Anyone knows how to execute burnup process in MCNP5?

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SUMMARY

The burnup process cannot be executed directly in MCNP5, as it lacks the necessary burn card feature available in MCNPX. However, users can utilize BURNCAL for depletion calculations with MCNP5, and there are established methodologies for coupling MCNP with other codes like ORIGEN through MCODE. Understanding burnup is crucial for reactor design, especially for long-term operations without refueling. The discussion emphasizes the importance of burnup in economic output and reactor design considerations.

PREREQUISITES
  • Familiarity with MCNP5 and MCNPX software
  • Understanding of nuclear reactor design principles
  • Knowledge of fuel depletion and burnup concepts
  • Experience with code coupling techniques in nuclear research
NEXT STEPS
  • Research the use of BURNCAL for depletion calculations in MCNP5
  • Explore the MCNPX Manual for detailed burnup process instructions
  • Investigate the coupling of MCNP with ORIGEN using MCODE
  • Study case examples of burnup impact on reactor economic output
USEFUL FOR

Nuclear engineers, reactor designers, and researchers involved in fuel depletion analysis and reactor performance optimization will benefit from this discussion.

dongge
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Hi for everyone, anyone knows whether it is possible to execute burnup process in MCNP5? I want to get K-inf as fuel depletion goes on. Thanks
 
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dongge said:
Hi for everyone, anyone knows whether it is possible to execute burnup process in MCNP5? I want to get K-inf as fuel depletion goes on. Thanks
Burn-up (MCNPX Manual Page: 5-77)
Burn-up of the fuel is one of the most important aspects for determining economic output of the reactor,
and could be a driving factor in the design of some of the new reactors that have to be run in remote areas
for 10+ years without a refueling. These constraints of the project are heavily emphasized in the initial design
phase and the final phase of gathering results, but burn-up does not play a large role in the central design portion
of the project. For this reason I have included burn-up as the last thing to cover in this tutorial. By this
point you should have a fully working model and realize what all of the information you are getting out pertains
too. The burn card can only be implemented with MCNPX, and is not in MCNP5.
http://homepages.cae.wisc.edu/~bohm/neep412/lucasMCNPTutorialspring2010.pdf

It's not clear to me, but perhaps BURNCAL can be used for depletion calculations with MCNP5. There are numerous examples of coupled codes in nuclear R&D/industry.

http://prod.sandia.gov/techlib/access-control.cgi/2002/023868.pdf

Here is another example of coupling MCNP with ORIGEN in a code called MCODE.
http://dspace.mit.edu/bitstream/handle/1721.1/16603/55011734.pdf?sequence=1
 
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Astronuc said:
Burn-up (MCNPX Manual Page: 5-77)

http://homepages.cae.wisc.edu/~bohm/neep412/lucasMCNPTutorialspring2010.pdf

It's not clear to me, but perhaps BURNCAL can be used for depletion calculations with MCNP5. There are numerous examples of coupled codes in nuclear R&D/industry.

http://prod.sandia.gov/techlib/access-control.cgi/2002/023868.pdf

Here is another example of coupling MCNP with ORIGEN in a code called MCODE.
http://dspace.mit.edu/bitstream/handle/1721.1/16603/55011734.pdf?sequence=1
thanks a lot. and I am going to try mcnpx and mcode.:D
 
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