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Rabat, Morocco

Nacir B.,Universite Ibn Tofail | Boulaich Y.,CEN Maamora | Chakir E.,Universite Ibn Tofail | El Bardouni T.,Abdelmalek Essaadi University | And 2 more authors.
Annals of Nuclear Energy | Year: 2014

The Moroccan TRIGA MARK II reactor core is loaded with 8.5% in weight of uranium standard fuel elements. Additional fresh fuel elements must periodically be added to the core in order to remedy the observed low power and to return to the initial reactivity excess at the End Of Cycle. 12%-uranium fuel elements are available to relatively improve the short fuel lifetime associated with standard TRIGA elements. These elements have the same dimensions as standards elements, but with different uranium weight. The objective in this study is to demonstrate that the Moroccan TRIGA reactor could safely operate, around 2 MW power, with new configurations containing these 12% fuel elements. For this purpose, different safety related thermal-hydraulic parameters have been calculated in order to ensure that the safety margins are largely respected. Therefore, the PARET model for this TRIGA reactor that was previously developed and combined with the MCNP transport code in order to calculate the 3-D temperature distribution in the core and all the most important parameters like the axial distribution of DNBR (Departure from Nucleate Boiling Ratio) across the hottest channel. The most important conclusion is that the 12% fuel elements utilization will have no influence on the safety of the reactor while working around 2 MW power especially for configurations based on insertions in C and D-rings. © 2013 Elsevier Ltd. All rights reserved.

Boulaich Y.,Abdelmalek Essaadi University | El Bardouni T.,Abdelmalek Essaadi University | Erradi L.,Mohammed V University | Chakir E.,Universite Ibn Tofail | And 6 more authors.
Nuclear Engineering and Design | Year: 2011

In the present work, we analyze the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA/Cadarache, was mainly dedicated to the RTC studies for both UO2 and UO2-PuO 2 PWR type lattices covering the whole temperature range from 20 °C to 300 °C. We have developed an accurate 3D model of the EOLE reactor by using the MCNP5 Monte Carlo code which guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII, JEFF3.1 and JENDL3.3 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values; the discrepancy is less than 0.23 pcm/°C. Whereas, when using ENDF-BVII evaluation for UOX configuration, this discrepancy reaches a value of 0.63 pcm/°C. In order to specify the source of this relatively large discrepancy in the case of ENDF-BVII nuclear data evaluation, the keff discrepancy between ENDF-BVII and JENDL3.3 was decomposed by using sensitivity and uncertainty analysis technique. The thermal energy range of 238U(n,γ) absorption cross section was found to contribute to the major part of the observed keff discrepancy between ENDFB-VII and JENDL3.3 evaluations. © 2011 Elsevier B.V. All rights reserved.

Boulaich Y.,CEN Maamora | Boulaich Y.,Abdelmalek Essaadi University | Nacir B.,CEN Maamora | El Bardouni T.,Abdelmalek Essaadi University | And 9 more authors.
Nuclear Engineering and Design | Year: 2011

The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maâmora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor. © 2010 Elsevier B.V. All rights reserved.

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