Sharif Digital Repository

This paper proposes a new method for solving the time-dependent neutron transport equation based on nodal discretization using the MCNPX code. Most valid nodal codes are based on the diffusion theory with differences in approximating the leakage term until now. However, the Monte Carlo (MC) method is able to estimate transport parameters without approximations usual in diffusion method. Therefore, improving the nodal approach via the MC techniques can substantially reduce the errors caused by diffusion approximations. In the proposed method, the reactor core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates and leakage ratio are... 

In this paper, localization of a noise source from limited neutron detectors sparsely distributed throughout the core of a typical VVER-1000 reactor is investigated. For this purpose, developing a 2-D neutron noise simulator for hexagonal geometries based on the 2-group diffusion approximation, the reactor dynamic transfer function is calculated. The boxscheme finite difference method is first developed for hexagonal geometries, to be used for spatial discretisation of both 2-D 2-group static and noise diffusion equations. The dynamic state is assumed in the frequency domain which leads to discarding of the time disrcetisation. The developed 2-D 2- group neutron noise simulator calculates...