Thanks alot Alex,Alex A said:Most MCNP runs are time independent because most particle transport is time independent. Two neutrons passing close by do not 'see' one another so one neutron only exists at any one time per thread. Two electrons would affect each other in the real world but this is not simulated.
All the tallies will produce an answer that is per source particle. More source particles simulated will make the answer more trustworthy. The statistics on the numbers will be better as they get closer to the 'true' result.
To simulate 1e18 you could enter nps 1e6, and then multiply the tally at the end by 1e18. The same answer would be true for a source twice as strong, except you would multiply by twice as much. 1e6 might be enough for many problems but too few for some.
The ps in nps does not mean 'per second' btw. I am not sure what it actually stands for but it is the number of particle 'histories' (MCNP speak for seeing what happens to a source particle) to run before it stops. It has nothing to do with the strength of the source.
Just for fun, try writing that as 10000000000 not 1E10. It probably won't make any difference, but it might. I don't recall exactly what the limit on an integer is for MCNP 5, but I seem to recall it was bigger than 1E9.Salman Khan said:Can any one please explain if I want to run mcnp 5 input file of certain geometry for flux calculation on different surfaces. So far as I know If I increase the NPS (number of particles) it wll give more accurate result but when I increase NPS from 10e9, input file do not run and close within a second ?
To set up the geometry in an MCNP 5 input file, you need to define the cells and surfaces. Cells are regions of space that are bounded by surfaces. Use the cell cards to describe the material and density of each cell, and surface cards to define the geometric boundaries. Ensure that the surfaces are properly defined using the appropriate surface types (e.g., planes, cylinders, spheres) and that the cells correctly reference these surfaces.
Materials in MCNP 5 are defined using the material cards (M cards). Each material is assigned a unique number, followed by the isotopic composition and atomic densities. The format is M followed by the material number, then the isotopes with their respective atomic fractions or densities. For example, M1 92235 0.02 92238 0.98 defines a material with 2% U-235 and 98% U-238.
The source definition in MCNP 5 is specified using the SDEF card. This card allows you to define the source type, energy, position, and direction. For example, SDEF POS=0 0 0 ERG=1.0 PAR=1 defines a point source at the origin with an energy of 1 MeV and particle type 1 (neutron). Ensure that the source parameters match the requirements of your simulation.
To calculate flux in MCNP 5, you need to use the F4 tally, which calculates the flux averaged over a cell. The F4 tally card should be followed by the cell numbers for which you want to calculate the flux. For example, F4:N 1 2 3 specifies a neutron flux tally for cells 1, 2, and 3. Additionally, you can use the FM card to apply multipliers or energy bins to the tally.
To run the MCNP 5 input file, use the command `mcnp5 i=inputfile o=outputfile r=runfile`. This command executes the simulation using the specified input file and generates output and run files. After the run completes, open the output file to find the results. The flux values will be listed under the tally section, showing the calculated flux for each specified cell. Ensure to check the statistical errors and convergence of the results.