openmc-feedbacks

An importable wrapper around OpenMC depletion. The package accepts a prepared OpenMC Model or a directory containing OpenMC XML inputs and runs depletion with one function. Provide the reactor geometry, materials, temperatures, volumes, cross-section data, and depletion chain for your calculation.

Install

Install OpenMC and its nuclear data according to the OpenMC documentation, then install this package:

pip install .

Install the released package with:

pip install openmc-feedbacks

Documentation

The full user guide can be found here: openmc-feedbacks documentation

Ordinary depletion

import openmc
from openmc_feedbacks import run_depletion

run_depletion(
    model="model",  # directory containing geometry.xml, materials.xml, settings.xml, etc.
    timesteps=[1, 7, 30, 90],
    timestep_units="d",
    power=160e6,  # watts
    chain_file="data/chain.xml",
    output_dir="depletion-results",
)

Soluble-species feedback

Provide a named material containing the soluble species at the stated reference concentration. The controller scales the listed nuclide atom densities together, leaving every other nuclide in the material untouched. Species concentration units are intentionally user-defined, so the same controller can represent ppm boron, ppm gadolinium, or another soluble absorber when the reference material has been prepared consistently.

from openmc_feedbacks import FeedbackControl, run_depletion

control = FeedbackControl(
    material_name="Primary coolant",
    nuclides=("B10", "B11"),
    reference_concentration=1000,  # ppm represented by the model densities
    minimum=0,
    maximum=1000,
    target_keff=1.005,
    search_particles=5000,
    search_batches=30,
    search_inactive=5,
)
run_depletion(
    model="model",  # directory containing the OpenMC XML inputs
    timesteps=[1, 7, 30],
    power=160e6,
    chain_file="data/chain.xml",
    control=control,
    output_dir="controlled-results",
)

Citation

Citation metadata is provided in CITATION.cff.