Re: MCNP help

[Charles Kelsey <ckelsey@csn.net>]

Robert Boston:

To tally neutron energy spectra, I recommend using a flux tally at a 
point or ring detector or averaged over either a cell or a surface and in 
any case supplying additionally a tally energy card.

For example, if you have a surfaces 1,2, and 3 defining a cylinder where 1 
is the side and 2 and 3 are the ends, a tally to determine flux averaged 
across the surfaces in each of three energy bins and a total for all 
energies the following tally cards would be used:

f2:n 1 2 T
e2   .1 1 10

The resulting tallies will be surface flux tallies for the side, an end, 
and the whole, each divided into 4 energy bins; from cutoff to .1 MeV, 
from .1 to 1 MeV, from 1 to 10 MeV, and all energies.

Regarding pulse height tallies (quoted from MCNP4A manual):

"Problems that give correct pulse height tallies are severely limited.  
The only possible variance reduction scheme is biasing the source 
itself.  CAUTION: The pulse height tally does not work well with neutrons 
and is discouraged because of the nonanalog nature of neutron transport 
that departs from microscopic realism at every turn."


On Tue, 28 Nov 1995, FS BOSTON ROBERT DENNIS wrote:

> For all you MCNP experts out there I have a question.
> 
>  I am trying to determine the neutron spetrum of several items (a research reactor 
> and several neutron sources).  My question is what is the exact syntax for 
> the F8:N command (the detector cell particle energy deposition) 
> Alternately can/should one use the PTRAC command to determine 
> particle energy spectrums.
> 
> Thanx in advance.
> ..
> Robert Boston
> Box 8106
> Idaho State University
> Pocatello, Id 83209-8106
> 208/236-2311
> bostrobe@fs.isu.edu
> 
> The standard disclaimer applies
> 
> 


                         | Charles T. Kelsey, IV     |
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