Does MCNP5 account for gamma decay from fission products in reactor cores?

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MAAQ
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Hi ,
This is my first post in this forum, I am new and happy to be in this forum :)

My question is, during the calculation of neutron and photon of a single-point reactor core, does MCNP5 taking into account the gamma decay? because during fission process, fission product can emit gamma. Does MCNP consider that?

thank you
 
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MAAQ said:
My question is, during the calculation of neutron and photon of a single-point reactor core, does MCNP5 taking into account the gamma decay? because during fission process, fission product can emit gamma. Does MCNP consider that?
Yes it can. Here is an example - http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/35/106/35106353.pdf

As I understand it, MCNP is primarily a particle (neutron and gamma) transport code, however, it can be coupled to other codes, e.g., ORIGEN, to simulate/calculate decay of radionuclides, or depletion. It all depends on how the source is defined.
 
hi Astronuc,
I am really glad that you answered my question. I used to see your comments long time ago before starting to use this forum. Thank you very much :)

Let's make the question more clear, if I am using MCNP5 alone without coupling it. Does it enough to track all gamma?
Another thing, code like Geant 4 "I haven't used it yet" but can it follow gamma decay.
 
MAAQ said:
hi Astronuc,
I am really glad that you answered my question. I used to see your comments long time ago before starting to use this forum. Thank you very much :)

Let's make the question more clear, if I am using MCNP5 alone without coupling it. Does it enough to track all gamma?
Another thing, code like Geant 4 "I haven't used it yet" but can it follow gamma decay.
It appears that one can address a source with a source card, but it is limited to particularly sources. See the SDEF card, and also, SUR for a surface source and CELL for a volume source.

https://canteach.candu.org/Content Library/20043507.pdf

The sources seem rather limited.

In order to do a time dependent sources, e.g., fissions of a fuel rod or assembly, I believe one has to couple a depletion module, e.g., CINDER, ORIGEN, to MCNP.
For example - https://mcnp.lanl.gov/pdf_files/la-ur-12-00676.pdf

I understand that MCNP 6.2 is out now, and that has combined features from MCNP and MCNPX.
 
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