Understanding F6 Tallies in MCNP Simulations

In summary, the F6 and +F6 tallies in MCNP simulations allow for material density and are normalized to the source-particles in the chosen cells. When calculating the MeV/gm energy density, it should be multiplied by the amount of source-particles in the chosen cells. Additional calculations, such as incorporating volume, density, and converting to joules, may also be necessary.
  • #1
Andrev
17
0
Hi,

I'm working on a MCNP simulation where I have to use F6 tallies. According to the manual: "In the F6 and +F6 tallies, material density is available for the chosen cells, and normalization is MeV/gm/source-particle."

To which source-particles is this value normalized: the source-particles in the chosen cell or the source-particles of the whole simulation?

(Example: Let's say I have two cells (cell1 and cell2) in the simulation, both tallied. If I would like to get cell1's MeV/gm energy density with which amount of source-particles do I have to multiply it with: those of cell1 or those of cell1+cell2?)
 
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  • #2
it is per simulation source particle

You should multiply it by the real world source strength.

For example in a fusion reactor you might get 1e21 neutrons per second , so you would take the tally results and multiply it be 1e21 neutrons to get using of MeV per gram

I typically go a little further and include volume, density and convert to joules in my calculations (I wish MCNP offered some choices about the units)
 
  • #3
Thank you very much!
 

Related to Understanding F6 Tallies in MCNP Simulations

What is the purpose of F6 tallies in MCNP simulations?

F6 tallies in MCNP simulations are used to calculate the energy deposition and flux in different regions of a system. This information is important for understanding the effects of radiation on materials and determining the shielding requirements for nuclear facilities.

How are F6 tallies defined in MCNP simulations?

F6 tallies are defined by specifying the energy bins and the regions in which the tallies will be calculated. This allows for the calculation of energy deposition and flux at specific energies and in specific areas of the system.

What is the difference between surface and cell F6 tallies?

Surface F6 tallies calculate the energy flux on the surface of a specified region, while cell F6 tallies calculate the energy deposition within a specified region. Surface tallies are typically used for radiation streaming or dose calculations, while cell tallies are used for material damage and activation calculations.

How can F6 tallies be used to optimize shielding design?

F6 tallies can be used to calculate the energy deposition and flux in different materials and regions, which can then be used to determine the most effective shielding materials and thicknesses for a given system. By adjusting the shielding design, the F6 tallies can be re-run to ensure that the desired level of protection is achieved.

Are F6 tallies affected by statistical noise?

Yes, like all other MCNP tallies, F6 tallies are affected by statistical noise due to the random nature of particle transport. To reduce this noise, it is recommended to increase the number of particles in the simulation or to run multiple simulations and average the results.

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