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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.

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Title: 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.
Authors: Fayziev, T. B.1 (AUTHOR) temurfayziyev@outlook.com, Baytelesov, S. A.1 (AUTHOR) baytel@inp.uz, Kungurov, F. R.1 (AUTHOR) fkungurov@inp.uz, Alikulov, Sh. A.1 (AUTHOR) alikulov@inp.uz, Tadjibaev, D. P.1 (AUTHOR) tadjibaev@inp.uz
Source: Interactions (30050731). 5/1/2026, Vol. 247 Issue 1, p1-9. 9p.
Subjects: Nuclear cross sections, Nuclear fuels, Neutron capture, Neutron transport theory, Nuclear energy safety measures, Neutron absorbers
Abstract: 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. [ABSTRACT FROM AUTHOR]
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Database: Engineering Source
Description
Abstract: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. [ABSTRACT FROM AUTHOR]
ISSN:30050731
DOI:10.1007/s10751-026-02474-1