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通过合理简化一体化模块式先进反应堆(SMART)螺旋管式蒸汽发生器,建立螺旋管单元管模型,采用两流体模型和非平衡过冷沸腾模型,在均匀热流密度下对螺旋管内流体进行不同参数下流动与传热数值模拟。结果表明:摩擦压降数值计算结果与陈学俊经验公式最为接近;曲率从0.04降至0.012时,摩擦压降明显下降,曲率继续下降,摩擦压降不变;加速压降几乎不受曲率影响;螺旋升角为3°~8.6°时,计算摩擦压降可不考虑螺旋升角的影响;雷诺数越大,总压降和摩擦压降均变大,摩擦压降梯度也明显增大。 相似文献
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以BW公司实际运行的直流蒸汽发生器为原型,采用两流体三流场数学模型数值模拟直流蒸汽发生器二次侧单相液对流换热区、核态沸腾区、蒸干后区域、过热区(即全范围)的流动沸腾换热现象。结果表明:本文采用的模型可较好地预测直流蒸汽发生器全范围的流动与传热特性,数值模拟结果与运行数据符合较好;蒸干后质量含汽率低于热平衡质量含汽率,结合壁温变化趋势表明直流蒸汽发生器运行过程中蒸干后区域存在一定程度的热力非平衡;随着流动换热的发展,表面传热系数迅速上升,蒸干的发生导致传热性能急剧下降,在蒸干后区域及过热蒸汽区表面传热系数缓慢上升。 相似文献
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建立与氦气对流换热的并联螺旋管蒸汽发生器数值模型,分别采用一维飘移流模型和一维可压缩流动模型描述水侧和氦气侧的流动。在此基础上研究了球床模块式高温气冷堆核电站螺旋管蒸汽发生器内的流量漂移不稳定性。动态计算结果表明,在一定条件下蒸汽发生器内有可能发生流量漂移,不同传热管流量可相差几倍,而出口温度则相差几百度。通过对质量流速-压降曲线的分析,发现热负荷对稳定性起主导作用,热负荷越大越易发生流量漂移,且边界质量流量随热负荷呈线性增长。增大入口节流阻力和过冷度可以在一定程度上避免流量飘移。最后给出了蒸汽发生器流量飘移的稳定边界。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(7):765-770
In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor was operated smoothly at the designated parameters. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. Design of the SG includes hydraulics, heat transfer and stability designs. Based on the design requirement, it is necessary to ensure sufficient heat removal from the reactor in order to maintain stable operation. In order to confirm the thermal hydraulic reliability of the SG, a series of experiments had been carried out. The purpose of this paper is to introduce the design features and experimental verification of HTR-10 SG, and the research results of small bending radius helical coil-pipe used in HTR-10, for example, the heat transfer coefficient of water, superheat steam and the two phase flow in the helical tube, the heat transfer coefficient of the He flow across the helical tube, and the centrifugal force effect on the heat transfer for the helical tube. In the paper, some operational experimental data of the HTR-10 SG have been presented. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(7):704-711
In this study, we carried out the numerical simulation of gas flow across a helically coiled tube bundle, which is fitted in the annular space between a central spine and an outer casing. We focused on the effect of clearances at the inner and outer radial boundaries on the gas flow distribution and thermal mixing in the tube bundle. The conservation equations of mass, momentum and energy were numerically solved by computational fluid dynamics. It was demonstrated from the numerical simulation results that (i) the gas flow distribution in the tube bundle is very sensitive to clearances at the inner and outer radial boundaries, and even a very small variation of clearances would lead to a considerable change in the average velocity profile, which correspondingly leads to a great deviation of the gas flow distribution away from the original one; (ii) the clearance effect on thermal mixing is not as strong as that on gas flow distribution;(iii) the Reynolds number has little effect on the gas flow distribution and thermal mixing, the latter of which is in agreement with the results of Achenbach experiments. 相似文献
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以大亚湾核电站蒸汽发生器为研究对象,建立了基于漂移流理论的蒸汽发生器一维动态数学模型及传热管泄漏模型,并进行了蒸汽发生器不同工况下的稳态仿真。在验证所建立漂移流模型和传热管泄漏模型的基础上,研究了不同工况下传热管泄漏位置及泄漏流量对蒸汽发生器关键参数的影响。研究结果表明,所建立的漂移流模型和传热管泄漏模型能准确反映不同泄漏情况下蒸汽发生器质量含汽率及蒸汽压力等关键参数的变化规律,泄漏发生在热端沸腾段入口处时各参数变化最显著,泄漏量为冷却剂流量的5%时出口质量含汽率由0.261降到0.163。基于漂移流理论传热管泄漏对蒸汽发生器动态特性影响的成功预测,为蒸汽发生器传热管泄漏事故的监测与防范措施的制定提供一定参考。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(4):524-528
