Impact of spread in minor actinide data from ENDF/B-VII.0, ENDF/B-VI.8, JENDL-3.3 and JEFF-3.0 on an IAEA-CRP FBR benchmark for MA incineration

Pandikumar, G.; Gopalakrishnan, V.; Mohanakrishnan, P.

doi:10.1016/j.anucene.2008.01.004

Cited by 9 publications

(1 citation statement)

References 6 publications

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“…Although the nuclear data and computational tools used for the present nuclear reactors have been extended to a high level of progress, the advanced types of reactors still require a further improvement in those tools and data, especially for employing unconventional nuclear fuel such as plutonium and minor actinides as a part of nuclear waste management and future design of nuclear systems. Addressing the above issues, several studies have been going on to test and validate nuclear data libraries such as ENDF/B-VI.0, ENDF/B-VII.0, ENDF/B-VII.1, JENDL-3.3, JEFF-3.0, and JEFF-3.1 using Monte Carlo transport codes MCNP, MCNPX, and others [1][2][3][4][5][6]. The experimental programs in MINERVE reactor [7][8][9] are designed to develop the integral absorption cross-sections of heavy isotopes and actinides.…”

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confidence: 99%

Evaluation of ENDF/B-VI.8 and ENDF/B-VII.0 Nuclear Data Libraries using OSMOSE Samples of MINERVE Reactor

Shahbunder

2019

Arab Journal of Nuclear Sciences and Applications

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The evaluation of nuclear data libraries is an important tool to develop nuclear technology. To achieve this goal, OSMOSE program utilized samples that are composed of separated actinides which contain isotopes of americium, plutonium, uranium, and thorium. By inserting those samples in the core configuration of MINERVE reactor, they can be used for improvement of nuclear cross-section databases by cross checking their calculated reactivity worth per each sample using several nuclear data library versions and experimental measurements. The present study investigated the effect of the used nuclear data library on the calculated reactivity of MINERVE reactor with several OSMOSE samples employed in R1-UO2, R1-UO2, and R1-MOX core configurations. The study performed a comparison between the calculated effective multiplication factor using MCNPX code with ENDF/BVI.8 and ENDF/BVII.0 nuclear data libraries. Moreover, the relative difference between the calculated reactivity worth by ENDF/BVI.8 and ENDF/BVII.0 was examined and discussed using available experimental results of the relevant core.

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