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