Flux Perturbations by Thermal Neutron Detectors

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Abstract
The transport equation which describes the thermal neutron population in and around a neutron detector is converted to an iterative integral equation. This integral equation is then solved for a wide range of specific cases using a digital computer.Using this method of calculation, the effects upon neutron density of nonisotropic scatter in the surrounding medium, of finite detector dimensions and of scatter by the detector are calculated to an accuracy of better than 1%. Detailed maps of the scalar neutron density in and around finite detectors are available from the calculations. The problem of nonisotropic, non-uniform initial neutron density is formulated using the integral method.

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