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快堆单个燃料组件完全堵流事故的建模及其验证 被引量：2

Modeling of Accident Scenario Caused by Total Blockage of Single Subassembly in a Fast Reactor and its Validation
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摘要 为了预测正常功率下快堆单个燃料组件入口完全堵流所导致的事故序列,根据SCARABEE-N系列实验建立了相关的计算模型。冷却剂的沸腾及其两相流动的描述采用两流体模型;包壳的流动、燃料的熔化及其塌陷采用类似SURFASS程序的简单方法处理。对于事故后期形成的UO2-钢混合沸腾池,采用一维半经验模型描述,即:用漂移速度模型来预测空泡份额分布;用修正后的Greene关系式计算沸腾池和壁面之间的传热系数;用焓方法(enthalpy method)求解包裹沸腾池的固化壳的温度场及厚度。为了验证本文建立的模型,对SCARABEE BE+1实验结果进行了校核计算,其结果与实验结果基本吻合。 In order to predict the accident scenario caused by total inlet blockage of single subassembly at full power in a fast reactor, a calculational model was developed according to SCARABEE-N subassembly melting and propagation tests. In the model, sodium thermal-hydraulics was described by a two-fluid subchannel model. The motion of melting cladding and melting fuel was dealt with very simple approach similar to that of SURFASS code. The behavior of UO2-steel mixed boiling pool was determined by a one dimensional semi-empirical model, in which the distribution of void fraction was calculated by drift flux model, heat transfer coefficient between the boiling pool and the wall was determined by modified Greene correlation, the temperature field and thickness of solidified UO2 crust were obtained by solving enthalpy model. Calculation for SCARABEE BE＋1 experiment was performed to validate the model, the results approximately agree with the experimental observation.
作者 石晓波 罗锐 王洲
机构地区 清华大学热能工程系
出处 《核动力工程》 EI CAS CSCD 北大核心 2006年第1期37-42,共6页 Nuclear Power Engineering
关键词 快堆 完全堵流 体热源沸腾池 Fast reactor, Total blockage, Volume heated boiling pool
分类号 TL433 [核科学技术—核技术及应用]
  • 相关文献

参考文献9

  • 1Kayser Gaston, Charpenel Jean, Jamond Claude. Summary of the SCARABEE-N Subassembly Melting and Propagation Tests with an Application to a Hypothetical Total Instantaneous Blockage in a Reactor[J]. Nuclear Science and Engineering, 1998, 128(2): 144 - 185.
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同被引文献29

  • 1石晓波,罗锐,王洲,杨献勇.钠冷快堆单个燃料组件冷却剂沸腾的数值模拟[J].核科学与工程,2004,24(4):312-317. 被引量：4
  • 2石晓波,罗锐,王洲.体热源沸腾池的建模及其验证[J].核科学与工程,2005,25(1):55-60. 被引量：1
  • 3HEUSENER G, COWKING C B, PENET F, et al. Fast reactor safety research and development in France DEBENE (Germany, Belgium and the Netherlands), Italy and the UK [C]// Proceedings of 1986 International Fast Reactor Safety Meeting: Vol 1. London: British Nuclear Energy Society, 1986: 1-8.
  • 4LEIGH K M. Fast reactor safety: Progress through evolution [C]// Proceedings of 1990 International Fast Reactor Safety Meeting: Vol 1. Snowbird, Utah, USA: American Nuclear Society, 1990: 75-84.
  • 5KAYSER G, CHARPENEL J. Summary of the SCARABEE-N subassembly melting and propagation tests with an application to a hypothetical total instantaneous blockage in a reactor [J]. Nuclear Science and Engineering, 1998, 128(2): 144-185.
  • 6LIVOLANT M, DADILLON J, KAYSER G, et al. SCARABEE: A test reactor and programme to study fuel melting and propagation in connexion with local faults, objectives and results [C]// Proceedings of 1990 International Fast Reactor Safety Meeting: Vol 2. Snowbird, Utah, USA: American Nuclear Society, 1990: 177-186.
  • 7SCHLEISIEK K, ABERLE J, HOMANN C, et al. The Mol-7c in-pile local blockage experiments: Main results, conclusions, and extrapolation to reactor conditions[J]. Nuclear Science and Engineering, 1998, 128: 93-143.
  • 8DICKERMAN C E, ROTHMAN A B, KLICKNAN A E, et al. Summary of TREAT in-pile experiments on LMFBR response to hypothetical core disruptive accidents[C]// Symposium on the Thermal and Hydraulic Aspects of Nuclear Reactor Safety: Vol 2, Liquid Metal Fast Breeder Reactors. Atlanta, Georgia, USA: The Heat Transfer Division, ASME, 1977: 19-50.
  • 9柏云清,汪卫华,蒋洁琼,等.加速器驱动核废料嬗变堆CLEAR概念设计[C]∥第五届反应堆物理与核材料学术研讨会、第二届核能软件自主化研讨会会议摘要集.[出版地不详]:[出版者不详],2011.
  • 10WU Yican, BAI Yunqing,SONG Yong,et al.Development strategy and conceptual design ofChina Lead-based Research Reactor for ADStransmutation system [J]. Applied Energy (inPress).

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核动力工程
2006年 第1期

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