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  • 标题:Burnup simulations of different fuel grades using the MCNPX Monte Carlo code
  • 本地全文:下载
  • 作者:Asah-Opoku Fiifi ; Liang Zhihua ; Huque Ziaul
  • 期刊名称:Nuclear Technology and Radiation Protection
  • 印刷版ISSN:1451-3994
  • 出版年度:2014
  • 卷号:29
  • 期号:4
  • 页码:259-267
  • DOI:10.2298/NTRP1404259A
  • 出版社:VINČA Institute of Nuclear Sciences
  • 摘要:

    Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX), uranium oxide fuel (UOX), and commercially enriched uranium or uranium metal (CEU) - are used in this simulation and their impact on the effective multiplication factor (Keff) and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  • 关键词:MCNPX code; burnup calculation; criticality calculation; radionuclide inventory
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