A Monte Carlo Method For Analyzing Mixed-Lattice Substitution Experiments Using MCNP

Abstract

Critical experiments involving a small region of test fuel substituted into a reference lattice have traditionally been analyzed using diffusion codes to extract lattice physics parameters of the test fuel such as the critical buckling and the associated bias in the calculation of keff. A method that was first developed in 2006 uses a version of MCNP5 that was modified to allow the analyst to selectively change fission neutron production in various parts of the model. This paper describes the modification made to MCNP5, demonstrates how the substitution experiment analysis is done through several examples using data from the ZED critical facility, and finally, quantifies the expected uncertainties in the method.