Problem with F5, FT5 and FU5 card in MCNP

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Discussion Overview

The discussion centers around the use of the MCNP code for calculating neutron flux in an ex-core detector, specifically focusing on the implementation of the F5 card with options FT5 and FU5 in the context of a full-scale PWR model.

Discussion Character

  • Technical explanation
  • Homework-related

Main Points Raised

  • One participant is attempting to calculate neutron flux using the F5 card with FT5 ICD and FU5 options but is unsure if their setup will work, as they are receiving zero results in their full-scale model.
  • Another participant seeks advice on evaluating neutron flux from the fuel assembly in the ex-core detector without providing specific details about their methodology.
  • A participant notes the physical placement of the ex-core detector, which is approximately 1.5 meters away from the neutron source, indicating potential shielding effects from concrete.
  • There is a request for assistance from someone knowledgeable in MCNP, suggesting a need for expertise in the software to resolve the issues presented.

Areas of Agreement / Disagreement

Participants have not reached a consensus, and multiple questions and uncertainties remain regarding the correct setup and evaluation of neutron flux in the MCNP code.

Contextual Notes

Limitations include the lack of clarity on the specific configurations and parameters used in the MCNP code, as well as the potential impact of physical shielding on the results.

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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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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