Recent content by Alex A

I'm assuming since you get answers both materials are on your MAT card. Can you show us what isn't working by renaming it to a .txt and attaching it to a post please? If you are not allowed to share it can you cut down the problem to something simple that has the same behavior and share that...

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...

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...

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...

The high level program flow seems a bit broken. Interaction in cladding prints multiple times (and this is actually interaction in moderator mislabeled), but the interaction section doesn't run. Eventually it when it does run it reports a scattering event, calculates the new energy of the...

Hello @Joao Pedro, Welcome to PhysicsForums. If you run mcnp with options ip it will plot the geometry. Dotted lines tend to indicate geometry errors. As a shorthand for reading geometry +surface is usually above or outside, -surface is usually below or inside. You've added two nested boxes...

Hi @BaoNgoc, Welcome to PhysicsForums. I'm a bit confused by L=d10, but I'm fairly sure the answer is no. Random numbers picked independently would be wrong half or more of the time. There are advantages to doing the sources separately, it's simpler and more flexible. You could also define...

You will need to rename the .out file to .txt to attach it. Try giving it a different name like neutron400out.txt

The file you are posting is the input file. It's helpful to read the output file because it gives you errors and warnings that are sometimes helpful. All :p cards need to be :h cards ( or deleted). So f6:p needs to be f6:h, imp:p needs to be imp:h. First make the change. If the answer is...

f6:p is photon, not proton. Change to h.

When run MCNP can create a lot of output files, when next run it typically alters a letter and retries. Eventually, 26 letters later it can give up and error. outp, mdata, runtpe, comout and other files depending on what the program is asked for. Move these to a fresh directory or delete if...

There seems to be at least one mistake in the input file. cora1 0.0 0.023 2.3 corb1 0 0.25 2.5 corc1 0 1 90 Is this intended to produce a mesh containing 8 or 90'000 elements (voxels)? For cylindrical geometry corc is the angle, it has to start with a 0, so this is implied. In MCNP this...

Hi @omarkhairallah, Welcome to PhysicsForums. I don't know what par=sa actually does. It's new in 6.1.1 and it looks up alpha decay information from decay tables and CINDER. I don't understand ACT either. When I use par=a erg=5.5 and remove ACT to make it run on my version the tally when...

I get no events in the ptrac file when I run that. Is that successful for you? I don't know ptrac. I would try to do this sort of thing using F8/F6 tallies and PHL. I don't fully understand that either, but you get coincidence and anticoincidence when done right and this can be used to plot...

The file you attached has this sdef line, sdef pos=0.5 0 0 erg=2.45 The file you ran has this sdef line, sdef pos=0 0 0 erg=2.45 If a source is on a surface the code can make mistakes. 0 0 0 is on surface 8 (px 0). If 0.5 is wrong, only a tiny change is needed to not be 0 so 0.001 should work...