How to Resolve MCNP Depletion Code Syntax Errors?

[shakystew]

Hello,

I am a graduate student attempting to run evaluate the depletion of a ceramic film attached to the moderator-side of the fuel clad. I am having some issues with my MCNP syntax/code and I was wondering if one could assist.

My input file is attached. I am not looking for someone to fully analyze the issue, just wanted to see if anyone notices anything wrong. I am getting a burnup of 0.000E+00 for the first step, then N/A for the remaining.

 

[QuantumPion]

I'm not familiar with MCNP's burnup function, what are the units for the burn time and power parameters?

 

[shakystew]

The burn time are in days and the power parameters represent the total recoverable fission system power (MW). (DEFAULT:POWER=1.0). I was forced to OMIT those isotopes in the respective material regions in order to not run into library issues.

 

[QuantumPion]

I'm just throwing out ideas here, but does MCNP let you deplete non-fuel materials (mat 4) like that? How it is defining the power of the cladding?

 

[Stephan_doc]

You have three burnup steeps 0 50 100 days. I know that also must be present PFRAC card ( PFRAC=1 1 1 ) in BURNUP
burnup=0 day*1MW/mass of heavy fuel = 0.0 MWD/MTU
your first time step is problem (must be different from 0)

 

[QuantumPion]

You have three burnup steeps 0 50 100 days. I know that also must be present PFRAC card ( PFRAC=1 1 1 ) in BURNUP
burnup=0 day*1MW/mass of heavy fuel = 0.0 MWD/MTU
your first time step is problem (must be different from 0)

I think PFRAC defaults to 1 for every step if you don't enter it at all. The N/A error leads me to believe it is having a problem with the depletion section somehow though.

 

[Stephan_doc]

I think PFRAC defaults to 1 for every step if you don't enter it at all.

Yes, you are right
Run my attached input
MCNPX will give materials concentrations for fresh fuel. (no burnup, 0 MWD/MTU)
Please read burn scheme from MCNCPX manual for a better understand

 

[shakystew]

It ran without errors, but I would like to see the isotropic depletion within each region (especially the film, to verify it will last an entire fuel cycle). The 'print table 210' only shows the library burnup.

 

[Stephan_doc]

Also change card MATVOL=11.02876358 192.687656 with MATVOL= 192.687656 11.02876358 for MAT=1 4. See concentrations for material 4 (ZrO2 Film) after burn steps requested.