超临界水堆核热耦合及系统安全特性研究

发布时间：2019-05-12 13:56

【摘要】：超临界水堆具有堆芯进出口温差大、冷却剂流量低和燃料棒间隙窄等特征,由此带来强烈的物理-热工反馈,以及比普通压水堆更高的冷却剂流量供应要求,从而影响其系统安全特性。在本文中,选用日本Super LWR为研究对象,开发了超临界水堆核热耦合分析程序与安全特性分析程序,展开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临界窄缝效应的概念,并且将核热耦合引入超临界水堆安全分析。首先,进行了超临界水堆稳态核热耦合特性分析,并结合超临界窄间隙和跨临界的设计特征进行窄缝效应研究：(1)对比了5%富集度的耦合计算结果与余弦曲线拟合轴向功率分布的非耦合计算结果,发现物理热工耦合导致内、外组件堆芯功率峰值因子沿轴向发生明显偏移,并且使得最高包壳温度降低。(2)进行了耦合条件下不同燃料棒间隙下的流动换热特性分析,发现燃料棒间隙减小,冷却剂换热系数明显增加,而最高包壳温度明显降低,但是变化幅度均较非耦合计算结果小。研究了不同流量时燃料棒间隙最大允许值,为设计优化提供理论参考。其次,进行了超临界水堆瞬态耦合特性分析及滑压启动特性研究：(1)通过冷却剂温度升高瞬态及慢化剂温度升高瞬态的特性分析,发现物理与热工之间的耦合作用将导致堆芯功率随时间明显降低,其中冷却剂通道入口温度升高引起功率降低幅度最明显。(2)通过堆芯功率升高瞬态的特性分析,发现冷却剂温度随功率升高呈现升高趋势,而物理热工的反馈作用机制抑制了最高包壳温度的升高幅度。(3)以滑压启动的功率上升过程为例,进行平均流量条件下不同组件的启动特性分析,提出堆芯组件冷却剂流量分配方案和滑压启动曲线优化方案。最后,进行了超临界水堆安全特性分析：(1)以主给水流量降低、温度降低和压力升高三种扰动为例,对比分析了不同扰动时控制参数及安全特性变化,发现主给水温度降低与流量降低导致更加显著的特性变化；(2)以5%流量丧失事件、单台冷却剂泵故障事件及丧失厂外电源事件为例,进行单通道安全特性分析及其敏感性分析,发现基于时空动力学耦合求解的最高包壳温度始终低于点堆方程求解的最高包壳温度；(3)以给水加热丧失事件和单台冷却剂泵故障事件为例,进行多通道安全特性分析。结果表明具有最大功率因子燃料组件的最高包壳温度峰值远远高于其它燃料组件,但是仍满足安全准则要求。
[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.
【学位授予单位】：华北电力大学
【学位级别】：博士
【学位授予年份】：2013
【分类号】：TL364

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