CINECA IRIS Institutional Research Information System

D'AURIA, FRANCESCO SAVERIO
 Distribuzione geografica
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Continente #
NA - Nord America 51.746
EU - Europa 28.395
AS - Asia 11.286
AF - Africa 378
SA - Sud America 308
OC - Oceania 146
Continente sconosciuto - Info sul continente non disponibili 42
Totale 92.301
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Nazione #
US - Stati Uniti d'America 49.260
IT - Italia 10.596
CN - Cina 6.159
SE - Svezia 4.420
BG - Bulgaria 3.294
CA - Canada 2.450
UA - Ucraina 2.244
DE - Germania 1.877
VN - Vietnam 1.379
GB - Regno Unito 1.280
TR - Turchia 1.124
CH - Svizzera 1.087
FI - Finlandia 952
FR - Francia 718
RU - Federazione Russa 716
HK - Hong Kong 639
IR - Iran 427
KR - Corea 412
IN - India 257
JP - Giappone 213
CI - Costa d'Avorio 203
AT - Austria 195
ES - Italia 195
NL - Olanda 170
SG - Singapore 164
BE - Belgio 159
BR - Brasile 138
AU - Australia 131
AR - Argentina 114
HU - Ungheria 92
RO - Romania 88
ID - Indonesia 87
CZ - Repubblica Ceca 86
PK - Pakistan 79
IL - Israele 56
TW - Taiwan 44
UZ - Uzbekistan 43
EU - Europa 40
NO - Norvegia 36
MY - Malesia 35
EG - Egitto 33
PL - Polonia 33
GR - Grecia 31
DZ - Algeria 29
MX - Messico 29
PH - Filippine 28
SN - Senegal 27
IE - Irlanda 25
JO - Giordania 23
SA - Arabia Saudita 21
ZA - Sudafrica 20
PE - Perù 19
BD - Bangladesh 17
DK - Danimarca 17
AE - Emirati Arabi Uniti 16
SK - Slovacchia (Repubblica Slovacca) 16
KZ - Kazakistan 14
NZ - Nuova Zelanda 14
PT - Portogallo 14
CO - Colombia 13
HR - Croazia 12
SI - Slovenia 12
CL - Cile 11
NG - Nigeria 11
SD - Sudan 10
GH - Ghana 9
SC - Seychelles 9
TH - Thailandia 9
AF - Afghanistan, Repubblica islamica di 7
KE - Kenya 7
LU - Lussemburgo 7
RS - Serbia 7
EC - Ecuador 6
GE - Georgia 6
LY - Libia 6
AM - Armenia 4
EE - Estonia 4
IQ - Iraq 4
IS - Islanda 4
AZ - Azerbaigian 3
LK - Sri Lanka 3
MO - Macao, regione amministrativa speciale della Cina 3
UG - Uganda 3
VE - Venezuela 3
BA - Bosnia-Erzegovina 2
BO - Bolivia 2
BY - Bielorussia 2
DM - Dominica 2
ET - Etiopia 2
HN - Honduras 2
KW - Kuwait 2
OM - Oman 2
PY - Paraguay 2
XK - ???statistics.table.value.countryCode.XK??? 2
AL - Albania 1
AO - Angola 1
BN - Brunei Darussalam 1
BT - Bhutan 1
BW - Botswana 1
CD - Congo 1
Totale 92.284
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Città #
Woodbridge 10.131
Ann Arbor 6.546
Houston 4.942
Milan 4.621
Sofia 3.286
Jacksonville 3.091
Chandler 3.052
Fairfield 2.596
Ottawa 2.253
Beijing 1.881
New York 1.828
Ashburn 1.655
Serra 1.637
Princeton 1.505
Wilmington 1.263
Lawrence 1.241
Nanjing 1.102
Seattle 1.048
Bern 1.030
Izmir 872
Cambridge 855
Des Moines 730
Medford 671
Dong Ket 632
Dearborn 614
Jüchen 548
Redwood City 439
Nanchang 419
Rome 419
Hong Kong 379
Boulder 375
London 256
Changsha 227
Kunming 223
Abidjan 203
Guangzhou 203
Shenyang 197
San Diego 189
Shanghai 185
Boardman 180
Hebei 172
Vienna 160
Düsseldorf 158
Tianjin 144
Los Angeles 137
Norwalk 135
Wuhan 127
Washington 125
Brussels 124
Hefei 123
Jiaxing 121
Ogden 118
Bremen 98
Gif-sur-yvette 98
Hangzhou 93
Kocaeli 92
Buenos Aires 81
Falls Church 79
Orange 76
Jinan 75
Fremont 71
Indiana 69
Montecatini Terme 69
Frankfurt am Main 67
Pisa 61
Verona 61
Central 60
Chengdu 60
Tokyo 60
Auburn Hills 58
San Jose 56
Helsinki 54
Chicago 53
Lanzhou 50
Central District 48
Xian 48
Zhengzhou 48
Barcelona 45
Istanbul 45
Paris 45
Singapore 43
Seoul 42
West Jordan 42
San Francisco 41
Chongqing 40
Grafing 40
Tappahannock 38
Moscow 37
