How to obtain MCNP tally data that is *not* normalized by nps

[Clarita]

Dear all,

This is my first post in this forum.

I would like to know how to obtain the result data of an MCNPX or MCNP6 tally for each simulated history, before the data of different histories is averaged and normalized by the total number of simulated source particles (nps).

I'm calculating the *F8 tally (energy deposition) in a sphere which contains an electron point source with a spectrum of energies. The spectrum of the source corresponds, in fact, to the electron emissions of Iodine-125. One decay of I-125 is equivalent to about 26 electrons (electron yield per decay).
In the end I would like to calculate the energy deposition due to each simulated decay (assumed to be a group of 26 consecutive source particles) in order to obtain a frequency spectrum of the possible energies deposited by one single decay. If I can get the values of the energy deposited by each source particle before all the results of all the simulated particles are pooled in an average value, then I could do the sum of the energies deposited by every consequtive 26 electrons by myself and from this calculate the frequency spectrum.

Does anyone know or has an idea on how to implement this in MCNPX or MCNP6?

Kind regards,

Clarita

 

[Astronuc]

In the following presentation is the statement: "All MCNP tallies (except in criticality problems) are normalized to one starting source particle." I'm not familiar enough with MCNP to know if the statement is still valid or if alternatives have been introduced.
http://cmpwg.ans.org/mcnp/workshop.pdf

I-125 decays by electron capture and transmutes to Te-125/-125m, so I'd expect the ~26 electrons to be secondary from Auger effect, Compton effect or photoelectron absorption. In contrast to beta decay, where there is a continuous spectrum of beta energy (partitioned with the anti-neutrino) and a maximum energy, there would be discrete energies from IT in Te-125m.
http://www.nndc.bnl.gov/chart/decaysearchdirect.jsp?nuc=125TE&unc=nds

 

[raining Pete]

Hello Clarita,

Welcome to the forum! It's great to have you here. I think I can help you out with your question.

To obtain the result data of an MCNPX or MCNP6 tally for each simulated history, you can use the "history" option in the tally card. This will give you the result data for each individual history before it is averaged and normalized by the total number of simulated source particles.

For your specific case, since you are calculating the *F8 tally for energy deposition, you can use the "history" option in the tally card and specify the "particle" option as "electron." This will give you the energy deposition data for each individual electron before it is averaged.

I hope this helps. Let me know if you have any further questions or if this solution works for you.