Hello everyone, I would like to ask a question. I built a model about the plate-type and found that the kinf of OpenMC and MCNP is different. I have checked many times but have not found any problems. Can you help me see the reason? My input card is shown in the figure.
Hi Lishuai, welcome to the openmc forum.
I think it is a good practice if you declare the density (also its unit) in the material description.
Regarding the difference, I think you are modeling a different problem since your MCNP input shows that you are using CZ (cylinder Z axis) while on your openmc input, you are using only planes. I am also seeing that you are using an array of 11 by 11 pins in your MCNP input while on your openmc it is only 1 by 11, so I think you are comparing a different model.
Also, regarding your openmc input, I think you forget to fill your clad_cell1 with your cladding material.
Finally, since the cells logic and universe logic used in openmc is quite similar to mcnp, then I think it would be okay if you use some surfaces multiple times. I see that you are declaring a z-plane with identical dimensions multiple times and using it on different cells. You could also use a simpler cell region definition as part of the universe which could be used on a different cell region later, i.e. while defining fuel meat and its cladding, you didn’t have to specify the z-axis region at first, since after you make it as a universe, you could fill it into a region which have a specific dimension as you could see in the plot geometry.
Hope it will help you to localize the problem.
Hi wahidluthfi,Thank you for your answer. I have identified my issue and will make corrections to see if the results are consistent.
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