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Modelling guidelines for core exit temperature simulations with system codes
Freixa, J., Martínez-Quiroga, V., Zerkak, O., & Reventós, F. (2015). Modelling guidelines for core exit temperature simulations with system codes. Nuclear Engineering and Design, 286, 116-129. https://doi.org/10.1016/j.nucengdes.2015.02.003
Assessment of OpenFOAM CFD toolbox for gravity driven mixing flows in a Reactor Pressure Vessel
Puragliesi, R., Zerkak, O., Pautz, A., & Zhou, Q. (2015). Assessment of OpenFOAM CFD toolbox for gravity driven mixing flows in a Reactor Pressure Vessel. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 952-965). American Nuclear Society.
Ranking of uncertain parameters for dynamic event tree analysis: a case study based on a Station Black Out scenario
Rahman, S., Karanki, D. R., Epiney, A., Zerkak, O., & Dang, V. N. (2015). Ranking of uncertain parameters for dynamic event tree analysis: a case study based on a Station Black Out scenario. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 5734-5747). American Nuclear Society.
A methodology for global sensitivity analysis of transient code output applied to a reflood experiment model using TRACE
Wicaksono, D., Zerkak, O., & Pautz, A. (2015). A methodology for global sensitivity analysis of transient code output applied to a reflood experiment model using TRACE. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 4862-4879). American Nuclear Society.
Review of multi-physics temporal coupling methods for analysis of nuclear reactors
Zerkak, O., Kozlowski, T., & Gajev, I. (2015). Review of multi-physics temporal coupling methods for analysis of nuclear reactors. Annals of Nuclear Energy, 84, 225-233. https://doi.org/10.1016/j.anucene.2015.01.019
Post-test analysis of OECD/NEA ROSA-2 test 4 using TRACE
Clifford, I., Zerkak, O., & Pautz, A. (2014). Post-test analysis of OECD/NEA ROSA-2 test 4 using TRACE (p. 1182 (12 pp.). Presented at the 10th international topical meeting on nuclear thermal-hydraulics, operation and safety (NUTHOS-10). .
Uncertainty- and sensitivity analysis of COBRA-TF for the simulation of selected OECD/NRC BFBT void experiments
Epiney, A., Zerkak, O., & Pautz, A. (2014). Uncertainty- and sensitivity analysis of COBRA-TF for the simulation of selected OECD/NRC BFBT void experiments (p. 1196 (13 pp.). Presented at the 10th international topical meeting on nuclear thermal-hydraulics, operation and safety (NUTHOS-10). .
Treatment of epistemic and aleatory uncertainties in DET simulations: computational framework with ADS-TRACE
Karanki, D. R., Zerkak, O., & Dang, V. N. (2014). Treatment of epistemic and aleatory uncertainties in DET simulations: computational framework with ADS-TRACE (p. 1173). Presented at the 10th international topical meeting on nuclear thermal-hydraulics, operation and safety (NUTHOS-10). .
Assessment of CFD uRANS models for buoyancy driven mixing flows based on ROCOM experiments
Puragliesi, R., Zerkak, O., & Pautz, A. (2014). Assessment of CFD uRANS models for buoyancy driven mixing flows based on ROCOM experiments (p. 1179). Presented at the 10th international topical meeting on nuclear thermal-hydraulics, operation and safety (NUTHOS-10). .
Exploring variability in reflood simulation results: an application of functional data analysis
Wicaksono, D., Zerkak, O., & Pautz, A. (2014). Exploring variability in reflood simulation results: an application of functional data analysis (p. 1208). Presented at the 10th international topical meeting on nuclear thermal-hydraulics, operation and safety (NUTHOS-10). .
Sensitivity analysis of a bottom reflood simulation using the Morris screening method
Wicaksono, D., Zerkak, O., & Pautz, A. (2014). Sensitivity analysis of a bottom reflood simulation using the Morris screening method (p. 1205). Presented at the 10th international topical meeting on nuclear thermal-hydraulics, operation and safety (NUTHOS-10). .
Thermal-hydraulic system code performance for SBLOCA phenomonelogy using different geometries and scales
Freixa, J., Belaid, S., & Zerkak, O. (2013). Thermal-hydraulic system code performance for SBLOCA phenomonelogy using different geometries and scales (pp. NURETH15-331 (19 pp.). Presented at the 15th international topical meeting on nuclear reactor thermal-hydraulics, NURETH-15. .
Post-analysis of a turbine trip test at a BWR/6 using the TRACE/S3K coupled code
Nikitin, K., Gudmundsson, T., Canepa, S., Zerkak, O., Ferroukhi, H., & Pautz, A. (2013). Post-analysis of a turbine trip test at a BWR/6 using the TRACE/S3K coupled code (pp. NURETH15-431 (12 pp.). Presented at the 15th international topical meeting on nuclear reactor thermal-hydraulics, NURETH-15. .
Simulation of international standard problem ISP-42 phase B using in-house coupled code GOTHIC-TRACE
Papini, D., Zerkak, O., & Prasser, H. (2013). Simulation of international standard problem ISP-42 phase B using in-house coupled code GOTHIC-TRACE. Presented at the 15th international topical meeting on nuclear reactor thermal hydraulics, NURETH-15. Pisa, Italy.
Application of power time-projection on the operator-splitting coupling scheme of the trace/S3K coupled code
Wicaksono, D., Zerkak, O., Nikitin, K., Ferroukhi, H., & Chawla, R. (2013). Application of power time-projection on the operator-splitting coupling scheme of the trace/S3K coupled code. In Vol. 3. International conference on mathematics and computational methods applied to nuclear science and engineering (M&C 2013) (pp. 1748-1760).
Analysis of the umsichtwater hammer benchmark experiment 329 using trace and RELAP5
Barten, W., Jasiulevicius, A., Zerkak, O., & Macian-Juan, R. (2011). Analysis of the umsichtwater hammer benchmark experiment 329 using trace and RELAP5. Multiphase Science and Technology, 23(1), 1-27. https://doi.org/10.1615/MultScienTechn.v23.i1.10
Vessel coolant mixing effects on a PWR Main Steam Line Break transient
Zerkak, O., & Ferroukhi, H. (2011). Vessel coolant mixing effects on a PWR Main Steam Line Break transient. Annals of Nuclear Energy, 38(1), 60-71. https://doi.org/10.1016/j.anucene.2010.08.016
Cross-section modelling effects on pressurised water reactor main steam line break analyses
Ferroukhi, H., Zerkak, O., & Chawla, R. (2009). Cross-section modelling effects on pressurised water reactor main steam line break analyses. Annals of Nuclear Energy, 36(8), 1184-1200. https://doi.org/10.1016/j.anucene.2009.04.006
Analysis of the capability of system codes to model cavitation water hammers: simulation of UMSICHT water hammer experiments with TRACE and RELAP5
Barten, W., Jasiulevicius, A., Manera, A., Macian-Juan, R., & Zerkak, O. (2008). Analysis of the capability of system codes to model cavitation water hammers: simulation of UMSICHT water hammer experiments with TRACE and RELAP5. Nuclear Engineering and Design, 238(4), 1129-1145. https://doi.org/10.1016/j.nucengdes.2007.10.004
Analysis of the Leibstadt power plant condensate and feedwater systems during selected operational transients
Zerkak, O., Coddington, P., & Eitschberger, H. (2007). Analysis of the Leibstadt power plant condensate and feedwater systems during selected operational transients. Nuclear Engineering and Design, 237(11), 1195-1208. https://doi.org/10.1016/j.nucengdes.2007.01.011