Home Technology Minor actinide burning in a CANDU thorium reactor
Article
Licensed
Unlicensed Requires Authentication

Minor actinide burning in a CANDU thorium reactor

  • S. Şahin , K. Yıldız , H. M. Şahin , A. Acır , N. Şahin and T. Altınok
Published/Copyright: March 26, 2013
Become an author with De Gruyter Brill
Submit Manuscript Author Information
Kerntechnik
From the journal Kerntechnik Volume 71 Issue 5-6

Abstract

Nuclear waste actinides can be used as a booster fissile fuel material in form of mixed fuel with thorium in a CANDU reactor in order to assure the initial criticality at startup. Two different fuel compositions have been found useful to provide sufficient reactor criticality over a long operation period: 1) 95% thoria (ThO2)+5% minor actinides MAO2 and 2) 90% ThO2+5% MAO2+5% UO2. The latter allows a higher degree of nuclear safeguarding through denaturing the new 233U fuel with 238U. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 10 years. The criticality starts by k∞>1.3 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2nd year and remains close to k∞ = ∼1.06 for ∼10 years. After the 2nd year, the CANDU reactor begins to operate practically as a thorium burner. Very high burn up could be achieved with the same fuel material (up to 200000 MW.D/MT), provided that the fuel rod claddings would be replaced periodically (after every 50000 or 100000 MW.D/MT). The reactor criticality can be maintained until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.

Kurzfassung

Aktiniden aus abgebranntem Kernbrennstoff können nach Mischung mit Thorium als nuklearer Spaltstoff in einem CANDU Reaktor eingesetzt werden, um die Überschußreaktivität beim Anfahren sicherzustellen. Zwei verschiedene Brennstoffzusammensetzungen wurden als wirksam gefunden, um für eine längere Betriebsdauer ausreichend Reaktivität bereitzustellen: 1) 95% Thoriumoxid+5% Minore Aktiniden und 2) 90% Thoriumoxid+5% Minore Aktiniden+5% Uranoxid. Der letztere Brennstoff ermöglicht einen höheren Grad an Schutz gegen Missbrauch durch die Verschlechterung des Uran-233 durch Uran-238. Die zeitliche Änderung der Kritikalität kunendl. und des Brennstoffabbrandes wurden für einen Leistungsbetrieb von 10Jahren berechnet. Die Kritikalität beginnt mit kunendl. >1,3 für beide Brennstoffzusammensetzungen. Eine starke Abnahme der Kritikalität wurde im ersten Betriebsjahr infolge des schnellen Plutoniumausbrandes im Aktinidenbrennstoff festgestellt. Die Kritikalität wird nach dem zweiten Jahr nahezu konstant und bleibt nahe bei kunendl. = 1,06 für etwa 10 Jahre. Nach dem zweiten Betriebsjahr wirkt der CANDU-Reaktor als Thorium-Brenner. Mit dem Brennstoff können sehr hohe Abbrandwerte von bis zu 200000 MWd/t erreicht werden, wenn das Hüllrohr regelmäßig ausgetauscht wird. Der Reaktor kann kritisch gehalten werden bis ein großer Teil des Thoriums verbrannt wurde. Damit könnten die Brennstoffkosten verringert und gleichzeitig die nuklearen Abfallmengen für die Endlagerung pro Energieerzeugung drastisch verringert werden.

References

1Šahin, S.; Calinon, R.: Criticality of Curium Assemblies. Atomkernenergie/Kerntechnik46 (1985) 45–49Search in Google Scholar

2Šahin, S.: Power Flattening in a Hybrid Blanket Using Nuclear Waste Actinides. Kerntechnik, 53 (1989) 285–290Search in Google Scholar

3Rubbia, C.: Neutrons in Highly Diffusive Transparent Medium: An Effective Neutron „Storage“ Device. Proceedings of The Ninth International Conference on Emerging Nuclear Energy Systems, p. 4, Tel-Aviv, Israel (June 28–July 2, 1998)Search in Google Scholar

4Ronen, Y.; Aboudy, M.; Regev, D.: A novel Method for Energy Production Using242m Am as a Nuclear Fuel. Proceedings of The Ninth International Conference on Emerging Nuclear Energy Systems, pp. 59–66, Tel-Aviv, Israel (June 28–July 2, 1998)Search in Google Scholar

