WJNST  Vol.4 No.2 , April 2014
Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor
ABSTRACT

The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) heat flux and DNB ratio, axial fuel centreline temperature and compared. The comparison indicates that the calculated values are in satisfactory agreement between the codes. The data obtained in this investigation are largely far to compromise safety of the reactor. The results can also be used to upgrade the current core configuration of the TRIGA reactor.


Cite this paper
Rahman, M. , Akond, M. , Basher, M. and Huda, M. (2014) Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor. World Journal of Nuclear Science and Technology, 4, 81-87. doi: 10.4236/wjnst.2014.42013.
References
[1]   Hossain, S.M., Zulquarnain, M.A., Kamal, I. and Islam, M.N. (2011) Current Status and Perspectives of Nuclear Reactor Based Research in Bangladesh. IAEA, Vienna, 7-14.

[2]   Huda, M.Q., Bhuiyan, S.I., Chakroborty, T.K., Sarker, M.M. and Mondal, M.A.W. (2001) Thermal Hydraulic Analysis of the 3 MW TRIGA MARK-II Research Reactor. Nuclear Technology, 135, 51-66.

[3]   FSAR (2006) Final Safety Analysis Report for 3 MW TRIGA MARK-II Research Reactor at AERE, Savar, Dhaka, Bangladesh. BAEC, Dhaka.

[4]   Rahman, M.M., Hussain, M.S., Imtiaz, M.A. and Huda, M.Q. (2012) Computational Analysis of Thermo-Hydraulic Behaviour of TRIGA Research Reactor. Annals of Nuclear Energy, 49, 48-56.
http://dx.doi.org/10.1016/j.anucene.2012.05.034

[5]   Kaminaga, M. (1994) COOLOD-N2: A Computer Code, for the Analyses of Steady-State Thermal Hydraulics in Research Reactors, JAERI Report. Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki.

[6]   Sudo, Y., Ikawa, H. and Kaminaga, M. (1985) Development of Heat Transfer Package for Core Thermal Hydraulics Design and Analysis of Upgradded JRR-3. Proceedings of the International Meeting of Reduced Enrichment for Research and Test Reactors, Petten, 14-16 October 1985, 363-372.

[7]   General Atomics (1979) 10 MW TRIGA-LEU Fuel and Reactor Design Description. General Atomics, San Diego.

[8]   Obenchain, C.F. (1969) PARET—A Program for the Analysis of Reactor Transients. IDO-17282, Idaho Atomic Energy Division, USAEC, Idaho.

[9]   Boulaich, Y., Nacir, B., El Bardouni, T., Zoubair, M., El Bakkari, B., Merroun, O., El Younoussi, C., Htel, A., Boukhal, H. and Chakir, E. (2011) Steady-State Thermal Hydraulic Analysis of the Moroccan TRIGA Mark-II Reactor. Nuclear Engineering and Design, 241, 270-273. http://dx.doi.org/10.1016/j.nucengdes.2010.10.033

 
 
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