Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

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Comparison of (Th-233U) O2 and (Th-235U) O2 fuel burn up into a thermal research reactor using MCNPX 2.6 code

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comparison of (th-233u) o2 and (th-235u) o2 fuel burn up into a thermal research reactor using mcnpx 2.6 code

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ژورنال

عنوان ژورنال: Nukleonika

سال: 2014

ISSN: 0029-5922

DOI: 10.2478/nuka-2014-0017