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Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

机译:加压和减压条件下被动冷却过程中模块化高温反应堆的热响应

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摘要

The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel). This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes. In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000℃) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000℃) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600℃ for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures, etc
机译:关于通过传导,辐射和自然对流进行被动衰减除热的模块化HTR设计固有安全特性的概念最早是在德国HTR-Module(卵石燃料)设计中引入的,随后在最近几年扩展到其他模块化HTR设计,例如PBMR(卵石燃料),GT-MHR(棱柱形燃料)和新一代反应堆V / HTR(棱柱形燃料)。本文介绍了使用热工代码THERMIX进行的V / HTR的数值模拟,该代码最初是为用卵石燃料分析HTR而开发的,并通过实验进行了验证,随后被用于棱柱形燃料的HTR中,并对照结果进行了验证。各国使用不同代码分析的CRP-3基准问题的解决方案。本文研究了在加压和减压的一次系统条件下,停机后被动冷却过程中V / HTR(入口/出口温度为490/1000℃)的热响应。还进行了其他研究以确定在降压冷却条件下其他进/出口工作温度(例如490 / 850、350 / 850或350/1000℃)对最高燃料和压力容器温度的影响。另外,还进行了一些灵敏度分析,以评估改变参数,即衰减热,石墨电导率,表面发射率等对最大燃料和压力容器温度的影响。结果表明,在所有这些情况下,燃料的名义峰值温度都保持在1600℃以下,这是与燃料释放放射性有关的极限温度。本文中提出的分析表明,代码THERMIX可以成功地用于棱柱形燃料HTR的热计算。该结果还为V / HTR的最大热功率,工作温度等的设计优化提供了一些基础信息。

著录项

  • 来源
    《Nuclear Engineering and Design》 |2006年第6期|p.475-484|共10页
  • 作者单位

    Framatome ANP GmbH, Freyesleben Strasse 1, 91058 Erlangen, Germany;

  • 收录信息 美国《科学引文索引》(SCI);美国《工程索引》(EI);美国《生物学医学文摘》(MEDLINE);
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类 原子能技术;
  • 关键词

  • 入库时间 2022-08-18 00:46:56

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