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We are collecting http://mcnpx.net/mcnp-manual-mcnpx-manual". If you need one, than have a look on what we have found.

Yes, it does. Georg, http://mcnpx.net"

If the value of TMP parameter (or its default value if not given explicitly) in a cell card differs from temperature of nuclide in cross-section file, then MCNP modifies the nuclide cross-section using free-gas treatment (you can see a warning message during execution and in output file). It...

There is a http://mcnpx.net/inputs/10-hplwr-fuel-scwr-model" if you are interested.

You can order the libraries from RSICC or NEA instead. Have a look on their websites for the package you want.

"Loss to fission weight" gives the number of fissions, however one should be careful with normalization. Also, be sure that you use the correct value of average energy per fission for your system - it can be tricky in some cases.

For this particular purpose, you may be more interested in a simpler tool. For example, have a look at this program http://www.oecd-nea.org/tools/abstract/detail/NEA-1840/

Looks reasonable. You can test your file w/ and w/o '#' to see the difference. The common sense is to design input file containing as simple cell definitions as possible to speed up the computation.

You need to remove KSRC card and provide additional parameter in the execution command: mcnp4c n=myinput srctp=mysrctp Some MCNP4C-based codes have SRCTP card which you can use instead of KSRC for the same thing. MCNPX has currently very confusing error messages if you just skip KSRC card and...