【實(shí)習(xí)課題 1】

 

實(shí)習(xí)題目：

對堆芯材料碳化硼與不銹鋼共晶行為機(jī)理的研究

Study on the mechanism of eutectic behavior of reactor core materials (boron carbide-stainless steel)

 

實(shí)習(xí)內(nèi)容：

反應(yīng)堆嚴(yán)重事故下碳化硼-不銹鋼共晶反應(yīng)會使控制棒過早失效，對反應(yīng)堆嚴(yán)重事故評估影響顯著，因此本實(shí)驗(yàn)旨在研究反應(yīng)堆中吸收棒的碳化硼與結(jié)構(gòu)材料不銹鋼間的共晶反應(yīng)機(jī)理，利用可視化高溫反應(yīng)爐探究反應(yīng)溫度、接觸面壓強(qiáng)、接觸面幾何形狀等不同因素對于碳化硼-不銹鋼共晶反應(yīng)速率的影響。在得出實(shí)驗(yàn)數(shù)據(jù)后，利用移動粒子半隱式方法（MPS）對不同實(shí)驗(yàn)參數(shù)下的碳化硼-不銹鋼共晶反應(yīng)過程進(jìn)行數(shù)值模擬，驗(yàn)證MPS方法對于碳化硼-不銹鋼共晶反應(yīng)模擬的適用性并嘗試建立碳化硼-不銹鋼共晶反應(yīng)模型。

The eutectic reaction between boron carbide and stainless steel will cause premature failure of control rods under severe reactor accidents, which has a significant impact on the assessment of severe reactor accidents. Therefore, this experiment aims to study the eutectic reaction mechanism between absorber rod material (boron carbide) and structural material (stainless steel) in the reactor. A visualized high temperature reactor is used to investigate the effects of different factors such as reaction temperature, contact surface pressure, and contact geometry on the rate of eutectic reaction between boron carbide and stainless steel. After obtaining the experimental data, the numerical simulation of the eutectic reaction process between boron carbide and stainless steel under different experimental parameters is carried out by using the moving particle semi-implicit method (MPS) to verify the applicability of the MPS method to the simulation of boron carbide-stainless steel and to try to establish a model of the eutectic reaction between boron carbide and stainless steel.

 

企業(yè)導(dǎo)師：成松柏  教授

校內(nèi)導(dǎo)師：賀秀杰  副教授  hexiujie@mail.sysu.edu.cn

 

實(shí)習(xí)補(bǔ)貼：3000RMB/月

 

 

【實(shí)習(xí)課題 2】

 

實(shí)習(xí)題目：

液態(tài)金屬池中放射性裂變產(chǎn)物氣溶膠洗滌研究

Study on Aerosol Scrubbing of Radioactive Fission Products in Liquid Metal Pools

 

實(shí)習(xí)內(nèi)容：

在液態(tài)金屬堆中，若出現(xiàn)燃料棒失效等嚴(yán)重事故，可能會導(dǎo)致固體裂變產(chǎn)物夾帶在氣體中一起被釋放到冷卻劑池中。載有放射性氣溶膠的氣泡在冷卻劑池中不斷上升至池表面覆蓋氣體區(qū)域，這一過程會發(fā)生復(fù)雜的熱工水力學(xué)現(xiàn)象，同時一部分氣溶膠會被去除。該研究旨在利用python等編程軟件，實(shí)現(xiàn)對液態(tài)金屬堆池洗放射性氣溶膠效果的數(shù)值模擬計(jì)算與理論分析，并通過與真實(shí)實(shí)驗(yàn)數(shù)據(jù)對比驗(yàn)證，實(shí)現(xiàn)對最貼近實(shí)際情況的氣溶膠池洗模型的建立。

In liquid metal reactors, serious accidents such as fuel rod failure may result in solid fission products being entrained in the gas and released into the coolant pool. Bubbles containing radioactive aerosols continuously rise in the coolant pool to the gas covered area on the pool surface, which leads to complex thermodynamic and hydraulic phenomena, and some aerosols are removed. The purpose of this study is to use programming software such as Python to achieve numerical simulation and theoretical analysis of the effectiveness of liquid metal reactor pool scrubbing radioactive aerosols. By comparing and verifying with real experimental data, the establishment of the aerosol pool scrubbing model that is closest to the actual situation is achieved.

