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 specify the geometry in the MCNP 5 input file, you will need to use the appropriate MCNP 5 geometry commands. These commands include SURFACE, CELL, and MATERIAL cards, which allow you to define surfaces, cells, and materials, respectively. You can also use transformation and lattice commands to create more complex geometries.
The input file in MCNP 5 contains all the information necessary to run a simulation, including the geometry, materials, and other parameters. The output file, on the other hand, contains the results of the simulation, such as flux calculations and particle tracks. Both files are essential for running and analyzing MCNP 5 simulations.
Yes, MCNP 5 is a versatile simulation code that can handle various types of geometries, including simple and complex geometries. However, it is essential to ensure that your input file accurately represents the geometry you want to simulate to obtain accurate flux calculations.
There are several ways to validate the results of your flux calculations in MCNP 5. One way is to compare your results with experimental data or other simulation codes. You can also perform sensitivity analyses by varying input parameters and observing the effect on the flux calculations.
Yes, you can run multiple simulations with different input files in MCNP 5. This can be useful for comparing the results of different geometries or input parameters. However, it is essential to keep track of the input files and their corresponding output files to avoid confusion and ensure accurate analysis of the results.