On the use of delta-tracking and the collision flux estimator in the Serpent 2 Monte Carlo particle transport code

The Serpent Monte Carlo code was originally developed for the purpose of spatial homogenization and other computational problems encountered in the field of reactor physics. However, during the past few years the implementation of new methodologies has allowed expanding the scope of applications to new fields, including radiation transport and fusion neutronics. These applications pose new challenges for the tracking routines and result estimators, originally developed for a very specific task. The purpose of this paper is to explain how the basic collision estimator based cell flux tally in Serpent 2 is implemented, and how it is applied for calculating integral reaction rates. The methodology and its limitations are demonstrated by an example, in which the tally is applied for calculating collision rates in a problem with very low physical collision density. It is concluded that Serpent has a lot of potential to expand its scope of applications beyond reactor physics, but in order to be applied for such problems it is important that the code users understand the underlying methods and their limitations.