3D SN and Monte Carlo calculations of the Utah TRIGA reactor core using PENTRAN and MCNP6

• Published in: Annals of nuclear energy Vol. 155; p. 108158
• Main Authors: Wang, Meng-Jen, Sjoden, Glenn E., Foley, Amanda, Mohanty, Swomitra K.
• Format: Journal Article
• Language: English
• Published: Elsevier Ltd 01.06.2021
• Subjects: 3-D Core flux distribution; Criticality; Eigenvalue; MCNP6; PENTRAN; UUTR; Eigenvalue; UUTR; MCNP6; 3-D Core flux distribution; PENTRAN; Criticality
• ISSN: 0306-4549; 1873-2100
• DOI: 10.1016/j.anucene.2021.108158

•3D SN Method TRIGA Reactor Core Calculation.•Criticality Problem Energy Group Collapsing and Geometry Homogenization.•Computational Verification of TRIGA Core Calculation.•Experimental Validation of 3D SN Method TRIGA Reactor Core Calculation.•High Performance Parallel Computation. We present a systematic and detailed approach to simulate the University of Utah TRIGA Reactor (UTR) in support of core criticality calculations to profile the entire reactor in detail. In performing this work, we utilized both a 3-D Cartesian SN deterministic code, PENTRAN, and a Monte Carlo code, MCNP6, to calculate complimentary, high accuracy 3-D transport derived neutron flux distributions at reactor full power. For deterministic models, our study shows that a 14-group CONTRIBUTON weighted multi-group cross-section library with up-scattering compared extremely well with continuous energy Monte Carlo results. We also modeled results to confirm the activity of an activated arsenic sample placed in the heavy water moderated thermal irradiation chamber of the UUTR (University of Utah TRIGA Reactor). A 138 pcm difference in the system eigenvalue for the full 3-D core PENTRAN discrete ordinates (SN) UUTR model utilizing 2×109 equations, in 14 energy groups, is compared to the reference case of a similar full core MCNP6 Monte Carlo calculation with continuous energy cross-sections. On average, deterministic and Monte Carlo core neutron group flux values differ by less than 1%, with some local maximum relative differences between 10% and 20% for the 3-D core flux distribution in some energy groups are observed. The differences of neutron fluxes from group 3/14 to group 10/14, with energies spanning from 1.11 eV to 13.8 MeV, are comparable to within 5% in the UUTR core active fuel region. For energy groups with energies greater than 13.8 MeV, MCNP6 is not able to resolve the flux distributions with reasonable statistical uncertainties due to the extremely low sampling probability for neutrons in those energies; orders of magnitude more histories would be required to do so. In contrast, the PENTRAN 3D SN calculation shows a very detailed flux distribution for energies larger than 13.8 MeV, highlighting the complimentary utility of applying both methods. Detailed approaches for the PENTRAN calculation, including energy group collapsing, homogenization of fuel + gap + cladding, and analysis are presented in the narrative.