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Simulation of steam generator plugging tubes in a PWR to analyze the operating impact

机译:在压水堆中模拟蒸汽发生器堵塞管以分析运行影响

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摘要

A number of nuclear power plants (NPPs) with pressurized water reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems. Several methods were attempted to correct the defects of the tubes, but eventually the only permanent solution was to plug them. The consequences of plugging the tubes are the decrease of heat transfer surface, the reduction of the flow area and subsequent reduction of the primary system mass flow and for a fraction of plugged tubes higher than a given value, the reduction of reactor output and economic losses. The objective of this paper is to analyze whether steam generator tube plugging has an impact in the effectiveness of accident management actions. An analysis with Relap5 Mod 3.3 patch03 for the Spanish reactor Asc6-2, a 3-loop 2940.6 MWth Westinghouse PWR, in which plugging of steam generator tubes are simulated, is presented in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. To complete the study two events, in which the steam generators are used to cooldown the plant, were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the emergency operating procedures (EOPs) to handle this kind of events. The selected events have been the rupture of one tube of a SG (SGTR) and a small break LOCA (SBLOCA). Two transient cases with no plugging at all (0%) and 12% SG tube plugging were performed. The actions of the corresponding EOPs for SG tube rupture and SBLOCA were coded in the input deck. The results show no significant impact on operator actions. (C) 2016 Elsevier B.V. All rights reserved.
机译:由于不同问题导致SG管的退化,世界上许多带有压水堆(PWR)的核电站(NPP)都已更换了它们的蒸汽发生器(SG)。尝试了几种方法来纠正管子的缺陷,但最终唯一的永久解决方案是将其堵塞。堵塞管子的后果是减少了传热面积,减小了流通面积并随后降低了主系统的质量流量,并且当堵塞的管子部分高于给定值时,反应器的输出量减少且经济损失。本文的目的是分析蒸汽发生器管道堵塞是否对事故管理措施的有效性产生影响。提出了使用Relap5 Mod 3.3 patch03对西班牙反应堆Asc6-2(3回路2940.6 MWth Westinghouse PWR)进行的分析,在其中模拟了蒸汽发生器管道的堵塞,以便找到工厂正常运行的极限。通过对一次到二次传热表面的减小以及在导管中一次冷却剂质量流动面积的减少进行建模,对不同比例的SG管堵塞进行了几种稳态计算。分析结果表明,堵塞12%的SG管在最佳反应器运行的极限附近。为了完成研究,模拟了两个事件,其中使用了蒸汽发生器来冷却工厂,以找出SGs管的堵塞是否会影响应急操作程序(EOP)中描述的操作员操作效率,以解决此问题。一种事件。选择的事件是SG的一根管破裂(SGTR)和LOCA的小断裂(SBLOCA)。进行了两个根本没有堵塞(0%)和SG管堵塞12%的瞬时情况。 SG管破裂和SBLOCA的相应EOP动作在输入平台中进行了编码。结果表明对操作员的行为没有重大影响。 (C)2016 Elsevier B.V.保留所有权利。

著录项

  • 来源
    《Nuclear Engineering and Design》 |2016年第8期|132-145|共14页
  • 作者单位

    European Commiss, Joint Res Ctr, Inst Energy & Transport, Nucl Reactor Safety Assessment Unit, Westerduinweg 1, NL-1755 Petten, Netherlands;

    Tech Univ Catalonia UPC, Barcelona, Spain;

    European Commiss, Joint Res Ctr, Policy Support Coordinat, Nucl Safety & Secur Coordinat Unit, Brussels, Belgium;

    Asociac Nucl Asco Vandellos II ANAV, Tarragona, Spain;

    European Commiss, Joint Res Ctr, Inst Energy & Transport, Nucl Reactor Safety Assessment Unit, Westerduinweg 1, NL-1755 Petten, Netherlands;

  • 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
  • 关键词

  • 入库时间 2022-08-18 00:41:54

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