 展开更多
关键词 高温气冷堆 概率安全评价 压水反应堆 框架分析 安全评估 氟盐 快中子反应堆 物理设计
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Corrosion of candidate materials for supercritical water-cooled reactor
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作者 ZHANG Lefu~(1)),BAO Yichen~(1)) and TANG Rui~(2)) 1) School of Nuclear Sci.&Eng,Shanghai Jiao Tong Univ.,Shanghai 200240,China 2) National Key Laboratory for Nuclear Fuel and Materials,Nuclear Power Institute of China,Chengdu 610041,China 《Baosteel Technical Research》 CAS 2010年第S1期71-,共1页
Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages... Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on. However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge. Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR) in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP). Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M) steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials. 展开更多
关键词 supercritical water cooled reactor cladding material CORROSION protective oxide film
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Primary Breeding Ratio Analysis of an Improved Supercritical Water Cooled Fast Reactor
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作者 Zijing Liu Jinsen Xie Lihua He 《World Journal of Nuclear Science and Technology》 2015年第4期253-264,共12页
The purpose of the study is to analyze the breeding ratio of a supercritical water cooled fast reactor (SCFR) and to increase the breeding core of SCFR. The sensitivities of assembly parameters, core arrangements and ... The purpose of the study is to analyze the breeding ratio of a supercritical water cooled fast reactor (SCFR) and to increase the breeding core of SCFR. The sensitivities of assembly parameters, core arrangements and fuel nuclide components to the breeding ratio are analyzed. In assembly parameters, the seed fuel rod diameter has higher sensitivities to the conversion ratio (CR) than the coolant tube diameter in blanket. Increasing heavy metal fraction is good to CR improvement. The CR of SCFR also increases with a reasonable core arrangement and Pu isotope mass fraction reduction in fuel, which can achieve more negative coolant void reactivity coefficient at the same time. The breeding ratio of SCFR is 1.03128 with a new core arrangement. And the coolant void reactivity coefficient is negative, which achieves a fuel breeding in initial fuel cycle. 展开更多
关键词 SUPERCRITICAL Water cooled Fast reactor BREEDING Ratio coolANT VOID COEFFICIENT
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Analysis of fixation method of fuel assembly for lead-alloy cooled reactor
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作者 韩骞 吴庆生 +2 位作者 陈建伟 梅华平 黄群英 《Nuclear Science and Techniques》 SCIE CAS CSCD 2015年第5期107-111,共5页
