Thermal neutron absorption cross sections of medical isotopes

Geometry and Materials
1

Hi
I aim to compare the experimentally measured thermal neutron absorption cross sections of medical isotopes, obtained through irradiation at research reactors, with theoretical values calculated via simulations. Is it feasible to perform these simulations using the FLUKA software?
Thank you

3

Dear Amna,

thank you for your question.
I will come back to you with an answer as soon as possible.

Best,
Giuseppe

1 Like
4

Dear Amna,

in FLUKA, for low-energy (< 20 MeV) neutron, cross sections come from distributed external libraries. On the other hand, cross sections for all other kinds of particles are embedded in the code, which calculates them internally.

Therefore, if your interest is to compare measured to theoretical microscopic cross sections, using the data from the libraries might give you a better reference, since in any case they are the ones used in FLUKA.

Additionally, understand that FLUKA was not designed to distribute microscopic cross sections, rather to use them to compute macroscopic quantities. Thus, for example, it is possible to compare the measured and simulated yield of particles in the specific conditions of your experimental setup. This clearly requires knowledge of different aspects of the experiment (just to name a few: the neutron source distribution, source intensity, material irradiated, geometry) to correctly describe the FLUKA simulation but still the comparison is feasible.

Hope this helps,
Giuseppe

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5

Dear Giuseppe,

Thank you for your detailed response. I will keep this in mind.
Actually, second part of my project includes calculation of simulated yields. I will design the core geometry and will get back to you for further guidance.
Is it possible to apply the same methodology for core design and material definition used in OpenMC?

Sincere regards,

Amna

6

Dear Amna,

focusing on the geometry, for both FLUKA and OpenMC, you can build it with the constructive solid geometry (CSG) approach. Please, refer to the lessons of Geometry I and Geometry editor for additional details on how to define a geometry in FLUKA. Some differences between OpenMC and FLUKA might be present (e.g. the negative and positive half-space defined by a sphere)

Additionally, consider the possibility of the last FLUKA feature of importing Unstructured Meshes (starting from a CAD file) as geometry of your simulation. This could give the possibility to use the same CAD, if available, for both OpenMC and FLUKA calculations. Beware of the more advanced use of this feature which would require additional processing. For more information, refer to the UM-guide.

Finally for the material definition, refer once again to the FLUKA beginner course (slide).

Hope this helps you,
Giuseppe