A Few Comments

Please take note of the fact that we shall not use MCNP itself. This is because there are several major difficulties that prevent the direct inclusion of MCNP into SNOMAN as a kind of 'black-box' routine. The most significant difficulty is that the MCNP code handles all of its own geometry, taking the necessary details from an input file. In order to use the MCNP code directly, all the geometry that is currently specified for the SNOMAN program would have to be specified for the MCNP code. Any future change in the geometry would have to be made to the inputs to both programs. The input specifications for the two codes are sufficiently different to make these tasks difficult to carry out and impossible to verify. Therefore MCNP will not be directly utilised in SNOMAN.

Note also that while many neutron interactions are possible, (n,$2$n), (n,$3$n), (n,$\gamma$), (n,$\alpha$), (n,fission) (n,n$^{\prime}$), to name but a few, MCNP deals only with the transport of neutrons. The same approach will be adopted in SNOMAN. This means that the only indication that a neutron has been absorbed during an (n,$\alpha$) reaction will be the interaction code, and not the subsequent propagation of an $\alpha$ particle.