Council Bluffs 36
Mumbai 36
Brisbane 35
Paks 35
Amsterdam 34
Madrid 33
Turin 33
Phoenix 32
Islamabad 31
Changchun 30
Harbin 30
Dakar 27
Totale 67.703
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Nome #
An overview of Thorium Utilization in Nuclear Reactors and Fuel Cycle 1.032
Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation. IAEA Safety Report Series 592
Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power Plants 428
Scaling in System Thermal-Hydraulics Applications to Nuclear Reactor Safety and Design: a State-of-the-Art Report. 306
Deterministic Safety Analysis for Nuclear Power Plants. IAEA Specific Safety Guide 232
Insights into BEPU 216
OECD/NEA/CSNI/WGAMA PERSEO benchmark: main outcomes and conclusions 216
A general method of predicting critical heat flux in Advanced Water Cooled Reactors 192
Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP3: Achievements and Recommendations Report 189
Numerical Simulation of Free Surface Flows With Heat and Mass Transfer 183
Post-BEMUSE Reflood Model input uncertainty methods (PREMIUM) Benchmark Phase II: identification of influential Parameters 177
"Blowdown experiments from a pressure vessel with internal structures" (in Italian), University of Pisa Report, IIN - RL 317(78), Pisa (I), 174
A proposed Methodology for the Analysis of a Phenomenon in Separate Effects and Integral Test Facilities 173
Methodology for the reliability evaluation of a passive system and its integration into a Probabilistic Safety Assessment 171
A procedure to optimize the timing of operator actions of accident management procedures 171
Methodology for the evaluation of the reliability of passive systems 171
A model for the analysis of pump start-up transients in Tehran Research Reactor 167
"Proposal of planning hydrogen related experiences in PIPER-ONE facility", University of Pisa Report, DCMN - NT 001(82), Pisa (I) 164
OECD CSNI ISP 15: pre-test analysis of a split break transient in FIX-II loop performed at Pisa University by RELAP4/MOD6 code 163
Analysis by RELAP4/MOD6 code of the reference transient in the design of PIPER-ONE loop" (in Italian), University of Pisa Report, IIN - RP 492(81), Pisa (I), 163
Thermal-hydraulic performance of primary system of RBMK in case of accidents 162
Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation 160
PIPER-ONE loop: simulation of heat transfer between fluid and structures during a SBLOCA in a BWR" (in Italian) 159
Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled Thermal-Hydraulic 3D Kinetics code 157
Validation of NEPTUNE CFD Module with Data of a Plunging Water Jet Entering a Free Surface 157
Blowdown two-phase flowrate evaluation from pressure and thrust measurements 156
LOBI test B-302: electric power evaluation and pre-test analysis performed by RELAP4/MOD6 computer code", University of Pisa Report, DCMN - RP 002(82), Pisa (I) 155
MTR benchmark static calculations with MCNP5 code 154
"Thermo-mechanical design of the core of PEC Fast Reactor" (in Italian) - CNEN-DRV Report VT.CC.00065, Bologna (I), Apr. 1978 154
Scaling of Natural Circulation in PWR Systems 153
Evaluation of uncertainties in system thermal-hydraulic calculations and key applications by the CIAU method 153
Uncertainties in Predictions by Thermal-Hydraulic Codes: Approaches and Results 152
The Individual Channel Monitoring (ICM) proposal to improve the safety performance of RBMK 152
Assessment of RELAP5/MOD2 Code on the Basis of Experiments Performed in LOBI Facility 151
Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code 150
Deterministic Safety Technology for RBMK Reactors 150
Analyses of pressure perturbation events in boiling water reactor 148
Flowrate and Density Oscillations During Two-Phase Natural Circulation in PWR Typical Conditions 147
Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code 147
Assessment of 12 CHF Prediction Methods, for an Axially Non-Uniform Heat Flux Distribution, with the RELAP5 Computer Code 147
A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP 146
Critical Flow Modelling in Nuclear Safety 145
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures 145
Post-test analysis of LOBI test SD-SL-03 performed by RELAP4/MOD6 code 145
Density Wave Instabilities in Steam Generators 144
Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme 144
Italian point of view in assessment and validation of large thermal-hydraulic computer codes", University of Pisa Report, DCMN - RL 070(83), Pisa (I), Dec. 1983, 3rd Meet. of the CSNI SACTE Task Group, Paris (F), Nov. 30-Dec. 3, 1983 144
Overview of accident analysis in nuclear research reactors 143
V & V in System Thermal-Hydraulics 143
Natural Circulation Phenomena and Modeling for Advanced Water Cooled Reactors 143
FONESYS : The Forum & Network of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics 143
Assessment study of the coupled code Relap5/Parcs against the Peach Bottom BWR turbine trip test 141
Validation of Neptune CFD Module With Data of a Plunging Water Jet Entering a Free Surface 141
Safety Analysis for Research Reactors 141
Loss of Coolant Accident: parametric analysis of the initial period of the depressurization of a BWR vessel with the use of the RELAP4/MOD5 code" (in Italian) 141
A Procedure for Characterizing the Range of Input Uncertainty Parameters by the Use of FFTBM 141
Problems in the evaluation of thrusts during a Loss of Coolant Accident in a Light Water Reactor" (in Italian) 140
Scaling of complex phenomena in System Thermalhydraulics 139
Application of an optimized AM procedure following a SBO in a VVER-1000 139
Assessment of RELAP5/MOD2 Code on the Basis of Experiments performed in LOBI Facility 138
Advancements in Evaluating Accuracy of Thermalhydraulic Codes Calculations 137
OECD CSNI ISP 15: post-test analysis of a split break transient in FIX-II loop performed at Pisa University by RELAP4/MOD6 and RELAP5/MOD1 codes", University of Pisa Report, DCMN - RL 067(83), Pisa (I), Dec. 1983, OECD CSNI Workshop on ISP 15, Nykoping (S), Dec. 15-16, 1983 137
Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP1: Data Requirements and Databases Needed for Transient Simulations and Qualification 136
Use of coupled code technique for Best Estimate safety analysis of nuclear power plants 136
ATLAS Program & Physical Models of SPACE, Critical evaluation of current status 136
Determination of code accuracy in predicting small break LOCA experiment 135
Sensitivity analysis of the Peach Bottom Turbine Trip 2 experiment 135
The BEMUSE Programme: Results of the first part concerning the LOFT L2-5 test 135
Deterministic accident analysis for RBMK 135
Feasibility analysis of PIPER-ONE loop: a system to simulate SBLOCA in BWRs" (in Italian), University of Pisa Report, IIN - RP 416(80), Pisa (I), Sept. 1980 135
"Determination of electrical power to be supplied to LOBI heaters", University of Pisa Report, DCMN - NT 011(80), Pisa (I), June 1980 135
Application of CFX-10 to the investigation of RPV coolant mixing in VVER reactors 134
