MCNP FMESH for Plotting power distribution

In summary, to use FMESH to get power distribution for a core geometry, you need to be careful about normalization and convergence, and read carefully in the FMESH manual about using reflective surfaces and setting up the FMESH. You will also need to calculate the neutron flux or get it from the reactor or model it. FMESH isn't going to work because the geometry is hexagonal. You will have to transfer the power to the CFD calculation from F6 tallies for fuel rods or downscatter and gamma capture in the moderator.
  • #1
AlexFi
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TL;DR Summary
Not familiar with MCNP, trying to get power distribution plot for 1/12 core slice
Hello
I'm trying to use FMESH command to get power distribution of this core geometry.
I want to use xyz coordinate in a 1/12 slice of a core so I could use the output of the MCNP sim for a CFD input
How should I approach this?
Thank you
 

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  • #2
Some generic comments first. I notice you have a code with 10000 per cycle, 50 skipped cycles, and 200 cycles total. It is unlikely you will have good convergence. Check carefully what is happening in the output to see if you are getting good stats. You might want to increase the particle per cycle to 1E5 or even more. And you might want to increase the skipped cycles and total cycles as well. Do some testing and check the results for convergence. MCNP reports a variety of things to guide you about the stats.

For FMESH, you will want to read carefully in the manual about using reflective surfaces. Your surfaces 500 and 501 are reflective. You can do it, you just need to be careful about normalization and such.

You will also want to be reading carefully in the manual about setting up your FMESH. It imposes a tally on top of the model. You set up a grid in x-y-z that does not depend on the cell boundaries. Basically, you specify to the FMESH card the origin of your mesh, the x-axis of the mesh, and the intervals to mesh over in each of x, y, and z.

By default, FMESH is going to give you the neutron flux in each mesh region. The normalization is going to be "per particle started." You can do neutrons or photons.

You can add multipliers to produce various things. For example, you can use a multiplier that is the cross section for various interactions such that you get things like neutron capture, neutron scattering, and so on. You can then get the heat deposited in each grid. This will produce something like MeV per neutron started.

To use this in CFD you are going to need to normalize it to give power, that is, energy per second. That means you will need the number of neutrons per second.

Your reactor may provide you with this information. Or you may need to calculate it. For example: Suppose your reactor is operating at 1 kilo-Watt (1 kW). Then you need to work out the average energy per fission, typically in the range of 200 MeV, depending on design. You can get MCNP to work that out for you for a kcode calculation, check the manual. There are some steps involved.

So you divide 1 kW by 200 MeV, convert the units, and it gives you neutrons per second. Keep in mind that's for the full core and you modeled 1/12. You multiply the value the FMESH gave you by the number of neutrons per second, and it converts it to a value per-second.
 
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  • #3
You asked about FMESH, so my other answer talked about that. For CFD you might be able to use that. Or you might want the heat deposited in components. Check out the F6 tally and the F7 tally in the user manual. They might be more what you want. Again, the normalization will be "per particle started."
 
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  • #4
Grelbr42 said:
You asked about FMESH, so my other answer talked about that. For CFD you might be able to use that. Or you might want the heat deposited in components. Check out the F6 tally and the F7 tally in the user manual. They might be more what you want. Again, the normalization will be "per particle started."
Thank you
Correct me if I'm wrong, but I thought that F6 and F7 tally calculates total energy depositions in a cell. I wanted to map its radial and axial energy distribution to create a surface plot, which is why I use FMESH.
 
  • #5
I don't think FMESH is going to work for you because FMESH uses an x-y-z grid and your reactor geometry is hexagonal. Even if your reactor was on a square grid, it still uses circular fuel rods, so FMESH isn't going to work.

What are you transferring to the CFD calculation? As @Grelbr42 suggested, if you are transferring the fission power from the rods, you need to define F6 tallies for each of your fuel rods. You also need to be very careful with the normalization. Make sure you are getting all the reactions (flux times fission times energy per fission) and the volume in the tally is correct.

If you are transferring energy from downscatter and gamma capture in the moderator, you will need to define F6 tallies for these regions as well.

You will have to manually map the power from each tally number to the correct region(s) in your CFD model.
 
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  • #6
rpp
I'm transferring the power distribution (volumetric heat flux?) to star ccm. Planning to get it plotted, get a best-fit equation, and use that as a heat source in Star CCM+.
I'm very new to MCNP. Can you explain what normalization is and how I do it in MCNP?
 
  • #7
One other thing, is there an mcnp command that will automatically calculate total power output of a reactor?
 
  • #8
At criticality a reactor's power output is whatever you want it to be. If you know the total fission rate you can multiply by 200 MeV or so to get power as said above but usually you know the design goal power and everything follows from that.
 
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  • #9
AlexFi said:
One other thing, is there an mcnp command that will automatically calculate total power output of a reactor?

No. It's not possible. You need to normalize the tallies to reactor power. The tallies are all "per particle started." You need to determine how many particles per second are started. MCNP can't tell what power you are operating at.

