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<p>This study analyzes the variation of macroscopic neutron cross sections for IRT-4&#xa0;M fuel assemblies during U-235 burnup, a key parameter for the safe and efficient operation of research reactors like the VVR-SM. Using the WIMS code with the ENDF/B-VI nuclear data library, cell-averaged two-group macroscopic absorption (Σₐ) and fission (νΣբ) cross sections were calculated for IRT-4&#xa0;M fuel with 19.75% U-235 enrichment. The calculations, performed for burnup levels from 0% to 60%, reveal distinct behavioral trends. The macroscopic absorption cross section Σₐ exhibits a slight initial increase up to ~ 10% burnup, attributed to the rapid accumulation of fission products with high absorption cross-sections (e.g., Xe-135, Sm-149). Beyond 10% burnup, Σₐ decreases monotonically as U-235 depletion and the saturation of short-lived poisons become dominant. In contrast, the macroscopic fission cross section νΣբ shows a steady, nearly linear decline throughout the burnup period, directly reflecting the progressive loss of the fissile U-235 inventory. The results underscore that accurate reactor core modeling requires burnup-dependent cross-section libraries to account for the competing effects of fissile material depletion and fission product poisoning, ensuring reliable reactivity margin and safety assessments.</p>

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Analysis of the dependence of macroscopic thermal neutron absorption and fission cross sections on U-235 burnup in IRT-4 M fuel using the WIMS code

  • 期刊论文

  • 发表日期：

    2026年5月1日

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  • T. B. Fayziev,
  • S. A. Baytelesov,
  • F. R. Kungurov,
  • Sh. A. Alikulov,
  • D. P. Tadjibaev

摘要

This study analyzes the variation of macroscopic neutron cross sections for IRT-4 M fuel assemblies during U-235 burnup, a key parameter for the safe and efficient operation of research reactors like the VVR-SM. Using the WIMS code with the ENDF/B-VI nuclear data library, cell-averaged two-group macroscopic absorption (Σₐ) and fission (νΣբ) cross sections were calculated for IRT-4 M fuel with 19.75% U-235 enrichment. The calculations, performed for burnup levels from 0% to 60%, reveal distinct behavioral trends. The macroscopic absorption cross section Σₐ exhibits a slight initial increase up to ~ 10% burnup, attributed to the rapid accumulation of fission products with high absorption cross-sections (e.g., Xe-135, Sm-149). Beyond 10% burnup, Σₐ decreases monotonically as U-235 depletion and the saturation of short-lived poisons become dominant. In contrast, the macroscopic fission cross section νΣբ shows a steady, nearly linear decline throughout the burnup period, directly reflecting the progressive loss of the fissile U-235 inventory. The results underscore that accurate reactor core modeling requires burnup-dependent cross-section libraries to account for the competing effects of fissile material depletion and fission product poisoning, ensuring reliable reactivity margin and safety assessments.