超臨界水堆核熱耦合及系統(tǒng)安全特性研究
發(fā)布時間：2019-05-12 13:56
【摘要】：超臨界水堆具有堆芯進(jìn)出口溫差大、冷卻劑流量低和燃料棒間隙窄等特征,由此帶來強(qiáng)烈的物理-熱工反饋,以及比普通壓水堆更高的冷卻劑流量供應(yīng)要求,從而影響其系統(tǒng)安全特性。在本文中,選用日本Super LWR為研究對象,開發(fā)了超臨界水堆核熱耦合分析程序與安全特性分析程序,展開了耦合特性分析及安全特性分析。在耦合特性分析中率先提出了超臨界窄縫效應(yīng)的概念,并且將核熱耦合引入超臨界水堆安全分析。 首先,進(jìn)行了超臨界水堆穩(wěn)態(tài)核熱耦合特性分析,并結(jié)合超臨界窄間隙和跨臨界的設(shè)計特征進(jìn)行窄縫效應(yīng)研究：(1)對比了5%富集度的耦合計算結(jié)果與余弦曲線擬合軸向功率分布的非耦合計算結(jié)果,發(fā)現(xiàn)物理熱工耦合導(dǎo)致內(nèi)、外組件堆芯功率峰值因子沿軸向發(fā)生明顯偏移,并且使得最高包殼溫度降低。(2)進(jìn)行了耦合條件下不同燃料棒間隙下的流動換熱特性分析,發(fā)現(xiàn)燃料棒間隙減小,冷卻劑換熱系數(shù)明顯增加,而最高包殼溫度明顯降低,但是變化幅度均較非耦合計算結(jié)果小。研究了不同流量時燃料棒間隙最大允許值,為設(shè)計優(yōu)化提供理論參考。 其次,進(jìn)行了超臨界水堆瞬態(tài)耦合特性分析及滑壓啟動特性研究：(1)通過冷卻劑溫度升高瞬態(tài)及慢化劑溫度升高瞬態(tài)的特性分析,發(fā)現(xiàn)物理與熱工之間的耦合作用將導(dǎo)致堆芯功率隨時間明顯降低,其中冷卻劑通道入口溫度升高引起功率降低幅度最明顯。(2)通過堆芯功率升高瞬態(tài)的特性分析,發(fā)現(xiàn)冷卻劑溫度隨功率升高呈現(xiàn)升高趨勢,而物理熱工的反饋作用機(jī)制抑制了最高包殼溫度的升高幅度。(3)以滑壓啟動的功率上升過程為例,進(jìn)行平均流量條件下不同組件的啟動特性分析,提出堆芯組件冷卻劑流量分配方案和滑壓啟動曲線優(yōu)化方案。 最后,進(jìn)行了超臨界水堆安全特性分析：(1)以主給水流量降低、溫度降低和壓力升高三種擾動為例,對比分析了不同擾動時控制參數(shù)及安全特性變化,發(fā)現(xiàn)主給水溫度降低與流量降低導(dǎo)致更加顯著的特性變化；(2)以5%流量喪失事件、單臺冷卻劑泵故障事件及喪失廠外電源事件為例,進(jìn)行單通道安全特性分析及其敏感性分析,發(fā)現(xiàn)基于時空動力學(xué)耦合求解的最高包殼溫度始終低于點堆方程求解的最高包殼溫度；(3)以給水加熱喪失事件和單臺冷卻劑泵故障事件為例,進(jìn)行多通道安全特性分析。結(jié)果表明具有最大功率因子燃料組件的最高包殼溫度峰值遠(yuǎn)遠(yuǎn)高于其它燃料組件,但是仍滿足安全準(zhǔn)則要求。
[Abstract]:The Supercritical Water reactor has the characteristics of large temperature difference between the inlet and outlet of the core, low coolant flow rate and narrow fuel rod gap, which leads to strong physical-thermal feedback and higher coolant flow supply requirements than ordinary PWR. As a result, the security characteristics of the system are affected. In this paper, Super LWR, Japan, is selected as the research object, and the nuclear-thermal coupling analysis program and safety characteristic analysis program of Supercritical Water reactor are developed, and the coupling characteristic analysis and safety characteristic analysis are carried out. In the analysis of coupling characteristics, the concept of Supercritical narrow slit effect is first put forward, and the nuclear-thermal coupling is introduced into the safety analysis of Supercritical Water reactor. Firstly, the steady-state nuclear-thermal coupling characteristics of Supercritical Water reactor (SCR) are analyzed. Combined with the design characteristics of supercritical narrow gap and transcritical, the narrow slit effect is studied: (1) the coupling calculation results of 5% enrichment and the uncoupled calculation results of axial power distribution fitted by cosine curve are compared. It is found that the peak power factor of the inner and outer core of the inner and outer components is obviously shifted along the axial direction, and the maximum shell temperature is decreased. (2) the flow heat transfer characteristics under different fuel rod gaps under the coupling condition are analyzed. It is found that the heat transfer coefficient of coolant increases obviously with the decrease of fuel rod gap, while the maximum shell temperature decreases obviously, but the variation range is smaller than that of uncoupled calculation. The maximum allowable value of fuel rod clearance at different flow rates is studied, which provides a theoretical reference for design optimization. Secondly, the transient coupling characteristics of Supercritical Water reactor (SCR) and the start-up characteristics of sliding pressure are studied: (1) the transient characteristics of coolant temperature increase and moderator temperature rise are analyzed. It is found that the coupling between physics and thermal engineering will lead to the decrease of core power with time, among which the increase of inlet temperature of coolant channel leads to the decrease of power. (2) through the analysis of transient characteristics of core power increase, It is found that the coolant temperature increases with the increase of power, and the feedback mechanism of physical heat suppresses the increase of the highest shell temperature. (3) take the power rise process of sliding pressure start as an example. The start-up characteristics of different components under the condition of average flow rate are analyzed, and the coolant flow distribution scheme and sliding pressure start-up curve optimization scheme of core assembly are put forward. Finally, the safety characteristics of Supercritical Water reactor (SCR) are analyzed: (1) taking three disturbances: the decrease of main water flow rate, the decrease of temperature and the increase of pressure, the changes of control parameters and safety characteristics under different disturbances are compared and analyzed. It is found that the decrease of main water supply temperature and flow rate lead to more significant characteristic changes. (2) taking 5% flow loss event, single coolant pump failure event and out-of-plant power loss event as examples, the single channel safety characteristic analysis and sensitivity analysis are carried out. It is found that the maximum shell temperature based on the coupling of space-time dynamics is always lower than that of the point pile equation. (3) taking the loss of feed water heating and the fault event of single coolant pump as examples, the multi-channel safety characteristics are analyzed. The results show that the peak temperature of the maximum shell temperature of the fuel assembly with the maximum power factor is much higher than that of the other fuel assemblies, but it still meets the requirements of the safety criteria.
