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Assessment of TRACE code for modeling of passive safety system during long transient SBO via PKL/SACO facility
Al-Yahia, O. S., Clifford, I., & Ferroukhi, H. (2024). Assessment of TRACE code for modeling of passive safety system during long transient SBO via PKL/SACO facility. Nuclear Engineering and Technology. https://doi.org/10.1016/j.net.2024.02.050
The influence of droplet breakup model on the prediction of reactor core parameters during reflood conditions
Al-Yahia, O. S., Bernard, M., Clifford, I., Perret, G., Bajorek, S., & Ferroukhi, H. (2024). The influence of droplet breakup model on the prediction of reactor core parameters during reflood conditions. Nuclear Engineering and Design, 416, 112815 (16 pp.). https://doi.org/10.1016/j.nucengdes.2023.112815
A systematic approach for the adequacy analysis of a set of experimental databases: Application in the framework of the ATRIUM activity
Baccou, J., Glantz, T., Ghione, A., Sargentini, L., Fillion, P., Damblin, G., … Adorni, M. (2024). A systematic approach for the adequacy analysis of a set of experimental databases: Application in the framework of the ATRIUM activity. Nuclear Engineering and Design, 421, 113035 (16 pp.). https://doi.org/10.1016/j.nucengdes.2024.113035
Benchmark exercise on ELSMOR passive heat removal system
Bersano, A., Lombardo, C., Alblouwy, F., Karppinen, I., Silde, A., Nikitin, K., … Weyermann, F. (2024). Benchmark exercise on ELSMOR passive heat removal system. Nuclear Engineering and Design, 419, 112961 (14 pp.). https://doi.org/10.1016/j.nucengdes.2024.112961
P2M simulation exercise on past fuel melting irradiation experiments
D’Ambrosi, V., Sercombe, J., Bejaoui, S., Chaieb, A., Baurens, B., Largenton, R., … Peltonen, J. (2024). P2M simulation exercise on past fuel melting irradiation experiments. Nuclear Technology, 210(2), 189-2015. https://doi.org/10.1080/00295450.2023.2194270
Hydrogen enhanced localized plasticity in zirconium as observed by digital image correlation
Fagnoni, F., Kursun, E. .C., Busi, M., Konarski, P., Yetik, O., Spolenak, R., … Duarte, L. I. (2024). Hydrogen enhanced localized plasticity in zirconium as observed by digital image correlation. Journal of Nuclear Materials, 590, 154873 (10 pp.). https://doi.org/10.1016/j.jnucmat.2023.154873
Development and testing of the hydrogen behavior tool for Falcon - HYPE
Konarski, P., Cozzo, C., Khvostov, G., & Ferroukhi, H. (2024). Development and testing of the hydrogen behavior tool for Falcon - HYPE. Nuclear Engineering and Technology, 56(2), 728-744. https://doi.org/10.1016/j.net.2023.11.012
Minimal requirements to thermal-hydraulics/3D-neutron kinetics clustering for BWR transient analysis
Nikitin, K., Clifford, I., & Ferroukhi, H. (2024). Minimal requirements to thermal-hydraulics/3D-neutron kinetics clustering for BWR transient analysis. Annals of Nuclear Energy, 195, 110147 (12 pp.). https://doi.org/10.1016/j.anucene.2023.110147
Experimental study for high Reynolds’ gas-liquid two-phase flow through rod bundle using Wire Mesh Sensor (WMS)
Al-Yahia, O. S., Clifford, I., Bissels, W. M., Suckow, D., & Ferroukhi, H. (2023). Experimental study for high Reynolds’ gas-liquid two-phase flow through rod bundle using Wire Mesh Sensor (WMS). Nuclear Engineering and Design, 405, 112224 (23 pp.). https://doi.org/10.1016/j.nucengdes.2023.112224
Implementation of a three-field framework in TRACE to improve the prediction of reactor core reflood conditions part II
Al-Yahia, O. S., Bernard, M., Clifford, I., Perret, G., Bajorek, S., & Ferroukhi, H. (2023). Implementation of a three-field framework in TRACE to improve the prediction of reactor core reflood conditions part II. In 20th international topical meeting on nuclear reactor thermal hydraulics (NURETH-20) (pp. 110-122). https://doi.org/10.13182/NURETH20-40133
