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The OFFBEAT multi-dimensional fuel behavior solver
Scolaro, A., Clifford, I., Fiorina, C., & Pautz, A. (2020). The OFFBEAT multi-dimensional fuel behavior solver. Nuclear Engineering and Design, 358, 110416 (16 pp.). https://doi.org/10.1016/j.nucengdes.2019.110416
First assessments of the dynamic gap conductance model in TRACE
Clifford, I., Cozzo, C., & Ferroukhi, H. (2019). First assessments of the dynamic gap conductance model in TRACE. In 18th international topical meeting on nuclear reactor thermal hydraulics (NURETH-18). La Grange Park, USA: American Nuclear Society (ANS).
Studies on the effects of local power peaking on heat transfer under dryout conditions in BWRs
Clifford, I., Pecchia, M., Mukin, R., Cozzo, C., Ferroukhi, H., & Gorzel, A. (2019). Studies on the effects of local power peaking on heat transfer under dryout conditions in BWRs. Annals of Nuclear Energy, 130, 440-451. https://doi.org/10.1016/j.anucene.2019.03.017
Screening analysis for pressurized thermal shock (PTS) transient scenarios
Mukin, R., Clifford, I., Costa Garrido, O., Mora, D. F., Niffenegger, M., Niceno, B., & Ferroukhi, H. (2019). Screening analysis for pressurized thermal shock (PTS) transient scenarios. In 18th international topical meeting on nuclear reactor thermal hydraulics (NURETH-18). La Grange Park, USA: American Nuclear Society (ANS).
PMSYS: plant management system as a part of Swiss Simulation Platform
Nikitin, K., Clifford, I., & Ferroukhi, H. (2019). PMSYS: plant management system as a part of Swiss Simulation Platform. In 18th international topical meeting on nuclear reactor thermal hydraulics (NURETH-18). LaGrange Park, USA: American Nuclear Society (ANS).
Global sensitivity and registration strategy for temperature profile of reflood experiment simulations
Perret, G., Wicaksono, D., Clifford, I. D., & Ferroukhi, H. (2019). Global sensitivity and registration strategy for temperature profile of reflood experiment simulations. Nuclear Technology, 205(12), 1638-1651. https://doi.org/10.1080/00295450.2019.1591154
Comparison of computational fluid dynamics and subchannel numerical solutions of fuel assemblies characterized by bowing
Puragliesi, R., Mukin, R., Clifford, I., Ferroukhi, H., & Seidl, M. (2019). Comparison of computational fluid dynamics and subchannel numerical solutions of fuel assemblies characterized by bowing. In 18th international topical meeting on nuclear reactor thermal hydraulics (NURETH-18). La Grange Park: American Nuclear Society (ANS).
Cladding plasticity modeling with the multidimensional fuel performance code OFFBEAT
Scolaro, A., Clifford, I., Fiorina, C., & Pautz, A. (2019). Cladding plasticity modeling with the multidimensional fuel performance code OFFBEAT. In Global/Top Fuel 2019. LaGrange Park, IL, USA: American Nuclear Society (ANS).
Coupling methodology for the multidimensional fuel performance code OFFBEAT and the Monte Carlo neutron transport code Serpent
Scolaro, A., Robert, Y., Fiorina, C., Clifford, I., & Pautz, A. (2019). Coupling methodology for the multidimensional fuel performance code OFFBEAT and the Monte Carlo neutron transport code Serpent. In Global/Top Fuel 2019. LaGrange Park, IL, USA: American Nuclear Society (ANS).
Simulation of the microfluidic mixing and the droplet generation for 3D printing of nuclear fuels
Shama, A., Pouchon, M. A., & Clifford, I. (2019). Simulation of the microfluidic mixing and the droplet generation for 3D printing of nuclear fuels. Additive Manufacturing, 26, 1-14. https://doi.org/10.1016/j.addma.2018.12.011
On the characteristics of the flow and heat transfer in the core bypass region of a PWR
Clifford, I., Pecchia, M., Puragliesi, R., Vasiliev, A., & Ferroukhi, H. (2018). On the characteristics of the flow and heat transfer in the core bypass region of a PWR. Nuclear Engineering and Design, 330, 117-128. https://doi.org/10.1016/j.nucengdes.2018.01.039
Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE
Mukin, R., Clifford, I., Zerkak, O., & Ferroukhi, H. (2018). Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE. Nuclear Engineering and Technology, 50(3), 356-367. https://doi.org/10.1016/j.net.2017.12.005
Pressurized Thermal Shock (PTS) transient scenarios analysis with TRACE
Mukin, R., Clifford, I., Ferroukhi, H., & Niffenegger, M. (2018). Pressurized Thermal Shock (PTS) transient scenarios analysis with TRACE. In Proceedings of the 26th international conference on nuclear engineering (ICONE26). Thermal hydraulics and safety analyses. https://doi.org/10.1115/ICONE26-81749
Sensitivity of CTF solution to subchannel window size
Mukin, R., Seidl, M., Clifford, I., & Ferroukhi, H. (2018). Sensitivity of CTF solution to subchannel window size. In Proceedings of the 26th international conference on nuclear engineering (ICONE26). Thermal hydraulics and safety analyses. https://doi.org/10.1115/ICONE26-82546
Computational fluid dynamics as a tool for deriving subchannel model parameters - The PSBT case study
Puragliesi, R., Mukin, R., Clifford, I., Ferroukhi, H., & Seidl, M. (2018). Computational fluid dynamics as a tool for deriving subchannel model parameters - The PSBT case study. In Vol. 8. Proceedings of the 26th international conference on nuclear engineering (ICONE26). Computational fluid dynamics (CFD); nuclear education and public acceptance. https://doi.org/10.1115/ICONE26-81743
First steps towards the development of a 3D nuclear fuel behaviour solver with OpenFOAM
Scolaro, A., Clifford, I., Fiorina, C., & Pautz, A. (2018). First steps towards the development of a 3D nuclear fuel behaviour solver with OpenFOAM. In Vol. 3. Proceedings of the 26th international conference on nuclear engineering (ICONE26). Nuclear fuel and material, reactor physics, and transport theory. https://doi.org/10.1115/ICONE26-82381
Thermal hydraulic analysis of PWR assembly bowing using subchannel code COBRA-TF
Mukin, R., Clifford, I., Ferroukhi, H., & Seidl, M. (2017). Thermal hydraulic analysis of PWR assembly bowing using subchannel code COBRA-TF. Presented at the 17th international topical meeting on nuclear reactor thermal hydraulics (NURETH 2017). Xi'an, China.
System code validation series based on a consistent plant nodalisation of the ROSA/LSTF integral test facility using TRACE v5.0 Patch 4
Clifford, I., Zerkak, O., Pautz, A., Yao, H., & Freixa, J. (2016). System code validation series based on a consistent plant nodalisation of the ROSA/LSTF integral test facility using TRACE v5.0 Patch 4. Presented at the 11th international topical meeting on nuclear reactor thermal hydraulics, operation and safety (NUTHOS-11). Gyeongju, Korea.
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). LaGrange Park, IL, USA: American Nuclear Society.
GeN-Foam: A novel OpenFOAM<sup>®</sup> based multi-physics solver for 2D/3D transient analysis of nuclear reactors
Fiorina, C., Clifford, I., Aufiero, M., & Mikityuk, K. (2015). GeN-Foam: A novel OpenFOAM® based multi-physics solver for 2D/3D transient analysis of nuclear reactors. Nuclear Engineering and Design, 294, 24-37. https://doi.org/10.1016/j.nucengdes.2015.05.035