Direct discrete method and its application to neutron transport problems

• Main Authors: Vosoughi Naser, Salehi Akbar Ali, Shahriari Majid, Tonti Enzo
• Format: Article
• Language: English
• Published: VINCA Institute of Nuclear Sciences 2003-01-01
• Series: Nuclear Technology and Radiation Protection
• Subjects: neutrons; transport equation; direct diserete method; ANISN code; MCNP-4B code
• Online Access: http://www.doiserbia.nb.rs/img/doi/1451-3994/2003/1451-39940302012V.pdf

The objective of this paper is to introduce a new direct method for neutronic calculations. This method, called direct discrete method, is simpler than the application of the neutron transport equation and more compatible with the physical meanings of the problem. The method, based on the physics of the problem, initially runs through meshing of the desired geometry. Next, the balance equation for each mesh interval is written. Considering the connection between the mesh intervals, the final discrete equation series are directly obtained without the need to pass through the set up of the neutron transport differential equation first. In this paper, one and multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without the associated clad and the coolant regions each with two different external boundary conditions. The validity of the results from this new method is tested against the results obtained by the MCNP-4B and the ANISN codes.