Problem with F5, FT5 and FU5 card in MCNP

  • Thread starter NuclearPhysicist
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In summary, the speaker is seeking advice on how to calculate neutron flux in an ex-core detector using MCNP Code. They are unsure if the record of flux registration they have created will work and are hoping for guidance on how to organize it better. They have attempted to calculate it, but have only received non-zero results for a model task, not for the full-scale model. They are also asking for someone who is knowledgeable in MCNP to provide assistance.
  • #1
NuclearPhysicist
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Dear colleagues,

I'm trying to make calculation of flux in ex-core detector. I have to evaluate neutron flux from the part of fuel assembly in full-scale model of PWR.
I don't know very well MCNP Code, so I decided to use F5 card with option FT5 ICD and FU5 671 number of cells, where the part of fuel assembly are placed.
Example:
F5:n 284 -126 0 10.5
FT5 ICD
FU5 671 672 673 674 675 676 677 678 679 680
Will this record of flux registration work?
Could you give me advice how to organize it better?
I tried to calculate it, but all funclionals are zero. I received non-zero results only for model task, but not for full-scale model.
I'm looking forward for your reply.
Thank you a lot in advance!
 
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  • #2
Dear colleagues,

Could you give advice how to evaluate neutron flux from the part of fuel assemply in the ex-core detector?
I'm looking forward for your reply!
 
  • #3
Ex-core detector placed from the neutron source (first fuel assembly) in the distance about 1,5 m in concrete physical shielding.
 
  • #4
Is there anyone here, who know MCNP very well?
 

FAQ: Problem with F5, FT5 and FU5 card in MCNP

1. What are F5, FT5, and FU5 cards in MCNP?

F5, FT5, and FU5 cards are input cards used in the Monte Carlo N-Particle (MCNP) code, a widely used software for simulating and analyzing nuclear and radiation processes. These cards are used to specify the type, energy, and direction of particles in a simulation.

2. What is the problem with F5, FT5, and FU5 cards in MCNP?

The problem with these cards is that they can sometimes lead to inaccurate results or errors in the simulation. This can be due to incorrect input parameters or other issues within the code itself.

3. How can I troubleshoot problems with F5, FT5, and FU5 cards in MCNP?

If you are experiencing issues with these cards in your MCNP simulation, it is important to carefully review your input parameters and ensure they are correct. You can also consult the MCNP user manual or seek help from experienced users or the MCNP support team for further troubleshooting.

4. Are there any alternatives to using F5, FT5, and FU5 cards in MCNP?

Yes, there are alternative methods for specifying particle parameters in MCNP, such as using the SDEF card or the SOURCE subroutine. These methods may be more complex, but they can also provide more flexibility and accuracy in certain simulations.

5. How can I learn more about using F5, FT5, and FU5 cards in MCNP?

There are many resources available for learning about the proper use of F5, FT5, and FU5 cards in MCNP. These include the MCNP user manual, online tutorials, and workshops or training courses. It is also helpful to consult with experienced users or seek help from the MCNP support team.

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