Araştırma Makalesi

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

Cilt: 10 Sayı: 1 11 Mayıs 2026
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Study of Thermal Hydraulics Parameters of TRIGA Research Reactor under Natural Convection Mode of Coolant Flow using NCTRIGA Computer Code

Öz

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.

Anahtar Kelimeler

Kaynakça

  1. [1] S. M. Hossain et al., Current Status and Perspectives of Nuclear Reactor Based Research in Bangladesh, IAEA-TECDOC-1659, Vienna, 2011.
  2. [2] R. S. Smith, Comparison of NCTRIGA Results to GA Data, NCTRIGA Input Format and Revision of NATCON, Argonne National Laboratory, 1992. [3] M. Q. Huda and S. I. Bhuiyan, Thermal-hydraulic analysis of TRIGA Mark-II research reactor under natural convection cooling mode, Annals of Nuclear Energy, 33, 1258–1268, 2006.
  3. [4] M. M. Rahman, M. Q. Huda, and S. I. Bhuiyan, Thermal-hydraulic analysis of TRIGA research reactor using NCTRIGA code under steady-state natural convection cooling, Annals of Nuclear Energy, 45, 138–146, 2012.
  4. [5] M. M. R. Sarker, M. Q. Huda, and S. I. Bhuiyan, Neutronic analysis of TRIGA Mark II research reactor using SRAC code system, Annals of Nuclear Energy, 35, 2222–2228, 2008.
  5. [6] M. Q. Huda et al., Modelling of the TRIGA Mark II Research Reactor using the Monte Carlo Technique, INST-57/RPED-13, AERE, Bangladesh, 1997.
  6. [7] J. F. Briesmeister, MCNP — The Monte Carlo N-Particle Transport Code System, Version 4C, LA-12625-M, 2000.
  7. [8] K. Okumura, SRAC — A Comprehensive Neutronics Calculation Code System, Japan Atomic Energy Agency, 2007.
  8. [9] B. T. Rearden, Engineering Analysis of a Power Upgrade for the Texas A&M Nuclear Science Centre Reactor, Texas A&M University, 1995.

Ayrıntılar

Birincil Dil

İngilizce

Konular

Nükleer Teknoloji

Bölüm

Araştırma Makalesi

Yazarlar

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

Yayımlanma Tarihi

11 Mayıs 2026

Gönderilme Tarihi

18 Ocak 2026

Kabul Tarihi

29 Mart 2026

Yayımlandığı Sayı

Yıl 2026 Cilt: 10 Sayı: 1

Kaynak Göster

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, ve 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 (01 Mayıs 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 ve 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, c. 10, sy 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 (01 Mayıs 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, ve 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, c. 10, sy 1, Mayıs 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. 01 Mayıs 2026;10(1). doi:10.59474/nuclear.2023.67