Problem with F5, FT5 and FU5 card in MCNP

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The discussion focuses on calculating neutron flux in an ex-core detector using the MCNP code, specifically with the F5 card and FT5 ICD option. The user is attempting to evaluate flux from a fuel assembly in a full-scale PWR model but is encountering issues with obtaining non-zero results, as calculations return zero for the full-scale model. They seek advice on improving their flux registration setup and organizing their MCNP input more effectively. Additionally, they inquire about the impact of the ex-core detector's placement, which is 1.5 meters from the neutron source and shielded by concrete. Expert guidance on MCNP usage and flux evaluation techniques is requested.
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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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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!
 
Ex-core detector placed from the neutron source (first fuel assembly) in the distance about 1,5 m in concrete physical shielding.
 
Is there anyone here, who know MCNP very well?
 

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