Uncertainty in cross-section calculations for reactions induced by neutrons with energy above 0.1 MeV
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C. H. M. Broeders
, A. Yu. Konobeyev and L. Mercatali
Abstract
The uncertainty in the calculation of neutron induced reaction cross-sections using modern nuclear models and codes has been investigated. The cross-sections have been calculated with the help of the TALYS code and the modified ALICE code using different models for the calculation of nuclear level density. The experimental data from EXFOR for neutron induced reactions for nuclei from 27Al to 209Bi and incident neutron energies above 0.1 MeV have been used for the comparison with calculations. The results obtained give the possibility to find the best approaches for the cross-section calculation for nuclei from different mass ranges.
Kurzfassung
Die Unsicherheit bei der Berechnung von Neutronen-induzierten Reaktionsquerschnitten mit Hilfe moderner Kernmodelle und Rechencodes wurde untersucht. Die Wirkungsquerschnitte wurden berechnet mit Hilfe des TALYS Codes und des modifizierten ALICE Codes unter Verwendung verschiedener Modelle zur Berechnung der Kernniveaudichte. Die experimentellen Daten von EXFOR für Neutronen-induzierte Reaktionen bei 27Al und 209Bi Kernen und einfallenden Energien von über 0,1 MeV wurden verwendet für den Vergleich mit den Berechnungen. Mit Hilfe der erhaltenen Ergebnisse können die besten Ansätze für die Berechnung von Wirkungsquerschnitten für Kerne verschiedener Massen gefunden werden.
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Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- The multipurpose thermalhydraulic test facility TOPFLOW: an overview on experimental capabilities, instrumentation and results
- Uncertainty in cross-section calculations for reactions induced by neutrons with energy above 0.1 MeV
- Determination of the exposure build-up factor in a slab using the LTSN method
- Lumped parameters analysis of the IAEA research reactor benchmark problem
- Assessment of the look-up table using the tubular and bundle CHF data and modification of the bundle correction factor
- A finite element model for static strength analysis of CANDU fuel bundle
- Effect of packing fraction variations on the multiplication factor in pebble-bed nuclear reactors
- Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects
- Radiological and thermal characteristics of CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks for spent nuclear fuel storage at Ignalina Nuclear Power Plant
- Analyses of severe accident scenarios in RBMK-1500