本文關(guān)鍵詞： 壓水核反應(yīng)堆 燃料棒 數(shù)值計(jì)算 (火用)分析　出處：《南京理工大學(xué)》2017年碩士論文　論文類型：學(xué)位論文

【摘要】：隨著核電事業(yè)的高速發(fā)展,針對(duì)先進(jìn)壓水核反應(yīng)堆堆芯熱工水力優(yōu)化設(shè)計(jì)成為核電領(lǐng)域的研究熱點(diǎn)。在核反應(yīng)堆熱工設(shè)計(jì)中,安全性能是第一要素,必須保證任何情況下反應(yīng)堆的安全運(yùn)行。本文基于非結(jié)構(gòu)化網(wǎng)格中的有限容積法,采用全隱格式離散非穩(wěn)態(tài)導(dǎo)熱方程,通過可視化語(yǔ)言Fortran90/95進(jìn)行數(shù)值計(jì)算,分析堆芯燃料棒傳熱溫度和(火用)損失分布規(guī)律以及不同運(yùn)行工況下熱安全特性。首先,對(duì)單通道模型穩(wěn)態(tài)傳熱工況進(jìn)行計(jì)算分析。通過理論推導(dǎo)獲得核燃料棒包殼內(nèi)外壁面、燃料芯塊表面和中心溫度場(chǎng)分布的解析解,并將精確解的結(jié)果與定熱導(dǎo)率和熱導(dǎo)率隨溫度變化兩種情況下的數(shù)值解對(duì)比分析。兩者結(jié)果吻合良好,并且后者誤差更小。研究發(fā)現(xiàn)正常運(yùn)行工況下高溫區(qū)在燃料棒中心偏上,向兩端逐漸減小;在芯塊區(qū)和氣隙區(qū)有較大溫降,且氣隙區(qū)溫降速率最大。包殼外壁面溫度均勻分布,最大溫度值出現(xiàn)在熱通道中心和出口之間。其次,以秦山二期工程壓水堆為原型,討論常見的三種事故工況下燃料棒傳熱和溫度分布規(guī)律。當(dāng)有反應(yīng)性引入時(shí),溫度反饋效應(yīng)可使急劇上升的溫度逐漸趨于穩(wěn)定,負(fù)溫度系數(shù)更有利于調(diào)節(jié)堆芯功率,和確保反應(yīng)堆安全運(yùn)行;同時(shí)燃料棒上下壁面邊界條件的改變對(duì)于溫度場(chǎng)影響很小。當(dāng)燃料芯塊發(fā)生偏置時(shí),燃料棒整體溫度有所下降,高溫區(qū)向芯塊偏置反方向偏移;偏心率越大,偏移程度越大,并且包殼外壁面溫度分布不再均勻,最大溫度值也越高。當(dāng)發(fā)生失流事故(LOFA)時(shí),對(duì)流換熱系數(shù)沿?zé)嵬ǖ雷兓群苄?對(duì)流換熱系數(shù)越小,芯塊中心和包殼溫升越快。尤其在小于20 kW/(m2·℃C)時(shí)影響越大,此時(shí)包殼外壁面最高溫度出現(xiàn)的位置越接近燃料棒中心點(diǎn),但是熱流密度值變化很小。最后,結(jié)合熱力學(xué)第一定律和熱力學(xué)第二定律,基于燃料棒溫度場(chǎng)采用(火用)分析法計(jì)算分析反應(yīng)堆內(nèi)能量轉(zhuǎn)換,以及熱量傳遞過程中(火用)損失分布規(guī)律。研究結(jié)果發(fā)現(xiàn),核燃料溫度越高,傳熱溫差越小,(火用)損失越小。但二者又相互制約,在總(火用)不變情況下內(nèi)熱源分布的改變對(duì)總(火用)損影響很小,只是改變?nèi)剂习?火用)損失分布。該分析方法可為優(yōu)化反應(yīng)堆熱工設(shè)計(jì)和可靠性評(píng)估提供有益參考。
[Abstract]:With the rapid development of nuclear power industry, the advanced pressurized water reactor core thermal hydraulic optimization design has become a research hotspot in the field of nuclear power. In the thermal design of nuclear reactor, safety is the first element, must ensure safe operation of the reactor under any situation. The finite volume method based on unstructured grids, by a full implicit to discretize the unsteady heat conduction equation, through the visual language Fortran90/95 numerical calculation, analysis of reactor core fuel rod heat transfer temperature and heat loss (exergy) distribution and safety characteristics under different operating conditions. Firstly, calculation and analysis of the single channel model for steady heat transfer conditions. Access to nuclear fuel rod cladding wall through theoretical derivation. The analytical solution distribution temperature of the fuel pellet surface and center, and the exact solution and the thermal conductivity and thermal conductivity change with temperature under two conditions The comparative analysis of the numerical solutions. The results are in good agreement, and the latter error is smaller. The study found that under normal operating conditions in the high temperature zone of fuel rod centre, gradually reduced to the ends of the core block; and the air gap area has a larger temperature drop, and the air gap region temperature drop rate. Maximum cladding surface temperature distribution and the maximum temperature value between the thermal channel center and export. Secondly, in the two phase of the project of Qinshan PWR as the prototype, discusses three common accident conditions of fuel rod heat transfer and temperature distribution. When the reaction temperature is introduced, the feedback effect can make a sharp rise in temperature gradually stabilized, negative temperature the coefficient is more conducive to regulating core power, and ensure the safe operation of the reactor; at the same time under the wall boundary condition on the fuel rod changes for temperature effects is very small. When the fuel pellets bias, the overall temperature under the fuel rods Drop, high temperature area to the core block bias against the direction of migration; the larger the eccentricity ratio, the greater the degree of deviation, and shell surface temperature distribution is not uniform, the maximum temperature is higher. When the loss of flow accident (LOFA), the convective heat transfer coefficient along the hot channel variation is very small; convective heat transfer the coefficient is small, the core block center and cladding temperature rise fast. Especially in less than 20 kW/ (M2 C C) when the greater the impact, the shell wall of the highest temperature is closer to the center of the fuel rods, but the heat flux changes little. Finally, according to the first law of thermodynamics and thermodynamics the second law, the fuel rod temperature field based on the calculation and analysis of energy (exergy) conversion reactor analysis method, and the heat transfer process (exergy loss distribution). The results of the study showed that nuclear fuel temperature is high, the heat transfer temperature difference is small (exergy) losses is small. But the two and restrict each other, Under the condition of total exergy invariance, the change of heat source distribution has little effect on total exergy loss, but only changes the exergy loss distribution of fuel rods. This analysis method can provide a useful reference for optimizing reactor thermal design and reliability evaluation.

