Development and Validation of OpenMC Model for Criticality Safety Analysis of an MNSR-type Research Reactor

Authors

  • A. Asuku Centre for Energy Research and Training, Ahmadu Bello University, Zaria.
  • J. Simon Department of Physics, National Open University of Nigeria, Abuja, Nigeria.
  • Y. V. Ibrahim Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria.
  • S. A. Jonah Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria.
  • B. Yahaya Department of Physics, Ahmadu Bello University, Zaria, Nigeria.
  • Y. Jibrin Department of Physics, Ahmadu Bello University, Zaria, Nigeria.
  • B. D. Jatau Department of Physics, Ahmadu Bello University, Zaria, Nigeria.
  • D. Sani Department of Physics, Ahmadu Bello University, Zaria, Nigeria.
  • I. B. Kachalla Department of Physics, Ahmadu Bello University, Zaria, Nigeria.
  • B. Bulus Department of Physics, Ahmadu Bello University, Zaria, Nigeria.
  • A. S. Umar Department of Physics, Ahmadu Bello University, Zaria, Nigeria.

DOI:

https://doi.org/10.63746/njtd.v23i2.4304

Keywords:

OpenMC, Nigeria Research Reactor-1, Miniature Neutron Source Reactor, Neutronics, Excess Reactivity, Control Rod Worth

Abstract

The accurate prediction of criticality and neutronic safety parameters is essential for the safe operation, licensing, and continued utilization of research reactors. There is limited published work assessing the performance of the open-source Monte Carlo code OpenMC for compact-core Miniature Neutron Source Reactors (MNSR) such as the Nigeria Research Reactor-1 (NIRR-1). This study aimed to develop and validate a high-fidelity OpenMC model of the NIRR-1 for predicting key criticality and neutronic safety parameters. A detailed three-dimensional OpenMC model of the NIRR-1 was developed using the ENDF/B-VII.1 nuclear data library. The reactor geometry was explicitly modelled to include the active fuel pins, dummy pins, tie rods, control rod, beryllium reflectors, irradiation channels, fission chamber, reactor vessel, moderator, and pool water. Material compositions, geometry, simulation settings, and tally structures were implemented through the OpenMC Python API. Criticality calculations employed 100,000 neutron histories per batch over 100 batches, including 20 inactive batches to ensure source convergence. The validated model was used to determine the effective multiplication factor for fully withdrawn and fully inserted control rod positions, from which the core excess reactivity, control rod worth, shutdown margin, and effective delayed neutron fraction were calculated. The calculated core excess reactivity, control rod worth, shutdown margin, and effective delayed neutron fraction were 3.78 mk, 6.82 mk, 3.04 mk, and 7.99 mk, respectively. These values agreed with commissioning measurements within a deviation of less than 5% and satisfied the neutronic safety requirements for MNSRs. The study demonstrates that OpenMC provides an accurate and reliable open-source platform for predicting the neutronic safety characteristics of the NIRR-1 and supports its application in reactor safety analysis, operational support, and future studies of MNSR-type research reactors

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Published

2026-06-30

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Vol. 23 No. 2 (2026): Vol. 23 No. 2 (2026): Volume 23, Issue 2, June 2026.

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