When a neutron source interacts with any material, there are reactions that produce photons, namely photons from inelastic reactions and photons from capture reactions. I am counting these photons, I just wanted to separate the capture and inelastic ones into different tallies. To calculate the...
I don't think the reaction number is 102. I simulated it and got this error: "fatal error. Illegal photon reaction number, 102, on FM 24 card."
Is there a reference to where I can find this table?
Got it. I'm using tally F4. My intention is to calculate the sigma of a given material. I need to separate only the part of the spectrum that comes from capture reactions. In this case, the detector isn't the target material; I need to calculate the material that interacts with the beam, and...
Hello everyone. I have a 14 MeV neutron source and a gamma photon detector at a distance. This detector is for counting photons from inelastic reactions and capture.
How do I configure a tally to output only inelastic and capture reactions separately?
use the formula to reverse engineer this and get this answer, because just looking at the data like this, I can't tell. Do you have the density of this material?
Do I need to put the pos inside the cell of interest? Because the cell I want to make the source is well above the position 0 0 0, but mcnp understands by the exclusion method that I only want the voxels of cell 132?
I'm very confused with this configuration, I set the pos to 0.001 0.001 0.001...
To calculate the atomic density of a material, follow these steps
1 - add up all the atomic masses
2 - know the density of the material
3 - calculate the mass fraction for each radionuclide
4 - calculate the atomic density with the formula
((density*mass fraction*avogadro number)/total atomic...
I would like some help to configure the SDEF card for a lattice cell.
I am configuring it but I am getting the error "fatal error. Cell 889 in SDEF CEL path not at lev=0"
I have tried several configurations but I am not having any success, please help me. I want to set cell 132 as my source...
The srun is the cluster system scheduler, used to configure the number of nodes and CPUs to be used. The MCNP manual states that the command to use multiple CPUs is 'mpirun -np X', where X is the number of CPUs, followed by mcnp6.mpi (MCNP6 compiled for MPI) and input/output file parameters. I...
Hi Pete94857.
Thank you for the help. I've seen some MPI script examples similar to mine, but they didn't generate this many runtpe files. I thought I might be doing something wrong or missing a command.
Hi DAntanov.
The book "Radiation Problems: From Analytical to Monte Carlo Solutions" has many examples, from the simplest to the most complex. I recommend reading it, you can learn a lot. I don't know if I can upload books in PDF format here, but if a moderator allows it, I can upload them.
When I increase the number of nodes in the simulation (e.g., -N 3), the number of runtpe files and outputs generated equals the number of nodes; in this case, three runtpe files and three outputs appear. Each output has a different simulation time, and sometimes even different results. Is this...
Hello everyone. I am simulating a Cesium-137 source with an energy of 0.662 MeV and an activity of 225 mCi. When I use the "T: tally time bins" card, for example:
F24:P 1
E24 20
T24 0 1000 25I 3600 196I 200600
I understand that I am asking the program to give me the average flux in this cell...
You opened my mind, thank you very much for the ideas. I will definitely use it. When I complete it, I'll post it here for anyone who wants to use the input with a pinpoint chambers.