Active Filters

  • (-) PSI Groups = 4103 System Behaviour
Search Results 1 - 20 of 50
Select Page
Measurement of the gas velocity in a water-air mixture in CROCUS using neutron noise techniques
Hursin, M., Pakari, O., Perret, G., Frajtag, P., Lamirand, V., Pázsit, I., … Pautz, A. (2020). Measurement of the gas velocity in a water-air mixture in CROCUS using neutron noise techniques. Nuclear Technology. https://doi.org/10.1080/00295450.2019.1701906
Analytical criteria for fuel fragmentation and burst FGR during a LOCA
Khvostov, G. (2020). Analytical criteria for fuel fragmentation and burst FGR during a LOCA. Nuclear Engineering and Technology. https://doi.org/10.1016/j.net.2020.03.009
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
Applying SHARK-X to perform data assimilation with the LWR-PROTEUS Phase II integral experiments
Siefman, D., Hursin, M., Perret, G., & Pautz, A. (2020). Applying SHARK-X to perform data assimilation with the LWR-PROTEUS Phase II integral experiments. Progress in Nuclear Energy, 121, 103245 (9 pp.). https://doi.org/10.1016/j.pnucene.2020.103245
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
Measurement of the gas velocity in a water-air mixture in crocus by neutron noise technique
Hursin, M., Pakari, O., Perret, G., Frajtag, P., Lamirand, V., Pázsit, I., … Pautz, A. (2019). Measurement of the gas velocity in a water-air mixture in crocus by neutron noise technique. Presented at the International conference on mathematics and computational methods applied to nuclear science and engineering (M&C 2019). Portland, OR, USA: American Nuclear Society.
Uncertainty quantification of LWR-proteus phase II experiments using CASMO-5
Hursin, M., Park, J., Kim, W., Siefman, D., Perret, G., Vasiliev, A., … Lee, D. (2019). Uncertainty quantification of LWR-proteus phase II experiments using CASMO-5. Presented at the International conference on mathematics and computational methods applied to nuclear science and engineering (M&C 2019). Portland, OR, USA.
Numerical simulation of the effects of localized cladding oxidation on LWR fuel rod design limits using a SLICE-DO model of the FALCON code
Khvostov, G. (2019). Numerical simulation of the effects of localized cladding oxidation on LWR fuel rod design limits using a SLICE-DO model of the FALCON code. Nuclear Engineering and Technology. https://doi.org/10.1016/j.net.2019.07.010
An experimental programme optimized with uncertainty propagation: PETALE in the CROCUS reactor
Lamirand, V., Laureau, A., Rochman, D., Perret, G., Gruel, A., Leconte, P., … Pautz, A. (2019). An experimental programme optimized with uncertainty propagation: PETALE in the CROCUS reactor. In O. Serot & A. Chebboubi (Eds.), EPJ web of conferences: Vol. 211. WONDER-2018 - 5th international workshop on nuclear data evaluation for reactor applications. https://doi.org/10.1051/epjconf/201921103003
BWR-4 ATWS modeling with RELAP5-S3K
Nikitin, K., Mueller, P., Bruder, A., Walser, S., Judd, J., & Hiltbrand, D. (2019). BWR-4 ATWS modeling with RELAP5-S3K. Nuclear Engineering and Design, 344, 38-45. https://doi.org/10.1016/j.nucengdes.2019.01.023
Uncertainty quantification of LWR-PROTEUS Phase II experiments using CASMO-5
Park, J., Kim, W., Hursin, M., Perret, G., Vasiliev, A., Rochman, D., … Lee, D. (2019). Uncertainty quantification of LWR-PROTEUS Phase II experiments using CASMO-5. Annals of Nuclear Energy, 131, 9-22. https://doi.org/10.1016/j.anucene.2019.03.023
Benchmark Monte Carlo calculations with ENDF/B-VIII.0 and JEFF-3.3 libraries for LWR criticality safety assessments
Pecchia, M., Vasiliev, A., Perret, G., & Ferroukhi, H. (2019). Benchmark Monte Carlo calculations with ENDF/B-VIII.0 and JEFF-3.3 libraries for LWR criticality safety assessments. Presented at the ICNC 2019 - 11th international conference on nuclear criticality safety. Paris, France.
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
Assessment of a URANS CFD model for gravity driven flows: a comparison with OECD/PKL2 ROCOM experiments
Puragliesi, R. (2019). Assessment of a URANS CFD model for gravity driven flows: a comparison with OECD/PKL2 ROCOM experiments. Nuclear Engineering and Design. https://doi.org/10.1016/j.nucengdes.2019.110365
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
Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes – PREMIUM benchmark
Skorek, T., de Crécy, A., Kovtonyuk, A., Petruzzi, A., Mendizábal, R., de Alfonso, E., … Pautz, A. (2019). Quantification of the uncertainty of the physical models in the system thermal-hydraulic codes – PREMIUM benchmark. Nuclear Engineering and Design, 354, 110199 (23 pp.). https://doi.org/10.1016/j.nucengdes.2019.110199
First tests of a gamma-blind fast neutron detector prototype based on ZnS and wavelength-shifting fibers
Wolfertz, A., Adams, R., & Perret, G. (2019). First tests of a gamma-blind fast neutron detector prototype based on ZnS and wavelength-shifting fibers. Presented at the ANS annual meeting 2019. The value of nuclear. Minneapolis, USA.