Hello everyone
I am trying to obtain the integral neutron flux based on the results obtained with MCNP (neutron spectrum calculation) for each energy bin (51 neutron energy bins). I have seen in many papers the calculation of the differential neutron flux multiplying the neutron flux results of...
Hello, I have seen also in some papers Integral neutron flux. I am trying to obtain it just as unfolding codes as SANDP or STAYS´L do. If somebody knows how to do that calculations I will appreciate it. The behavior in the integral neutron flux is also different from those shown previously...
Hello
I am a lower-intermediate user of MCNP and I do not know how to obtain the diffusion coefficient (or maybe the angle of scattering) using tallies. I also have read a paper (Multigroup Scattering Matrix Generation Method using Weight-to-Flux Ratio Based on a Continous Energy Monte Carlo...
No, I don´t use it. In a TRIGA reactor core simulation I used the same information to move the mesh (I only used "origin" to place the mesh in other location) and it result. In this case I don´t know what's is happening.
Hi everyone:
Im trying to use the FMESH card of MCNP5 to obtain axial flux. I simulated an hexagonal reactor core having different enrichments of uranium (see uploaded file). When I run the program, I obtain a good value of keff (1.03941) but the meshtal file have all the values equal to zero...