Research Article

Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code

Volume: 10 Number: 1 May 11, 2026
EN TR

Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code

Abstract

A thermal-hydraulic analysis of the TRIGA Mark II Research Reactor operating at 500 kW under natural convection conditions was conducted using the NCTRIGA code utilizing SRAC2006 neutronics data, and results were benchmarked against MCNP4C calculations. The study focused on evaluating key thermal-hydraulic parameters along the axial length of the hottest fuel rod (C4) to ensure the reactor's safety. The Reynolds number exhibited a consistent increase with axial height. Conversely, heat flux, heat transfer coefficient, and fuel centerline temperature demonstrated a similar trend: increasing from the top of the core, peaking near the midpoint, and subsequently decreasing. Notably, fuel centerline temperatures remained significantly below established safety limits. Fuel surface temperatures remained relatively constant, while coolant temperature demonstrated a slow, incremental increase along the axial length from the top. While minor discrepancies were observed between the SRAC2006 and MCNP4C datasets, the peak values and their locations remained consistent across both.

Keywords

References

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Details

Primary Language

English

Subjects

Nuclear Technology

Journal Section

Research Article

Authors

M. J. H. Khan
0000-0003-4856-4350
Bangladesh

Publication Date

May 11, 2026

Submission Date

January 18, 2026

Acceptance Date

March 29, 2026

Published in Issue

Year 2026 Volume: 10 Number: 1

APA
Hossen, M. A., & Khan, M. J. H. (2026). Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code. Journal of Nuclear Sciences, 10(1). https://doi.org/10.59474/nuclear.2023.67
AMA
1.Hossen MA, Khan MJH. Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code. Journal of Nuclear Sciences. 2026;10(1). doi:10.59474/nuclear.2023.67
Chicago
Hossen, Md Altaf, and M. J. H. Khan. 2026. “Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow Using NCTRIGA Computer Code”. Journal of Nuclear Sciences 10 (1). https://doi.org/10.59474/nuclear.2023.67.
EndNote
Hossen MA, Khan MJH (May 1, 2026) Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code. Journal of Nuclear Sciences 10 1
IEEE
[1]M. A. Hossen and M. J. H. Khan, “Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code”, Journal of Nuclear Sciences, vol. 10, no. 1, May 2026, doi: 10.59474/nuclear.2023.67.
ISNAD
Hossen, Md Altaf - Khan, M. J. H. “Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow Using NCTRIGA Computer Code”. Journal of Nuclear Sciences 10/1 (May 1, 2026). https://doi.org/10.59474/nuclear.2023.67.
JAMA
1.Hossen MA, Khan MJH. Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code. Journal of Nuclear Sciences. 2026;10. doi:10.59474/nuclear.2023.67.
MLA
Hossen, Md Altaf, and M. J. H. Khan. “Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow Using NCTRIGA Computer Code”. Journal of Nuclear Sciences, vol. 10, no. 1, May 2026, doi:10.59474/nuclear.2023.67.
Vancouver
1.Md Altaf Hossen, M. J. H. Khan. Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code. Journal of Nuclear Sciences. 2026 May 1;10(1). doi:10.59474/nuclear.2023.67