As a potential candidate for generation IV reactors, lead-alloy cooled reactor has attracted much attentions in recent years. The China LEAd-based research Reactor(CLEAR) is proposed as the primary choice for the acce... As a potential candidate for generation IV reactors, lead-alloy cooled reactor has attracted much attentions in recent years. The China LEAd-based research Reactor(CLEAR) is proposed as the primary choice for the accelerator driven subcritical system project launched by Chinese Academy of Sciences. Lead-bismuth eutectic(LBE) is selected as the coolant of CLEAR owing to its efficient heat conductivity properties and high production rate of neutrons. In order to compensate the buoyancy due to the high density of lead-alloy, fixation methods of fuel assembly(FA) have become a research hotspot worldwide. In this paper, we report an integrated system of ballast and fuel element for CLEAR FA. It guarantees the correct positioning of each FA in normal and refueling operations. Force calculation and temperature analysis prove that the FA will be stable and safe under CLEAR operation conditions. 展开更多
关键词 研究反应堆 固定方法 燃料组件 铅合金 冷却 次临界系统 中国科学院 操作条件
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Materials R & D for sodium-cooled fast reactor in China
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作者 XIE Chuchunn 《Baosteel Technical Research》 CAS 2010年第S1期73-,共1页
The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China... The study gives a brief introduction on development of innovated nuclear system in China,mainly focus on the materials R&D status for the sodium cooled fast reactor.With the high speed development of economy,China needs a huge energy supply;at same time a more cleaning energy to reduce the carbon release is demanded.The nuclear energy is the most cleaning energy at present time,especially the innovated nuclear system which is so-called GenerationⅣpower plants has got its prior development due to its safety, economical and little fission production produced.Fast breeder reactor,as the priority development reactor type in the Gen-Ⅳnuclear system,is the key to the advanced closed fuel cycle technologies.China experimental fast reactor(CEFR ) has been completed the design,construction the synthesis system commissioning and reached its physical criticality on July 21,2010.At China Institute of Atomic Energy,the CEFR and other research facilities have been established,and extensive studies are planning to carry out in the areas of fuel and materials development.This will laid the foundation for the design and development of the future's CFR—900(China Demonstration Fast Reactor) and CCFR(China Commercial Fast Reactor). Highlights of some of materials R&D studies are discussed in this paper. 展开更多
关键词 CEFR sodium-cooled fast reactor sodium compatibility irradiation property mechanical property
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Status of a Sodium Cooled Fast Reactor Technology Development Program in Korea
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作者 Chungho Cho Younggyun Kim Jinwook Chang Sang-Ji Kim Chan-Bock Lee Seong-O Kim Jong-Bum Kim Hae-Yong Jeong Yong-Bum Lee Yeong-Il. Kim 《Journal of Energy and Power Engineering》 2012年第9期1379-1397,共19页
Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. ... Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER (Korea advanced liquid metal reactor), the KAERI (Korea Atomic Energy Research Institute) is currently developing advanced SFR (sodium cooled fast reactor) design concepts that can better meet the Gen IV (Generation IV) technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper. 展开更多
关键词 Sodium cooled fast reactor BURNER metal fuel pyroprocess.
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Experimental study on the mechanism of flow blockage formation in fast reactor
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作者 Wen-Hui Jin Song-Bai Cheng Xiao-Xing Liu 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2023年第6期171-182,共12页
Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structura... Various sources of solid particles might exist in the coolant flow of a liquid metal cooled fast reactor(e.g.,through chemical interaction between the coolant and impurities,air,or water,through corrosion of structural materials,or from damaged/molten fuel).Such particles may cause flow blockage accidents in a fuel assembly,resulting in a reduction in coolant flow,which potentially causes a local temperature rise in the fuel cladding,cladding failure,and fuel melt.To understand the blockage formation mechanism,in this study,a series of simulated experiments was conducted by releasing different solid particles from a release device into a reducer pipe using gravity.Through detailed analyses,the influence of various experimental parameters(e.g.,particle diameter,capacity,shape,and static friction coefficient,and the diameter and height of the particle release nozzle)on the blockage characteristics(i.e.,blockage probability and position)was examined.Under the current range of experimental conditions,the blockage was significantly influenced by the aforementioned parameters.The ratio between the particle diameter and outlet size of the reducer pipe might be one of the determining factors governing the occurrence of blockage.Specifically,increasing the ratio enhanced blockage(i.e.,larger probability and higher position within the reducer pipe).Increasing the particle size,particle capacity,particle static friction coefficient,and particle release nozzle diameter led to a rise in the blockage probability;however,increasing the particle release nozzle height had a downward influence on the blockage probability.Finally,blockage was more likely to occur in non-spherical particles case than that of spherical particles.This study provides a large experimental database to promote an understanding of the flow blockage mechanism and improve the validation process of fast reactor safety analysis codes. 展开更多
关键词 Liquid metal cooled fast reactor Flow blockage Granular jamming Experimental study
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池式钠冷快堆堆内自然循环余热排出设计研究
