Burnup Output

Discussion on physics, models and calculation methods

Burnup Output

Postby tim wyant » Tue Jul 19, 2011 6:03 pm

I'm working on a comparison study using SERPENT 1.1.14 and SCALE6.0. I'm basing my comparison on the report "Burnup Calculation Capability in the PSG2/SERPENT Monte Carlo Reactor Physics Code" by Jaakko Leppanen and Maria Pusa.

As part of the study, I am trying to compare the mass of various isotopes such as PU-239, Xe-135, Zp-237, ect. The isotopic mass is being compared as burnup increases from 0-51000MWd. My difficulty is determining the form of output in the SERPENT file. The output "<input>_dep.m" lists a MAT_fuelp1r1_ADENS and MAT_fuelp1r1_MDENS matrix. It also shows a formula for total mass as being TOT_MASS(:,j) = TOT_MASS(:,j) + MAT_fuelp1r1_VOLUME(j).*MAT_fuelp1r1_MDENS(:,j);

So for each burnup step, do I determine the total mass of the nuclide by multiplying the volume (cm^3) with the mass density (gm/cm^3)? This does not seem to give the result I am looking for.

Thanks for your help.

Respectfully,

Tim Wyant
tim wyant
 
Posts: 20
Joined: Tue Jul 19, 2011 5:47 pm

Re: Burnup Output

Postby Jaakko Leppänen » Tue Jul 19, 2011 7:56 pm

Tim,

Table MAT_xxx_MDENS gives the mass densities of nuclides in units g/cm3 and MAT_xxx_VOL material volume in cm3 (3D-problem) or cross-sectional area in cm2 (2D-problem). When your read the file in matlab, TOT_VOLUME is calculated automatically by summing the products over all materials, so the result should be total nuclide masses in grams (3D problem) or grams per axial length (2D problem).

If your results do not make sense, the first thing to do is to check your volumes. Serpent tries to calculate some cell volumes automatically, but the routine is not fool-proof. If the volumes are wrong, it also affects the flux normalization in the burnup calculation, and the depletion rate is not what it should be. You can run a built-in volume-checker routine by using command-line parameter "sss -checkvolumes N <input>", and the code will sample N random points in the geometry and determine cell and material volumes from the results. The values used in the calculation are given in the resulting output for comparison.

Also note that the unit for burnup is MWd per kgU, not per tonne, so if you use numerical value of 51000, you are off by a factor of 1000.

I hope this helps. Let me know if you cannot find the problem.
- Jaakko
User avatar
Jaakko Leppänen
Site Admin
 
Posts: 1927
Joined: Thu Mar 18, 2010 10:43 pm
Location: Espoo, Finland

Re: Burnup Output

Postby tim wyant » Wed Jul 20, 2011 10:46 pm

Jaakko,

I went back and looked at your "Burnup Calculation Capability..." paper from 2009 and realized your isotopic comparison graphs were done using atomic densities. Since the SERPENT dep ourput already lists isotopic atomic densities for each requested timestep, it was easiest to go back and rerun the SCALE6.0 and get atom densities in the OPUS output. This allows for direct comparison with no additional calculations and conversions.

Many thanks for your help.

Tim
tim wyant
 
Posts: 20
Joined: Tue Jul 19, 2011 5:47 pm

Re: Burnup Output

Postby Anikin » Thu Sep 08, 2011 1:33 pm

Can anybody tell me in what units decay heat is written in burnup output file? Is it Watts/cm3 or Watts/g???
Anikin
 
Posts: 6
Joined: Fri Mar 26, 2010 3:17 pm
Location: SECNRS, Moscow, Russian Federation

Re: Burnup Output

Postby Jaakko Leppänen » Fri Sep 09, 2011 12:48 am

Variable TOT_DECAY_HEAT in the _res.m output file has the same units as power, so it should be in watts (or watts per cm in 2D problems).

Decay heat data per material is printed in MAT_<mat>_H in the _dep.m output, and the units are the same.
- Jaakko
User avatar
Jaakko Leppänen
Site Admin
 
Posts: 1927
Joined: Thu Mar 18, 2010 10:43 pm
Location: Espoo, Finland

Re: Burnup Output

Postby roskofnj » Tue Jan 09, 2018 11:58 pm

Hi Jakko,

I know this is an old thread, but I have a question on the value output in the burnup calculation 'TOT_ADENS'. In the manual it states this value is an average atomic density. Does this mean it is just a simple averaged value, or is it a flux-weighted value?

My problem is a 17x17 2D, infinite assembly burnup calculation where is specify 39 different pins within one octant and then reflect these to fill the entire assembly. I can then obtain pin-wise material/source information, but Im curious what exactly the TOT_ADENS values represents. I would like to obtain the flux-weighted average atomic density.

Best regards,

-Nate
roskofnj
 
Posts: 10
Joined: Wed Sep 06, 2017 10:23 pm

Re: Burnup Output

Postby Jaakko Leppänen » Thu Jan 11, 2018 3:54 pm

It's simple average (i.e. average over volumes). To get flux-weighted average, you'll need to calculate flux in each material using detectors.
- Jaakko
User avatar
Jaakko Leppänen
Site Admin
 
Posts: 1927
Joined: Thu Mar 18, 2010 10:43 pm
Location: Espoo, Finland


Return to Methods

Who is online

Users browsing this forum: No registered users and 1 guest