OpenMC Fusion Depletion Chain Generator

Creates a reduced OpenMC depletion chain for fusion blanket materials. The generator reads the nuclides in an OpenMC materials.xml, checks that each nuclide has neutron data in the selected cross_sections.xml, and combines them with the lithium breeder reaction paths.

The generated chain includes:

• Li-6 (n,t) production of H-3
• Li-7 MT=205 (n,Xt) tritium production
• H-3 beta-minus decay to He-3
• All nuclides found in the supplied materials file and their complete decay/reaction daughter closure

Requirements

• OpenMC 0.16 or newer
• Python with the OpenMC package available
• A base OpenMC depletion-chain XML file
• An OpenMC materials.xml file
• An OpenMC cross_sections.xml file

Installation

Install the released package with:

pip install openmc-fusion-chain-generator

For development from a local checkout:

pip install -e .

OpenMC is installed separately because it is distributed through conda-forge rather than the PyPI package index used for this project:

conda install -c conda-forge openmc

After installation, the generator can be imported from any working directory:

import openmc_fusion_chain_generator

Python API

The main user-facing function is build_chain:

from pathlib import Path
from openmc_fusion_chain_generator import build_chain

build_chain(
    base_file=Path("/path/to/chain_endfb81_fast.xml"),
    materials_file=Path("/path/to/materials.xml"),
    cross_sections_file=Path("/path/to/cross_sections.xml"),
    output_file=Path("chain_fusion_breeder.xml"),
    include_all_nuclides=False,
)

To retain every nuclide from the base chain, independently of the nuclides listed in materials.xml, use:

build_chain(
    base_file=Path("chain_endfb81_fast.xml"),
    materials_file=Path("materials.xml"),
    cross_sections_file=Path("cross_sections.xml"),
    output_file=Path("chain_all_nuclides.xml"),
    include_all_nuclides=True,
)

The helper functions are also available:

from openmc_fusion_chain_generator import (
    read_material_nuclides,
    read_neutron_library,
    check_neutron_data,
)

nuclides = read_material_nuclides(Path("materials.xml"))
library = read_neutron_library(Path("cross_sections.xml"))
check_neutron_data(nuclides, Path("cross_sections.xml"))

build_chain recursively includes every decay daughter, neutron-reaction target, and tracked light-particle secondary reachable from the material nuclides. It validates neutron-data availability for every retained nuclide that has neutron reactions before writing the output. It raises RuntimeError with the missing nuclide or data-file paths when the selected library is incomplete.

The equivalent command-line option is:

openmc-fusion-chain-generator \
  --materials /path/to/materials.xml \
  --base /path/to/chain.xml \
  --cross-sections /path/to/cross_sections.xml \
  --output chain_all_nuclides.xml \
  --include-all-nuclides

Using the generated chain

MT=205 is represented by the custom (n,Xt) reaction. Import the generator module before creating the depletion operator so that this reaction is registered with OpenMC:

import sys
sys.path.insert(0, "/path/to/OpenMC-Fusion-Depletion-Chain-Generator")
import openmc_fusion_chain_generator  # registers (n,Xt)

operator = openmc.deplete.CoupledOperator(
    model,
    "chain_fusion_breeder.xml",
    normalization_mode="energy-deposition",
)

The generator does not modify the input materials, cross-section library, or base chain files.

Citation

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