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  • (-) Publication Year = 2006 - 2019
  • (-) Journals = 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16)
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Computational fluid dynamics analysis of the fluid flow and heat transfer in the core bypass region of a PWR
Clifford, I., Vasiliev, A., Zerkak, O., Ferroukhi, H., & Pautz, A. (2015). Computational fluid dynamics analysis of the fluid flow and heat transfer in the core bypass region of a PWR. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 7806-7819). American Nuclear Society.
TRACE/SIMULATE-3K analysis of the NEA/OECD Oskarshamn-2 stability benchmark
Dokhane, A., Zerkak, O., Ferroukhi, H., Gajev, I., Judd, J., & Kozlowski, T. (2015). TRACE/SIMULATE-3K analysis of the NEA/OECD Oskarshamn-2 stability benchmark. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 4757-4770). American Nuclear Society.
Towards a consolidated approach for the validation of plant system codes and models: case study for a BWR fast depressurisation event
Epiney, A., Canepa, S., Zerkak, O., & Ferroukhi, H. (2015). Towards a consolidated approach for the validation of plant system codes and models: case study for a BWR fast depressurisation event. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 2157-2170). American Nuclear Society.
Analysis of Fukushima Daiichi NPP unit 3 with Melcor_2.1
Fernandez-Moguel, L., Rydl, A., & Jaeckel, B. (2015). Analysis of Fukushima Daiichi NPP unit 3 with Melcor_2.1. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 6401-6411). American Nuclear Society.
Numerical study of heat diffusion controlled phase growth in a pressurized liquid
Giustini, G., Murallidharan, J., Sato, Y., Niceno, B., Badalassi, V., & Walker, S. (2015). Numerical study of heat diffusion controlled phase growth in a pressurized liquid. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 4069-4082). American Nuclear Society.
Performance of hydrogen mitigation systems for a scaled accident scenario: overview of ERCOSAM project experimental results for the PANDA facility
Mignot, G., Paranjape, S., Kapulla, R., & Paladino, D. (2015). Performance of hydrogen mitigation systems for a scaled accident scenario: overview of ERCOSAM project experimental results for the PANDA facility. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 5456-5469). American Nuclear Society.
Interface tracking based evaluation of bubble growth rates in high pressure pool boiling conditions
Murallidharan, J., Giustini, G., Sato, Y., Ničeno, B., Badalassi, V., & Walker, S. P. (2015). Interface tracking based evaluation of bubble growth rates in high pressure pool boiling conditions. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 6530-6542). American Nuclear Society.
Combined effects of cooler and spray activation on the hydrogen distribution in the presence of a jet flow
Paladino, D., Kapulla, R., Mignot, G., & Paranjape, S. (2015). Combined effects of cooler and spray activation on the hydrogen distribution in the presence of a jet flow. In Vol. 2. 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 1502-1515). American Nuclear Society.
Parametric study on density stratification erosion caused by a horizontal steam jet interacting with a vertical plate obstruction
Paranjape, S., Kapulla, R., Mignot, G., & Paladino, D. (2015). Parametric study on density stratification erosion caused by a horizontal steam jet interacting with a vertical plate obstruction. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 4490-4503). American Nuclear Society.
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.
Melt-concrete interface heat transfer models and coolability models: PWR analyses with MELCOR/CORCON and CORQUENCH
Rýd, A., Jäckel, B., Klügel, J. U., & Steiner, P. (2015). Melt-concrete interface heat transfer models and coolability models: PWR analyses with MELCOR/CORCON and CORQUENCH. In 16th international topical meeting on nuclear reactor thermal hydraulics (NURETH-16) (pp. 484-498). 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.