5Boczar, P. G.; Chan, P. S. W.; Dyck, G. R.; Ellis, R. J.; Jones, R. T.; Sullivan, J. D.; Taylor, P.: Thorium Fuel-Cycle Studies for CANDU Reactors, Thorium Fuel Utilization: Options and Trends. Proceedings of three IAEA meetings held in Vienna in 1997, 1998 and 1999, IAEA TECDOC-1319, 25–41, International Atomic Energy Agency, Vienna, Austria, (2002)Search in Google Scholar

6Boczar, P. G.; Dyck, G. R.; Chan, P. S. W.; Buss, D. B.: Recent Advances in Thorium Fuel Cycles for CANDU Reactors, Thorium Fuel Utilization: Options and Trends. Proceedings of three IAEA meetings held in Vienna in 1997, 1998 and 1999, IAEA TECDOC-1319, 104–120, International Atomic Energy Agency, Vienna, Austria, (2002)Search in Google Scholar

7Critoph, E.: Prospects for Self–sufficient Equilibrium Thorium Cycles in CANDU Reactors. Report AECL-5501, Atomic Energy of Canada Ltd, Canada, (1976)Search in Google Scholar

8Jagannathan, V.; UshaP.; Karthikeyan, R.; Ganesan, S.; Jain, R. P.; Kamat, S. U.: ATBR-A Thorium Breeder Reactor Concept for Early Induction of Thorium in an Enriched Uranium Reactor. Nuclear Technology133 (2001) 1–31Search in Google Scholar

9Loewen, E. P.; Wilson, R. D.; Hohorst, J. K.; Kumar, A. S.: Preliminary Frapcon-3th Steady-State Fuel Analysis of ThO2 and UO2 Fuel Mixtures. Nuclear Technology136 (2001) 261–277Search in Google Scholar

10Šahin, S.; Šahin, H. M.; Alkan, M.; Yıldız, K.: An Assessment of Thorium and Spent LWR-Fuel Utilization Potential in CANDU Reactors. Energy Conversion and Management45 (2004) 1067–1085Search in Google Scholar

11Šahin, S.; Yıldız, K.; Acır, A.: Power Flattening in the Fuel Bundle of a CANDU Reactor. Nuclear Engineering and Design232 (2004) 7–18Search in Google Scholar

12Alkan, M.: Reutilization of Nuclear Fuel Wastes. PhD Thesis in Turkish, Gazi University, Institute of Science and Technology, Ankara, Turkey, (2003)Search in Google Scholar

13Lee, J. S.; Song, K. C.; Yang, M. S.; Chun, K. S.; Rhee, B. W.; Hong, J. S.; Park, H. S.; Rim, C. S.; Keil, H.: Research and Development Program of KAERI for DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors). Proc. Int. Conf. Technology Exhibition on Future Nuclear Systems: Emerging Fuel Cycles and Waste Disposal Options (GLOBAL'93), Seattle, Washington, American Nuclear Society, September12–17, (1993) 733Search in Google Scholar

14YangM. S.; Lee, Y. W.; Bae, K. K., Na, S. H.: Conceptual Study on the DUPIC Fuel Manufacturing Technology. Proc. Int. Conf. Technology Exhibition on Future Nuclear Systems: Emerging Fuel Cycles and Waste Disposal Options (GLOBAL'93), Seattle, Washington, American Nuclear Society, September 12–17, (1993) 740Search in Google Scholar

15Choi, H. B.; Rhee, B. W.; Park, H. S.: Physics Study on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU (DUPIC). Nuclear Science and Engineering126 (1997) 80Search in Google Scholar

16Choi, H. B.; Rhee, B. W.; Park, H. S.: Fuel Management and DUPIC Fuel Performance. Trans. Am. Nucl. Soc.73 (1995) 162Search in Google Scholar

17Na, M. G.; Ko, W. I.; Choi, H.: Spent Fuel Combination Method for Minimizing Composition Variability of DUPIC Fuel. Nuclear Science and Engineering142 (2002) 315–326Search in Google Scholar