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作者 周志伟 薛秀丽 +3 位作者 林超 余新太 杨勇 杨红义 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第9期1817-1824,I0001,共9页
基于装量功率比约为1 t/MW的较典型池式钠冷大功率快堆的堆内自然循环设计,采用ERAC程序,对两组4种不同事故余热排出系统(DHRS)输入条件下,反应堆在紧急停堆后堆内的自然循环余热排出特性进行分析。结果表明,从DHRS启用到其对堆芯产生... 基于装量功率比约为1 t/MW的较典型池式钠冷大功率快堆的堆内自然循环设计,采用ERAC程序,对两组4种不同事故余热排出系统(DHRS)输入条件下,反应堆在紧急停堆后堆内的自然循环余热排出特性进行分析。结果表明,从DHRS启用到其对堆芯产生显著冷却效应,需要较长时间,在千秒量级。在该段时间内,堆芯余热的排出依靠反应堆固有的热工流体安全特性。随后,在堆内关键温度上升到限值之前启用DHRS带出池内热量,使堆内关键温度处于下降趋势即可满足安全要求;相比将独立热交换器(DHX)布置在冷池,将其布置在热池时,热池温度及主容器壁温相对要低,这有利于主容器的温度控制,其效果要优于布置于冷池。另外,不同布置会对堆芯盒内、盒间流流量产生影响,但总体上对堆芯的冷却效应影响不大;池式钠冷快堆余热排出设计中,要充分利用固有热工流体安全特性,降低对DHRS的时效性要求。可以考虑将全部的DHX都布置在热池,并缩小设备体积、降低散热功率设计值,或在不降低安全性的前提下选用其他更经济便捷的有效方式等,以此大幅降低余热排出设备投入成本,降低反应堆运行成本,提高经济性。本文研究结果可为我国后续的商用快堆、一体化快堆等池式液态金属堆的堆内自然循环设计提供重要参考。 展开更多
关键词 大功率快堆 钠冷快堆 自然循环 余热排出 固有安全 热工流体安全特性 盒间流
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钠基纳米流体中钠原子吸附行为特性模拟计算
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作者 朴君 李春晖 +3 位作者 阿不都赛米·亚库甫 张智刚 王荣东 矫彩山 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第9期1912-1919,共8页
钠基纳米流体利用过渡金属纳米颗粒吸附钠原子的性质,有效降低了钠原子的反应性,进而有效抑制钠火及钠水事故严重性,但目前相关吸附行为及规律尚不明确。研究基于密度泛函理论和电子结构分析,计算分析了钛、铁和铜纳米团簇(TM_(n),TM=Ti... 钠基纳米流体利用过渡金属纳米颗粒吸附钠原子的性质,有效降低了钠原子的反应性,进而有效抑制钠火及钠水事故严重性,但目前相关吸附行为及规律尚不明确。研究基于密度泛函理论和电子结构分析,计算分析了钛、铁和铜纳米团簇(TM_(n),TM=Ti、Fe、Cu,n=2~13)及其与钠原子间形成复合物(Na-TM_(n))的结构和性质,分析了TM_(n)的稳定性以及其与Na原子间相互作用。结果表明,Ti_(n)具有最高的稳定性,但其吸附钠原子的能力低于Fe_(n)和Cu_(n)。钠原子主要通过范德华作用吸附于TM_(n)表面,且两者间的电荷转移行为使得TM_(n)带负电荷。 展开更多
关键词 钠冷快堆 纳米流体 过渡金属 计算化学 团簇
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钠冷快堆小栅板联箱压降对组件流量分配影响研究
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作者 林超 高鑫钊 +1 位作者 周志伟 余新太 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第9期1859-1865,共7页
钠冷快堆堆芯采用大栅板联箱、小栅板联箱和组件的三级流量分配方式,小栅板联箱的压降影响组件的流量分配,进而影响堆芯的安全,因此进行钠冷快堆小栅板联箱压降对组件流量分配影响研究有重要意义。根据小栅板联箱压降造成组件流量分配... 钠冷快堆堆芯采用大栅板联箱、小栅板联箱和组件的三级流量分配方式,小栅板联箱的压降影响组件的流量分配,进而影响堆芯的安全,因此进行钠冷快堆小栅板联箱压降对组件流量分配影响研究有重要意义。根据小栅板联箱压降造成组件流量分配偏差的机理,提出了理论计算模型和堆芯组件优化设计的方法,并针对中国实验快堆(CEFR)堆芯进行了组件压降的优化设计,通过优化设计降低了CEFR燃料组件流量分配负偏差。结果表明,在进行钠冷快堆堆芯热工水力设计时,需要结合实际堆芯布置分析组件压降设计值的优化方向,并进行敏感性分析,以确定组件的最优设计压降,将小栅板联箱压降对组件流量分配影响降低到最低程度。本文结果可为钠冷快堆堆芯热工水力设计提供参考。 展开更多
关键词 钠冷快堆 堆芯 小栅板联箱 热工水力 流量分配
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熔盐堆下舱室非能动冷却系统的优化设计
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作者 梅牡丹 周翀 +2 位作者 傅瑶 邹杨 王纳秀 《核技术》 EI CAS CSCD 北大核心 2024年第8期107-118,共12页
熔盐堆下舱室非能动冷却系统是确保反应堆安全运行的重要保障,其结构设计是热工水力设计中的重要一环,其功能是保证熔盐堆下堆舱所有设备在反应堆正常运行时不超温,同时在事故工况下,能够最大程度地导出堆芯衰变热。基于一种热功率为153... 熔盐堆下舱室非能动冷却系统是确保反应堆安全运行的重要保障,其结构设计是热工水力设计中的重要一环,其功能是保证熔盐堆下堆舱所有设备在反应堆正常运行时不超温,同时在事故工况下,能够最大程度地导出堆芯衰变热。基于一种热功率为153 MWt的百兆瓦级熔盐堆的概念设计,建立了熔盐堆下堆舱的1/4结构模型,使用ANSYS FLUENT 20.1软件进行下堆舱三维流场与温度场的数值模拟,通过优化下舱室非能动冷却系统的结构布局、空气环腔的结构尺寸、隔热板上保温棉厚度以及进风管的入口位置,使得下舱室内双通道非能动空冷系统的热屏蔽效果最好,且在事故工况下导出堆芯衰变热最多。结果表明:改变空冷系统中空气环腔的结构尺寸对下堆舱热屏蔽结果的影响很小;在空冷系统的中间隔板上增加保温棉可以显著降低侧面混凝土墙的温度;冷却系统的进风管入口位置距离空冷环腔顶端越近热屏蔽效果越好。据此最终设计出了一种新型的下舱室内双通道非能动空冷系统,达到了153 MWt熔盐堆下堆舱的屏蔽冷却的设计要求。为未来大功率熔盐堆下舱室内非能动余热排出系统的工程优化设计提供了重要参考。 展开更多
关键词 熔盐堆下舱室 双通道非能动冷却系统 热屏蔽设计 结构优化 计算流体力学
