Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)

[HEU_LL]

TL;DR

OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) : mcnp6 input file

Hi everyone,
I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start.
I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through this specific benchmark using MCNP6?
If so, would you be willing to share your MCNP input file for it? Seeing an actual working example would be incredibly helpful for my learning.
I'd be really grateful for any guidance or code you could share!
Thank you so much in advance!

 

[Alex A]

Welcome to PhysicsForums @HEU_LL,

I don't really know burnup, but the geometry is fairly simple. Try writing one for a single cell with reflecting edges, make the fuel material 1 and put it in the center. Then try these lines for a "Case A" burn,

ksrc 0 0 0
kcode 10000 1 50 300
BURN TIME=306 71 381.7 83.1 466 85 461.1 1870
     MAT= 1
     OMIT=-1,7,6014,7016,8018,9018,90234,91232,66159
     POWER= 4.46272126E-2
     PFRAC= 0.887287699 0 1 0 0.876994339 0 0.749871333 0

I simplified a few things and my run did not give a good answer, e.g. 3.9mg 94239 when the right answer is 4.2mg, but I figure the burn lines are not very wrong.

If you get stuck doing the geometry I'm happy to help, but I don't want to put all the effort in.

 

[HEU_LL]

Welcome to PhysicsForums @HEU_LL,

I don't really know burnup, but the geometry is fairly simple. Try writing one for a single cell with reflecting edges, make the fuel material 1 and put it in the center. Then try these lines for a "Case A" burn,

ksrc 0 0 0
kcode 10000 1 50 300
BURN TIME=306 71 381.7 83.1 466 85 461.1 1870
     MAT= 1
     OMIT=-1,7,6014,7016,8018,9018,90234,91232,66159
     POWER= 4.46272126E-2
     PFRAC= 0.887287699 0 1 0 0.876994339 0 0.749871333 0

I simplified a few things and my run did not give a good answer, e.g. 3.9mg 94239 when the right answer is 4.2mg, but I figure the burn lines are not very wrong.

If you get stuck doing the geometry I'm happy to help, but I don't want to put all the effort in.

Thank you so much for your enthusiastic reply, which has been very encouraging. Based on your suggestions, I carefully studied the usage rules of the BURN card, but I did not find data for Am241 and Am243 in the output results. How should this issue be addressed?

 

[Alex A]

Adding a token quantity of Am and Pu isotopes to the material being burned seems to work (0 might work too?). A listing in a material will also load the cross sections into memory rather than rely on the multigroup data in CINDER (I think this is right). Additional isotopes will be tracked according to some simple rules but IIRC their cross sections don't get loaded. I added some of the weirder ones to the OMIT line.

 

[HEU_LL]

Adding a token quantity of Am and Pu isotopes to the material being burned seems to work (0 might work too?). A listing in a material will also load the cross sections into memory rather than rely on the multigroup data in CINDER (I think this is right). Additional isotopes will be tracked according to some simple rules but IIRC their cross sections don't get loaded. I added some of the weirder ones to the OMIT line.

Adding a token quantity of Am and Pu isotopes to the material being burned seems to work (0 might work too?). A listing in a material will also load the cross sections into memory rather than rely on the multigroup data in CINDER (I think this is right). Additional isotopes will be tracked according to some simple rules but IIRC their cross sections don't get loaded. I added some of the weirder ones to the OMIT line.

Thank you very much for your patient guidance. Actually, I added Am241 and Am243 to the initial composition of the burnup materials and set them to a minute quantity of 1.0E-36, but the code failed to run and reported the following errors:
“ fatal error. cross-section tables missing for zaid = 95240. c
fatal error. cross-section tables missing for zaid = 97245. c
fatal error. cross-section tables missing for zaid = 97246. c
fatal error. cross-section tables missing for zaid = 97247. c ”
So, it should be an issue with my mcnp data library?

 

[Alex A]

The automatic search can produce isotopes that are not tabulated. There are multiple ways to deal with that, one is to use an OMIT line, (I didn't need to deal with 97247)
OMIT=-1,11,6014,7016,8018,9018,90234,91232,66159,95240,95244,97245,97246
-1 means all materials being burned, 11 is how many isotopes are in the list. So using that and adding the missing one is one option.

A second method is to turn models on in a BOPT card.

 

[HEU_LL]

The automatic search can produce isotopes that are not tabulated. There are multiple ways to deal with that, one is to use an OMIT line, (I didn't need to deal with 97247)
OMIT=-1,11,6014,7016,8018,9018,90234,91232,66159,95240,95244,97245,97246
-1 means all materials being burned, 11 is how many isotopes are in the list. So using that and adding the missing one is one option.

A second method is to turn models on in a BOPT card.

Thank you very much for your enthusiasm and patient guidance. I deeply admire your excellent professional knowledge. With the two approaches you mentioned above, I have successfully resolved the issue. Once again, I sincerely appreciate your help. Wishing you all the best in your endeavors.