PORTO @ Archivio Istituzionale della Ricerca

ABRATE, NICOLO'
 Distribuzione geografica
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Continente #
EU - Europa 217
NA - Nord America 18
AS - Asia 6
AF - Africa 2
Totale 243
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Nazione #
IT - Italia 169
IE - Irlanda 18
US - Stati Uniti d'America 18
FR - Francia 12
DE - Germania 9
CH - Svizzera 4
CN - Cina 4
CZ - Repubblica Ceca 2
GB - Regno Unito 2
IN - India 2
ZA - Sudafrica 2
NO - Norvegia 1
Totale 243
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Città #
Turin 84
Dublin 18
Torino 18
Milan 6
Bremen 5
Napoli 4
Bottanuco 3
Gif-sur-yvette 3
Houston 3
Parma 3
Somma Vesuviana 3
Abingdon 2
Andover 2
Arenzano 2
Asti 2
Brescia 2
Bryanston 2
Mumbai 2
Munich 2
Naples 2
Omegna 2
Piacenza 2
Rivalta di Torino 2
Varese 2
Ann Arbor 1
Augusta 1
Beinasco 1
Binasco 1
Canegrate 1
Carentino 1
Darien 1
Dronero 1
Geneva 1
Gissi 1
Grazzano Badoglio 1
Milpitas 1
Ne 1
Palo Alto 1
Piscataway 1
Rome 1
Rothrist 1
San Donato Milanese 1
San Giuliano Milanese 1
Stavanger 1
Verzuolo 1
Totale 197
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Nome #
Neutronic benchmark of the FRENETIC code for the multiphysics analysis of lead fast reactors, file e384c431-fb78-d4b2-e053-9f05fe0a1d67 48
STUDY OF THE EIGENVALUE SPECTRA OF THE NEUTRON TRANSPORT PROBLEM IN PN APPROXIMATION, file e384c432-67e0-d4b2-e053-9f05fe0a1d67 33
Coupled modelling of the EBR-II SHRT-45R including photon heat deposition, file e384c432-3398-d4b2-e053-9f05fe0a1d67 23
Methods for safety and stability analysis of nuclear systems, file 30949454-6667-40e1-9473-f0be044b797e 19
Methods for safety and stability analysis of nuclear systems, file 42333a58-ccc7-4fd5-b704-9a685a41ae4a 9
A non-intrusive reduced order model for light water reactor core stability analysis, file e384c434-ba9d-d4b2-e053-9f05fe0a1d67 9
Formulation of the Density Eigenvalue Problemin Neutron Transport for Relevant Engineering Applications, file af06dd72-942e-4d22-b47e-848bb757e932 7
Convergence acceleration aspects in the solution of the PN neutron transport eigenvalue problem, file 390249dd-4b6e-48db-b17f-820bfe738c98 6
Genetic algorithm-based optimisation of the few-group structure for lead fast reactor analysis, file 6f9f7fc4-2e5e-467a-ba78-8135016fb45d 6
Neutronic Analysis of the Fusion Reactor ARC: Monte Carlo Simulations with the Serpent Code, file e384c434-8e3a-d4b2-e053-9f05fe0a1d67 6
Modelling the PbLi flow including tritium transport and permeation with GETTHEM, file f45af58c-ec11-45e8-9577-63e1fbe47cba 6
A non-intrusive reduced order model for neutronic transient analyses of the ALFRED reactor, file 95fcbc20-560b-4cfa-beef-1860883e4712 5
Monte Carlo modelling of the ARC Breeding Blanket with the Serpent code , file 12692cad-26c9-4a3a-8808-0f13e6d189a2 4
A non-intrusive reduced order model for neutronic transient analyses of the ALFRED reactor, file 230f3ca9-5ab5-4b1e-82c6-ea3353e511cc 4
Genetic algorithm-based optimisation of the few-group structure for lead fast reactor analysis, file 482b8abe-02e9-4ed0-9643-7956ba43c784 4
Code-to-code SIMMER/FRENETIC comparison for the neutronic simulation of lead-cooled fast reactors, file 8e25a240-307f-4808-8335-7022d7ed5fef 4
null, file e285d67b-1bc8-4043-80d1-099e98143758 4
Neutronic benchmark of the FRENETIC code for the multiphysics analysis of lead fast reactors, file e384c431-d0c1-d4b2-e053-9f05fe0a1d67 4
Formulation of the Density Eigenvalue Problemin Neutron Transport for Relevant Engineering Applications, file 1d271de8-4983-47d8-8750-10e386341e80 3
Nuclear data uncertainty quantification in the ARC fusion reactor, file 4f3d3d8b-b173-44a4-89f4-0a477330fcf3 3
