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Need help configuring f4:e,p with fm4 to accurately calculate deposited dose for both photons and electrons. The problem is that i can't save inside vised my imput file, i have to open my imput file in an editor text an then reopen this file in vised. So i kinda stuck when i tried to run my code in mcnp6 because the output keep showing me bad trouble in subroutine source of mcrun you need a source subroutine. while im sure i already put my kcode and ksrc in my code (on the picture below)

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Could anyone help me where i should start looking for the mistake I'm beginning in vised for mcnp6 and i want to modify my imput file inside vised to understand how when i changed cards this affects the simulation My input deck is given below.

It is suggested that performing separate runs for each source may simplify the process and help identify errors

The discussion includes technical details on how to extend source definitions and dependent variables for the simulation Additionally, the user mentions challenges with liquid. But i do see xsdir in the mcnp_data directory so i am a bit confused about that I have upload the output file with the errors in case i have not properly explained what's going on.

I am beginner in mcnp Could someone please show me how to read the output file when using tally f5 I saw that the result involves collided and uncollided photon flux, so which part needs to be chosen for calculating Therefore mcnp cannot distinguish which size he has to assign to the surfaces

I want to calculate the dose (with an f2 tally) at the entrance and at the exit of the concrete, which means through the surface 22.2 and 22.3, whose area is not automatically calculated

I could use the sd card to pass the correct area to the code. The discussion revolves around using the fmesh command in mcnp to obtain power distribution for a reactor core geometry, specifically for cfd input Users are advised to ensure good statistical convergence by increasing particle counts and cycles in their simulations The fmesh requires careful setup, particularly regarding normalization and grid configuration, as it outputs neutron flux.

If you have access to the codes there are primers available online by simply searching for mcnp primer on google From there you can learn the basics quite easily The biggest problem is probably getting access to the codes since the new ones are personally licensed and there is a screening process.