Data for: Estimation of monoenergetic neutron source distribution for a prescribed power density generation in subcritical X,Y-geometry fission-chain reacting systems

Any sub-critical system can be driven by time-independent interior sources of neutrons. Thus, we present a methodology to determine the intensities of uni- form and isotropic sources of neutrons that must be added inside a sub-critical system so that it becomes stabilized, generating a prescribed distribution of power. To accomplish this, we use the time-independent, monoenergetic, X,Y-geometry neutron transport equa- tion for the forward transport problem and the equation which is adjoint to it for the adjoint transport problem. The well-known reciprocity relation is used to correlate these two prob- lems, yielding an explicit relation between interior sources and the power generated by the fuel regions. The discrete ordinates (SN) formulation is applied to the forward and adjoint transport problems with the level-symmetric angular quadratures for X,Y-geometry calcu- lations. The adjoint SN problems are solved by the adjoint response matrix-constant nodal (RM†-CN) method with the adjoint partial one-node block inversion iterative scheme.

任何亚临界系统均可由与时间无关的内源中子驱动。因此，我们提出了一种确定必须添加至亚临界系统内部以实现稳定并产生预定功率分布的均匀各向同性中子源强度的方法。为实现此目标，我们采用了针对正向传输问题的时不变、单能、X,Y几何中子传输方程及其伴随传输问题的伴随方程。利用已知的互易关系将这两个问题联系起来，从而得出内部源与燃料区域产生的功率之间的显式关系。对于正向和伴随传输问题，我们应用了离散坐标（SN）公式，并使用X,Y几何计算的级对称角 quadratures。伴随SN问题通过伴随响应矩阵-常节点（RM†-CN）方法和伴随部分单节点块逆迭代方案求解。