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Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure. 相似文献
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新型氮气稳压器系统稳态和瞬态特性研究 总被引:1,自引:0,他引:1
根据氮气稳压器系统的基本理论模型,分析了氮气稳压器的稳态和瞬态运行特性,得到了两种不同波动流量工况下,稳压器压力、水位、水区焓、水区质量、氮气温度及氮气体积随时间的变化特性.结果表明:当波动流量为正波动时,稳压器的压力、水区质量、水区焓、水位、氮气温度均呈上升趋势,氮气的体积降低;而当波动流量为负时,各参数变化规律相反.研究表明,氮气稳压器的响应特性较好.两种工况下主要参数的变化趋势与理论分析相一致,但对该模型的实验验证以及控制研究仍需在将来的工作中进行. 相似文献
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稳压器安全阀用于核电站一回路系统和设备的超压保护,如果发生故障卡开,将造成冷却剂丧失事故(LOCA)。本文使用机理性分析程序对三门核电厂1号机组进行建模,并对稳压器安全阀误开启导致的LOCA事故进行模拟分析,研究在稳压器水位较高的情况下,非能动安全设施对LOCA事故的响应情况。之后,为验证三门核电站对类似三哩岛事故的应对能力,假设丧失给水叠加稳压器安全阀卡开事故并进行相应事故分析。通过以上两个事故的分析表明,三门核电厂的非能动安全设计能够应对稳压器安全阀故障造成的LOCA事故,保证对一回路补水,不会造成非常严重的事故后果。 相似文献
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利用MELCOR程序对小型船用堆稳压器喷雾除气过程及停堆过程进行建模,进而模拟核动力装置从功率运行至降功率除气,以及除气结束后停堆消除稳压器气腔的全部物理过程。通过对反应堆关键运行参数变化趋势的仿真分析,验证了模拟的物理过程的合理性。结合建立的除气及停堆仿真模型,计算分析了包壳破损状态下,稳压器喷雾除气、停堆过程对稳压器内惰性气体含量的影响,评估了稳压器高点放气和喷雾除气对放射性物质的去除作用。研究结果能为小型堆包壳破损状态下放射性安全管理策略提供指导和帮助。 相似文献
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通过分析相间的传热传质过程以及非凝性气体存在时壁面蒸汽冷凝过程,建立了汽 气稳压器模型,研究了非凝性气体对稳压过程的影响,描述了稳压器的稳压特性,并将模型计算结果与MIT稳压器实验数据进行了对比。结果表明:当不含非凝性气体时,计算精度高,相对偏差在0.8%内,压力峰值为0.647 MPa;当非凝性气体含量从0增至20%时,计算精度相对减小,最高相对偏差为15.4%;压力峰值从0.647 MPa增至1.02 MPa。研究表明非凝性气体对稳压器稳压过程具有重要影响作用,随着非凝性气体的种类和含量的变化,稳压器内稳压过程发生显著变化。 相似文献
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