TRACE code simulation of the interaction between reactor coolant system and containment building with passive heat removal system
Al-Yahia, O. S., Clifford, I., Nikitin, K., Liu, P., & Ferroukhi, H. (2023). TRACE code simulation of the interaction between reactor coolant system and containment building with passive heat removal system. Nuclear Engineering and Design, 406, 112234 (21 pp.). https://doi.org/10.1016/j.nucengdes.2023.112234
FEM heat transfer modelling with tomography-based SiC<sub>f</sub>/SiC unit cell
Cavaliere, A., Marone, F., Cozzo, C., Buchanan, K., Lorrette, C., & Pouchon, M. A. (2023). FEM heat transfer modelling with tomography-based SiCf/SiC unit cell. High Temperatures-High Pressures, 52(2), 95-110. https://doi.org/10.32908/hthp.v52.1281
State of the art for thermal-hydraulic analysis of pressurised thermal shock scenarios
Clifford, I., Kral, P., Vyskocil, L., Pistora, V., Trewin, R., Filonova, Y., … Roy, J. (2023). State of the art for thermal-hydraulic analysis of pressurised thermal shock scenarios. In 20th international topical meeting on nuclear reactor thermal hydraulics (NURETH-20) (pp. 5778-5790). https://doi.org/10.13182/NURETH20-40247
TRACE investigation on the performance of passive safety condenser as ultimate heat sink
Clifford, I., Al-Yahia, O. S., & Ferroukhi, H. (2023). TRACE investigation on the performance of passive safety condenser as ultimate heat sink. In Proceedings of Saudi international conference on nuclear power engineering (SCOPE) (p. 23220 (6 pp.). SCOPE.
Assessment of the influence of scaling on turbulent mixing in downcomer and core-inlet flow distribution
Fogliatto, E., Puragliesi, R., Clifford, I., & Ferroukhi, H. (2023). Assessment of the influence of scaling on turbulent mixing in downcomer and core-inlet flow distribution. Annals of Nuclear Energy, 185, 109715 (15 pp.). https://doi.org/10.1016/j.anucene.2023.109715
CFD simulations of the UPTF-TRAM test C1
Fogliatto, E., & Clifford, I. (2023). CFD simulations of the UPTF-TRAM test C1. In 20th international topical meeting on nuclear reactor thermal hydraulics (NURETH-20) (pp. 606-617). https://doi.org/10.13182/NURETH20-40211
Assessment of the choked flow model of RELAP5 for the application of inverse quantification methods
Freixa, J., Martínez-Quiroga, V., & Perret, G. (2023). Assessment of the choked flow model of RELAP5 for the application of inverse quantification methods. In 20th international topical meeting on nuclear reactor thermal hydraulics (NURETH-20) (pp. 5426-5439). https://doi.org/10.13182/NURETH20-40674
Analysis of thermal fuel behaviour under steady-state irradiation using selected cases of the IAEA CRP FUMEX
Khvostov, G. (2023). Analysis of thermal fuel behaviour under steady-state irradiation using selected cases of the IAEA CRP FUMEX. Journal of Nuclear Materials, 584, 154589 (8 pp.). https://doi.org/10.1016/j.jnucmat.2023.154589
Calibration and validation of thermal fuel behaviour models based on the first case of the first IAEA CRP FUMEX
Khvostov, G. (2023). Calibration and validation of thermal fuel behaviour models based on the first case of the first IAEA CRP FUMEX. Journal of Nuclear Materials, 584, 154588 (11 pp.). https://doi.org/10.1016/j.jnucmat.2023.154588
Insights into fuel behaviour during relatively fast thermal transients based on calculations for two tests of the Halden IFA-507 experiment
Khvostov, G. (2023). Insights into fuel behaviour during relatively fast thermal transients based on calculations for two tests of the Halden IFA-507 experiment. Nuclear Engineering and Technology, 55(10), 3801-3807. https://doi.org/10.1016/j.net.2023.06.045
 

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