MCNP5, dose calculation, subcritical system

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SUMMARY

The discussion focuses on calculating dose from a subcritical system using MCNP5, specifically involving a neutron source surrounded by natural uranium. Participants emphasize the need to incorporate dose contributions from fissions in uranium alongside the neutron source. Key considerations include determining the energy of neutrons, gammas, and beta particles, as well as the fission rate. The correct tallying methods in MCNP5 for accurately modeling these interactions are also highlighted.

PREREQUISITES
  • Understanding of MCNP5 version 5 for radiation transport simulations
  • Knowledge of neutron and gamma radiation interactions with matter
  • Familiarity with fission processes in uranium
  • Basic principles of dose calculation in radiation physics
NEXT STEPS
  • Research how to define SOURCE TERM in MCNP5 for mixed radiation sources
  • Learn about tallying methods in MCNP5 for neutron and gamma dose calculations
  • Study the effects of fission rates on dose calculations in subcritical systems
  • Explore the energy spectrum of neutrons and gammas in radiation transport
USEFUL FOR

This discussion is beneficial for nuclear engineers, radiation physicists, and students working on radiation dose calculations in subcritical systems using MCNP5.

Kashif
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Greetings to all
need forum members help regarding MCNP5 college assignment.
The task is to calculate dose from a system comprising of a neutron source surrounded by natural uranium. the system is subcritical.
The problem is how to calculate dose due to fissions in the uranium because dose from neutron source can be calculated separately but i don't know how to incorporate dose due to fissions.
which tally and how i should use?
 
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Kashif said:
Greetings to all
need forum members help regarding MCNP5 college assignment.
The task is to calculate dose from a system comprising of a neutron source surrounded by natural uranium. the system is subcritical.
The problem is how to calculate dose due to fissions in the uranium because dose from neutron source can be calculated separately but i don't know how to incorporate dose due to fissions.
which tally and how i should use?
Is one talking about dose from neutrons, or neutrons, gammas and beta particles. One simply needs to know the energy of the neutrons (and gammas and betas, if including those), and the fission rate.
 
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Astronuc said:
Is one talking about dose from neutrons, or neutrons, gammas and beta particles. One simply needs to know the energy of the neutrons (and gammas and betas, if including those), and the fission rate.
basically the dose from neutron plus gammas is required. like we have to calculate dose a the surface of sphere filled with enriched uranium plus a point neutron source at center. in this case how we should define SOURCE TERM in MCNP5.
 

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