AP1000非能动余热排出系统误动作及ADS误动作事故分析
发布时间：2018-07-15 13:02
【摘要】：本文基于最佳估算热工水力程序RELAP5/MOD3.3,针对AP1000核电厂进行系统建模,主要对核电厂一回路系统堆芯压力容器、冷却剂管道、稳压器、反应堆冷却剂泵;二回路系统的蒸汽发生器、蒸汽出口管道;非能动堆芯冷却系统中非能动安全注入系统、非能动余热排出系统、自动降压系统进行建模。在建立的AP1000系统模型基础上进行程序稳态调试工作,即核电厂各项参数最终达到稳定状态,并且符合AP1000核电厂初始条件参数的参考范围。在AP1000系统程序稳态调试完成之后,引入ADS误动作事故与非能动余热排出系统误动作事故模型,对这两种事故工况进行瞬态计算。ADS误动作事故瞬态计算结果中重要参数符合验收准则的要求,验证了 AP1000核电厂在此事故工况下可以导出堆芯衰变热,不会导致严重事故。非能动余热排出系统误动作事故瞬态计算结果中重要参数符合验收准则的要求,结果证明在反应堆冷却剂泵不惰转与反应堆不停堆的情况下,AP1000核电厂在此工况下不会导致严重事故。之后通过改变安全壳内置换料水箱内水温度及PRHR热交换器换热面积两种方式对PRHR热交换器误动作事故进行敏感性分析,分析结果表明,改变内置换料水箱内水温度范围不大情况下,对PRHR热交换器内自然循环影响不大,改变热交换器换热面积对自然循环影响较大,敏感性分析计算结果中重要参数均符合验收准则要求,不会导致严重事故。
[Abstract]:Based on the optimal estimation of thermohydraulic program RELAP 5 / MOD 3.3, this paper models the AP1000 nuclear power plant system, mainly to the core pressure vessel, coolant pipeline, regulator, reactor coolant pump, steam generator of the secondary loop system of the nuclear power plant primary circuit system, the reactor core pressure vessel, the coolant pipeline, the stabilizer, the reactor coolant pump, and the secondary loop system steam generator. Steam outlet pipeline, passive core cooling system, non-active safety injection system, inactive residual heat removal system, automatic depressurization system are modeled. Based on the established AP1000 system model, the program steady-state debugging is carried out, that is, the parameters of the nuclear power plant finally reach the stable state, and accord with the reference range of the initial condition parameters of the AP1000 nuclear power plant. After the steady-state debugging of AP1000 system program is completed, the models of ads maloperation accident and inactive residual heat ejection system maloperation accident are introduced. The results of transient calculation. Ads misoperation accident transient calculation results meet the requirements of acceptance criteria. It is verified that AP1000 nuclear power plant can derive core decay heat under this accident condition and will not cause serious accidents. The important parameters in the transient calculation results of the maloperation accident of inactive residual heat removal system meet the requirements of acceptance criteria. The results show that the AP1000 nuclear power plant will not cause serious accidents under the condition that the reactor coolant pump does not turn inert and the reactor does not stop reactor. After that, the sensitivity of the misoperation accident of PRHR heat exchanger is analyzed by changing the water temperature and the heat exchange area of the PRHR heat exchanger. The results show that, When the range of water temperature in the built-in refueling tank is small, the natural circulation in the PRHR heat exchanger is not affected much, but the change of the heat exchange area of the heat exchanger has a great influence on the natural circulation. The important parameters of sensitivity analysis and calculation all meet the requirements of acceptance criteria and will not lead to serious accidents.
【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
本文编号：2124161
[Abstract]:Based on the optimal estimation of thermohydraulic program RELAP 5 / MOD 3.3, this paper models the AP1000 nuclear power plant system, mainly to the core pressure vessel, coolant pipeline, regulator, reactor coolant pump, steam generator of the secondary loop system of the nuclear power plant primary circuit system, the reactor core pressure vessel, the coolant pipeline, the stabilizer, the reactor coolant pump, and the secondary loop system steam generator. Steam outlet pipeline, passive core cooling system, non-active safety injection system, inactive residual heat removal system, automatic depressurization system are modeled. Based on the established AP1000 system model, the program steady-state debugging is carried out, that is, the parameters of the nuclear power plant finally reach the stable state, and accord with the reference range of the initial condition parameters of the AP1000 nuclear power plant. After the steady-state debugging of AP1000 system program is completed, the models of ads maloperation accident and inactive residual heat ejection system maloperation accident are introduced. The results of transient calculation. Ads misoperation accident transient calculation results meet the requirements of acceptance criteria. It is verified that AP1000 nuclear power plant can derive core decay heat under this accident condition and will not cause serious accidents. The important parameters in the transient calculation results of the maloperation accident of inactive residual heat removal system meet the requirements of acceptance criteria. The results show that the AP1000 nuclear power plant will not cause serious accidents under the condition that the reactor coolant pump does not turn inert and the reactor does not stop reactor. After that, the sensitivity of the misoperation accident of PRHR heat exchanger is analyzed by changing the water temperature and the heat exchange area of the PRHR heat exchanger. The results show that, When the range of water temperature in the built-in refueling tank is small, the natural circulation in the PRHR heat exchanger is not affected much, but the change of the heat exchange area of the heat exchanger has a great influence on the natural circulation. The important parameters of sensitivity analysis and calculation all meet the requirements of acceptance criteria and will not lead to serious accidents.
【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
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