Generation of Transport Equivalent Multi-Group Cross Sections and Diffusion Coefficients for Neutronic Analysis


TOMBAKOĞLU M. , GÜRDAL Ş. O.

25th International Conference Nuclear Energy for New Europe (NENE), Portoroz, Slovenia, 5 - 08 September 2016 identifier

  • Publication Type: Conference Paper / Full Text
  • Volume:
  • City: Portoroz
  • Country: Slovenia

Abstract

In this study, generation of transport equivalent assembly averaged macroscopic cross section set using Monte Carlo technique is discussed for graphite and light water moderated reactors. One of the contributions of this study is demonstration of cell averaging technique to find an expression for direction dependent diffusion coefficient using the simulation results of MCNP6 code with analytical results obtained by using diffusion theory. It should be noted that, reaction rates and flux shapes obtained by using diffusion theory become equivalent to transport theory results for two group transport equivalent cross section set and diffusion parameters. The results are also compared with the lattice cell code results for benchmark problems defined in literature.