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AN EXPERIMENTAL STUDY OF THERMAL NON EQUILIBRIUM CONVECTIVE BOILING IN POST-CRITICAL-HEAT-FLUX REGION IN ROD BUNDLES.

机译:棒束临界后热通量区域热非平衡对流沸腾的实验研究。

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摘要

Prediction of heat transfer beyond critical-heat-flux (CHF) is an important aspect for boiling systems, particularly for nuclear reactors. Significant differences occur when these correlations are used in advanced computer programs to estimate the thermal-hydraulic conditions in a rod bundle assembly. In order to develop more accurate correlations, there is a need for a full size test set up and measurements of nonequilibrium vapor temperatures, wall temperatures and wall heat fluxes.; In this study, an experimental investigation of convective heat transfer beyond CHF has been performed with a full size test set up. Post-CHF experiments with simultaneous measurements of nonequilibrium vapor temperature at two axial locations has been successfully performed for a 3 x 3 rod bundle array, under stabilized post-CHF conditions. In addition, some transient (slowly moving CHF) experiments have also been conducted. Total of 374 data points, 98 steady-state with fixed CHF point and 276 transient with moving CHF point, have been obtained. The range of the data included conditions pertinent to reflood and quench phase of light water reactors.; Very high vapor and wall superheats were measured. Typical vapor superheat encountered in this research was 85 percent of corresponding wall superheat, and was in the range of 400 to 600(DEGREES)C. The experimental results indicated the existence of two different regions downstream from the CHF point, namely near region and far region. Modeling of heat transfer process in convective film boiling must account for these two regions of behavior. Significant variations in radial vapor temperature, up to 150(DEGREES)C, was found. It showed that use of one dimensional models for prediction of thermal-hydraulic conditions in a rod bundle is not proper. Available models, developed from single tube data base, showed no reasonable prediction of heat transfer for the new rod bundle data.
机译:预测超过临界热通量(CHF)的传热是沸腾系统(尤其是核反应堆)的重要方面。当在高级计算机程序中使用这些相关性来估计杆束组件中的热工条件时,会出现显着差异。为了建立更精确的相关性,需要建立一个全尺寸的测试装置并测量不平衡蒸气温度,壁温和壁热通量。在这项研究中,已经通过全尺寸测试装置对超过CHF的对流换热进行了实验研究。在稳定的CHF后条件下,对于3 x 3棒束阵列,已经成功进行了CHF后实验,同时测量了两个轴向位置的非平衡蒸气温度。此外,还进行了一些瞬态(CHF缓慢移动)实验。总共获得了374个数据点,具有固定CHF点的98个稳态和具有移动CHF点的276个瞬态。数据范围包括与轻水反应堆的再灌注和骤冷阶段有关的条件。测量到非常高的蒸气和壁过热。在这项研究中遇到的典型的蒸气过热是相应的壁过热的85%,并且在400至600(摄氏度)C的范围内。实验结果表明,CHF点下游存在两个不同的区域,即近端区域和远端区域。对流膜沸腾过程中的传热过程建模必须考虑这两个行为区域。发现径向蒸气温度的显着变化高达150(DEGREES)C。结果表明,使用一维模型预测棒束中的热工条件是不合适的。从单管数据库开发的可用模型没有显示出对新棒束数据传热的合理预测。

著录项

  • 作者

    UNAL, CETIN.;

  • 作者单位

    Lehigh University.;

  • 授予单位 Lehigh University.;
  • 学科 Engineering Mechanical.
  • 学位 Ph.D.
  • 年度 1986
  • 页码 359 p.
  • 总页数 359
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类 机械、仪表工业;
  • 关键词

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