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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]
Copyright of Interactions (30050731) is the property of Springer Nature and its content may not be copied or emailed to multiple sites without the copyright holder's express written permission. Additionally, content may not be used with any artificial intelligence tools or machine learning technologies. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)
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DbLabel: Engineering Source
An: 193456707
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  Data: 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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  Data: <searchLink fieldCode="JN" term="%22Interactions+%2830050731%29%22">Interactions (30050731)</searchLink>. 5/1/2026, Vol. 247 Issue 1, p1-9. 9p.
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  Data: <searchLink fieldCode="DE" term="%22Nuclear+cross+sections%22">Nuclear cross sections</searchLink><br /><searchLink fieldCode="DE" term="%22Nuclear+fuels%22">Nuclear fuels</searchLink><br /><searchLink fieldCode="DE" term="%22Neutron+capture%22">Neutron capture</searchLink><br /><searchLink fieldCode="DE" term="%22Neutron+transport+theory%22">Neutron transport theory</searchLink><br /><searchLink fieldCode="DE" term="%22Nuclear+energy+safety+measures%22">Nuclear energy safety measures</searchLink><br /><searchLink fieldCode="DE" term="%22Neutron+absorbers%22">Neutron absorbers</searchLink>
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  Data: 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]
– Name: AbstractSuppliedCopyright
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  Data: <i>Copyright of Interactions (30050731) is the property of Springer Nature and its content may not be copied or emailed to multiple sites without the copyright holder's express written permission. Additionally, content may not be used with any artificial intelligence tools or machine learning technologies. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract.</i> (Copyright applies to all Abstracts.)
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