MCNP6.2 - How to plot normalized tallies with MCNP6.2?

In summary, the conversation discusses using MCNP to plot tallies and the possibility of using a tool like python or matlab for this purpose. It is suggested to use the FM multiplier subcard in the input program to directly process the data and plot it. There is a question about whether the FM card can take in a continuously updated parameter like k_eff or only a constant.
  • #1
19matthew89
47
12
TL;DR Summary
Newbie with MCNP here: how to plot (normalized tallies)?
Hi everyone,
I'm really new to MCNP here and I'm "playing" around trying to understand what is going on.

I'd like to plot my tallies (F2, F4 and F6). Is there any tool (e.g. python or matlab package) you recommend?

I know that the internal plot MCPLOT is available but I'm wondering how you normalize the tallies there. In a postprocess data treatment you'd multiply, for instance the F4 tallies, times nu power and divide by energy per fission and k_eff. But I was wondering, can you have that done by MCPLOT so to properly plot the "real" tally?

Thanks a lot in advance
 
Last edited:
Engineering news on Phys.org
  • #2
Perhaps you can try to use the FM multiplier subcard in the input program to process the data directly, then plot.
 
  • #3
Hi,
Yeah. I had thought of that. But as you need to divide by the k_eff which is found in the code itself, can you feed in the FM card a parameter like the k_eff which is continuously updated? Or can you only multiply by a constant?
 

FAQ: MCNP6.2 - How to plot normalized tallies with MCNP6.2?

How do I normalize tallies in MCNP6.2?

To normalize tallies in MCNP6.2, you need to divide the tally results by the total source strength or the total number of source particles. This can be done manually after running the simulation by using the tally multiplier options in the input file or by post-processing the output data.

What is the procedure for plotting normalized tallies in MCNP6.2?

To plot normalized tallies in MCNP6.2, you first need to ensure that your tallies are normalized as described above. Then, you can use the MCNP6.2 built-in plotting tools or external software like Python with matplotlib, MATLAB, or Excel to visualize the normalized tally data.

Can I use the FM card to normalize tallies directly in MCNP6.2?

Yes, the FM (Fission Multiplier) card can be used to apply normalization factors directly within the MCNP6.2 input file. The FM card allows you to specify a multiplier for each tally, which can be used to normalize the results according to your specific requirements.

How do I interpret the normalized tally results in the MCNP6.2 output file?

In the MCNP6.2 output file, normalized tally results will appear similar to non-normalized results but scaled according to the normalization factor you applied. You should check the normalization factor you used and ensure that you understand its impact on the results to correctly interpret the normalized data.

What tools can I use to post-process and plot normalized tallies from MCNP6.2?

Several tools can be used to post-process and plot normalized tallies from MCNP6.2, including Python with libraries like NumPy and matplotlib, MATLAB, Excel, and specialized nuclear data visualization software like Vised or VTK. These tools allow for flexible and detailed analysis and visualization of the normalized tally data.

Similar threads

How does MNCP calculate an F6 Tally?
Replies
1
Views
598
If MCNP tallies are normalized, shouldn't they be <1?
Replies
5
Views
2K
How to determine the minimum grid length | Numerical Plasma Physics
Replies
7
Views
2K

Hot Threads

Recent Insights

Back
Top