Tallies are quantities that are stored during an MCNP run. There are several tally types depending on what one wants to store (current, flux, …). Here, we will present only one type of tally and associated ‘modifier’: this tally is the one which calculates the neutron flux in a cell (or, when used with modifiers, counting rate, mean cross-section and so on). They are always normalized by the number of source neutrons and the volume (or surface for surface tallies) of the cell.

Tallies are defined in the Data cards block. The keyword for tally declarations is ‘F’ followed by a number (three digits maximum). The last digit corresponds to the tally basic type.*

For example, the neutron fluence in cell 1 is written

F4:n 1

* This last digit ranges from 1 (number of neutrons integrated over a surface) to 7 (fission energy deposition) for neutrons. For our purpose, tally basic type 4 (fluence over a cell) is the most useful; then, one can choose for this tally type F4, F14, …, F994, i. e. a total of 100 flux tallies in the same MCNP run.

where ‘:n’ indicates that neutrons are concerned. It is possible to calculate a given quantity in more than one cell; for example, the neutron flux in cell 1 and 2 will be

F14:n 1 2

The result of this tally will consist of four numbers, the flux value and its statistical error in each of the two cells, whereas the line

F24:n (1 2)

is the mean flux in cells 1 and 2 (thus two numbers, the flux and its error).