18Choi, H.; Ko, W. I.; Yang, M. S.: Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors-I: DUPIC Fuel Fabrication Cost. Nuclear Technology134 (2001) 110Search in Google Scholar

19Choi, H.; Ko, W. I.; Yang, M. S.; Namgung, I.; Na, B. G.: Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors-II: DUPIC Fuel-Handling Cost. Nuclear Technology134 (2001) 130Search in Google Scholar

20Ko, W. I.; Choi, H.; Roh, G.; Yang, M. S.: Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors-III: DUPIC Fuel Disposal Cost. Nuclear Technology134 (2001) 149Search in Google Scholar

21Jeong, C. J.; Choi, H. B.: Compatibility Analysis on Existing Reactivity Devices in CANDU 6 Reactors for DUPIC Fuel Cycle. Nuclear Science and Engineering134 (2000) 265Search in Google Scholar

22Ko, W. I.; Choi, H.; Yang, M. S.: Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors-IV: DUPIC Fuel Cycle Cost. Nuclear Technology134 (2001) 167Search in Google Scholar

23Ko, W. I.; Kim, H. D.; Yang, M. S.: Advantages of Irradiated DUPIC Fuels from the Perspective of Environmental Impact. Nuclear Technology138 (2002) 167Search in Google Scholar

24Woodhead, L. W.; Ingolfsrud, L. J.: Performance of Canadian Commercial Nuclear Units and Heavy Water Plants. Paper presented at European Nuclear Conference, 21 April 1975, Paris, France. Summary in Transaction American Nuclear Society, 20: 160Search in Google Scholar

25Altınok, T.: Neutronic Analysis of CANDU Reactors Operating with Rejuvenated Spent Fuel in Catalysed Fusion Reactors. PhD thesis in Turkish, Gazi University, Institute of Science and Technology, Ankara, Turkey, (1993)Search in Google Scholar

26IAEA: Status and Prospects of Thermal Breeders and their Effect on Fuel Utilization, Technical Report Series No. 195, International Atomic Energy Agency, Vienna, Austria, (1979)Search in Google Scholar

27Petrie, L. M.; Fox, P. B.; Lucius, K.: Standard Composition Library, NUREG/CR-0200, Revision 6, 3, Section M8, ORNL/NUREG/CSD-2/Volume III/Revision 7, Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A, (2004)Search in Google Scholar

28Ma, B. M.: Nuclear Reactor Materials and Applications. Van Nostrand Reinhold Company Limited, Molly Millars Lane, Wokingham, Berkshire, England, (1983)Search in Google Scholar

29Lamarsh, J. R.; Baratta, A. J.: Introduction to Nuclear Engineering. Third Edition, Prentice Hall, Upper Saddle River, New Jersey 07458, (2001)Search in Google Scholar

30Manson, B.; Pigford, T. H.; Levi, H. W.: Nuclear Chemical Engineering, New York: McGraw-Hill, (1981)Search in Google Scholar

31Petrie, L. M.: SCALE5-Scale System Driver, NUREG/CR-0200, Revision 7, Volume III, Section M1, ORNL/NUREG/CSD-2/V3/R7, Oak Ridge National Laboratory, (2004)Search in Google Scholar

32GreeneN. M.; PetrieL. M.: XSDRNPM, A One-Dimensional Discrete-Ordinates Code For Transport Analysis, NUREG/CR-0200, Revision 7, 2, Section F3, ORNL/NUREG/CSD-2/V2/R7, Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A, (2004)Search in Google Scholar

33JordanW. C.; Bowman, S. M.; Hollenbach, D. F.: Scale Cross-Section Libraries, NUREG/CR-0200, Revision 7, 3, Section M4, ORNL/NUREG/CSD-2/V3/R7, Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A, (2004)Search in Google Scholar

34LandersN. F.; Petrie, L. M.; Hollenbach, D. F.: CSAS, Control Module for Enhanced Criticality Safety Analysis Sequences, NUREG/CR-0200, Revision 6, 1, Section C4, ORNL/NUREG/CSD-2/V1/R7, Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A, (2004)Search in Google Scholar

35GreeneN. M.: BONAMI, Resonance Self-Shielding by the Bondarenko Method, NUREG/CR-0200, Revision 6, 2, section F1, ORNL/NUREG/CSD-2/V2/R7, Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A, (2004)Search in Google Scholar

