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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
Burn-up (MCNPX Manual Page: 5-77)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
http://homepages.cae.wisc.edu/~bohm/neep412/lucasMCNPTutorialspring2010.pdfBurn-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.
thanks a lot. and I am going to try mcnpx and mcode.:DAstronuc 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
To prepare the input files for the burnup process in MCNP5, you will need to define the material compositions, initial conditions, and depletion parameters. This can be done through the use of MCNP5's input file format, which is a plain text file that can be edited using a simple text editor or a graphical user interface.
The choice of nuclear data library for the burnup process in MCNP5 depends on the specific application and the desired level of accuracy. The most commonly used libraries are ENDF/B-VI, ENDF/B-VII, and JENDL. It is recommended to consult the MCNP5 manual or seek advice from experienced users to determine the best library for your particular case.
MCNP5 does have built-in capabilities for performing burnup calculations, but they are limited to a few predefined fission product chains. For more complex depletion scenarios, it is recommended to use a dedicated depletion code such as ORIGEN or CINDER. The output from these codes can then be used as input for MCNP5's transport calculations.
To verify the results of the burnup process in MCNP5, it is important to compare them with experimental data or other validated codes. It is also recommended to perform sensitivity and uncertainty analyses to assess the impact of various parameters on the results. Additionally, benchmark calculations with known solutions can be used to verify the accuracy of the burnup calculations.
While MCNP5 is a widely used and well-tested code, there are some limitations and known issues with its burnup capabilities. These include limitations in the number of isotopes that can be handled and potential inaccuracies in certain nuclear data libraries. It is important to carefully review the results and consult the MCNP5 manual or experienced users to ensure the validity of the burnup calculations.