【學(xué)位授予單位】：華北電力大學(xué)
【學(xué)位級別】：博士
【學(xué)位授予年份】：2013
【分類號】：TL364
本文編號：2475443
[Abstract]:The Supercritical Water reactor has the characteristics of large temperature difference between the inlet and outlet of the core, low coolant flow rate and narrow fuel rod gap, which leads to strong physical-thermal feedback and higher coolant flow supply requirements than ordinary PWR. As a result, the security characteristics of the system are affected. In this paper, Super LWR, Japan, is selected as the research object, and the nuclear-thermal coupling analysis program and safety characteristic analysis program of Supercritical Water reactor are developed, and the coupling characteristic analysis and safety characteristic analysis are carried out. In the analysis of coupling characteristics, the concept of Supercritical narrow slit effect is first put forward, and the nuclear-thermal coupling is introduced into the safety analysis of Supercritical Water reactor. Firstly, the steady-state nuclear-thermal coupling characteristics of Supercritical Water reactor (SCR) are analyzed. Combined with the design characteristics of supercritical narrow gap and transcritical, the narrow slit effect is studied: (1) the coupling calculation results of 5% enrichment and the uncoupled calculation results of axial power distribution fitted by cosine curve are compared. It is found that the peak power factor of the inner and outer core of the inner and outer components is obviously shifted along the axial direction, and the maximum shell temperature is decreased. (2) the flow heat transfer characteristics under different fuel rod gaps under the coupling condition are analyzed. It is found that the heat transfer coefficient of coolant increases obviously with the decrease of fuel rod gap, while the maximum shell temperature decreases obviously, but the variation range is smaller than that of uncoupled calculation. The maximum allowable value of fuel rod clearance at different flow rates is studied, which provides a theoretical reference for design optimization. Secondly, the transient coupling characteristics of Supercritical Water reactor (SCR) and the start-up characteristics of sliding pressure are studied: (1) the transient characteristics of coolant temperature increase and moderator temperature rise are analyzed. It is found that the coupling between physics and thermal engineering will lead to the decrease of core power with time, among which the increase of inlet temperature of coolant channel leads to the decrease of power. (2) through the analysis of transient characteristics of core power increase, It is found that the coolant temperature increases with the increase of power, and the feedback mechanism of physical heat suppresses the increase of the highest shell temperature. (3) take the power rise process of sliding pressure start as an example. The start-up characteristics of different components under the condition of average flow rate are analyzed, and the coolant flow distribution scheme and sliding pressure start-up curve optimization scheme of core assembly are put forward. Finally, the safety characteristics of Supercritical Water reactor (SCR) are analyzed: (1) taking three disturbances: the decrease of main water flow rate, the decrease of temperature and the increase of pressure, the changes of control parameters and safety characteristics under different disturbances are compared and analyzed. It is found that the decrease of main water supply temperature and flow rate lead to more significant characteristic changes. (2) taking 5% flow loss event, single coolant pump failure event and out-of-plant power loss event as examples, the single channel safety characteristic analysis and sensitivity analysis are carried out. It is found that the maximum shell temperature based on the coupling of space-time dynamics is always lower than that of the point pile equation. (3) taking the loss of feed water heating and the fault event of single coolant pump as examples, the multi-channel safety characteristics are analyzed. The results show that the peak temperature of the maximum shell temperature of the fuel assembly with the maximum power factor is much higher than that of the other fuel assemblies, but it still meets the requirements of the safety criteria.
【學(xué)位授予單位】：華北電力大學(xué)
【學(xué)位級別】：博士
【學(xué)位授予年份】：2013
【分類號】：TL364
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