Meaning of "Average" Flux Tallies in MCNP

  • #1
a1234
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Hello, I've been working with MCNP on and off for a few years now, but just recently realized that I don't entirely understand how tallies are actually calculated in MCNP, and what they signify.

Taking the example of the F2 tally, the user manual (Section 3.3.5.1) states that F2 is the "flux averaged over a surface." I understand that the F2 tally takes the number of particles incident on a surface and divides it by the surface area. This value is multiplied by the source strength/flux multiplier card to obtain the true value of the flux on the surface.

I don't fully understand how this is an "average" flux. Is it simply an average in the sense that it is divided by the total surface area? And if so, how is the standard deviation (which is used to find the relative error) calculated, and what does this value represent physically?
 
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  • #2
It's an average because different parts of a surface usually have different flux, and this is the total crossings over the total area. I would expect something similar to Poisson statistics, if 100 particles go through a surface then the uncertainty is SQR(100)=10 (to within a certain sigma).

I found F4 tallies to be stranger. They are the total path length of all particles in a cell divided by the volume. Dimensionally it works and it is right, and probably a better way of doing it, but it still feels all kinds of weird.
 
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Related to Meaning of "Average" Flux Tallies in MCNP

What is the meaning of "average" flux tally in MCNP?

The "average" flux tally in MCNP refers to the mean value of the neutron or photon flux over a specified region or volume within the simulation geometry. It is calculated by averaging the flux values obtained from multiple particle histories or events, providing a statistical representation of the flux distribution.

How is the average flux tally calculated in MCNP?

The average flux tally in MCNP is calculated by summing the contributions of particle flux in a given region or volume over all simulated particle histories and then dividing by the number of histories. This results in an average value that represents the expected flux in that region or volume.

What units are used for average flux tallies in MCNP?

The units for average flux tallies in MCNP depend on the type of particles being simulated. For neutron flux, the units are typically particles per square centimeter per source particle (particles/cm²/source particle). For photon flux, similar units are used, but they may also be presented in terms of energy flux depending on the specific tally and application.

How can I improve the statistical accuracy of my average flux tally in MCNP?

To improve the statistical accuracy of your average flux tally in MCNP, you can increase the number of particle histories or events in your simulation. This reduces the statistical uncertainty and provides a more precise estimate of the average flux. Additionally, using variance reduction techniques such as importance sampling or weight windows can help focus computational effort on regions of interest, further improving accuracy.

What are some common sources of error in average flux tallies in MCNP?

Common sources of error in average flux tallies in MCNP include insufficient number of particle histories, leading to high statistical uncertainty, and improper use of variance reduction techniques, which can bias the results. Additionally, errors in the geometry or material definitions, as well as incorrect tally specifications, can also lead to inaccurate flux calculations. It is important to carefully validate and verify the input deck and simulation setup to minimize these errors.

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