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Experimental Studies of Heat Transfer Characteristics and Properties of the Cross-Flow Pipe Flow Melt Lead 被引量:1
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作者 Alexandr Viktorovich Beznosov Mikhail Vladimirovich Yarmonov +3 位作者 Artyom Dmitrievich Zudin Alexey Sergeevich Chernysh Olga Olegovna Novogilova Tatyana Alexsandrovna Bokova 《Open Journal of Microphysics》 2014年第4期54-65,共12页
The process of heat transfer in a HLMC cross-flow around heat-transfer tubes is not yet thoroughly studied. Therefore, it is of great interest to carry out experimental studies for determining the heat transfer charac... The process of heat transfer in a HLMC cross-flow around heat-transfer tubes is not yet thoroughly studied. Therefore, it is of great interest to carry out experimental studies for determining the heat transfer characteristics in a lead coolant cross-flow around tubes. It is also interesting to explore the velocity and temperature fields in a HLMC flow. To achieve this goal, experts of the NNSTU performed the work aimed at the experimental determination of the temperature and velocity fields in high-temperature lead coolant cross-flows around a tube bundle. The experimental studies were carried out in a specially designed high-temperature liquid-metal facility. The experimental facility is a combination of two high-temperature liquid-metal setups, i.e., FT-2 with a lead coolant and FT-1 with a lead-bismuth coolant, united by an experimental site. The experimental site is a model of the steam generator of the BREST-300 reactor facility. The heat-transfer surface is an in-line tube bank of a diameter of 17 × 3.5 mm, which is made of 10H9NSMFB ferritic-martensitic steel. The temperature of the heat-transfer surface is measured with thermocouples of a diameter of 1 mm being installed in the walls of heat-transfer tubes. The velocity and temperature fields in a high-temperature HLMC flow are measured with special sensors installed in the flow cross section between the rows of heat-transfer tubes. The characteristics of heat transfer and velocity fields in a lead coolant flow were studied in different directions of the coolant flow: The vertical (“top-down” and “bottom-up”) and the horizontal ones. The studies were conducted under the following operating conditions: The temperature of lead was t = 450°C - 5000°C, the thermodynamic activity of oxygen was a = 10-5 - 100, and the lead flow through the experimental site was Q = 3 - 6 m3/h, which corresponds to coolant velocities of V = 0.4 - 0.8 m/s. Comprehensive experimental studies of the characteristics of heat transfer in a lead coolant cross-flow around tubes have been carried out for the first time and the dependences for a controlled and regulated content of the thermodynamically active oxygen impurity and sediments of impurities have been obtained. The effect of the oxygen impurity content in the coolant and characteristics of protective oxide coatings on the temperature and velocity fields in a lead coolant flow is revealed. This is because the presence of oxygen in the coolant and oxide coatings on the surface, which restrict the liquid-metal flow, leads to a change in the characteristics of the wall-adjacent region. The obtained experimental data on the distribution of the velocity and temperature fields in a HLMC flow permit studying the heat-transfer processes and, on this basis, creating program codes for engineering calculations of HLMC flows around heat-transfer surfaces. 展开更多