Code-to-code SIMMER/FRENETIC comparison for the multiphysics simulation of the ALFRED core, file 5c8fe165-843a-417b-b968-2bbb7d58fbc5 3
Uncertainty quantification in FRENETIC calculations of ALFRED lead-cooled fast reactor, file 9f55998d-9a41-4672-9ef2-de9184b71c69 3
Genetic algorithm-based optimisation of the few-group structure for lead fast reactor analysis, file d638031b-be7a-4d69-bf25-09a6b14945b4 3
Convergence acceleration aspects in the solution of the PN neutron transport eigenvalue problem, file ff5be89a-ba63-4f00-a1cd-0faaaea9d0e0 3
Code-to-code SIMMER/FRENETIC comparison for the neutronic simulation of lead-cooled fast reactors, file 0ef2798c-9fe3-4985-960f-3b893a922aba 2
A Nonintrusive Nuclear Data Uncertainty Propagation Study for the ARC Fusion Reactor Design, file 862462e3-0ef5-4697-a6a8-f027305949fc 2
A novel approach combining bootstrapped non-intrusive reduced order models and unscented transform for the robust and efficient CFD analysis of accidental gas releases in congested plants, file 9d94657e-c330-4026-8493-13d72aa00f39 2
Spectrum features of the eigenvalue formulations in neutron transport, file da130096-0e89-4d5b-8816-d2b18a9b59bf 2
A generalised eigenvalue formulation for core-design applications, file dee30e1b-53fd-4fae-a81d-91dfcae02e20 2
NUCLEAR DATA UNCERTAINTY QUANTIFICATION IN MOLTEN SALT REACTORS WITH XGPT, file e384c432-4592-d4b2-e053-9f05fe0a1d67 2
Generalized perturbation techniques for uncertainty quantification in lead-cooled fast reactors, file e384c434-331a-d4b2-e053-9f05fe0a1d67 2
A Nonintrusive Nuclear Data Uncertainty Propagation Study for the ARC Fusion Reactor Design, file 0a6c66b7-5040-45ad-841a-3c4a4333a94a 1
A generalised eigenvalue formulation for core-design applications, file 2aea360c-2176-42e0-8b5f-cdf1c74da465 1
Alternative Weighting Functions for Group Collapsing of Nuclear Data in Lead Fast Reactors Time-dependent Analyses, file 309a8330-c9a6-46d1-9f8a-b80e5aaf46fc 1
Assessment of numerical methods for the evaluation of higher-order harmonics in diffusion theory, file 379a27cb-23ee-47a6-99cd-fb73d94918ab 1
Non-intrusive Uncertainty Propagation in the ARC Fusion Reactor through the nemoFOAM Multi-physics Tool, file 3bb86094-b62b-42e7-834b-3d0a6416e81b 1
Nuclear Data Uncertainty Propagation for the Molten Salt Fast Reactor Design, file 47447643-6917-4560-b52c-1aa86059f986 1
Nuclear Data Uncertainty Propagation for the Molten Salt Fast Reactor Design, file afe51021-6251-4285-ac48-bef992262907 1
null, file b9b66cef-238e-48e9-9304-a3a73ceb5fa3 1
Neutron multiplication and fissile material distribution in a nuclear reactor, file e384c430-c6a9-d4b2-e053-9f05fe0a1d67 1
Eigenvalue Formulations for the PN Approximation to the Neutron Transport Equation, file e384c432-be5b-d4b2-e053-9f05fe0a1d67 1
On some features of the eigenvalue problem for the PN approximation of the neutron transport equation, file e384c434-5ec8-d4b2-e053-9f05fe0a1d67 1
Neutronic Analysis of the Fusion Reactor ARC: Monte Carlo Simulations with the Serpent Code, file e384c434-709a-d4b2-e053-9f05fe0a1d67 1
Methods for safety and stability analysis of nuclear systems, file e62713e6-8834-4ded-a045-b78362ad11d1 1
Totale 246
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Categoria #
all - tutte 516
article - articoli 146
book - libri 0
conference - conferenze 239
curatela - curatele 0
other - altro 0
patent - brevetti 0
selected - selezionate 0
volume - volumi 0
Totale 901
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Totale Lug Ago Sett Ott Nov Dic Gen Feb Mar Apr Mag Giu
2019/20201 0 0 0 0 1 0 0 0 0 0 0 0
2020/202131 2 0 4 7 1 5 2 4 0 1 3 2
2021/202255 6 1 4 6 5 5 3 6 8 1 7 3
2022/2023104 5 5 5 35 10 12 12 9 5 2 3 1
2023/202455 0 2 0 3 2 10 14 1 18 5 0 0
Totale 246