Introduction

IF YOU ONLY READ ONE SECTION ABOUT NEUTRONS, MAKE IT THE LAST SECTION ON SPECIAL FEATURES! READ SPECIAL FEATURES!

The SNOMAN program models all of its physics in discrete stages - each neutron history can be broken down into tracks and vertices. The tracks represent rectilinear motion through space between two interactions, the vertices represent the interactions. The neutron life cycle can be broken into these stages:

$\bullet$ Calculate the neutron's mean free path in the medium, $\lambda$. 

$\bullet$ Randomly sample the neutron's step length to the nextinteraction, $S$. 

$\bullet$ Find the distance to the next media boundary, $D$. 

$\bullet$ Propagate the neutron the lesser of the two distances. 

$\bullet$ IF( $S < D$ )THEN 

$\bullet$ Account for the interaction. 

$\bullet$ ELSE 

$\bullet$ Account for the boundary. 

$\bullet$ ENDIF

The requirements for neutron transport are therefore twofold - to be able to calculate the neutron's mean free path in a medium, and to choose and resolve any interactions that might occur. Both requirements are difficult to fulfil, and to do so we turn to a well-known transport code developed at the Los Alamos National Laboratory, MCNP, which describes itself as follows:

"MCNP is a general-purpose Monte CArlo N-Particle (MCNP) code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport ... Pointwise cross-section data are used. For Neutrons, all reactions given in a particular cross section evaluation (such as ENDF/B-VI) are accounted for. Thermal neutrons are described by both the free gas and S($\alpha$,$\beta$) models."

Whilst the electron and photon transport capabilities of MCNP will not be used by us in modelling the SNO detector (it has been superceded by another transport code, EGS4, as described by Lay), MCNP does provide an excellent source from which to draw both the algorithms and the data tables required to solve the neutron transport problem.