Отправить статью по E-mail Neutronic Study of Fuel Depletion for the MNSR Research Reactor Using DRAGON5 Code
- Авторы: Al Zain J.1,2, El Hajjaji O.1, El Bardouni T.1
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Учреждения:
- Radiations and Nuclear Systems Laboratory
- Physics Department
- Выпуск: Том 74, № 6 (2019)
- Страницы: 706-709
- Раздел: Engineering Physics
- URL: https://journals.rcsi.science/0027-1349/article/view/165279
- DOI: https://doi.org/10.3103/S0027134919060031
- ID: 165279
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Аннотация
The aim of this work is to study the fuel depletion of 30 kW miniature neutron source reactor (MNSR). Under operating conditions of two hours per day for five days a week at a peak thermal neutron flux of 1.0E+12 n/cm2 s, the estimated core life is 10 years (200 days). The DRAGON5 code is utilizing to generate the fuel group constants and the infinite multiplication factor versus the MNSR reactor operating time, at different burn-up values. The amounts of uranium burnt up and plutonium produced in the reactor core are calculated using the DRAGON5 transport lattice code. The results are in good agreement with previous studies.
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Об авторах
Jamal Al Zain
Radiations and Nuclear Systems Laboratory; Physics Department
Автор, ответственный за переписку.
Email: jalzain@uae.ac.ma
Марокко, Tetouan; Sana’a
O. El Hajjaji
Radiations and Nuclear Systems Laboratory
Email: jalzain@uae.ac.ma
Марокко, Tetouan
T. El Bardouni
Radiations and Nuclear Systems Laboratory
Email: jalzain@uae.ac.ma
Марокко, Tetouan
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