AP1000核電廠冷卻劑流量喪失事故分析
發(fā)布時間：2018-11-25 09:36
【摘要】：AP1000是我國引進(jìn)的由美國西屋公司設(shè)計研發(fā)的第三代先進(jìn)壓水堆核電系統(tǒng),與傳統(tǒng)壓水堆相比,AP1000最大的特點(diǎn)是采用了大量的非能動安全系統(tǒng),這使安全系統(tǒng)的配置得到大幅簡化。失流事故是指當(dāng)一回路系統(tǒng)中的主泵由于機(jī)械故障卡死,或者喪失動力電源而停止運(yùn)轉(zhuǎn)的情況下,反應(yīng)堆冷卻劑流量急劇降低甚至中斷的事故。本文以AP1000核電廠為原型,利用系統(tǒng)分析程序RELAP5對其進(jìn)行建模,針對三種失流事故工況,即部分失流、全失流和冷卻劑泵卡軸工況進(jìn)行計算和分析。對部分失流和全失流事故的短期過程進(jìn)行計算,重點(diǎn)關(guān)注事故發(fā)生后短時間內(nèi)一回路各參數(shù)的變化特點(diǎn)。分析結(jié)果表明:AP1000在這兩類失流事故瞬態(tài)過程中,堆芯最小DNBR高于安全分析限值,滿足偏離泡核沸騰(DNB)設(shè)計基準(zhǔn)。通過改變主泵的轉(zhuǎn)動慣量并進(jìn)行計算后,發(fā)現(xiàn)較大的轉(zhuǎn)動慣量可以使一回路系統(tǒng)維持較高的冷卻劑慣性流量,利于堆芯冷卻,避免發(fā)生DNB。對廠外電有效和無效情況下的反應(yīng)堆冷卻劑泵卡軸事故短期過程進(jìn)行計算,分析結(jié)果表明:在事故瞬態(tài)過程中燃料包殼峰值溫度沒有超過驗收準(zhǔn)則中的規(guī)定限值,保證了燃料包殼的完整性。最后對喪失廠外電源情況下反應(yīng)堆冷卻劑泵卡軸事故的長期過程進(jìn)行計算,分析結(jié)果表明:在該事故工況的長期冷卻階段,AP1000的非能動安全系統(tǒng)能夠有效導(dǎo)出堆芯余熱,維持堆芯冷卻,保證反應(yīng)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.
【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
[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.
【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
【參考文獻(xiàn)】
相關(guān)期刊論文 前10條
1 左學(xué)兵;陳晶晶;張金東;代帥;鄭東宏;;AP1000反應(yīng)堆冷卻劑系統(tǒng)主要設(shè)備安裝技術(shù)[J];壓力容器;2013年11期
2 楊萍;賈紅軼;王U，
本文編號：2355637
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