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Czasopismo
2015 | 60 | 3 | 571-580
Tytuł artykułu

Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

Treść / Zawartość
Warianty tytułu
Języki publikacji
EN
Abstrakty
EN
In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB) in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238) and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242). The main results were presented as a calculated-to-experimental ratio (C/E) for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55). The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.
Wydawca

Czasopismo
Rocznik
Tom
60
Numer
3
Strony
571-580
Opis fizyczny
Daty
wydano
2015-09-01
otrzymano
2014-09-24
zaakceptowano
2015-05-20
online
2015-09-25
Twórcy
  • Department of Nuclear Energy, Faculty of Energy and Fuels, AGH University of Science and Technology, 30 Mickiewicza Ave., 30-059 Krakow, Poland, Tel.: +48 12 617 5186, Fax: +48 12 617 4547, moettin@agh.edu.pl
autor
  • Department of Nuclear Energy, Faculty of Energy and Fuels, AGH University of Science and Technology, 30 Mickiewicza Ave., 30-059 Krakow, Poland, Tel.: +48 12 617 5186, Fax: +48 12 617 4547
Bibliografia
  • 1. Cetnar, J., Gudowski, W., & Wallenius, J. (1999). MCB: A continuous energy Monte Carlo burn-up simulation code. In Proceedings of Actinide and Fission Product Partitioning and Transmutation. (EUR 18898 EN, OECD/NEA 523).
  • 2. Suyama, K., Murazaki, M., Ohkubo, K., Nakahara, Y., & Uchiyama, G. (2011). Re-evaluation of assay data of spent nuclear fuel obtained at Japan Atomic Energy Research Institute for validation of burnup calculation code systems. Ann. Nucl. Energy, 38, 930-941. DOI: 10.1016/j.anucene.2011.01.025.[Crossref][WoS]
  • 3. http://www.kepco.co.jp
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  • 5. http://www.jaea.go.jp
  • 6. Adachi, T., Nakahara, Y., Kohno, N., Gunji, K., Suzuki, T., Sonobe, T., Onuki, M., Kato, K., & Tachikawa, E. (1994). Comparison of calculated values with measured values on the amount of TRU and FP nuclide accumulated in gadolinium bearing PWR spent fuels, J. Nucl. Sci. Technol., 31(10), 1119-1129.
  • 7. American Society for Testing and Materials. (2012). Standard Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Neodymium-148 Method). U.S.A. (ASTM E321-96).
  • 8. X-5 Monte Carlo Team. (2003). MCNP-A General Monte Carlo N-Particle Transport Code, Version 5. Los Alamos National Laboratory. (LA-UR-03-1987).
  • 9. Cetnar, J. (2006). General solution of Bateman equations for nuclear transmutations. Ann. Nucl. Energy, 33, 640-645. DOI: 10.1016/j.anucene.2006.02.004.[Crossref]
  • 10. Koning, A., Forrest, R., Kellett, M., Mills, R., Henriksson, H., & Rugama, Y. (2006). The JEFF-3.1 Nuclear Data Library. OECD Nuclear Energy Agency. (OECD JEFF Report 21).
  • 11. Firestone, R., Shirley, V., Baglin, C., Chu, S., & Zipkin, J. (1996). Table of Isotopes 8E. New York: John Wiley & Sons, Inc.
  • 12. IAEA. (1996). The Basic Safety Standards. Vienna: International Atomic Energy Agency. (Safety Series No. 115).
  • 13. Croff, A. (1980). A User’s Manual for the ORIGEN2 Computer Code. Oak Ridge National Laboratory. (ORNL/TM 7157).
  • 14. Mann, F., & Schenter, R. (1977). Calculated neutron capture cross sections to the americium ground and isomer states. Nucl. Sci. Eng., 63, 242-249.
  • 15. Wahl, A. (1985). Nuclear-charge distribution near symmetry for thermal-neutron-induced fission of 238U. Phys. Rev. C, 32, 184-195.[Crossref]
  • 16. Canadian Nuclear Safety Commission. (2003). Reactor physics. CNSC Science and Reactor Fundamentals - Reactor Physics Technical Training Group.
  • 17. Lewins, J., & Becker, M. (2002). Advances in nuclear science and technology (Vol. 25). New York: Kluwer Academic Publishers.
  • 18. OECD Nuclear Energy Agency. (2011). Spent nuclear fuel assay data for isotopic validation. (Nuclear Science NEA/NSC/WPNCS/DOC(2011)5).
  • 19. Harężlak, D., Kasztelnik, M., & Pawlik, M. (2015). RIMROCK Robust Remote Process Controller. Retrieved 21 March 2015, from https://submit.plgrid.pl/
  • 20. Konsorcjum PL-Grid. (2015). Retrieved 21 March 2015, from http://www.plgrid.pl/projekty/ng.
Typ dokumentu
Bibliografia
Identyfikatory
Identyfikator YADDA
bwmeta1.element.-psjd-doi-10_1515_nuka-2015-0102
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