If it's a reactor, you start with reactor power. Then you divide this by the energy per fission. Typically that's around 200 MeV, but it depends on the details of your reactor design, the isotopes, and few other things. You *can* get MCNP to tell you that. Divide the power by the energy per fission to get the fissions per second.

Then you need the average neutrons per fission. You multiply fissions per second by this number to get neutrons per second.

An alternative way to normalize is the following. Total the heat deposition (using F6 tallies) in *every* component of the reactor, which will be energy per particle started. Then set this equal to the reactor power. That will let you convert the tally to a number of neutrons per second.
 
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  • #10
About FMESH vs other tallies.

By default, FMESH tallies flux. That is not usually directly useable in a thermal hydraulics type code. Such codes usually want a heat deposition rate.

You have a couple choices.

One choice is to use multipliers. You can multiply the FMESH tally by the cross section for a reaction. This means you get the flux times the cross section, meaning you effectively tally the reaction. So, for example, if you tally capture, you get capture events. There's a multiplier for heat deposition due to scattering, one for capture, one for fission, and some others. You need to read up about the multiplier cards. You might need the latest version of MCNP (6.2) to get the full effect here. You want to read about FM and UM cards.

The other choice is to divide your problem into the portions you want to report the energy deposition into. Then use the F6 type tally for heat, and F7 for fission power.

You still need normalization. Check my previous answer in this thread.
 
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  • #11
I end up using F6 just to see my MCNP power output match the power output listed in the reference paper.
MCNP give me an error every time I put cell 4 or 1 in the tally even though the fuel is located in those cells.

Secondly, can someone help me interpret the tally output. As far as I can understand, I need to multiply the number of particles with the those tally numbers to get energy in MeV per gram of material, is this correct? And because cell 2 and 3 don't have any U235 in them, these value are just from collision heating?
 

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  • #12
The fatal error is that the program can't determine the volume of the cell automatically. So you need to specify the volume, for example with a vol= on the cell definition line for every cell you want to tally that has a zero value in the cell volume table (there will be a note saying why it failed to calculate).

It may help to split the model in the z direction, because the flux in the middle will be different to the flux at the top and bottom. There are a few ways to do this, including using the lattice to tile it for you, but it's not something to worry about until the input file works.
 
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  • #13
Alex A said:
The fatal error is that the program can't determine the volume of the cell automatically. So you need to specify the volume, for example with a vol= on the cell definition line for every cell you want to tally that has a zero value in the cell volume table (there will be a note saying why it failed to calculate).

It may help to split the model in the z direction, because the flux in the middle will be different to the flux at the top and bottom. There are a few ways to do this, including using the lattice to tile it for you, but it's not something to worry about until the input file works.
How should I define the volume? Should I put the volume of individual fuel elements, all fuel elements in the slice combined, or the volume of the slice itself?
 
  • #14
That is actually a more involved question than I was prepared for. I'd put them in cell definitions 3 and 4. So long as your thermal model matches I don't think you then need any corrections for the partial rods, you just need to remember that the total power of the simulation is 1/12th of the design goal.
 
  • #15
Alex A said:
That is actually a more involved question than I was prepared for. I'd put them in cell definitions 3 and 4. So long as your thermal model matches I don't think you then need any corrections for the partial rods, you just need to remember that the total power of the simulation is 1/12th of the design goal.
Sorry I didn't quite understand your answer. Yes, I will add volume definition in cell 3 and 4 in the code.
My question is, how should I calculate the volume of each cell? For example, in cell 3, the coolant channels, should I put the volume of a single coolant channel or should I put the volume of all coolant channels in the reactor slice?
 
  • #16
Single cell volumes in 3 and 4 and then see if that fixes all the errors.

The center rod is only 1/12th exposed to flux so it's tally is 1/12th what the real rod is exposed to, other rods are half exposed or fully exposed. So long as the geometry matches the thermal code this should not be a problem. If the thermal code isn't the same 1/12th segment then some multiplying needs to be done.
 

1. What is MCNP FMESH?

MCNP FMESH is a software package used for plotting power distribution in nuclear reactors. It is a part of the Monte Carlo N-Particle (MCNP) code system, which is widely used for radiation transport simulations.

2. How does MCNP FMESH work?

MCNP FMESH uses a combination of deterministic and stochastic methods to calculate the power distribution in a nuclear reactor. It takes into account the geometry, material properties, and neutron interactions to simulate the movement of neutrons and their resulting power distribution.

3. What are the benefits of using MCNP FMESH?

MCNP FMESH allows for accurate and detailed simulations of power distribution in nuclear reactors. It can handle complex geometries and materials, and provides valuable information for reactor design and safety analysis.

4. Can MCNP FMESH be used for all types of nuclear reactors?

Yes, MCNP FMESH can be used for a wide range of nuclear reactors, including thermal and fast reactors, as well as different fuel types. It is a versatile software package that can be adapted for various reactor designs.

5. How can MCNP FMESH results be interpreted?

The power distribution results from MCNP FMESH are typically presented in the form of 2D or 3D plots, which show the neutron flux or power density distribution within the reactor. These results can be analyzed to evaluate the performance and safety of the reactor design.

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