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2014 | 59 | 4 | 129-136
Tytuł artykułu

Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Treść / Zawartość
Warianty tytułu
Języki publikacji
EN
Abstrakty
EN
Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th) O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
Słowa kluczowe
Wydawca

Czasopismo
Rocznik
Tom
59
Numer
4
Strony
129-136
Opis fizyczny
Daty
wydano
2014-12-01
otrzymano
2013-08-27
zaakceptowano
2014-08-08
online
2014-12-30
Twórcy
  • Faculty of Engineering, Univesidad de Talca, 2 Norte 685 Talca, Chile, Tel.: +56 071 201 702, Cadmium_109@yahoo.com
  • Department of Radiation Application, Shahid Beheshti University, G. C., Tehran, Iran
  • Nuclear Science & Technology Research Institute, Atomic Energy Organization of Iran (AEOI), G. C., Tehran, Iran
  • Nuclear Science & Technology Research Institute, Atomic Energy Organization of Iran (AEOI), G. C., Tehran, Iran
  • Faculty of Engineering, Univesidad de Talca, 2 Norte 685 Talca, Chile and Department of Energy Science, Sungkyunkwan University, 300 Cheoncheon-dong, Suwon, Korea
  • Department of Physics, Firoozkooh Branch, Islamic Azad University, Firoozkooh, Iran
Bibliografia
  • 1. IAEA. (2005). Thorium fuel cycle - potential benefi ts and challenges. Vienna: International Atomic Energy Agency. (IAEA-TECDOC-1450).
  • 2. Weaver, K. D., & Herring, J. S. (2002). Performance of thorium-based mixed oxide fuels for the consumption of plutonium in current and advanced reactors. In International Congress on Advanced Nuclear Power Plants (ICAPP). ANS Annual Meeting, 9-13 June 2002, Hollywood, Florida, USA.
  • 3. Lung, M., & Gremm, O. (1998). Perspectives of the thorium fuel cycle. Nucl. Eng. Des., 180, 133-146.
  • 4. Usha, S., Ramanarayanan, R. R., Mohanakrishnan, P., & Kapoor, R. P. (2006). Research reactor KAMINI. Nucl. Eng. Des., 236, 872-880.
  • 5. Kumar, A., Srivenkatesan, R., & Sinha, R. K. (2009). On the physics design of advanced heavy water reactor (AHWR). In International Conference on Opportunities and Challengers for Water Cooled Reactors in the 21st Century, 27-28 October 2009 (pp. 84-85). Vienna: International Atomic Energy Agency. (IAEA-CN-164).
  • 6. Maitra, R. (2005) Thorium: Preferred nuclear fuel of the fuel. Sci. Technol., 18, 64-71.
  • 7. Sasidharan, K., & Chafale, S. B. (2012). New reactor concepts. BARC Highlights - Reactor Technology and Engineering, from http://barc.gov.in/publications/eb/golden/reactor/toc/chapter9/9.pdf.
  • 8. Pelowitz, D. B. (2008). MCNPX2.6.0 user manual. Los Alamos: Los Alamos National Laboratory (LA- -CP-07-1473).
  • 9. Thorium high temperature reactor (THTR), from paksnuclearpowerplant.com.
  • 10. Fensin, M. L. (2008). Development of the MCNPX depletion capability: A Monte Carlo depletion method that automates the coupling between MCNPX and CINDER90 for high fi delity burn-up calculations. Doctoral dissertation, University of Florida.
  • 11. Persson, C. -M. (2005). Reactivity determination and Monte Carlo simulation of the subcritical reactor experiment - “Yalina”. Master of Science Thesis, Department of Nuclear and Reactor, Physics Royal Institute of Technology, Stockholm, from http://neutron. kth.se/publications/library/CalleMSc.pdf.
  • 12. Hassanzadeh, M., Feghhi, S. A. H., & Khalafi , H. (2013). Calculation of kinetic parameters in an accelerator driven subcritical TRIGA reactor using MCNIC method. Ann. Nucl. Energy, 59, 188-193.[WoS]
  • 13. Westlen, D. (2007). Why faster is better - on minor actinide transmutation in hard neutron spectra. Doctoral dissertation, Division of Reactor Physics, University of Stockholm, from http://neutron.kth.se/publications/PhDtheses.shtml
  • 14. Snoj, L., & Ravnik, M. (2006). Calculation of power density with MCNP in TRIGA reactor. In Proceedings of the International Conference on Nuclear Energy for New Europe, 12-15 September 2006 (Paper no. 109, pp. 1-6). Portoroz, Slovenia.
  • 15. Shultis, J. K., & Faw, R. E. (2011) An MCNP primer. Department of Mechanical and Nuclear Engineering, Kansas State University, from http://krex.ksu.edu.
  • 16. El Bakkari, B., El Bardouni, T., Merroun, O., El Younoussi, Ch., Boulaich, Y., & Chakir, E. (2009). Development of an MCNP-tally based burn-up code and validation through PWR benchmark exercises. Ann. Nucl. Energy, 36, 626-633.[WoS]
  • 17. Marin, T. W., Takahashi, K., & Bartels, D. M. (2006). Temperature and density dependence light and heavy water ultraviolet absorption edge. Chem. Phys., 125, 1-11.
  • 18. Kazimi, M. S., Czerwinski, K. R., Driscoll, M. J., Hejzlar, P., & Meyer, J. E. (1999). On the use of thorium in light water reactors. Department of Nuclear Engineering, Massachusetts Institute of Technology. (MIT-NFC-TR-016), from http://www.ltbridge.com/ assets/15.pdf.
Typ dokumentu
Bibliografia
Identyfikatory
Identyfikator YADDA
bwmeta1.element.-psjd-doi-10_2478_nuka-2014-0017
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