## Abstract

One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as `acceleration of synthetic diffusion` which has been applied to solve the neutron transport equation with `classical schemes of spatial integration` obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author).

## Citation Formats

Valdes Parra, J J.
Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry.; Aplicacion del metodo de difusion sintetica en la solucion numerica de las ecuaciones de transporte de neutrones en geometria plana..
Mexico: N. p.,
1986.
Web.

Valdes Parra, J J.
Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry.; Aplicacion del metodo de difusion sintetica en la solucion numerica de las ecuaciones de transporte de neutrones en geometria plana..
Mexico.

Valdes Parra, J J.
1986.
"Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry.; Aplicacion del metodo de difusion sintetica en la solucion numerica de las ecuaciones de transporte de neutrones en geometria plana."
Mexico.

@misc{etde_10108591,

title = {Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry.; Aplicacion del metodo de difusion sintetica en la solucion numerica de las ecuaciones de transporte de neutrones en geometria plana.}

author = {Valdes Parra, J J}

abstractNote = {One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as `acceleration of synthetic diffusion` which has been applied to solve the neutron transport equation with `classical schemes of spatial integration` obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author).}

place = {Mexico}

year = {1986}

month = {Dec}

}

title = {Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry.; Aplicacion del metodo de difusion sintetica en la solucion numerica de las ecuaciones de transporte de neutrones en geometria plana.}

author = {Valdes Parra, J J}

abstractNote = {One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as `acceleration of synthetic diffusion` which has been applied to solve the neutron transport equation with `classical schemes of spatial integration` obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author).}

place = {Mexico}

year = {1986}

month = {Dec}

}