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Descripción
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| The assessment of the accuracy of parameters related to the reactor core performance (e.g., ke?) and fuel cycle (e.g., isotopic evolution/transmutation) due to the uncertainties in the ba- sic nuclear data (ND) is a critical issue. Di?erent error propagation techniques (adjoint/forward sensitivity analysis procedures and/or Monte Carlo technique) can be used to address by computa- tional simulation the systematic propagation of uncertainties on the ?nal parameters. To perform this uncertainty assessment, the ENDF covariance ?les (variance/correlation in energy and cross- reactions-isotopes correlations) are required. | |
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Internacional
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Si |
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JCR del ISI
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Si |
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Título de la revista
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Journal of the Korean Physical Society, |
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ISSN
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0374-4884 |
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Factor de impacto JCR
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0 |
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Información de impacto
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Volumen
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59 |
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DOI
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Número de revista
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2 |
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Desde la página
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1268 |
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Hasta la página
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1271 |
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Mes
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AGOSTO |
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Ranking
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