Mentioning Materials in material.xml

User Support
1

Hello Everyone,

Namaskar.!!

I have four questions:

  1. Is it necessary to mention all fission product in fuel material element of materials.xml?
  2. from where I can get knowledge of fission product build in fuel?
  3. Can I calculate burn up using openmc?
  4. Can I calculate power distribution ?

Thank you for the help.

Regards,
Darpan

2

Hi Darpan,

To answer your questions with a general response with regards to depletion, it is in the works and will hopefully be released soon. All relevant features are currently under development and so are therefore unavailable. Unfortunately, I think this means the answers are 1) no, 2) DNE, 3) no, 4) no. If I am wrong, hopefully another developer with different knowledge can correct me.

Best,
Derek

3

Thank you Very much Mr. Darek, I get the answers. I do not know what do you mean by DNE. Thank you once again.

4

Does not exist.

Derek

5

I also want to know about “is there any development for more lattice type like hexagonal, circular etc?” Can I expect for it?

6

There is ongoing development for hexagonal lattices. They will hopefully be included in release 0.7

As for your earlier questions: I believe you can use a fission tally with a mesh to calculate the power distribution in the reactor. Perhaps you can manually predict burnup from that information.

7

While OpenMC currently does not perform gamma transport and, thus, cannot provide an energy deposition tally of the type you may want for a power distribution calculation, fission rate and kappa-fission tallies can be used to obtain the fission rate distribution and energy production rate distribution, respectively.

If you want to run a simulation of a depleted core in OpenMC, your best bet, at this point, is to obtain the isotopic composition of the burned fuel from another code that performs depletion calculations and use those nuclide density values in materials.xml.

Jon

8

Thank you Mr. Harper. Write now I am modelling Candu geometry with manually calculated positions of fuel rod.

9

Thank you Mr. Walsh, I get your answer.
Due to there is not Gamma Transport in OpenMC, we can not perform Shielding Efffects . Write?

Thank you once again for your answer.

10

In the context of a reactor physics simulation to obtain a power distribution, the inability to simulate gamma transport means that one cannot accurately calculate the gamma heating of the fuel and, thus, the true energy deposition/power distribution. You can use the kappa-fission tally, though, to record the energy production distribution.

Jon