Recent content by seedsluis

Sure, Thanks for your help. I am using the most time-assuming way to do so, I think. Maybe later I will find a better method, *.*

OK, since scattering will not be an issue. Now I'm using F2 tally, since I should divide the whole surface into many segments, and I need the flux through individual segment, I try to use FSn to get it, I still try to find an way to calculate it in one time, since now I only can get one flux for...

Thanks so much for your respond. I know I should not care the magnitude about the value. I think F4 tally is not so good in my case, 'cause usually the sum of all value should be 1, but in my case it is far more than one, there should be some where I should modify in my input, F4 gives flux...

Thanks, but if I do not care much about interactions, for example, I define a void cell, what is the difference between imp:n=1 and imp:n=0?

hello, I am new to MCNP, could somebody tell me how to use imp:n, what is imp:n=0 means, if neutron importance is 0 in one cell, why the F4 tally is 0 in this cell? how about imp:n=1 or some large number? Thanks for all.

hello, everybody, I try to use mcnp to calculate view factor of radiation heat transfer, could somebody give me some advice? I want to use surface source, but I do not know how to get the number of the particles emitted from the source from the output file? and I use F4 tally, I do not...