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Input Files for "Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP"

Javier, Ruiz-Pineda; Romero-Barrientos, Jaime

Abstract

# Input Files for "Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP" **Author:** J. Ruiz-Pineda **Corresponding author:** Jaime Romero-Barrientos **E-mail:** [email protected] **ORCID:** [0000-0002-3256-8181](https://orcid.org/0000-0002-3256-8181) ## Description This dataset contains the input files used in the article _"Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP"_. The intention of sharing these files is to promote transparent research practices and facilitate reproducibility and benchmarking efforts in the nuclear simulation community. ## Case Description The included input files correspond to the "Normal Operation" environment. This configuration models a spent fuel canister placed in a borehole drilled into a tunnel, following the KBS-3 concept similar to that used by Posiva Oy. The fuel modeled corresponds to pressurized water reactor (PWR) spent nuclear fuel. Cell, material, and surface definitions were aligned across OpenMC and MCNP to the extent allowed by each code. ## Files Included - OpenMC input files: - `settings.xml` - `geometry.xml` - `materials.xml` - `NormalOperation_SpentFuel_mcnp` – MCNP input file- `isotopic_composition_v1.0.0.pdf` - List of isotopes for the Spent and Fresh fuel used in the simulations. ## Simulation Parameters - 50,000 particles per batch - 3,000 total batches - 300 inactive batches ## Results - **OpenMC**: *k_eff* = 0.19556 ± 0.00002 - **MCNP**: *k_eff* = 0.19588 ± 0.00002 ## Code Versions - OpenMC (TD) version 2.1 - MCNP version 6.1 ## Nuclear Data Library - ENDF/B-VIII.0 ## License This dataset is distributed under the [Creative Commons Attribution 4.0 International (CC BY 4.0)](https://creativecommons.org/licenses/by/4.0/) license. ## Embargo Notice This dataset is currently under embargo and will be made publicly available upon acceptance of the associated article. In the meantime, it can be accessed upon reasonable request. Please contact the corresponding author: - Jaime Romero-Barrientos - Email: [email protected] For any questions or further information, please contact the corresponding author.

Full text

Isotopic Lists for Spent and Fresh Nuclear Fuel Supplementary Dataset for: “Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP” J. Ruiz-Pineda et al. September 12, 2025 Description This document provides the isotopic composition used in the simulations associated with the manuscript by J. Ruiz-Pineda et al.,“Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP”, submitted to Nuclear Engineering and Technology (2025). The content is intended for archival on Zenodo as a self-contained reference for reproducibility. Metadata •Context: KBS-3–type systems; comparison between OpenMC and MCNP. •Spent fuel (PWR): initial enrichment 4.5% U-235; burnup 50 GWd/tHM. •Total number density (spent fuel): 0.07133315757 atoms/barn/cm. •Fresh fuel (UO2, 4.5 at.% U-235): density 10.97 g/cm3(if included). Spent fuel isotopic vector Number densities are reported in atoms/barn/cm. Nuclide Number density [atoms/barn/cm] O16 4.792730E-02 U238 2.194897E-02 U235 2.495224E-04 Pu239 1.834560E-04 U236 1.504535E-04 Cs137 7.497633E-05 Pu240 7.286789E-05 Cs133 6.988027E-05 Tc99 6.673897E-05 Ru101 6.562089E-05 Continued on next page 1 Nuclide Number density [atoms/barn/cm] Zr93 6.364271E-05 Mo95 6.080845E-05 Sr90 4.849071E-05 Pu241 4.780428E-05 Nd143 4.657118E-05 Nd145 3.805241E-05 Rh103 3.470511E-05 Cs135 3.462911E-05 Np237 1.989078E-05 Pu242 1.900670E-05 Pd107 1.850466E-05 Sm150 1.625546E-05 I129 1.025192E-05 Pu238 9.345469E-06 Pm147 7.325843E-06 Eu153 6.625394E-06 Sm152 6.322567E-06 Ag109 6.073445E-06 Am243 5.609559E-06 Sm147 4.709022E-06 U234 4.573309E-06 Cm244 2.346898E-06 Am241 2.231002E-06 Ru103 1.381040E-06 Sn126 1.285315E-06 Nb95 1.057503E-06 Cl36 1.000090E-06 Ca41 1.000089E-06 Ni59 1.000089E-06 Sm151 9.711471E-07 Cm242 5.252722E-07 Se79 5.058653E-07 Eu155 3.404694E-07 Pr143 1.466691E-07 Cm245 1.395316E-07 Sm149 1.380724E-07 Nd147 3.658855E-08 Cm243 2.199518E-08 Cm246 1.389708E-08 U237 1.217488E-08 Xe133 6.668318E-09 Gd155 5.462690E-09 Continued on next page 2 Nuclide Number density [atoms/barn/cm] I133 5.388112E-09 Eu152 3.228557E-09 Pu244 6.747405E-10 Np239 2.343782E-10 U233 8.708978E-11 Mo99 8.183378E-11 Fresh fuel isotopic vector (UO2, 4.5 at.% U-235) Number densities are reported in atoms/barn/cm. Bulk density of 10.97 g/cm3. Nuclide Number density [atoms/barn/cm] O17 1.855270E-05 O16 4.883524E-02 O18 9.792301E-05 U238 2.334593E-02 U235 1.114852E-03 U234 9.964707E-06 U236 5.106559E-06 Reproducibility & Versioning •Dataset version: 1.0.0. Citation & License If you use this dataset, please cite: •Dataset: “Isotopic Lists for Spent and Fresh Nuclear Fuel (Supplementary)” — Zenodo, DOI: to be assigned. •Article: J. Ruiz-Pineda et al., “Extending OpenMC Validation to Spent Fuel Canisters: A Criticality Benchmark Against MCNP,” submitted to Nuclear Engineering and Technology (2025). This dataset is released under the CC BY 4.0 license. 3