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Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

机译:马鞍山压水堆厂延长停电期间NSSS的热工水力分析和安全壳响应

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摘要

A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS pressures will remain high in conjunction with the higher RCP seal leakage flow. The accumulators will not be available due to high RCS pressure and the reactor core will eventually become uncovered since there is no any water make-up. In the aspect of the containment response, the high-energy RCP seal leakage fluid continues flowing into the containment and heats up the containment. The containment pressure and temperature will increase to high values, respectively. There exists no clear relationship between the available TDAFWS time and the maximum containment pressure and temperature. The response of the containment temperature is much worse than that of the containment pressure. The most severe containment temperature response occurs for the case with no secondary depressurization and the calculated maximum containment temperature is 336.8 degrees F, which exceeds the design temperature of 300 degrees F but is still below the inside-containment safety-related equipment environmental qualification temperature of 450 degrees F. (C) 2015 Elsevier B.V. All rights reserved.
机译:在马鞍山压水堆厂的24 h延长停电(SBO)期间,对核蒸汽供应系统(NSSS)的响应和安全壳进行了热工水力分析。马鞍山电厂是西屋公司的三回路PWR设计,额定核心热功率为2822 MWt。 NSSS和安全壳中的分析分别基于RELAP5-3D和GOTHIC模型。在各个模型中考虑了工厂针对SBO的重要设计特征,例如,蒸汽发生器PORV,涡轮驱动辅助给水系统(TDAFWS),蓄能器,反应堆冷却剂泵(RCP)密封设计,安全壳内的各种热结构在分析中,假定每个RCP中的轴密封由于密封件冷却损失而失效,并且RCS流体直接流到安全壳。从RELPA5-3D模型计算出的一些参数输入到安全壳GOTHIC模型中,包括RCS平均温度以及RCP密封泄漏流量和焓。 RCS的平均温度用于驱动显热传递到安全壳。发现事件的严重性主要取决于次级侧是否减压。如果次级侧及时降压(在SBO后1小时内)并且TDAFWS可用时间超过19小时,则在整个SBO持续时间内,反应堆堆芯将被水覆盖,从而确保了反应堆堆芯的完整性。相反,如果次级侧未减压,则RCS压力将与较高的RCP密封泄漏流量一起保持较高状态。由于RCS压力高,将无法使用蓄能器,并且由于没有任何补水,反应堆堆芯最终将被发现。在安全壳响应方面,高能RCP密封泄漏流体继续流入安全壳并加热安全壳。密闭压力和温度将分别增加到很高的值。可用的TDAFWS时间与最大安全壳压力和温度之间没有明确的关系。安全壳温度的响应远比安全壳压力的响应差。最严重的安全壳温度响应发生在没有二次降压的情况下,计算出的最高安全壳温度为336.8华氏度,超过了设计温度300华氏度,但仍低于安全壳内安全相关设备的环境认证温度450度(C)2015 Elsevier BV保留所有权利。

著录项

  • 来源
    《Nuclear Engineering and Design》 |2015年第7期|11-18|共8页
  • 作者单位

    Atom Energy Council, Inst Nucl Energy Res, Longtan Township 32546, Taoyuan County, Taiwan;

    Atom Energy Council, Inst Nucl Energy Res, Longtan Township 32546, Taoyuan County, Taiwan;

    Atom Energy Council, Inst Nucl Energy Res, Longtan Township 32546, Taoyuan County, Taiwan;

  • 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
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