Request for Guidance on microscopic depletion calculation scheme implemented throught OpenMC and a nodal code

Hello everyone,

I am writing to seek your assistance with a project that involves the utilization of OpenMC and a nodal code for a series of tasks.

The tasks I aim to accomplish are as follows:

1.Lattice Calculations with OpenMC: Performing lattice calculations by using OpenMC to generate homogenized microscopic cross sections.

2.Homogenized Macroscopic Cross Sections: Microscopic cross sections will be multiplied by the respective nuclide densities to obtain homogenized macroscopic cross sections of each nuclide. Finally, these macroscopic cross sections are added together to obtain the lattice homogenized macroscopic cross section.

3.Steady-State Core Calculation: The lattice homogenized macroscopic cross sections are utilized for a steady-state core calculation using a nodal code, aiming to obtain the lattice neutron flux distribution.

4.Depletion Calculations: Furthermore, I plan to use the nuclide densities, microscopic cross sections, and lattice neutron flux to conduct depletion calculations with the purpose of updating nuclide densities. It’s important to note that I intend to keep the microscopic cross sections obtained in step 1 unchanged, step 2 to 4 will be repeated for the next timestep.

I am pleased to inform you that I have successfully completed steps 1 to 3 of the project. However, I am now facing some uncertainty regarding step 4. I am unsure about the best approach and would greatly appreciate any guidance or insights you can provide.

Your expertise in this field would be immensely valuable in helping me move forward with this project. If you have the time to provide some assistance or recommendations, I would be most grateful.

Thank you very much for considering my request. I look forward to your response and any advice you can offer.

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In this scheme, we assume that the microscopic cross sections of each nuclide do not change significantly with burnup.