MCNP Gamma Decay: Neutron & Photon Calculation in Reactor Cores

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Discussion Overview

The discussion revolves around the capabilities of the MCNP5 code in calculating neutron and photon interactions in a single-point reactor core, specifically regarding the consideration of gamma decay during the fission process. Participants explore the potential for coupling MCNP5 with other codes for enhanced simulation accuracy.

Discussion Character

  • Technical explanation
  • Debate/contested

Main Points Raised

  • One participant inquires whether MCNP5 accounts for gamma decay during neutron and photon calculations in a reactor core, particularly due to gamma emissions from fission products.
  • Another participant asserts that MCNP can account for gamma decay, mentioning its primary function as a particle transport code and its ability to couple with other codes like ORIGEN for radionuclide decay simulations.
  • A participant seeks clarification on whether MCNP5 alone is sufficient to track all gamma emissions without coupling to other codes.
  • There is a mention of Geant 4 as another code that may have capabilities for tracking gamma decay, although the participant has not used it yet.
  • Concerns are raised about the limitations of source definitions in MCNP5, specifically regarding the SDEF card for defining sources and the need for coupling with depletion modules for time-dependent sources.
  • It is noted that MCNP 6.2 has been released, which combines features from both MCNP and MCNPX, potentially affecting its capabilities.

Areas of Agreement / Disagreement

Participants express differing views on the sufficiency of MCNP5 alone for tracking gamma decay, with some suggesting that coupling with other codes is necessary for comprehensive simulations. The discussion remains unresolved regarding the extent of MCNP5's capabilities without additional modules.

Contextual Notes

Participants highlight limitations in source definitions within MCNP5, indicating that certain types of sources may not be adequately addressed without coupling to other codes. There is also mention of the need for additional modules for time-dependent source calculations.

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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