# Recent content by rpp

1. ### Specifying temperatures in MCNP

How about "Specifying temperatures in MCNP"?
2. ### Specifying temperatures in MCNP

I also suggest that you change the name of the post to something that has to do with specifying temperatures in MCNP. Your question has nothing to do with supercritical water reactors.
3. ### Specifying temperatures in MCNP

In the cell cards, you specify the temperature in units of MeV by multiplying the temperature in Kelvin by the Boltzmann constant 8.617e-11 MeV/K. For example, 600K would be "tmp=5.17020E-08" However, this is not the only place you need to worry about temperatures. In the material cards...
4. ### 3D model of the VVER reactor

That is really nice, a lot of detail!
5. ### Neutron quantity normalization in an eigenvalue computation

You are correct that it is an eigenvalue problem and the eigenvalue is k-eff. If you write the equation in matrix form ($Ax=\lambda x$) you can see that the magnitude of the flux ($x$) is arbitrary. You can multiply the flux by any non-zero value, and it will still be a valid solution to the...
6. ### I want to know how I can plot Nuclear reaction rates when the equation depends on the fusion cross section of the reaction

I just realized my last post was incorrect, and I don't see a way to edit it. The cross sections are always going to be evaluated at "E", you should never have to divide by "delta-E". It is the flux that can be a function of E. Sorry for the confusion. Mentors: is there a way to edit my...
7. ### I want to know how I can plot Nuclear reaction rates when the equation depends on the fusion cross section of the reaction

Do you have a plot of the cross sections, or a table of the cross section values? You will need the table of values. If you only have the plot, check the website again to see if there is a way to download the raw data. Next, check to see if the cross sections are evaluated at "E", are are...

If you are really interested in fusion, mechanical engineering is not the right degree for you. You should consider switching to engineering physics, electrical engineering, or even physics. Prepare to go all the way with a PhD. As a freshmen, take as many math and physics courses that you...
9. ### Scaling nuclear reactors

I think it might be easier if you describe exactly what equations you are trying to scale. Are you looking at the diffusion equations? 6-factor formulas? transport equations? something else? You ask about the number of scattering collisions to slow down a neutron, but this is never used as...
10. ### Inventory of Cs-137

I just noticed that the original post was from 2011. I assume he eventually figured it out :)
11. ### MCNP model

Are you asking what a spacer grid looks like? or setting up the geometry in MCNP? What work have you done so far?
12. ### Inventory of Cs-137

You are on the right path. However, I don't know what the application is so I can't tell you if your assumptions are correct (is this a HW problem, or a real application?). Some things to think about 1. You are assuming a constant fission yield of Cs-137. This may be appropriate, but a...
13. ### MCNP summary table

This table provides the "neutron balance" for the system. The left hand is all of the events that create a neutron and the right side is the events where the neutrons are lost. The main source of neutrons is prompt fission, with smaller contributions from delayed fission, (n,xn), and...
14. ### Result is zero flux for MCNP6 *F4 tally

Then the warning is correct. You only have a vacuum, so the code is giving you a warning that there are no materials (or cross sections). The code is very general and models a lot of different physics, so there is a fairly steep learning curve to it. However, I would not say that it isn't...
15. ### Result is zero flux for MCNP6 *F4 tally

You do not have any materials defined. Each cell should have a material number as the second number, but all of your cells are void (material 0). No materials, no cross sections.