Neutronic calculations for CANDU thorium systems usingMonte Carlo techniques

  • In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium (CANDU) reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.
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  • [1] Şahin S, Şahin H M, Acir A. Nuclear Engineering and Design, 2010, 240: 2066-2074[2] Şahin S, Acir A. Energy Conversion and Management, 2006, 47: 1661-1675[3] Şahin S, Yildiz K, Şahin H M, Şahin N, Acir A. Nuclear Engineering and Design, 2006, 236(17): 178-179[4] Busse M. Optimization Of Thorium-Based Seed-Blanket Fuel Cycles For Nuclear Power Plants. M.S., Nuclear Engineering Instituto Balseiro, University of Cuyo, Argentina, 1995[5] Canerl M, Dugan E T. Annals of Nuclear Energy, 2000, 27: 759-770[6] Chanyun L. Design and Neutronic Evaluation of Thorium Fuel in Pressurized Water Reactors. Master of Science Thesis Reactor Physics Department Royal Institute of Technology, Stockholm, Sweden, 2008[7] LONG Y. Modeling the Performance of High Burn-up Thoria and Urania PWR Fuel. Department of Nuclear Engineering, B.S., Tsinghua University, Beijing, China, 1995[8] IAEA. Thorium Based Fuel Options for the Generation of Electricity: Development in the 1990s, IAEA-TECDOC-1155, 2000[9] Margeanuand C A, Rizoiu A C. Thorium-Based Fuels Preliminary Lattice Cell Studies For CANDU Reactors. Proceedings of the 7th Conference on Nuclear and Particle Physics. Sharm El-Sheikh, Egypt, 2009[10] Moosakhani A, Nasrabadi M N, Timuri B. Nuclear Engineering and Design, 2011, 241: 1459-1462[11] Oak Ridge National Laboratory, Monte Carlo N-Particle Transport Code System. Los Alamos National Laboratory, Los Alamos, New-Mexico, 2000."MCNP4C manual"[12] Snoj L, Ravnik M. Calculation of Power Density with MCNP in TRIGA Reactor. Nuclear Energy for New Europe, International Conference, Portorož, Slovenia, 2006[13] Eshghi Yaraziz M, Shayesteh M. Golobal J. P A Sci. and Tech., 2011, v01: 65-72[14] Eshghi M, Shayesteh M. Archives Des Sciences, 2012, 65(4): 1[15] Lamarsh J R. Introduction to Nuclear Engineering. Second ed. Addison-Wesley Publishing Co., USA, 1983
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M. Saldideh, M. Shayesteh and M. Eshghi. Neutronic calculations for CANDU thorium systems usingMonte Carlo techniques[J]. Chinese Physics C, 2014, 38(8): 088201. doi: 10.1088/1674-1137/38/8/088201
M. Saldideh, M. Shayesteh and M. Eshghi. Neutronic calculations for CANDU thorium systems usingMonte Carlo techniques[J]. Chinese Physics C, 2014, 38(8): 088201.  doi: 10.1088/1674-1137/38/8/088201 shu
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Received: 2013-09-11
Revised: 1900-01-01
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Neutronic calculations for CANDU thorium systems usingMonte Carlo techniques

    Corresponding author: M. Saldideh,
    Corresponding author: M. Eshghi,

Abstract: In this paper, we have investigated the prospects of exploiting the rich world thorium reserves using Canada Deuterium Uranium (CANDU) reactors. The analysis is performed using the Monte Carlo MCNP code in order to understand how much time the reactor is in criticality conduction. Four different fuel compositions have been selected for analysis. We have obtained the infinite multiplication factor, k, under full power operation of the reactor over 8 years. The neutronic flux distribution in the full core reactor has already been investigated.

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