發(fā)布時(shí)間：2018-06-17 11:57

  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(zhǎng)期積累的運(yùn)行經(jīng)驗(yàn)的基礎(chǔ)上開(kāi)發(fā)出來(lái)的三代+壓水堆,它是一個(gè)革新性的設(shè)計(jì),符合美國(guó)核管會(huì)安全評(píng)審要求,并滿(mǎn)足先進(jìn)輕水堆用戶(hù)要求文件。AP1000是一個(gè)單堆布置的兩環(huán)路核電廠(chǎng),其凈電輸出功率為1117MWe。與傳統(tǒng)壓水堆核電廠(chǎng)相比,其最主要的特點(diǎn)就是使用了非能動(dòng)安全系統(tǒng),利用非能動(dòng)特性,如壓縮氣體儲(chǔ)能,重力勢(shì)能,自然循環(huán)等代替能動(dòng)設(shè)備如泵,交流電源等進(jìn)行驅(qū)動(dòng),從而使得電廠(chǎng)的安全性和可靠性得到大幅提升。AP1000的非能動(dòng)堆芯冷卻系統(tǒng)包括非能動(dòng)余熱排出系統(tǒng)和非能動(dòng)安全注射系統(tǒng),以及用于有效銜接高、中、低壓安注的自動(dòng)卸壓系統(tǒng)。AP1000核電站的整體系統(tǒng)結(jié)構(gòu),運(yùn)行模式和特點(diǎn)以及其非能動(dòng)設(shè)計(jì)理念與我國(guó)目前大量運(yùn)營(yíng)的反應(yīng)堆相比有較大不同。為了熟悉先進(jìn)壓水堆的系統(tǒng)結(jié)構(gòu),全面掌握其運(yùn)行特點(diǎn),充分理解其非能動(dòng)設(shè)計(jì)理念,并且對(duì)堆芯下降腔等具有典型多維流動(dòng)的部件進(jìn)行模擬,有必要使用具有多維組分模擬功能的RELAP-HD程序,對(duì)AP1000進(jìn)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(jìn)在于多維水力學(xué)部件和多維中子動(dòng)力學(xué)模型的引入。GSE公司將RELAP5-3D嵌入其實(shí)時(shí)仿真平臺(tái)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(shí)時(shí)地進(jìn)行電廠(chǎng)運(yùn)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(chǎng)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(dòng)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(jìn)行模擬。對(duì)該模型進(jìn)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(chǎng)額定值進(jìn)行比較,以驗(yàn)證穩(wěn)態(tài)模型的適用性。同時(shí),還在穩(wěn)態(tài)情況下對(duì)壓力容器下降通道和堆芯內(nèi)的多維流動(dòng)進(jìn)行了分析。利用AP1000核電廠(chǎng)對(duì)冷段10-in小破口失水事故的響應(yīng),對(duì)非能動(dòng)堆芯冷卻系統(tǒng)模型及控制系統(tǒng)進(jìn)行了驗(yàn)證。最后,使用經(jīng)過(guò)驗(yàn)證的模型,對(duì)壓水堆核電廠(chǎng)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(jìn)行了模擬,分析了 AP1000非能動(dòng)堆芯冷卻系統(tǒng)對(duì)非LOCA事故的響應(yīng),并對(duì)主蒸汽管道破裂事故中壓力容器內(nèi)的多維流動(dòng)和不對(duì)稱(chēng)現(xiàn)象進(jìn)行了分析。仿真結(jié)果表明,事故中非能動(dòng)堆芯冷卻系統(tǒng)都能自動(dòng)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(dòng)系統(tǒng)響應(yīng)的不對(duì)稱(chēng)性,壓力容器的下降通道和堆芯出口處也會(huì)有明顯的不對(duì)稱(chēng)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(jí)別】：碩士
【學(xué)位授予年份】：2014
【分類(lèi)號(hào)】：TM623;TL421.1

【參考文獻(xiàn)】

相關(guān)期刊論文 前10條

1 張亮;肖立新;龔宏偉;薛松;;核電站堆內(nèi)構(gòu)件堆芯圍筒制造技術(shù)分析與比較[J];裝備機(jī)械;2013年03期

2 張之華;葉茂;羅昕;錢(qián)達(dá)志;;日本福島核事故的思考與警示[J];原子能科學(xué)技術(shù);2012年S2期

3 莫小錦;莊亞平;佟立麗;曹學(xué)武;;AP1000主給水管道斷裂事故中PRHR系統(tǒng)冷卻能力分析[J];原子能科學(xué)技術(shù);2012年S1期

4 莫小錦;佟立麗;曹學(xué)武;;AP1000喪失正常給水事故PRHR冷卻能力研究[J];科技導(dǎo)報(bào);2012年21期

5 朱耀明;;福島核事故對(duì)發(fā)展核電的警示[J];海峽科學(xué);2012年06期

6 王偉偉;蘇光輝;田文喜;秋穗正;;AP1000非能動(dòng)余熱排出系統(tǒng)建模與瞬態(tài)數(shù)值分析[J];原子能科學(xué)技術(shù);2011年12期

7 王志;;AP1000非能動(dòng)堆芯冷卻系統(tǒng)及系統(tǒng)設(shè)計(jì)瞬態(tài)研究[J];中國(guó)核電;2011年03期

8 胡珊;;淺談我國(guó)核電發(fā)展的現(xiàn)狀與未來(lái)[J];科技資訊;2011年24期

9 嚴(yán)春;王建軍;閻昌琪;;非能動(dòng)余熱排出系統(tǒng)的穩(wěn)態(tài)特性研究[J];原子能科學(xué)技術(shù);2010年09期

10 薛若軍;鄧程程;彭敏俊;;非能動(dòng)余熱排出熱交換器數(shù)值模擬[J];原子能科學(xué)技術(shù);2010年04期

相關(guān)碩士學(xué)位論文 前2條

1 林支康;AP1000核電廠(chǎng)小破口失水事故RELAP5分析模式建立與應(yīng)用[D];上海交通大學(xué);2012年

2 李明巖;核電站非能動(dòng)余熱排出過(guò)程仿真研究[D];哈爾濱工程大學(xué);2009年

，

本文編號(hào)：2030957