本文關(guān)鍵詞：小型自然循環(huán)鉛冷快堆SNCLFR-100一回路主冷卻系統(tǒng)熱工安全分析　出處：《中國科學技術(shù)大學》2017年博士論文　論文類型：學位論文

  更多相關(guān)文章： 鉛冷快堆 自然循環(huán) 一回路主冷卻系統(tǒng) 熱工安全分析 程序開發(fā)

【摘要】：鉛冷快堆具備良好的增殖核燃料和嬗變核廢料潛力,擁有突出的經(jīng)濟性和固有安全性,被第四代國際核能論壇(GIF)視為有望成為首個實現(xiàn)工業(yè)示范化的第四代核能系統(tǒng)。小型自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)采用全自然循環(huán)驅(qū)動,可進一步簡化鉛冷快堆的系統(tǒng)設(shè)計,避免液態(tài)金屬泵制造和運行給鉛冷快堆技術(shù)研發(fā)帶來的一系列挑戰(zhàn),具有良好的發(fā)展前景。掌握自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)的熱工安全特性是研發(fā)該新型反應(yīng)堆亟需探索和研究的核心技術(shù)之一。本文針對自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)熱工安全分析的需求,從熱工設(shè)計與分析程序開發(fā)、三維穩(wěn)態(tài)熱工水力特性研究和瞬態(tài)熱工安全性能分析等方面開展100Wth小型自然循環(huán)鉛冷快堆SNCLFR-100—回路主冷卻系統(tǒng)的熱工安全分析研究,論文的主要工作包括:(1)針對小型自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)的設(shè)計特點,建立了物性、傳熱和壓降模型,單通道模型,閉式并聯(lián)多通道模型和最熱通道模型等熱工水力分析模型;基于上述模型開發(fā)了具備熱工水力設(shè)計和穩(wěn)態(tài)熱工性能分析雙重功能的STAC程序;并開展了程序相關(guān)模塊和主體功能的初步驗證,驗證結(jié)果表明,STAC程序的計算結(jié)果具有良好的準確性和可信度;最后基于STAC程序?qū)NCLFR-100—回路主冷卻系統(tǒng)溫度分布、壽期初和壽期末的堆芯溫度分布及—回路主冷卻系統(tǒng)自然循環(huán)能力進行了分析研究。(2)基于SNCLFR-100—回路主冷卻系統(tǒng)設(shè)計方案,建立了其四分之一整體CFD分析模型和全堆芯CFD分析模型,開發(fā)了穩(wěn)態(tài)燃料棒熱傳導模型和穩(wěn)態(tài)主熱交換器換熱模型,并進行了湍流模型選取、多孔介質(zhì)模型使用和堆芯釋熱方式應(yīng)用的數(shù)值模擬方案研究;利用ANSYS FLUENT開展了額定工況下整體三維熱工水力特性分析和全堆芯自然循環(huán)流量分配特性研究,對—回路主冷卻系統(tǒng)內(nèi)冷卻劑的流動和換熱現(xiàn)象、堆芯入口流量分配特性進行了較深入的分析,并對一回路主冷卻系統(tǒng)的設(shè)計提出了相應(yīng)的優(yōu)化建議;最后基于理論計算和CFD模擬,對全堆芯流量分配方案進行了初步優(yōu)化設(shè)計,實現(xiàn)了堆芯各組件流量份額和功率份額的匹配。(3)利用熱工水力系統(tǒng)安全分析程序ATHLET對SNCLFR-100—回路主冷卻系統(tǒng)的瞬態(tài)熱工水力特性進行了分析,重點研究了無保護超功率事故(UTOP)、無保護失熱阱事故(ULOHS)和無保護超功率疊加失熱阱事故(UTOP+ULOHS)3類嚴重事故下反應(yīng)堆—回路主冷卻系統(tǒng)的安全響應(yīng)特性。研究結(jié)果表明:SNCLFR-100具有良好的固有安全性,各事故下堆芯安全的最大挑戰(zhàn)來自于包殼材料無法承受過高的溫度而失效。(4)針對現(xiàn)有的系統(tǒng)安全分析程序和CFD程序在分析小型自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)內(nèi)復雜三維流動現(xiàn)象時均存在局限性的問題,基于系統(tǒng)安全分析程序ATHLET和CFD程序ANSYS FLUENT提出了耦合三維特性的瞬態(tài)熱工安全分析方法,并基于該方法開展了 SNCLFR-100的全廠斷電事故(SBO)分析,重點研究了事故工況下上腔室的熱分層現(xiàn)象以及一維和三維熱工水力現(xiàn)象的耦合反饋。本文旨在研發(fā)自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)熱工水力設(shè)計、熱工安全分析工具,研究自然循環(huán)鉛冷快堆一回路主冷卻系統(tǒng)的穩(wěn)態(tài)和瞬態(tài)熱工安全特性,相關(guān)研究成果可進一步豐富鉛冷快堆熱工水力設(shè)計和熱工安全分析工具,掌握小型自然循環(huán)鉛冷快堆的熱工安全特性,具有一定的學術(shù)意義和工程應(yīng)用價值。
[Abstract]:With the proliferation of lead cooled fast reactor nuclear fuel and nuclear waste transmutation has good potential, the economy and the inherent safety of prominent, is the fourth generation of international nuclear Forum (GIF) as is expected to become the first to achieve industrial demonstration of fourth generation nuclear energy systems. The primary cooling system of small natural circulation lead cooled fast reactor by nature the driving cycle, the system design can be further simplified lead cooled fast reactor, to avoid a series of challenges of liquid metal pump manufacturing and operation to bring the lead cooled fast reactor technology research and development, with good prospects for development. Grasp the thermal safety characteristics of natural circulation lead cooled fast reactor primary cooling system is one of the core technology research and development of the new reactors need to explore and study. Based on the analysis of natural circulation thermal safety lead cooled fast reactor primary cooling system, the thermal design and analysis of program development, the three-dimensional steady state thermal Study on analysis of thermal safety 100Wth small natural circulation lead cooled fast reactor SNCLFR-100 circuit cooling system to carry out the research on hydraulic characteristics and transient thermal safety performance analysis, the main work of this paper includes: (1) for small natural circulation lead cooled fast reactor primary cooling system design features is built