Fixed source depletion

hrash47 · August 20, 2025, 2:08am

Hello everyone,

I’ve recently started studying OpenMC depletion and have been going through the available examples. Most of what I’ve found so far focuses on depletion simulations in eigenvalue mode, but I haven’t come across any examples for fixed-source mode.

Could anyone share a simple example or point me toward resources that demonstrate depletion in fixed-source mode? It would really help me get a better understanding of how to set it up.

Thank you in advance!

wahidluthfi · August 20, 2025, 9:18am

Hi Hrash47, welcome to the openmc community. I think some tasks from neutronic workshop can help you with fixed source depletion, such as this transmutation isotope build up

github.com/fusion-energy/neutronics-workshop

tasks/task_10_activation_transmutation_depletion/1_example_transmutation_isotope_build_up.ipynb

main

{
 "cells": [
  {
   "attachments": {},
   "cell_type": "markdown",
   "id": "cc29c481-66a0-4cb1-9bbd-2062e84ac1bb",
   "metadata": {},
   "source": [
    "# Nuclide Build Up During Depletion Simulation\n",
    "\n",
    "This example simulates the build up of activate products within a material under neutron irradiation. The subsequent decay of unstable isotopes is also simulated."
   ]
  },
  {
   "attachments": {},
   "cell_type": "markdown",
   "id": "d4ccbdf3-2acc-45e8-8ece-62e5efb08dd0",
   "metadata": {},
   "source": [
    "This first cell imports the packages needed, note the extra import openmc.deplete import"

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Or the shutdown dose rate case scenario

github.com/fusion-energy/neutronics-workshop

tasks/task_11_CSG_shut_down_dose_tallies/1_R2S_cell_shut_down_dose_rate.py

main

# This script simulates R2S method of shut down dose rate
# on a simple sphere model.

import numpy as np
import openmc
import openmc.deplete
from pathlib import Path
import math
from matplotlib.colors import LogNorm

# users might want to change these to use specific xml files to use particular decay data or transport cross sections
# the chain file was downloaded with
# pip install openmc_data
# download_endf_chain -r b8.0
# openmc.config['chain_file'] = Path.home() / 'nuclear_data' / 'chain-endf-b8.0.xml'
# openmc.config['cross_sections'] = Path.home() / 'nuclear_data' / 'cross_sections.xml'

# a few user settings
# Set up the folders to save all the data in
n_particles = 1_00000

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Fixed source depletion

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