• Title/Summary/Keyword: H.235

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Basic characterization of uranium by high-resolution gamma spectroscopy

  • Choi, Hee-Dong;Kim, Junhyuck
    • Nuclear Engineering and Technology
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    • v.50 no.6
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    • pp.929-936
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    • 2018

  • A basic characterization of uranium samples was performed using gamma- and X-ray spectroscopy. The studied uranium samples were eight types of certified reference materials with $^{235}U$ enrichments in the range of 1-97%, and the measurements were performed over 24 h using a high-resolution and high-purity planar germanium detector. A general peak analysis of the spectrum and the $XK_{\alpha}$ region of the uranium spectra was carried out by using HyperGam and HyperGam-U, respectively. The standard reference sources were used to calibrate the spectroscopy system. To obtain the absolute detection efficiency, an effective solid angle code, EXVol, was run for each sample. Hence, the peak activities and isotopic activities were determined, and then, the total U content and $^{234}U$, $^{235}U$, and $^{238}U$ isotopic contents were determined and compared with those of the certified reference values. A new method to determine the model age based on the ratio of the activities of $^{223}Ra$ and $^{235}U$ in the sample was studied, and the model age was compared with the known true age. In summary, the present study developed a method for basic characterization of uranium samples by nondestructive gamma-ray spectrometry in 24 h and to obtain information on the sample age.

  • https://doi.org/10.1016/j.net.2018.04.008 인용 PDF KSCI

Optimization of CANFLEX-RU Fuel Bundle for CANDU-6

  • Lee, Y. O.;C. J. Jeong;K. S. Sim;J. S. Jun;Park, G. S.;Kim, B. G.;Park, J. H.;H. C. Suk
    • Proceedings of the Korean Nuclear Society Conference
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    • 1995.10a
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    • pp.35-40
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    • 1995

  • Considering the higher discharge burnup, lower channel refuelling rate, lower linear element rating(LER), lower coolant void reactivity and axial power shape, CANFLEX-RU fuel bundle is optimized for CANDU-6 by grading the fissile composition in the ring-wise of the bundle and by applying fuel management scheme appropriately. The fissile composition of the fuel bundle is graded as the recovered uranium (0.9 w/o U-235) in the outer and intermediate elements, depleted Uranium (0.2 w/o U-235) in the center element, natural uranium (0.71 w/o U-235) in the inner elements. Enrichment is not required for these fuel. The fissile composition is optimized by lattice calculation and by time-averaged reactor simulation. CANFLEX-RU optimized for CANDU-6 resulted to be the 15% lower channel refuelling rate, acceptable axial power profile and power envelope, 70% higher discharge burnup, 15% lower LER and not increase coolant void reactivity compared with the 37-element natural uranium bundle for CANDU-6.

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Sensitivity and uncertainty quantification of neutronic integral data in the TRIGA Mark II research reactor

  • Makhloul, M.;Boukhal, H.;Chakir, E.;El Bardouni, T.;Lahdour, M.;Kaddour, M.;Ahmed, Abdulaziz;Arectout, A.;El Yaakoubi, H.
    • Nuclear Engineering and Technology
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    • v.54 no.2
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    • pp.523-531
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    • 2022

  • In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U235(n, f), $ U_{235}(n\bar{\nu})$ and H1(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.

  • https://doi.org/10.1016/j.net.2021.08.003 인용 PDF KSCI