In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor operated smoothlyqbthe design parameters were successfully attained. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. There are two thermal hydraulic requirements to be satisfied for the once-through steam generator: (1) enough heat transfer surface; (2) qualified steam can be produced under rated electrical generation power, and water-steam two phase flow un-stability can be avoided. In order to obtain the thermal hydraulic characteristics of the SG reliably, before design, a numerical code was developed for the design, and a full-scale test loop with two heating tubes as model was established, and series experiments had been carried out. The purpose of this paper is to introduce the design of SG and researches on the stability of small bending radius helical coil-pipe used in HTR-10, for exempla, the effects of outlet steam pressure, inlet water sub-cooling degree, thermal power and inlet throttling degree. Up to now, the SG has experienced full power operation smoothly, and approvingly reached its original design requirements. In the paper, some operational experimental data of the HTR-10 S.G have been presented. 相似文献
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清华大学核能与新能源技术研究院在建的250 MWt高温气冷堆核电站示范工程(HTR-PM)中蒸汽发生器二回路为亚临界水,由于反应堆能提供750℃的高温氦气,二回路水可提高到超临界压力和温度,采用多堆带一机方案可与超临界蒸汽透平机组匹配,因此研究超临界水在管内的流动、传热以及流动不稳定现象非常重要。本文通过使用RNGk-ε模型耦合强化壁面函数,发现模拟结果与Yamagata等的实验数据符合较好。基于此模型,分析了超临界流体流动时换热系数的变化规律,并采用瞬态计算方法,线性增大加热功率,分析了流动不稳定现象,发现流体一旦进入不稳定区,进出口流量的波动非常严重,甚至出现倒流,应尽可能避免此类现象。 相似文献
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为计算得到蒸汽发生器流量分配板及横向流的局部压力损失系数,应用三维稳态热工水力软件GENEPI,在无重力、单相条件下,对管束区入口到第2块管子支撑板下游进行三维流场模拟,计算得到了给定出口压力下入口静压,进而求出进出口压降,并通过沿程摩擦以及局部压力损失关系式等,减去两块管子支撑板的压力损失及沿程阻力,推导求出蒸汽发生器流量分配板及横向流的局部压力损失系数。为验证方法的正确性及可行性,以CPR1000-SG和EPR-SG为对象,计算这两个型号蒸汽发生器流量分配板及横向流的局部压力损失系数,并将计算结果与国外经验系数进行对比,结果表明:计算结果与国外经验系数接近,误差在可接受范围内。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(12):965-984
A heat transfer experiment was performed on steam-water two-phase flow in an annular flow path with a uniformly heated rod under the conditions of the mass flow rates from 0.2xlO6 to l.Ox 106 kg/m2-h, inlet qualities from 0.5 to 1.0, heat fluxes below 4.7x 105 W/m2 and pressure of 31 bar. Dryout of the heater rod surface was observed resulting in the sharp rise of the heater rod surface temperature. Measured heat transfer coefficients were compared with the several empirical and semi-empirical correlations with the emphasis on the applicability of the correlations to the present test conditions being important in the analysis of the thermal hydraulic behavior during a LOCA of a nuclear reactor. The measured heat transfer coefficient in the pre-dryout region is lower than the existing correlations. The cooling of the heat transfer surface by the liquid phase in the post-dryout region is significant, which is neglected in the existing correlations. The heat transfer coefficients calculated for the post-dryout region by the Groeneveld correlation show good agreement with the presently measured results within the accuracy of 0~27%. 相似文献
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HTGR蒸汽发生器螺旋管内壁温分布和传热恶化规律的研究 总被引:3,自引:0,他引:3
介绍了在西安交通大学高压汽水实验回路上进行的垂直螺旋管内汽水两相流传实验,得出了沿螺旋管长度方向及圆周方向的壁温分布特性,由此确定了发生传热恶化的临界含汽率及其位置,通过关联度实验数据,得到了计算临界含汽率的经验公式,为高温气冷堆(HTGR)螺旋管式蒸汽发生器的设计提供了必要的依据。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(2):286-294
One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. 相似文献
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在系统压力P=4~10MPa,质量流速G=450~1200kg/m~2·s,人口过冷度△T_(sub)=10~70℃,热负荷q=O~650kW/m~2的工况范围内,采用φ20×2mm的23Crl8Ni不锈钢U型管,在高压试验台上进行了密度波型脉动的研究;分析了系统压力、质量流速、入口过冷度、热负荷对它的影响。结果表明,随着压力的增加,系统稳定性增加;随着质量流速增加,界限热负荷增加,界限干度下降;进口过冷度对密度波型脉动呈现单值性影响,随着进口过冷度下降,界限热负荷降低。根据实验结果及量纲分析,得出了预测界限热负荷的无因次关联式,与试验值相比,误差在15%以内。 相似文献
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快堆燃料组件棒束通道内流动和传热现象分析与研究 总被引:3,自引:3,他引:0
利用三维计算流体力学软件CFX 12.0对由7根带螺旋状定位绕丝的燃料棒组成的快堆燃料组件典型棒束通道内的流动和传热现象进行了数值模拟。模拟得到不同Re下的压降系数曲线与Nu曲线,并将计算结果与经验公式的计算结果进行了比较,两者符合较好。研究了组件内3类典型子通道的横向流交混效应,分析了3类典型子通道的横向流分布特点,发现角子通道横向流交混强度沿轴向波动较大,而3类子通道的横向流交混强度均存在周期性。研究了中心燃料棒壁面上3个截面的局部换热效应,发现在燃料棒与绕丝接触处传热效果最差,在事故分析时应重点关注。 相似文献