【摘要】：本文以AP1000核電廠為對(duì)象,首先利用系統(tǒng)分析程序RELAP5進(jìn)行建模,參考西屋公司的SGTR事故進(jìn)程,設(shè)定了安全系統(tǒng)以及輔助系統(tǒng)的促發(fā)邏輯和順序,對(duì)AP1000 SGTR事故進(jìn)行模擬計(jì)算,將穩(wěn)態(tài)與瞬態(tài)(單根傳熱管破裂)計(jì)算結(jié)果和西屋公司SGTR事故分析結(jié)果進(jìn)行比對(duì)分析;然后在單根傳熱管破裂事故模型的基礎(chǔ)上,進(jìn)一步對(duì)SGTR事故進(jìn)行深入分析,研究了多根傳熱管破裂事故,為驗(yàn)證電廠非能動(dòng)安全特性,考察了廠外電是否有效和完好側(cè)大氣釋放閥是否開啟故障對(duì)事故后果的影響;最后針對(duì)破口模型和傳熱管節(jié)點(diǎn)劃分?jǐn)?shù)量進(jìn)行了敏感性分析。結(jié)果表明,本文基于RELAP5建立的模型和西屋公司LOFTTR2的計(jì)算結(jié)果具有較好的一致性,穩(wěn)態(tài)熱工參數(shù)比如壓力、溫度、流量等都吻合良好,在單根傳熱管破裂事故工況下,瞬態(tài)參數(shù)的變化趨勢(shì)大致相同,由于物理模型的不同,在數(shù)值上存在一定的差異,AP1000依靠非能動(dòng)余熱排出系統(tǒng)能夠避免破損SG發(fā)生滿溢,且有一定的裕量,多根傳熱管破裂工況下,堆內(nèi)有可能會(huì)出現(xiàn)兩相從而導(dǎo)致流動(dòng)不穩(wěn)定的現(xiàn)象出現(xiàn),需要加以關(guān)注,改變假設(shè)條件進(jìn)行特性分析及敏感性研究對(duì)事故后果都有不同程度的影響,采用不同的破口模型會(huì)使事故進(jìn)程及冷卻劑喪失質(zhì)量發(fā)生改變,節(jié)點(diǎn)劃分?jǐn)?shù)量的不同影響破口臨界噴放流量的大小,但系統(tǒng)響應(yīng)大致相同,破損側(cè)SG均沒有發(fā)生滿溢,更加驗(yàn)證了三代核電技術(shù)的安全性,其研究結(jié)果可進(jìn)一步支持AP1000相關(guān)的審評(píng)工作。
[Abstract]:In this paper, AP1000 nuclear power plant is taken as the object. Firstly, the system analysis program RELAP5 is used to model the model. Referring to the SGTR accident process of Westinghouse Company, the trigger logic and sequence of safety system and auxiliary system are set up, and the simulation calculation of AP1000 SGTR accident is carried out. The calculation results of steady state and transient state (single heat transfer tube rupture) and SGTR accident analysis results of Westinghouse Company are compared and analyzed. Then on the basis of the single heat transfer tube rupture accident model, the SGTR accident is further analyzed, and the multiple heat transfer tube rupture accident is studied, in order to verify the non-active safety characteristics of the power plant. The effect on the accident consequence of whether the power plant is effective or not and whether the air release valve of the intact side is open or not is investigated. Finally, sensitivity analysis is carried out for the break model and the number of heat transfer pipe nodes. The results show that the model based on RELAP5 is in good agreement with the calculation results of LOFTTR2 of Westinghouse, and the steady-state thermal parameters such as pressure, temperature and flow rate are in good agreement. The variation trend of transient parameters is roughly the same. Because of the difference of physical model, there are some differences in numerical value. AP1000 can avoid the overflowing of damaged SG and have a certain margin by relying on inactive residual heat discharge system. Under the condition of rupture of multiple heat transfer pipes, it is possible that two phases will appear in the reactor, which will lead to flow instability, which needs to be paid attention to. Changing the hypothetical conditions for characteristic analysis and sensitivity study have different degrees of influence on the consequence of the accident. Different fracture models will change the process of the accident and the quality of coolant loss, and the number of nodes will affect the critical discharge rate of the break, but the response of the system is roughly the same, and the SG of the damaged side is not overflowing. The safety of the third generation nuclear power technology is verified, and the research results can further support the AP1000 related review.
【學(xué)位授予單位】：哈爾濱工程大學(xué)
【學(xué)位級(jí)別】：碩士
【學(xué)位授予年份】：2014
【分類號(hào)】：TM623.4

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