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Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes

  • Z. Gholamzadeh and S. A. H. Feghhi
Published/Copyright: November 4, 2014
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Kerntechnik
From the journal Kerntechnik Volume 79 Issue 5

Abstract

Recently thorium-based fuel matrixes are taken into consideration for nuclear waste incineration because of thorium proliferation resistance feature moreover its breeding or convertor ability in both thermal and fast reactors. In this work, neutronic influences of adding Am to (Th-235U)O2 on effective delayed neutron fraction, reactivity coefficients and burn up of a fed CANDU core has been studied using MCNPX 2.6.0 computational code. Different atom fractions of Am have been introduced in the fuel matrix to evaluate its effects on neutronic parameters of the modeled core. The computational data show that adding 2% atom fraction of Am to thorium-based fuel matrix won't noticeably change reactivity coefficients in comparison with the fuel matrix containing 1% atom fraction of Am. The use of 2% atom fraction of Am resulted in a higher delayed neutron fraction. According to the obtained data, 32.85 GWd burn up of the higher Americium-containing fuel matrix resulted in 55.2%, 26.5%, 41.9% and 2.14% depletion of 241Am, 243Am, 235U and 232Th respectively. 132.8 kg of 233U fissile element is produced after the burn up time and the nuclear core multiplication factor increases in rate of 2390 pcm. The less americium-containing fuel matrix resulted in higher depletion of 241/243Am, 235U and 232Th while the nuclear core effective multiplication factor increases in rate of 5630 pcm after the burn up time with 9.8 kg additional 233U production.

Kurzfassung

In letzter Zeit werden Brennstoff-Matrizen auf Thoriumbasis zur Verbrennung nuklearer Abfälle in Betracht gezogen. In der vorliegenden Arbeit wird das Neutronenverhalten bei Zusatz von Am-Anteilen zu (Th-235U)O2, die Reaktivitätskoeffizienten und das Abbrandverhalten eines CANDU Kerns untersucht mit Hilfe des Rechencodes MCNPX 2.6.0. Die Rechenergebnisse zeigen, dass sich durch den Zusatz von 2% Am zur Thorium-basierten Brennstoff-Matrix die Reaktivitätskoeffizienten im Vergleich zu 1% Am nicht merklich ändern. Die Verwendung von 2% Am-Anteilen in der Thorium-basierten Brennstoff-Matrix führt aber zu einem höheren Anteil verzögerter Neutronen. Die Ergebnisse zeigen, dass 32.85 GWd Abbrand der Brennstoffmatrix mit dem höheren Am-Anteil zu 55.2%, 26.5%, 41.9% und 2.14% Abreicherung von 241Am, 243Am, 235U und 232Th führt. 132.8 kg spaltbares 233U fissile werden nach der Abbrandzeit erzeugt und der Multiplikationsfaktor erhöht sich auf 2390 pcm. Die Brennelement-Matrix mit dem geringeren Am-Anteil führt zu einer höheren Abreicherung von 241/243Am, 235U und 232Th, während der effective Multiplikationsfaktor sich auf 5630 pcm erhöht nach der Abbrandzeit bei einer Erzeugung von 9.8 kg 233U.

References

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Received: 2014-04-15
Published Online: 2014-11-04
Published in Print: 2014-11-30

© 2014, Carl Hanser Verlag, München

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https://doi.org/10.3139/124.110432

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  3. Summaries/Kurzfassungen
  4. Summaries
  5. Technical Contributions/Fachbeiträge
  6. Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor
  7. Flow accelerated corrosion study in feeder pipes
  8. Analysis of the flow instability among channels of the OTSG in the naval craft NPP
  9. Experimental investigation of the MSFR molten salt reactor concept
  10. Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes
  11. Atomistic nano-scale 3D simulations about effects of Cr percentage on the molecular dynamics parameters of Fe-9–12% Cr alloys at fusion reactor temperature conditions
  12. On an analytical evaluation of the flux and dominant eigenvalue problem for the steady state multi-group multi-layer neutron diffusion equation
  13. A finite volume approach to the problem of heat transfer in axisymmetric annulus geometry with internal heating element using local analytical solution techniques
  14. CO2 doped γ-irradiated hydroxyapatite for EPR dosimetry
  15. An investigation of the effect of the upper beryllium reflector on the moderator temperature coefficient of reactivity of miniature neutron source reactors
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