Recent content by Islam Nabil

No, the energy bins will be below the energy cutoff. 0.025 ev will not be detected. The cutoff or phys:n must be in the input file to handle the low energy bin E= 0.025ev

How can i detect the thermal neutron, E = 0.025 Ev, by MCNP using CUToff Or PHYS:N cards?

I have a problem with my MCNP5 cross-section library. Se and Te elements with atomic numbers; 34000 and 52000, respectively, have no neutron cross-section stored, either neutron contentious or neutron other, in the library, and so I can't run any input file that has any material of them because...

Need a the source file of a 3d scatter plot with connected lines like attached figure

The input file ... There is a lattice 16x16 in the center 0.0.0 and another big lattice ... If u make the cell 2 which fill the small lattice a univers of for ex u =5 and fill the big one by it ...Always there is a block message or a geometry error .... Try it

I will be more detailed ... To make this ... I have to make a 17x 17 rpp boxes manually and in the same way its surfaces and cells . Haven't I? I think there is another way ... !

May be the question not clear ... I ment need a input file which consists of lattice 16x16 inside a lattice 17x17 pitches ?? I will be more detailed ... How can i make a lattice 16x16 is a univers fill another lattice ... There is a block message always ...

need a MCNP simulated APR1400 input file? which consists of lattice 16x16 inside a lattice 17x17 ??

Thanks a lot doctor... King regards

Thank you, thank you very much, Doctor... The problem has already been resolved... Thank you...

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There is an input file for a simple 16 x 16 lattice fuel assembly. I have a message blocking the run of the code; "bad trouble in subroutine newcel of mcrun source particle no 1 random number 6647299061401 zero lattice element hit." What is wrong?