36GreeneN. M.; Petrie, L. M.; Westfall, R. M.: NITAWL-III, Scale System Module For Performing Resonance Shielding and Working Library Production, NUREG/CR-0200, Revision 6, 2, Section F2, ORNL/NUREG/CSD-2/V2/R7, Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A, (2004)Search in Google Scholar

37Leggertt, R. D.; Omberg, R. P.: Mixed Oxide Fuel Development, Proc. Int. Conf. Fast Breeder Systems: Experience Gained and Path to Economical Power Generation, Richland, Washington, September 13–17, American Nuclear Society, (1987)Search in Google Scholar

38Waltar, A. E.; Deitrich, L. W.: Status of research on Key LMR Safety Issues. Nuclear Safety29 (1988) 125Search in Google Scholar

39Šahin, S.; Yapıcı, H.: Investigation of the Neutronic Potential of Moderated and Fast (D, T) Hybrid Blankets for Rejuvenation of CANDU Spent Fuel. Fusion Technology16 (1989) 331–345Search in Google Scholar

40Šahin, S., Baltacıoğlu, E.; Yapıcı, H.: Potential of a Catalyzed Fusion Driven Hybrid Reactor for the Regeneration of CANDU Spent Fuel. Fusion Technology20 (1991) 26–39Search in Google Scholar

41Meyer, W.; Loyalka, S. K.; Nelson, W. E.; WilliamsR. W.: The Homemade Nuclear Bomb Syndrome. Nuclear Safety18 (1977) 427Search in Google Scholar

Received: 2006-5-2
Published Online: 2013-03-26
Published in Print: 2006-11-01

© 2006, Carl Hanser Verlag, München

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  3. Summaries/Kurzfassungen
  4. Summaries
  5. Technical Contributions/Fachbeiträge
  6. Minor actinide burning in a CANDU thorium reactor
  7. Modeling of the TOSQAN test facility with the lumped parameter code COCOSYS
  8. Thermal-hydraulic modeling of the onset of flow instability in MTR reactors
  9. 10.3139/124.100302
  10. 10.3139/124.100304
  11. Radiation transfer in inhomogeneous exponential media
  12. Neutron transport problems for extremely anisotropic scattering
  13. Milne problem for isotropic and linearly anisotropic scattering for specular and diffuse reflecting boundary conditions solved with the HN method
  14. ICDE-results on complete common cause failures in the light of results obtained with the POS model
  15. Modelling of the cumulative behaviour of Caesium and Strontium activities in nuclear fuel
  16. Analytical study of the closure flow inside the ETRR-2 core chimney
  17. Minimizing radiation exposure for the reactor staff during the dismantling of a TRIGA research reactor
https://doi.org/10.3139/124.100299

Articles in the same Issue

  1. Contents/Inhalt
  2. Contents
  3. Summaries/Kurzfassungen
  4. Summaries
  5. Technical Contributions/Fachbeiträge
  6. Minor actinide burning in a CANDU thorium reactor
  7. Modeling of the TOSQAN test facility with the lumped parameter code COCOSYS
  8. Thermal-hydraulic modeling of the onset of flow instability in MTR reactors
  9. 10.3139/124.100302
  10. 10.3139/124.100304
  11. Radiation transfer in inhomogeneous exponential media
  12. Neutron transport problems for extremely anisotropic scattering
  13. Milne problem for isotropic and linearly anisotropic scattering for specular and diffuse reflecting boundary conditions solved with the HN method
  14. ICDE-results on complete common cause failures in the light of results obtained with the POS model
  15. Modelling of the cumulative behaviour of Caesium and Strontium activities in nuclear fuel
  16. Analytical study of the closure flow inside the ETRR-2 core chimney
  17. Minimizing radiation exposure for the reactor staff during the dismantling of a TRIGA research reactor
Sign up now to receive a 20% welcome discount
Institutional Access
How does access work?
Have an idea on how to improve our website?
Please write us.
© 2025 De Gruyter Brill
Downloaded on 13.12.2025 from https://www.degruyterbrill.com/document/doi/10.3139/124.100299/html
Scroll to top button