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微型铅铋反应堆小型化与轻量化设计优化方法研究
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作者 刘紫静 赵鹏程 李琼 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第11期2299-2309,共11页
微型铅铋反应堆在实现核能综合利用方面具有独特优势,是未来可移动能源供给技术的重要选项,其小型化和轻量化是提高铅铋核动力装置总体性能的源头和关键。本文针对微型铅铋反应堆小型化和轻量化设计优化中的多物理、多变量、多约束耦合... 微型铅铋反应堆在实现核能综合利用方面具有独特优势,是未来可移动能源供给技术的重要选项,其小型化和轻量化是提高铅铋核动力装置总体性能的源头和关键。本文针对微型铅铋反应堆小型化和轻量化设计优化中的多物理、多变量、多约束耦合影响难题,首先通过分析燃料/冷却剂、固体慢化剂/反射层材料对堆芯临界尺寸及质量的影响开展了燃料/材料选型,然后采用自主开发的铅铋反应堆多物理智能设计优化平台DOPPLER开展了堆芯多因素协同优化设计,提出了一种小型化与轻量化的5 MWt微型铅铋反应堆概念设计方案MILLER-5,堆内装载核燃料139.8 kg,功率密度为114.8 W/cm^(3),换料周期为1000 d,堆芯具备平稳的反应性波动与平坦的功率分布,反应性系数均为负值,且稳态热工安全裕量较大。 展开更多
关键词 铅铋反应堆 小型化 轻量化 优化设计 物理热工特性
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热管堆单根热管失效事故瞬态数值分析研究
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作者 韩冶 杨思远 +2 位作者 文青龙 柴宝华 张亚坤 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第9期1920-1929,共10页
热管冷却反应堆(热管堆)如果发生单根热管失效事故,可能会超过热管最大允许温度和功率并出现级联失效。本文以热管堆堆芯为研究对象,通过建立单根热管失效事故瞬态计算模型,利用ANSYS Mechanical APDL程序对不同工况单根热管失效事故进... 热管冷却反应堆(热管堆)如果发生单根热管失效事故,可能会超过热管最大允许温度和功率并出现级联失效。本文以热管堆堆芯为研究对象,通过建立单根热管失效事故瞬态计算模型,利用ANSYS Mechanical APDL程序对不同工况单根热管失效事故进行了瞬态数值分析研究。结果表明:最恶劣工况是处于靠近外围热管失效的工况,基体最高温度为1314.16 K,芯块中心最高温度达到1352.49 K,热管最高工作温度为1149.84 K,均未超出容许工作温度限值,约123 s达到新的稳态;在最恶劣工况下,靠近中心的热管最高功率为83709.87 W,未超出热管传热极限范围,并能顺利达成新的稳态进行工作,不会有整体级联失效的风险。本文结果可为该堆型的设计提供热管失效事故的参考,并为堆芯结构设计奠定基础。 展开更多
关键词 热管冷却反应堆 热管失效 级联失效 瞬态数值分析
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热管冷却反应堆系统研究进展和挑战
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作者 田文喜 王成龙 +2 位作者 郭凯伦 秋穗正 苏光辉 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第S02期340-354,共15页
热管是一种高效的非能动热量传递元件,热管冷却反应堆核动力系统在多场景特种应用领域具备独特优势。本文概述了热管冷却反应堆特种核动力系统发展情况。首先介绍了热管冷却反应堆概念提出以及在海陆空天等领域的应用场景分析,同时总结... 热管是一种高效的非能动热量传递元件,热管冷却反应堆核动力系统在多场景特种应用领域具备独特优势。本文概述了热管冷却反应堆特种核动力系统发展情况。首先介绍了热管冷却反应堆概念提出以及在海陆空天等领域的应用场景分析,同时总结了国内外典型堆型的发展现状。其次探讨了当前热管冷却反应堆面临的关键技术挑战,包括高性能材料研究、高性能热管研制、高效能量转换技术研究、设计分析技术研究。最后对未来发展趋势进行了分析和展望,强调了整体系统一体化研制、发电器件特性研究以及智能自主控制技术在热管冷却反应堆领域的重要性。本文的系统性总结将推动热管冷却反应堆技术的进一步发展,为未来特种核动力系统的应用提供重要支持。 展开更多
关键词 热管 热管冷却反应堆 特种核动力系统 关键技术挑战
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基于FR-Sdaso程序对FFTF LOFWOS Test#13基准例题的热工水力分析
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作者 杨军 叶尚尚 王利霞 《原子能科学技术》 EI CAS CSCD 北大核心 2024年第9期1884-1892,共9页
FFTF LOFWOS Test#13是美国FFTF钠冷回路式快堆进行的无保护失流试验,目的是为了证明反应堆的固有安全特性。本文采用中国原子能科学研究院自主开发的FR-Sdaso程序参加了IAEA策划发起的基于该试验的基准例题项目。利用FR-Sdaso程序将一... FFTF LOFWOS Test#13是美国FFTF钠冷回路式快堆进行的无保护失流试验,目的是为了证明反应堆的固有安全特性。本文采用中国原子能科学研究院自主开发的FR-Sdaso程序参加了IAEA策划发起的基于该试验的基准例题项目。利用FR-Sdaso程序将一回路主泵转速、二回路流量和空气热交换器出口钠温作为边界条件,建立了FFTF基准例题模拟模型。计算结果与FFTF LOFWOS Test#13试验结果对比分析表明,FR-Sdaso程序能较好地预测无保护失流事故后反应堆功率以及一、二回路温度和流量的瞬态变化,自然循环阶段反应堆衰变功率计算值与试验值的最大相对偏差为−7.1%,一回路3个环路自然循环流量与初始稳态值的最大相对偏差为0.65%。对于第2排和第6排PIOTA组件,由于模拟中未考虑瞬态过程中堆芯功率分布变化和组件之间的传热,出口温度的计算值较试验测量值最大偏高25.5℃,计算结果更保守。FR-Sdaso程序对FFTF LOFWOS Test#13基准例题的分析初步验证了程序堆芯和一、二回路热工水力模型的正确性。 展开更多
关键词 钠冷快堆 FFTF基准例题 系统分析程序 FR-Sdaso 程序验证
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