 

企業(yè)導(dǎo)師：成松柏  教授

校內(nèi)導(dǎo)師：劉曉星  副教授  liuxx85@mail.sysu.edu.cn

 

實(shí)習(xí)補(bǔ)貼： 2000元/月

 

 

【實(shí)習(xí)課題 3】

 

實(shí)習(xí)題目：

基于CFD-DEM方法對鈉冷快堆嚴(yán)重事故下的碎片床堆積過程的數(shù)值模擬研究

Numerical simulation of debris bed accumulation process at severe accident of sodium-cooled fast reactor based on the CFD-DEM method

 

實(shí)習(xí)內(nèi)容：

鈉冷快堆堆芯解體事故發(fā)生后，堆芯內(nèi)的堆芯熔融物將會釋放出來，與冷卻劑發(fā)生相互作用，碎化、沉降和堆積在壓力容器下腔室或其上方的堆芯捕集器上，形成碎片床。如未及時排出碎片床的剩余衰變熱，碎片床發(fā)生重熔將進(jìn)一步熔穿下腔室，造成嚴(yán)重的后果。因此，有必要全面地研究碎片床的堆積機(jī)理以提升其冷卻能力的評估。本研究基于CFD-DEM流固耦合數(shù)值方法，對鈉冷快堆嚴(yán)重事故下的碎片床堆積過程進(jìn)行了數(shù)值模擬，驗(yàn)證CFD-DEM方法對鈉冷快堆嚴(yán)重事故下碎片床堆積行為模擬的適用性及預(yù)測模型的可靠性，如可能將對經(jīng)驗(yàn)?zāi)Ｐ瓦M(jìn)行修正或開發(fā)新的經(jīng)驗(yàn)?zāi)Ｐ汀?/p>

After the core disruptive accident of sodium-cooled fast reactor (SFR), the corium inside the core will be released, interact with the coolant, fragment, settle and accumulate in the lower plenum of the pressure vessel or on the core catcher, forming a debris bed. If the remaining decay heat in the debris bed is not discharged in time, remelting of the debris bed will further melt through the lower plenum, resulting in severe consequences. Therefore, it is necessary to comprehensively investigate the mechanism of the debris bed accumulation to improve the evaluation of its cooling capacity. Based on the fluid-structure coupling numerical method of CFD-DEM, this study simulates the debris bed accumulation process under severe accident of sodium-cooled fast reactor, and verifies the applicability of CFD-DEM method to simulate the debris bed accumulation behavior under severe accident of sodium-cooled fast reactor and the reliability of the prediction model. If possible, the empirical model will be modified or a new empirical model will be developed.

 

企業(yè)導(dǎo)師：成松柏  教授

校內(nèi)導(dǎo)師：楊衛(wèi)岐  副教授  yangweiqi@mail.sysu.edu.cn

 

實(shí)習(xí)補(bǔ)貼：3000元/月

 

 

【實(shí)習(xí)題目 4】

 

實(shí)習(xí)題目：

液態(tài)金屬燃料反應(yīng)堆流動傳熱特性的數(shù)值研究

Numerical study of flow and heat transfer characteristics of liquid metal reactor

 

實(shí)習(xí)內(nèi)容：

第四代核反應(yīng)堆如鉛冷快堆為了安全性考慮采用液態(tài)金屬流體進(jìn)行熱量的傳遞，但也給模擬計(jì)算帶來一定挑戰(zhàn)。本實(shí)習(xí)調(diào)研已有的模擬方法和開展的實(shí)驗(yàn)工作，擬采用直接數(shù)據(jù)模擬或者大渦模擬方法建立一些工況的可靠的計(jì)算數(shù)據(jù)庫，然后對傳熱湍流傳熱過程采用非線性湍流模式進(jìn)行相應(yīng)修改，建立更可靠的工程湍流模擬方法。

Fourth-generation nuclear reactors, such as lead-cooled fast reactors, use liquid metal fluids for heat transfer for safety considerations, but they also bring certain challenges to the calculation work. This internship investigates the current simulation methods and experimental work, and  use direct numerical simulation or large eddy simulation methods to establish a reliable database under certain working conditions. The data is used to modify the nonlinear algebraic turbulence mode in the heat transfer process accordingly to establish a more reliable engineering turbulence simulation method.

 

企業(yè)導(dǎo)師：程輝  副教授

校內(nèi)導(dǎo)師：李萬愛  副教授  liwai@mail.sysu.edu.cn

 

實(shí)習(xí)補(bǔ)貼：2000元 /月