本文選題：AP1000 + RELAP5-HD��； 參考：《哈爾濱工程大學(xué)》2014年碩士論文

【摘要】：AP1000是西屋公司在繼承傳統(tǒng)壓水堆成熟技術(shù),并吸取其長期積累的運行經(jīng)驗的基礎(chǔ)上開發(fā)出來的三代+壓水堆,它是一個革新性的設(shè)計,符合美國核管會安全評審要求,并滿足先進輕水堆用戶要求文件。AP1000是一個單堆布置的兩環(huán)路核電廠,其凈電輸出功率為1117MWe。與傳統(tǒng)壓水堆核電廠相比,其最主要的特點就是使用了非能動安全系統(tǒng),利用非能動特性,如壓縮氣體儲能,重力勢能,自然循環(huán)等代替能動設(shè)備如泵,交流電源等進行驅(qū)動,從而使得電廠的安全性和可靠性得到大幅提升。AP1000的非能動堆芯冷卻系統(tǒng)包括非能動余熱排出系統(tǒng)和非能動安全注射系統(tǒng),以及用于有效銜接高、中、低壓安注的自動卸壓系統(tǒng)。AP1000核電站的整體系統(tǒng)結(jié)構(gòu),運行模式和特點以及其非能動設(shè)計理念與我國目前大量運營的反應(yīng)堆相比有較大不同。為了熟悉先進壓水堆的系統(tǒng)結(jié)構(gòu),全面掌握其運行特點,充分理解其非能動設(shè)計理念,并且對堆芯下降腔等具有典型多維流動的部件進行模擬,有必要使用具有多維組分模擬功能的RELAP-HD程序,對AP1000進行建模仿真研究。RELAP5-3D是RELAP5系列程序的最新版本,與之前的RELAP5版本相比,RELAP5-3D最重要的改進在于多維水力學(xué)部件和多維中子動力學(xué)模型的引入。GSE公司將RELAP5-3D嵌入其實時仿真平臺SimExec上,形成了 RELAP5-HD,它可以在不損害RELAP5-3D最佳估算程序的完整性的前提下實時地進行電廠運行狀態(tài)的熱工水力求解。本文首先利用RELPA5-HD程序建立了 AP1000核電廠的模型,主要包括其壓力容器、蒸汽發(fā)生器、主管道、穩(wěn)壓器、非能動堆芯冷卻系統(tǒng),以及控制系統(tǒng)等。壓力容器內(nèi)的下降通道和堆芯用多維組分進行模擬。對該模型進行了穩(wěn)態(tài)調(diào)試,并將最終的穩(wěn)態(tài)結(jié)果與AP1000電廠額定值進行比較,以驗證穩(wěn)態(tài)模型的適用性。同時,還在穩(wěn)態(tài)情況下對壓力容器下降通道和堆芯內(nèi)的多維流動進行了分析。利用AP1000核電廠對冷段10-in小破口失水事故的響應(yīng),對非能動堆芯冷卻系統(tǒng)模型及控制系統(tǒng)進行了驗證。最后,使用經(jīng)過驗證的模型,對壓水堆核電廠內(nèi)的典型事故,如主給水喪失事故,主蒸汽管道破裂事故等進行了模擬,分析了 AP1000非能動堆芯冷卻系統(tǒng)對非LOCA事故的響應(yīng),并對主蒸汽管道破裂事故中壓力容器內(nèi)的多維流動和不對稱現(xiàn)象進行了分析。仿真結(jié)果表明,事故中非能動堆芯冷卻系統(tǒng)都能自動投入,有效導(dǎo)出堆芯余熱,確保反應(yīng)堆安全。在主蒸汽管道破裂事故中,由于環(huán)路以及非能動系統(tǒng)響應(yīng)的不對稱性,壓力容器的下降通道和堆芯出口處也會有明顯的不對稱現(xiàn)象。
[Abstract]:AP1000 is a third-generation PWR developed by Westinghouse on the basis of inheriting the mature technology of traditional PWR and absorbing its long accumulated operating experience. It is an innovative design that meets the requirements of the safety assessment of the American Nuclear Regulatory Commission. AP1000 is a two-loop nuclear power plant with a single reactor arrangement and its net output power is 1117MWe. Compared with the traditional PWR nuclear power plant, its main feature is the use of passive safety system, the use of inactive characteristics, such as compressed gas energy storage, gravity potential energy, natural circulation instead of active equipment such as pumps, AC power, etc. This greatly improves the safety and reliability of power plants. AP1000's inactive core cooling system includes inactive residual heat removal systems and inactive safety injection systems, and is used to effectively connect high, medium, The whole system structure, operation mode and characteristics of the automatic pressure relief system .AP1000, and its inactive design concept are quite different from those of a large number of reactors in our country at present. In order to be familiar with the system structure of the advanced PWR, fully grasp its operating characteristics, fully understand its inactive design concept, and simulate the typical multi-dimensional flow components such as the falling chamber of the reactor core, It is necessary to use the RELAP-HD program, which has the function of multi-component simulation, to model and simulate AP1000. RELAP5-3D is the latest version of the RELAP5 series. The most important improvement over previous RELAP5 versions is the introduction of multidimensional hydraulics components and multidimensional neutron dynamics models. GSE has embedded RELAP5-3D on its real-time simulation platform, SimExec. The RELAP5-HD is formed, which can be used to solve the thermohydraulic problems of power plant operation in real time without compromising the integrity of the RELAP5-3D optimal estimation program. In this paper, the model of AP1000 nuclear power plant is established by using RELPA5-HD program, including its pressure vessel, steam generator, main pipeline, voltage regulator, inactive core cooling system and control system. The drop channel and core in the pressure vessel are simulated with multi-dimensional components. The steady-state model is debugged and compared with AP1000 power plant rating to verify the applicability of the steady-state model. At the same time, the multi-dimensional flow in the drop channel and core of the pressure vessel is analyzed under steady state. The model and control system of inactive core cooling system are verified by the response of AP1000 nuclear power plant to the small break water loss accident of 10-in cold section. Finally, the typical accidents in PWR nuclear power plant, such as main water supply loss accident, main steam pipeline rupture accident and so on, are simulated by using the verified model, and the response of AP1000 inactive core cooling system to non-LOCA accident is analyzed. The multi-dimensional flow and asymmetry in the pressure vessel in the main steam pipeline rupture accident are analyzed. The simulation results show that the non-active core cooling system of the accident can be automatically put into operation, and the residual heat of the reactor core can be effectively derived to ensure the safety of the reactor. In the main steam pipeline failure due to the asymmetry of the loop and the response of the inactive system there will also be obvious asymmetry in the descending channel and the core outlet of the pressure vessel.
【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;TL421.1

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本文編號：2030957