up, and the heat transfer the pressure drop model, single channel model, closed parallel multi channel model and the hot channel model of thermal hydraulic analysis model; based on the above model is developed with the thermal hydraulic design and performance analysis of steady state thermal dual function STAC program; and to carry out a preliminary verification program module and the main function, the verification results show that the calculation of the STAC program the result has a good accuracy and reliability; finally, based on the STAC program for temperature distribution of SNCLFR-100 main loop cooling system, life at the beginning of the period and the ending of the core The temperature distribution and the main loop cooling system of natural circulation ability are analyzed. The design scheme of SNCLFR-100 (2) - the main loop cooling system based on a quarter of the overall CFD whole core analysis model and CFD analysis model, developed a steady fuel rod heat conduction model and steady-state main heat exchanger model, and the choice of turbulence model, scheme of numerical simulation of porous medium model and the core heat release application; carried out under the rated condition the whole three-dimensional thermal hydraulic characteristics analysis and whole core research on distribution characteristics of natural circulation flow rate by ANSYS FLUENT, the main loop in the cooling system of the coolant flow and heat transfer phenomena, pile the core entrance flow distribution characteristics are analyzed deeply, and puts forward corresponding suggestions on Design of primary cooling system; finally, based on theoretical calculation and CFD simulation On the whole, the core flow distribution scheme for a preliminary optimization design, implementation, core components and power flow share share. (3) the analysis of transient thermal hydraulic characteristics of ATHLET program on SNCLFR-100 - circuit cooling system with thermal hydraulic system safety are analyzed, focusing on the protection of non super power accident (UTOP), unprotected loss of heat sink accident (ULOHS) and unprotected loss of heat sink of super power superposition accident (UTOP+ULOHS) security response characteristics of reactor - the main cooling system of 3 kinds of circuit under severe accident. The results show that SNCLFR-100 has good inherent safety, the biggest challenge from the core safety accident in the cladding materials cannot withstand high temperature and failure. (4) according to the procedures and CFD procedures in the analysis of small natural circulation lead cooled fast reactor primary cooling system in complex three-dimensional flow analysis system of existing security The phenomenon are limitations, system security analysis program ATHLET and CFD program ANSYS FLUENT proposed a safety analysis method of 3D transient thermal coupling based on the characteristics, and based on this method to carry out a blackout accident of SNCLFR-100 (SBO) analysis, feedback coupling focuses on the thermal stratification of the upper chamber and accident conditions one dimensional and three-dimensional thermal hydraulic phenomena. This paper aims to develop the natural circulation lead cooled fast reactor thermal hydraulic design of primary cooling system, thermal safety analysis tools, research on natural circulation lead cooled fast reactor primary cooling system in the steady state and transient thermal safety characteristics, the research results can further enrich the lead cooled fast reactor thermal hydraulic design and thermal safety thermal safety analysis tools, master the characteristics of natural circulation of small lead cooled fast reactor, has a certain academic significance and engineering application value.

