# F5 tally

In Monte Carlo techniques, often one would like to determine the flux of particles at a small detector some distance from a radioactive source. Due to the finite number of particles used in the simulation, the distance between the source and detector, and/or the absorption/scatteing by shield materials, only very few particles reach the detector. Therefore to calculate the flux, dose rate, buildup factor, etc., large numbers of particles and large computing times are required to obtain good statistics.

These problems can be overcome by using, for example, the F5 tally (originally developed for the Monte Carlo code MCNP. The F5 tally works as follows:

Send a few particles and at each interaction (including at creation) compute

- the probability that a particle of that type is produced (or the original deviated) in the direction that it would reach the detector

- the probability that the produced particle pointing to the detector reaches it with no interaction

Multiply these two probabilities and add the result to the scoring

This type of tally makes use of what is sometime called a variance reduction technique, namely, use of the “next event estimator” (see the MCNP Primer below). For each source particle and each collision event, a deterministic estimate is made of the ﬂuence contribution at the detector point (or ring in an axisymmetric problem).

AN MCNP PRIMER

MCNP — A General Monte Carlo N-Particle Transport Code, Version 5,Volume I: Overview and Theory, see page 2-91 for a detailed description

Tallying in MCNP

Biasing and Scoring

Biasing

Event biasing