Neutron Transport

Note, again, that SNOMAN only transports photons / gammas, electrons/positrons, muons, and neutrons. Any other particle generated by a neutron collision (most probably an $\alpha$ particle) will not be transported. The only indication of the generation of such a particle will be the interaction code of the vertex where the particle generation took place.

Another important caveat attached to the neutron code is the generation of light coming from inelastic nuclear collisions in (n,n $^{\prime} \gamma$) reactions. MCNP does this independently of the interaction being considered - for a Monte Carlo of non-analogue type, this can make sense. In SNOMAN, the light from a vertex would have to be correlated with the interaction occurring there - this work has not been done, and at the moment, no light is generated from inelastic neutron collisions within SNOMAN.

Gammas from absorption interactions (reactions of the kind (n,$\gamma$)) are generated in the SNOMAN code, provided the target nucleus is one of the isotopes listed in section 13.3.4. No other absorption $\gamma$s are generated at all at this stage.

With the exception of the caveats supplied (above), the transport of neutrons throughout the SNO detector can be simulated with the SNOMAN code at this stage. The cross sections, partial cross sections, and energy-angle spectrums characteristic of neutron transport are simulated by SNOMAN in the same way as they are in MCNP. Elastic and inelastic interactions are included, and the thermal motion of the target nuclide is taken into account. MCNP in itself represents either the best, or one of the best neutron transport codes available at this time.

Please note that for many nuclides (see the list above), only a natural table is available within SNOMAN for simulating the neutron transport options. At the moment, the target list is isotope specific. Clearly these two facts are inconsistent, and when interpreting a target code for most neutron interactions with a nuclide for which a natural table is available, the user should realise that the target code arises as a result of simulating the averaged behaviour on neutrons interacting with a number of isotopes. The exception to this rule is in interpreting neutron absorption gamma production reactions. In this case, take the example of nickel, the initial target code placed by the neutron handling routines indicate the most abundant isotope ($^{58}$Ni), since a natural nickel table has been used. The gamma cascade routines then randomly select one of the nickel isotopes using the appropriate thermal capture cross sections, and modify the interaction code to reflect their selection, before going on to generate a gamma cascade. In this case, the target code for the interaction should be taken literally, as it should for all the isotope specific tables listed above.