Recent content by ethnscot

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    Efficient MCNP Lattice Source Help: Defining Universes and Tallies in Cell File

    I actually figured it out myself. I needed eff to be 1E-10. Now I have ***** lost particle in newcel - zero lattice element hit ***** as an error. Does anyone know how to fix this?
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    Revisiting MCNP: Refreshing Skills for a New Job in Nuclear Engineering

    Hello, After some time away I've gotten back into MCNP. I've been in the field of Nuclear Engineering for over ten years, but I recently changed jobs and need to use MCNP. I'm trying to get my skills back up, since I haven't been using it as much in my old job. Looking forward to some great...
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    Efficient MCNP Lattice Source Help: Defining Universes and Tallies in Cell File

    This is what I hate about MCNP, not a lot of documentation. How do I define all of a universe as a source and a tally? I have a lattice like the below code. How do I get this code to work with tallies for positions 1,2, and 3 in the lattice; and a source for the 2's. I get the error "sampling...