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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1. What is the meaning of "average" flux tallies in MCNP?

Average flux tallies in MCNP refer to the average number of particles (such as neutrons, photons, or electrons) passing through a specific region or surface over a given period of time. This tally is used to calculate the average flux of particles in a particular area, providing important information for radiation shielding and dose calculations.

2. How are average flux tallies calculated in MCNP?

MCNP uses a Monte Carlo method to simulate the transport of particles through a system. For average flux tallies, the program tracks the number of particles that pass through a specific region or surface and calculates the average based on the number of particles and the time interval specified by the user.

3. What are the units of average flux tallies in MCNP?

The units of average flux tallies in MCNP depend on the type of particle being tracked. For example, the units for neutron flux may be neutrons per square centimeter per second, while the units for photon flux may be photons per square centimeter per second. It is important to specify the correct units when setting up a tally in MCNP.

4. How are average flux tallies used in radiation transport calculations?

Average flux tallies provide important information about the intensity and distribution of radiation in a system. This data is used to calculate the dose received by different materials and to design effective shielding. Average flux tallies can also be used to verify the accuracy of a simulation by comparing the results to experimental data.

5. Can average flux tallies be used to track multiple types of particles at once?

Yes, MCNP allows for the tracking of multiple types of particles in a single simulation. This means that average flux tallies can be used to track the average flux of different particles simultaneously, providing a more comprehensive understanding of the radiation environment in a system.

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