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Forums
Engineering
Nuclear Engineering
MCNP6.2 - ENDF/B reaction numbers for tally multiplier
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[QUOTE="19matthew89, post: 6899693, member: 340324"] Ow! Thanks! Indeed I wanted to check the cross section. I know that in theory the MCNP plotter would allow you to plot both cross section and reaction rates but I have never managed to have it working. I can run the MCPLOT (i.e. the command "MCNP6 Z" works) but if I try to launch mcnp6 ixz i=inputfile that does not work. I think it's not clear to me which input file I have to give as it says that "Cross section plots cannot be made from a runtape or MCTAL file"...so what other file shall be given? [/QUOTE]
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Engineering
Nuclear Engineering
MCNP6.2 - ENDF/B reaction numbers for tally multiplier
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