Improved Estimation of Anisotropic Neutron Diffusion Coefficients for Neutron Streaming in Advanced Gas Cooled Reactors (AGRs): Application to Steam I

这个博士项目的目的是解决先进气冷反应堆（AGR）核电站（NPPs）中的数值模拟挑战。在AGR的反应堆物理分析中，需要确定AGR燃料组件（弦组）内的功率分布。为了确定这些功率分布，必须解出中子扩散或中子输运方程。如果求解中子扩散方程（中子输运方程的近似形式），则必须计算径向和轴向中子扩散系数以及其他核截面数据，并将其用作代码中的输入数据。这些径向和轴向均匀化的中子扩散系数必须加以修正，以考虑中子在AGR气体通道中流动的详细影响。这些径向和轴向中子扩散系数的校正因子对于AGR的几何形状非常重要；在10-30%的范围内，导致在整个堆芯分析模拟中采用径向和轴向不同的方向扩散系数。对于AGR正常运行，这些校正因子是相当恒定的，并且产生的扩散参数与测量结果非常吻合，例如扭曲的径向/轴向通量形状。目前对AGR的关注是蒸汽进入运行或关闭AGR堆芯的反应性影响。随着石墨失重量的增加，加氢的反应性效应可以从净负向净正转变。此外，氢填充团簇中的流动空隙，导致中子流动的损失，泄漏的减少和进一步的反应性增加。后一种效应仅在晶格物理计算中得到部分表现，因此目前使用了很大的不确定性来为这种效应提供数值界限。这个博士项目的目的是确定新的数学/计算方法来计算径向和轴向均匀中子扩散系数。与参考中子输运理论模拟相比，这些新的数学/计算方法旨在减少AGR反应堆物理模拟中的悲观情绪，以确定新开发方法的准确性。此外，这些方法可以重新应用于第四代核电站，如高温气冷堆（HTGCRs）和气冷快堆（GCFRs）。

The aim of this PhD project is to address numerical modelling challenges within Advanced Gas Cooled Reactor (AGR) Nuclear Power Plants (NPPs). In reactor physics analyses of AGRs one needs to determine power distribution within AGR fuel assemblies (stringers). In order to determine these power distributions one must solve the neutron diffusion or neutron transport equation. If one solves the neutron diffusion equation (an approximate form of the neutron transport equation) radial and axial neutron diffusion coefficients along with other nuclear cross-section data must be calculated and used as input data within the codes. These radial and axial homogenized neutron diffusion coefficients must be corrected to take into consideration the detailed effects of neutron streaming in AGR gas channels. The correction factors for these radial and axial neutron diffusion coefficients can be very significant for the AGR geometry; in the region of around 10-30%, and results in differing directional diffusion coefficients in the radial and axial directions being used in the whole core reactor analysis simulations. For AGR normal operation, these correction factors are fairly constant and produce diffusion parameters which agree well with measurements, e.g. of distorted radial/axial flux shapes. A current concern for AGRs is the reactivity effects of steam ingress into an operating or shutdown AGR core. As graphite weight loss increases, the reactivity effects of the hydrogen addition can change from net negative to net positive. In addition, the hydrogen fills the streaming voids in the cluster, resulting in a loss of neutron streaming, reduction in leakage and further reactivity addition. The latter effect is only partially represented in lattice physics calculations, and therefore a generous uncertainty is currently used to provide numerical bounds to this effect. The aim of this PhD project will be to determine new mathematical/computational methods for calculating the assembly homogenized radial and axial neutron diffusion coefficients. These new mathematical/computational methods will aim to reduce the pessimism in AGR reactor physics simulations, as compared against reference neutron transport theory simulations, in order to ascertain the accuracy of the newly developed methods. In addition these methods can be re-applied to generation IV nuclear power plants (NPPs) such as high temperature gas cooled reactors (HTGCRs) and gas cooled fast reactors (GCFRs).