【摘要】：本文基于最佳估算熱工水力程序RELAP5/MOD3.3,針對AP1000核電廠進行系統(tǒng)建模,主要對核電廠一回路系統(tǒng)堆芯壓力容器、冷卻劑管道、穩(wěn)壓器、反應(yīng)堆冷卻劑泵;二回路系統(tǒng)的蒸汽發(fā)生器、蒸汽出口管道;非能動堆芯冷卻系統(tǒng)中非能動安全注入系統(tǒng)、非能動余熱排出系統(tǒng)、自動降壓系統(tǒng)進行建模。在建立的AP1000系統(tǒng)模型基礎(chǔ)上進行程序穩(wěn)態(tài)調(diào)試工作,即核電廠各項參數(shù)最終達到穩(wěn)定狀態(tài),并且符合AP1000核電廠初始條件參數(shù)的參考范圍。在AP1000系統(tǒng)程序穩(wěn)態(tài)調(diào)試完成之后,引入ADS誤動作事故與非能動余熱排出系統(tǒng)誤動作事故模型,對這兩種事故工況進行瞬態(tài)計算。ADS誤動作事故瞬態(tài)計算結(jié)果中重要參數(shù)符合驗收準(zhǔn)則的要求,驗證了 AP1000核電廠在此事故工況下可以導(dǎo)出堆芯衰變熱,不會導(dǎo)致嚴重事故。非能動余熱排出系統(tǒng)誤動作事故瞬態(tài)計算結(jié)果中重要參數(shù)符合驗收準(zhǔn)則的要求,結(jié)果證明在反應(yīng)堆冷卻劑泵不惰轉(zhuǎn)與反應(yīng)堆不停堆的情況下,AP1000核電廠在此工況下不會導(dǎo)致嚴重事故。之后通過改變安全殼內(nèi)置換料水箱內(nèi)水溫度及PRHR熱交換器換熱面積兩種方式對PRHR熱交換器誤動作事故進行敏感性分析,分析結(jié)果表明,改變內(nèi)置換料水箱內(nèi)水溫度范圍不大情況下,對PRHR熱交換器內(nèi)自然循環(huán)影響不大,改變熱交換器換熱面積對自然循環(huán)影響較大,敏感性分析計算結(jié)果中重要參數(shù)均符合驗收準(zhǔn)則要求,不會導(dǎo)致嚴重事故。
[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.
【學(xué)位授予單位】：哈爾濱工程大學(xué)
【學(xué)位級別】：碩士
【學(xué)位授予年份】：2014
【分類號】：TM623

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