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RISS 인기검색어
- 검색키워드
- 저자 : H. Boukhal (검색결과 3 건)
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M. Makhloul,H. Boukhal,E. Chakir,T. El Bardouni,M. Lahdour,M. Kaddour,Abdulaziz Ahmed,A. Arectout,H. El Yaakoubi 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.2
In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, amodel of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP MonteCarlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor ofthis reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energygroups, 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 forthe 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 pcmrespectively for the reactions U235(n, f), U235(nn) and H1(n, g). On the other hand, these differences arevery 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 spectrapresent two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV
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El Ouahdani, S.,Erradi, L.,Boukhal, H.,Chakir, E.,El Bardouni, T.,Boulaich, Y.,Ahmed, A. Korean Nuclear Society 2020 Nuclear Engineering and Technology Vol.52 No.6
The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO<sub>2</sub> lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO<sub>2</sub> lattice configuration, we have also analyzed integral measurements in UO<sub>2</sub> clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.
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A. Arectout,I. Zidouh,Y. Sadeq,M. Azougagh,B. Maroufi,E. Chakir,H. Boukhal 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.1
This work aims to establish some X-ray qualities recommended by the International Standard Organization(ISO) using the half-value layer (HVL) and Hp(10) dosimetry approaches. The HVL values of thefollowing qualities N-60, N-80, N-100, N-150 and N-250 were determined using various attenuationlayers. The obtained results were compared to those of reference X-ray beam qualities and a goodagreement was found (difference less than 5% for all qualities). The GAMOS (Geant4-based Architecturefor Medicine-Oriented Simulations) radiation transport Monte Carlo toolkit was employed to simulatethe production of X-ray spectra. The characteristics HVLs, mean energy and the spectral resolution ofsimulated spectra have been calculated and turned out to be conform to the ISO reference ones (differenceless than the limit allowed by ISO). Furthermore, the conversion coefficients from air kerma topersonal dose equivalent for simulated and measured spectra were fairly similar (the maximum differenceless than 4.2%)
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