Evaluation of Helical-Cruciform Fuel Rod Assemblies for High-Power-Density LWRs
Evaluation of Helical-Cruciform Fuel Rod Assemblies for High-Power-Density LWRs
复制标题
DOI：
10.13182/nt13-104
复制
发表时间：
2014-11
影响因子：
1.5
通讯作者：
T. Conboy;T. J. McKrell;M. Kazimi
中科院分区：
文献类型：
--
作者：
T. Conboy;T. J. McKrell;M. Kazimi
摘要由于与新的核电建设相关的巨大的资本成本，在开发策略以提高现有轻水堆（LWR）的功率密度和提高下一代轻水堆设计的堆芯功率密度方面的兴趣仍然很高。为了实现这些目标，已经提出了螺旋十字形（HC）燃料棒组件。HC燃料棒组件是一种自支撑的核燃料结构，由四瓣形、轴向扭曲的燃料棒紧密排列成正方形阵列组成。相对于传统燃料几何形状的优点包括较大的表面积与体积比和改进的径向混合特性。组件的自支撑特性消除了对栅格板的需要，改善了堆芯水力学。过去的研究已经确定了HC燃料棒几何形状的这些和其他优点，并且已经使其形状和设计适应沸水反应堆（BWR）和压水反应堆（PWR）应用的LWR燃料组件。然而，由于缺乏合适的热工水力相关性，以捕捉HC棒束流行为，这项工作未能完成的潜力评估。最近在这方面已经取得了进展，由于经验的发展专门的液压和横向混合相关HC棒的几何形状。因此，利用HC燃料棒组件的先进轻水堆堆芯设计可以以更高的精度和置信度进行重新检查。对于使用HC棒组件的BWR堆芯，将新的HC棒束相关性应用于子通道模型，发现需要增加紧侧和拐角子通道的水力直径，以防止流量不足。在组装盒侧的轴向位置处添加小突起，对应于每个棒四分之一扭转，以用作垫片。这促使对棒横截面形状进行轻微重新设计。同样地，调整中心水棒区域以维持参考氢与铀原子比。通过这些变化，子通道模型预测，与使用传统燃料的参考沸水堆相比，200 cm扭距HC堆芯的允许功率提高了24%。假设堆芯功率流量比固定，则可实现功率提升。相比之下，模拟表明，采用HC燃料棒组件的PWR堆芯可以允许功率提升高达47%，功率流量比固定。与沸水堆情况的一个主要区别是，过冷临界热通量（CHF）水平随着冷却剂质量速度的增加而上升，与饱和CHF条件的趋势相反。然而，过冷CHF也被认为是更敏感的局部峰值热通量，这是没有明确地在这些模拟建模。这里声称的功率密度增益将最终取决于HC棒的扭曲破坏新生的汽袋的程度，作为过冷CHF极限接近，这种效果应进一步实验研究。
Abstract Because of the immense capital costs associated with new nuclear construction, interest remains high in developing strategies to uprate existing light water reactors (LWRs) for higher power density and in raising core power density for next-generation LWR designs. Toward these goals, the helical-cruciform (HC) fuel rod assembly has been proposed. The HC fuel rod assembly is a self-supporting nuclear fuel configuration consisting of four-petaled, axially twisted fuel rods closely packed in a square array. Advantages over traditional fuel geometry include a larger surface-to-volume ratio and improved radial mixing characteristics. The self-supporting nature of the assembly obviates the need for grid plates, improving core hydraulics. Past studies have identified these and other benefits of HC fuel rod geometry and have adapted its shape and design to LWR fuel assemblies for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. However, because of a lack of suitable thermal-hydraulic correlations to capture HC rod bundle flow behavior, this work fell short of a complete assessment of the potential. Recent progress has been made in this regard due to the empirical development of specialized hydraulic and lateral mixing correlations for HC rod geometry. As a result, advanced LWR core designs taking advantage of the HC fuel rod assembly can be reexamined with a greater degree of precision and confidence. For a BWR core using HC rod assemblies, applying the new HC rod bundle correlations to subchannel models uncovered a need to increase the hydraulic diameter of the tight side and corner subchannels, to prevent flow starvation. Small protrusions were added to the assembly box side at axial locations corresponding to each rod quarter-twist to act as spacers. This prompted a slight redesign of the rod cross-sectional shape. Likewise, the central water rod region was adjusted to maintain the reference hydrogen-to-uranium atom ratio. With these changes, subchannel models predicted a 24% allowable power uprate for the 200-cm twist pitch HC core, in comparison to a reference BWR with traditional fuel. The uprate is accomplished assuming a fixed-core power-to-flow ratio. In comparison, modeling showed that a PWR core employing HC fuel rod assemblies may allow power uprates up to 47%, for a fixed power-to-flow ratio. One major difference from the BWR case is that subcooled critical heat flux (CHF) levels rise with increasing coolant mass velocity, opposite the trend for saturated CHF conditions. However, subcooled CHF is also known to be more sensitive to locally peaked heat flux, which was not explicitly modeled in these simulations. Power density gains claimed here will be ultimately dependent on the degree to which the HC rod’s twist disrupts nascent pockets of vapor as subcooled CHF limits are approached; this effect should be further investigated experimentally.