Table of Contents Author Guidelines Submit a Manuscript
Science and Technology of Nuclear Installations
Volume 2017, Article ID 2615409, 10 pages
Research Article

Comparative Analysis of the Dalat Nuclear Research Reactor with HEU Fuel Using SRAC and MCNP5

1Faculty of Applied Sciences, Ton Duc Thang University, Ho Chi Minh City 700000, Vietnam
2Institute of Fundamental and Applied Sciences, Duy Tan University, 3 Quang Trung, Da Nang 550000, Vietnam
3Nuclear Research Institute, VINATOM, 01 Nguyen Tu Luc, Dalat, Lamdong 670000, Vietnam
4Institute for Nuclear Science and Technology, VINATOM, 179 Hoang Quoc Viet, Hanoi 100000, Vietnam
5Western University Hanoi, Yen Nghia Ward, Ha Dong District, Hanoi 100000, Vietnam

Correspondence should be addressed to Giang Phan; nv.ude.tdt@gnaignahp and Hoai-Nam Tran; nv.ude.utd@4maniaohnart

Received 7 June 2017; Revised 25 August 2017; Accepted 24 September 2017; Published 27 December 2017

Academic Editor: Arkady Serikov

Copyright © 2017 Giang Phan et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Linked References

  1. TRIGA History, URL:
  2. N. D. Nguyen, Ed., Safety Analysis Report for the Dalat Nuclear Research Reactor, Nuclear Research Institute, Vietnam Atomic Energy Commission, 2009.
  3. N. D. Nguyen, B. V. Luong, V. V. Le et al., “Results of Operation and Utilization of the Dalat Nuclear Research Reactor,” Nuclear Science and Technology, vol. 4, no. 1, pp. 1–9, 2014. View at Google Scholar
  4. M. Türkmen and Ü. Çolak, “Analysis of ITU TRIGA Mark II research reactor using Monte Carlo method,” Progress in Nuclear Energy, vol. 77, pp. 152–159, 2014. View at Publisher · View at Google Scholar · View at Scopus
  5. B. Q. Do and L. P. Nguyen, “Application of a genetic algorithm to the fuel reload optimization for a research reactor,” Applied Mathematics and Computation, vol. 187, no. 2, pp. 977–988, 2007. View at Publisher · View at Google Scholar · View at MathSciNet · View at Scopus
  6. K. Okumura, T. Kugo, K. Kaneko, and K. Tsuchihashi, SRAC2006: A Comprehensive Neutronics Calculation Code System, JAEA-Data/Code 2007-004, 2007.
  7. D. H. Pham, Q. H. Ngo, H. L. Vu, and K. M. Tran, Report startup of nuclear research reactor: Part 2 - Physics startup for core configuration with a neutron trap, Nuclear Research Institute, 1984.
  8. M. B. Chadwick, P. Obložinský, M. Herman et al., “ENDF/B-VII.0: next generation evaluated nuclear data library for nuclear science and technology. Nucl. Data Sheets,” Special Issue on Evaluated Nuclear Data File ENDF/B-VII.0, 2006. View at Google Scholar
  9. X-5 Monte Carlo Team, 2005. MCNP - A General Monte Carlo N-Particle Transport Code. Version 5. LA-UR-03-1987, April 24, 2003.