关键词 HEAVY liquid-Metal COOLANT lead lead-bismuth Fast Neutron reactors Heat-Exchange Wall Boundary Area
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气泡在液态铅铋金属中的运动特性及曳力系数模型研究
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作者 罗皓天 刘莉 +4 位作者 袁俊杰 包睿祺 田晓艳 李达 顾汉洋 《核技术》 EI CAS CSCD 北大核心 2024年第6期129-140,共12页
当铅铋快堆发生蒸汽发生器传热管破裂(Steam Generator Tube Rupture,SGTR)事故后,一回路高温液态铅铋合金(Lead-Bismuth Eutectic,LBE)与二回路高压过冷水相互作用产生大量蒸汽,这些气泡在LBE的携带作用下可能进入堆芯,引起局部传热恶... 当铅铋快堆发生蒸汽发生器传热管破裂(Steam Generator Tube Rupture,SGTR)事故后,一回路高温液态铅铋合金(Lead-Bismuth Eutectic,LBE)与二回路高压过冷水相互作用产生大量蒸汽,这些气泡在LBE的携带作用下可能进入堆芯,引起局部传热恶化和功率瞬变,严重影响反应堆的安全运行。掌握气泡在液态LBE中的运动特性及其动力学行为,开发适用于LBE中气泡迁移的曳力系数模型,是开展SGTR事故堆芯安全评估的基础。基于CLSVOF(Coupled Level-Set and Volume-Of-Fluid)方法建立了气泡在高温液态LBE中迁移运动的三维数值模型,通过分析气泡的运动轨迹、速度和粒径的变化规律,结合气泡受力平衡方程,获得了气泡曳力系数的模拟值。在此基础上,对比分析了现有曳力模型对LBE中气泡迁移的适用性,优选了最佳曳力系数模型并进行了进一步优化,优化后的模型对于液态LBE中气泡曳力系数的计算误差在15%之内。研究结果可为后续SGTR事故安全分析程序的开发提供理论支持。 展开更多
关键词 铅铋快堆 SGTR事故 气泡-液态金属两相流 气泡动力学 曳力系数
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小型可运输长寿命铅铋冷却快堆堆芯设计研究 被引量:9
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作者 雷驰 吴宏春 +2 位作者 曹良志 周生诚 邵一穷 《原子能科学技术》 EI CAS CSCD 北大核心 2019年第8期1451-1458,共8页
为满足偏远地区供电需求,提出了一种小型可运输长寿命铅铋冷却快堆(STLFR)堆芯设计方案,额定热功率为20 MW,在不换料条件下可运行18 EFPY(有效满功率年)。为减小堆芯体积,堆芯采用蜂窝煤型燃料组件,内设若干冷却剂管道,管外为燃料,实现... 为满足偏远地区供电需求,提出了一种小型可运输长寿命铅铋冷却快堆(STLFR)堆芯设计方案,额定热功率为20 MW,在不换料条件下可运行18 EFPY(有效满功率年)。为减小堆芯体积,堆芯采用蜂窝煤型燃料组件,内设若干冷却剂管道,管外为燃料,实现了较高的堆芯燃料体积占比。为展平堆芯径向功率分布,将堆芯燃料区沿径向划分为三区,分别采用不同的冷却剂管道尺寸。为降低堆芯高度,设计使用含高富集度6Li的液态锂作为吸收体的液态吸收体控制系统。为降低初始剩余反应性,在堆芯控制组件与安全组件中布置两组固定式可替换吸收体,分别在堆芯燃耗1/3和2/3寿期时替换为固定式反射体。提出的堆芯设计方案在整个运行寿期内满足热工设计限值,控制系统和安全系统能独立满足堆芯控制和停堆要求。采用准静态反应性平衡方法对5种典型无保护事故工况进行分析,初步证明了堆芯具有固有安全特性。 展开更多
关键词 小型可运输长寿命铅铋冷却快堆 蜂窝煤型燃料 液态吸收体控制系统 固定式可替换吸收体
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铅铋堆用T91钢耐腐蚀性能及脆化现象研究进展 被引量:1
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作者 李骥 何西扣 +3 位作者 许斌 魏学栋 吴冰洁 杨钢 《金属功能材料》 CAS 2021年第4期35-41,共7页
铅基反应堆结构的完整性和可靠性是反应堆服役期间安全运行的基础。结构材料的服役环境非常苟刻,会受到快中子辐照、高温液态金属的腐烛与冲刷及应力等的综合作用,因此对材料的性能要求非常高。T91钢具有良好的导热性能、低的膨胀系数... 铅基反应堆结构的完整性和可靠性是反应堆服役期间安全运行的基础。结构材料的服役环境非常苟刻,会受到快中子辐照、高温液态金属的腐烛与冲刷及应力等的综合作用,因此对材料的性能要求非常高。T91钢具有良好的导热性能、低的膨胀系数和良好的抗辐照性能,一直被认为是发展核电技术的首选结构材料。综合叙述了T91钢在铅铋合金液相容性研究现状,总结了耐腐蚀性能和腐蚀后脆化性能的研究进展,给出未来堆用T91钢的发展趋势。 展开更多
关键词 铅铋堆 T91 耐铅铋腐蚀 液态金属脆化
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ADS散裂靶与次临界堆的耦合传热研究
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作者 宋健 贾欢 +2 位作者 蔡汉杰 张勋超 何源 《原子核物理评论》 CAS CSCD 北大核心 2023年第4期651-659,共9页
对ADS液态铅铋靶与铅基次临界堆之间进行了热耦合,研究了在靶堆间是否加入阻热层、阻热层的热导率和厚度、加入气体阻热层时气体的压力对靶堆间热流量的影响。研究发现,加入阻热层可以减少耦合后靶堆间的热流量,气体阻热层有显著降低热... 对ADS液态铅铋靶与铅基次临界堆之间进行了热耦合,研究了在靶堆间是否加入阻热层、阻热层的热导率和厚度、加入气体阻热层时气体的压力对靶堆间热流量的影响。研究发现,加入阻热层可以减少耦合后靶堆间的热流量,气体阻热层有显著降低热流量的效果,允许靶和堆中铅铋的流速以及靶堆间的温差在更大范围内波动。靶堆间传递的热流量与阻热材料的热导率成正比,与阻热层厚度成反比,气体阻热层厚度可以选择在0.06到0.08 m之间。气体阻热层的压力在0.1到10 Pa区间内,热流量随压力的变化显著。因此,气体阻热层压力可以选择在0.1 Pa左右。 展开更多
关键词 加速器驱动次临界系统 液态铅铋散裂靶 液态铅铋反应堆 热耦合 气体阻热层
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