【學(xué)位授予單位】：南京理工大學(xué)
【學(xué)位級(jí)別】：碩士
【學(xué)位授予年份】：2017
【分類號(hào)】：TL329

【相似文獻(xiàn)】

相關(guān)期刊論文 前2條

1 ;化工上的應(yīng)用[J];生物技術(shù)通報(bào);1988年10期

2 ;[J];;年期

相關(guān)會(huì)議論文 前7條

1 譚昭怡;王子久;李燁;蘇容波;李兆岷;;水下等離子弧模擬切割300~#反應(yīng)堆堆芯部件研究[A];全國(guó)核與輻射設(shè)施退役學(xué)術(shù)研討會(huì)論文集[C];2007年

2 馬倉(cāng);S.DOUCE;;SAPHYR系統(tǒng)應(yīng)用于嶺澳核反應(yīng)堆堆芯計(jì)算介紹[A];中國(guó)核科學(xué)技術(shù)進(jìn)展報(bào)告——中國(guó)核學(xué)會(huì)2009年學(xué)術(shù)年會(huì)論文集（第一卷·第2冊(cè)）[C];2009年

3 范福平;陳明軍;;CANDU 6啟動(dòng)儀表的特點(diǎn)[A];全國(guó)第三屆核反應(yīng)堆用核儀器學(xué)術(shù)會(huì)議論文集[C];2003年

4 劉于瓏;徐明峰;朱立輝;;反應(yīng)堆水下異物吸取裝置研制及其應(yīng)用[A];中國(guó)核科學(xué)技術(shù)進(jìn)展報(bào)告（第二卷）——中國(guó)核學(xué)會(huì)2011年學(xué)術(shù)年會(huì)論文集第3冊(cè)（核能動(dòng)力分卷（下））[C];2011年

5 司勝義;;節(jié)塊內(nèi)嵌S_N方法的算法研究及程序開發(fā)[A];第十二屆反應(yīng)堆數(shù)值計(jì)算與粒子輸運(yùn)學(xué)術(shù)會(huì)議論文集[C];2008年

6 熊厚華;彭旦;蔡德富;李義國(guó);;微型反應(yīng)堆燃料低濃化可行性初步研究[A];第五屆北京核學(xué)會(huì)核技術(shù)應(yīng)用學(xué)術(shù)交流會(huì)論文集[C];2008年

7 張俊軍;盧磊;王國(guó)忠;曾勤;吳宜燦;FDS團(tuán)隊(duì);;TRIPOLI計(jì)算模型的三維交互式可視化實(shí)現(xiàn)及其應(yīng)用[A];第四屆全國(guó)反應(yīng)堆物理與核材料學(xué)術(shù)研討會(huì)論文集[C];2009年

相關(guān)碩士學(xué)位論文 前3條

1 張功偉;壓水核反應(yīng)堆堆芯燃料棒熱工數(shù)值計(jì)算與研究[D];南京理工大學(xué);2017年

2 范巖成;反應(yīng)堆堆芯核測(cè)量數(shù)據(jù)在線處理方法研究[D];哈爾濱工程大學(xué);2005年

3 何偉波;基于UH_x-H_2體系的反應(yīng)堆堆芯中子輸運(yùn)性能研究[D];中國(guó)工程物理研究院;2014年

，

本文編號(hào)：1505130