How to Calculate C Parameter in FMn Card for MCNP Code

  • Thread starter chivasorn
  • Start date
  • Tags
    Code Mcnp
I don't know how detailed you want to get.In summary, the conversation revolves around a question about the FMn card in mcnp code. The speaker is unsure of how to calculate the value of the parameter named C that comes before the FMn card. They are seeking help and there is information available in the literature, specifically for medical physics applications.
  • #1
chivasorn
22
0
Hi there

I have a question about FMn card in mcnp code.
there is a parameter (is named C) in front of FMn card, I can not understand to calculate the value of that parameter. please help me.

thanks alot.
 
Engineering news on Phys.org
  • #2


chivasorn said:
Hi there

I have a question about FMn card in mcnp code.
there is a parameter (is named C) in front of FMn card, I can not understand to calculate the value of that parameter. please help me.

thanks alot.
What is the context?

It the C on a separate line and in the first column? This could be a comment.
 
  • #3


I think you are referring to the first value after the FM, which in some literature and depending on what you are multiplying for is referred to as C. Depending on if you want reaction rate, heating and what sort of tally F1, F2, F4 etc then it is different.

There is good information in the literature particularly for medical physics
 

1. What is the purpose of calculating the C parameter in FMn card for MCNP code?

The C parameter in the FMn card for MCNP code is used to specify the number of energy groups in the neutron flux spectrum. This parameter is essential for accurate neutron transport calculations in MCNP.

2. How is the C parameter calculated in FMn card for MCNP code?

The C parameter is calculated by dividing the maximum neutron energy by the energy group width. This value is then rounded up to the nearest integer to determine the number of energy groups.

3. What is the maximum neutron energy and energy group width in MCNP?

The maximum neutron energy is determined by the user and can range from 1e-11 to 1e+7 MeV. The energy group width is also specified by the user and can range from 1e-11 to 1e+3 MeV. These values should be chosen carefully to ensure accuracy in the calculations.

4. How does the C parameter affect the accuracy of MCNP calculations?

The C parameter directly affects the number of energy groups used in MCNP calculations. A higher C parameter will result in more energy groups and therefore, a more accurate neutron flux spectrum. However, this will also increase the computational time and memory requirements of the simulation.

5. Can the C parameter be changed during the MCNP simulation?

No, the C parameter cannot be changed during the MCNP simulation. It must be specified in the FMn card before the simulation begins. If the user wishes to change the C parameter, they must restart the simulation with the new value.

Similar threads

Help with neutron spectroscopy experiments in MCNP code
    • Nuclear Engineering
Replies
19
Views
1K
Meaning of "Average" Flux Tallies in MCNP
    • Nuclear Engineering
Replies
1
Views
1K
If MCNP tallies are normalized, shouldn't they be <1?
    • Nuclear Engineering
Replies
5
Views
2K
MCNP - Tallies definition with "<"
    • Nuclear Engineering
Replies
2
Views
1K
Codes to calculate diffusion parameters of homogeneous reactor
    • Nuclear Engineering
Replies
3
Views
1K
Please help with this MCNP program -- in the output response I get too small a value
    • Nuclear Engineering
Replies
2
Views
1K
MCNP4C suppressing terminal/console window popping up?
    • Nuclear Engineering
Replies
3
Views
1K
Understanding MCNP Tally F5 Output: Tips for Beginners
    • Nuclear Engineering
Replies
1
Views
1K
MCNP Output Data: Tutorials & PDFs for SCWR Criticality Analysis
    • Nuclear Engineering
Replies
2
Views
2K
An extremely basic question on MCNP
    • Nuclear Engineering
Replies
2
Views
2K

Hot Threads

Recent Insights

Back
Top