【學位授予單位】：中國科學技術(shù)大學
【學位級別】：博士
【學位授予年份】：2017
【分類號】：TL433

【參考文獻】

相關(guān)期刊論文 前10條

1 駱鵬;王思成;胡正國;徐瑚珊;詹文龍;;加速器驅(qū)動次臨界系統(tǒng)——先進核燃料循環(huán)的選擇[J];物理;2016年09期

2 吳磊;賈海軍;劉洋;張濤;馬計中;楊星團;;自然循環(huán)系統(tǒng)流動-阻力特性理論和實驗研究[J];核動力工程;2016年03期

3 Doris Sung;;A New Look at Building Facades as Infrastructure[J];Engineering;2016年01期

4 中國科學院"未來先進核裂變能——ADS嬗變系統(tǒng)"戰(zhàn)略性先導科技專項研究團隊;;直面挑戰(zhàn) 追夢核裂變能可持續(xù)發(fā)展——“未來先進核裂變能——ADS嬗變系統(tǒng)”戰(zhàn)略性先導科技專項及進展[J];中國科學院院刊;2015年04期

5 任成;楊星團;劉志勇;姜勝耀;;一體化小型堆主回路自然循環(huán)穩(wěn)態(tài)特性實驗研究[J];原子能科學技術(shù);2014年S1期

6 吳宜燦;柏云清;宋勇;黃群英;劉超;王明煌;周濤;金鳴;吳慶生;汪建業(yè);蔣潔瓊;胡麗琴;李春京;高勝;李亞洲;龍鵬程;趙柱民;郁杰;FDS團隊;;中國鉛基研究反應(yīng)堆概念設(shè)計研究[J];核科學與工程;2014年02期

7 薛秀麗;付陟瑋;馮預(yù)恒;劉一哲;許義軍;楊紅義;;日本文殊原型快堆堆芯出口腔室熱分層現(xiàn)象數(shù)值模擬[J];原子能科學技術(shù);2013年10期

8 劉余;李峰;張虹;張渝;;三維物理-熱工耦合系統(tǒng)RECON的開發(fā)與驗證[J];原子能科學技術(shù);2012年10期

9 靖劍平;張春明;陳妍;孫微;莊少欣;;淺談核電領(lǐng)域中的熱工水力分析程序[J];核安全;2012年03期

10 喬雪冬;胡文軍;馮預(yù)恒;張春明;孫微;趙守智;;中國實驗快堆全廠斷電事故多維度熱工耦合計算[J];原子能科學技術(shù);2012年S1期

，

本文編號：1417741