AP1000核电厂冷却剂流量丧失事故分析
发布时间：2018-11-25 09:36
【摘要】：AP1000是我国引进的由美国西屋公司设计研发的第三代先进压水堆核电系统,与传统压水堆相比,AP1000最大的特点是采用了大量的非能动安全系统,这使安全系统的配置得到大幅简化。失流事故是指当一回路系统中的主泵由于机械故障卡死,或者丧失动力电源而停止运转的情况下,反应堆冷却剂流量急剧降低甚至中断的事故。本文以AP1000核电厂为原型,利用系统分析程序RELAP5对其进行建模,针对三种失流事故工况,即部分失流、全失流和冷却剂泵卡轴工况进行计算和分析。对部分失流和全失流事故的短期过程进行计算,重点关注事故发生后短时间内一回路各参数的变化特点。分析结果表明:AP1000在这两类失流事故瞬态过程中,堆芯最小DNBR高于安全分析限值,满足偏离泡核沸腾(DNB)设计基准。通过改变主泵的转动惯量并进行计算后,发现较大的转动惯量可以使一回路系统维持较高的冷却剂惯性流量,利于堆芯冷却,避免发生DNB。对厂外电有效和无效情况下的反应堆冷却剂泵卡轴事故短期过程进行计算,分析结果表明:在事故瞬态过程中燃料包壳峰值温度没有超过验收准则中的规定限值,保证了燃料包壳的完整性。最后对丧失厂外电源情况下反应堆冷却剂泵卡轴事故的长期过程进行计算,分析结果表明:在该事故工况的长期冷却阶段,AP1000的非能动安全系统能够有效导出堆芯余热,维持堆芯冷却,保证反应堆安全。
[Abstract]:AP1000 is the third generation advanced PWR nuclear power system designed and developed by Westinghouse Company in our country. Compared with traditional PWR, AP1000 is characterized by a large number of passive safety systems. This greatly simplifies the configuration of security systems. Loss of flow accident is an accident in which the coolant flow rate of reactor decreases sharply or even breaks down when the main pump in the primary circuit system is blocked by mechanical failure or the power supply is lost. In this paper, the AP1000 nuclear power plant is used as the prototype, and the system analysis program RELAP5 is used to model it. The calculation and analysis are carried out for three kinds of out-of-flow accident conditions, that is, partial loss of flow, total loss of flow and coolant pump shaft. In this paper, the short-term process of partial and total loss of flow is calculated, and the characteristics of the parameters of the primary circuit in a short time after the accident are emphasized. The results show that the minimum core DNBR of AP1000 is higher than the limit of safety analysis in the transient process of these two kinds of loss of flow accident, which meets the design standard of deviated bubble boiling (DNB). By changing the moment of inertia of the main pump and calculating, it is found that the larger moment of inertia can make the primary circuit system maintain a higher coolant inertial flow rate, which is conducive to core cooling and avoid the occurrence of DNB.. The short term process of reactor coolant pump shaft accident is calculated under the condition of effective and invalid power outside the plant. The results show that the peak temperature of fuel cladding does not exceed the prescribed limit of acceptance criterion in the transient process of the accident. The integrity of the fuel cladding is ensured. Finally, the long-term process of reactor coolant pump shaft accident under the condition of power loss outside the plant is calculated. The results show that in the long cooling stage of the accident condition, the passive safety system of AP1000 can effectively derive the residual heat of reactor core. Maintain core cooling to ensure reactor safety.
【学位授予单位】：哈尔滨工程大学
【学位级别】：硕士
【学位授予年份】：2014
【分类号】：TM623
[Abstract]:AP1000 is the third generation advanced PWR nuclear power system designed and developed by Westinghouse Company in our country. Compared with traditional PWR, AP1000 is characterized by a large number of passive safety systems. This greatly simplifies the configuration of security systems. Loss of flow accident is an accident in which the coolant flow rate of reactor decreases sharply or even breaks down when the main pump in the primary circuit system is blocked by mechanical failure or the power supply is lost. In this paper, the AP1000 nuclear power plant is used as the prototype, and the system analysis program RELAP5 is used to model it. The calculation and analysis are carried out for three kinds of out-of-flow accident conditions, that is, partial loss of flow, total loss of flow and coolant pump shaft. In this paper, the short-term process of partial and total loss of flow is calculated, and the characteristics of the parameters of the primary circuit in a short time after the accident are emphasized. The results show that the minimum core DNBR of AP1000 is higher than the limit of safety analysis in the transient process of these two kinds of loss of flow accident, which meets the design standard of deviated bubble boiling (DNB). By changing the moment of inertia of the main pump and calculating, it is found that the larger moment of inertia can make the primary circuit system maintain a higher coolant inertial flow rate, which is conducive to core cooling and avoid the occurrence of DNB.. The short term process of reactor coolant pump shaft accident is calculated under the condition of effective and invalid power outside the plant. The results show that the peak temperature of fuel cladding does not exceed the prescribed limit of acceptance criterion in the transient process of the accident. The integrity of the fuel cladding is ensured. Finally, the long-term process of reactor coolant pump shaft accident under the condition of power loss outside the plant is calculated. The results show that in the long cooling stage of the accident condition, the passive safety system of AP1000 can effectively derive the residual heat of reactor core. Maintain core cooling to ensure reactor safety.
【学位授予单位】：哈尔滨工程大学
【学位级别】：硕士
【学位授予年份】：2014
【分类号】：TM623
【参考文献】
相关期刊论文 前10条
1 左学兵;陈晶晶;张金东;代帅;郑东宏;;AP1000反应堆冷却剂系统主要设备安装技术[J];压力容器;2013年11期
2 杨萍;贾红轶;王U，
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