Application of REPAS methodology to assess the reliability of passive safety systems 134
Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems 134
Scaling Issues for the Experimental Characterization of Reactor Coolant System in Integral Test Facilities and Role of System Code as Extrapolation Tool 134
PERSEO Benchmark - UNIPI results 134
Fluiddynamic Analysis of Steam-Water Flows from a Pressure Vessel 133
International training program: 3D SUNCOP – Scaling, Uncertainty and 3D thermal-hydraulic / Neutron kinetics coupled codes Seminar 133
Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core 132
Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event 132
CFD Analysis of a Slug Mixing Experiment Conducted on a VVER-1000 Model 132
The utilization of thorium in Small Modular Reactors – Part I: Neutronic assessment 132
Study on The Relap5 Break Models Making Reference to the Lobi BL-30 Experiment 131
Capabilities of Transuranus Code in Simulating Power Ramp Tests from the IFPE Database 131
Density Measurements for Two-Phase Critical Flows" (in Italian), 38th Conf. Italian Thermo-technic Association (ATI), Bari (I), Sept. 28-30 1983 131
Proposed Set of Criteria in Designing Nuclear Power Plants Experimental Simulators 130
Effect of Steam Generator Heat Transfer upon Core Reflood in a PWR 130
Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis 130
Analysis of the influence of the upper downcomer nodalization on the LOBI test SD-SL-03 using RELAP4/MOD6 130
The REPAS study: Reliability evaluation of passive safety systems 129
Three-Dimensional Thermal-Hydraulics Analysis of ROCOM Mixing Experiment By Relap5-3d© Code 129
Thermo-Hydraulic Calculation and Safety Analysis of the Tehran Research Reactor 129
RBMK Fuel Channel Blockage Reactivity Analysis by MCNP5 and RELAP5-3D© codes 128
Main Results of Phase IV BEMUSE Project. Simulation of LBLOCA in a NPP 128
PREMIUM, a benchmark on the quantification of the uncertainty of the physical models in the system thermal-hydraulic codes: methodologies and data review 128
Thermal-hydraulic design of PIPER-ONE loop", University of Pisa Report, DCMN - RL 003(82), Pisa (I) 128
Report of the Uncertainty Methods Study for advanced best estimate thermalhydraulic code applications 127
Assessment of RELAP5 Model for the University of Massachusetts Lowell Research Reactor 127
Application of coupled code technique to a safety analysis of a standard MTR research reactor 127
Brotini P., Carbone C., D'Auria F., DeSanti G., Mazzini M., Oriolo F., "LOBI test A2-55: electric power evaluation and preliminary analysis of results obtained by RELAP4/MOD6 code", University of Pisa Report, IIN - RP 472(81), Pisa (I) 127
Totale 16.466
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Categoria #
all - tutte 203.235
article - articoli 0
book - libri 0
conference - conferenze 0
curatela - curatele 0
other - altro 0
patent - brevetti 0
selected - selezionate 0
volume - volumi 0
Totale 203.235
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Totale Lug Ago Sett Ott Nov Dic Gen Feb Mar Apr Mag Giu
2018/20191.845 0 0 0 0 0 0 0 0 0 0 1.074 771
2019/202013.383 2.041 1.549 1.620 619 1.147 1.113 1.347 595 1.203 735 1.215 199
2020/20218.578 824 219 755 339 860 524 759 650 622 840 447 1.739
2021/202213.257 305 1.035 324 924 2.700 2.150 305 661 513 322 789 3.229
2022/20239.882 1.801 810 501 948 1.392 1.452 266 754 1.249 105 406 198
2023/202413.976 2.625 1.920 1.945 1.255 2.283 2.310 338 407 291 565 37 0
Totale 93.962