CINECA IRIS Institutional Research Information System

D'AURIA, FRANCESCO SAVERIO
 Distribuzione geografica
Salva come PNG Salva come JPEG
Continente #
NA - Nord America 72.290
AS - Asia 50.411
EU - Europa 35.218
SA - Sud America 7.847
AF - Africa 1.130
OC - Oceania 291
Continente sconosciuto - Info sul continente non disponibili 53
Totale 167.240
Salva come PNG Salva come JPEG
Nazione #
US - Stati Uniti d'America 69.138
CN - Cina 14.310
SG - Singapore 12.933
IT - Italia 11.464
HK - Hong Kong 9.286
BR - Brasile 6.289
SE - Svezia 4.496
VN - Vietnam 4.221
BG - Bulgaria 3.315
CA - Canada 2.747
KR - Corea 2.575
UA - Ucraina 2.344
DE - Germania 2.307
FR - Francia 2.218
RU - Federazione Russa 1.989
TR - Turchia 1.932
GB - Regno Unito 1.910
FI - Finlandia 1.545
CH - Svizzera 1.273
JP - Giappone 1.115
IN - India 904
AR - Argentina 660
IR - Iran 591
BD - Bangladesh 420
ID - Indonesia 420
NL - Olanda 406
AT - Austria 338
ES - Italia 322
AU - Australia 260
EC - Ecuador 255
HU - Ungheria 255
PK - Pakistan 219
BE - Belgio 212
IQ - Iraq 210
CI - Costa d'Avorio 208
MX - Messico 201
ZA - Sudafrica 198
CO - Colombia 177
SA - Arabia Saudita 166
CZ - Repubblica Ceca 144
UZ - Uzbekistan 141
MY - Malesia 130
RO - Romania 112
MA - Marocco 110
EG - Egitto 109
PL - Polonia 109
VE - Venezuela 108
PH - Filippine 102
PY - Paraguay 101
CL - Cile 96
IL - Israele 88
BJ - Benin 85
TW - Taiwan 85
PE - Perù 84
AE - Emirati Arabi Uniti 83
KE - Kenya 80
JO - Giordania 72
DZ - Algeria 71
KZ - Kazakistan 66
NO - Norvegia 65
TH - Thailandia 57
IE - Irlanda 56
TN - Tunisia 50
GR - Grecia 49
NP - Nepal 48
UY - Uruguay 48
SN - Senegal 47
EU - Europa 40
JM - Giamaica 39
LT - Lituania 39
PT - Portogallo 39
SK - Slovacchia (Repubblica Slovacca) 32
AZ - Azerbaigian 31
DK - Danimarca 30
ET - Etiopia 29
NZ - Nuova Zelanda 25
OM - Oman 25
CR - Costa Rica 24
SI - Slovenia 24
GH - Ghana 22
LB - Libano 22
NG - Nigeria 22
RS - Serbia 22
BO - Bolivia 21
DO - Repubblica Dominicana 21
TT - Trinidad e Tobago 21
AL - Albania 19
BY - Bielorussia 18
HN - Honduras 18
SY - Repubblica araba siriana 18
GE - Georgia 17
HR - Croazia 16
PS - Palestinian Territory 16
AO - Angola 15
BB - Barbados 15
KW - Kuwait 15
PA - Panama 15
LY - Libia 13
BH - Bahrain 12
GA - Gabon 11
Totale 166.971
Salva come PNG Salva come JPEG
Città #
Woodbridge 10.134
Hong Kong 8.889
Singapore 7.505
Ashburn 6.569
Ann Arbor 6.546
Houston 5.003
Milan 4.781
Sofia 3.293
Jacksonville 3.104
San Jose 3.058
Chandler 3.052
Fairfield 2.600
Hefei 2.573
Dallas 2.522
Beijing 2.363
Shanghai 2.311
Ottawa 2.273
Seoul 2.033
New York 2.001
Serra 1.664
Boardman 1.528
Princeton 1.507
Santa Clara 1.300
Wilmington 1.276
Lawrence 1.243
Nanjing 1.110
Seattle 1.092
Lauterbourg 1.075
Ho Chi Minh City 1.047
Bern 1.030
Izmir 881
Cambridge 868
Los Angeles 842
Tokyo 823
Des Moines 744
Medford 672
Chicago 664
Hanoi 658
Istanbul 642
Dong Ket 632
Dearborn 615
Helsinki 563
Buffalo 550
Jüchen 548
Rome 506
São Paulo 493
Redwood City 440
Nanchang 424
Redondo Beach 411
Council Bluffs 393
Boulder 377
London 320
Guangzhou 303
Orem 298
Changsha 254
Vienna 237
Kunming 231
Rio de Janeiro 210
Abidjan 208
Shenyang 207
San Diego 192
Wuhan 173
Hebei 172
Düsseldorf 169
Tianjin 166
Paris 164
Zurich 157
Belo Horizonte 154
Frankfurt am Main 153
Brussels 150
Buenos Aires 147
Washington 144
Norwalk 138
Ogden 134
Da Nang 124
Jiaxing 124
Brasília 123
Munich 123
Haiphong 121
Manchester 119
Chennai 118
Hangzhou 110
Amsterdam 108
Tashkent 108
Porto Alegre 105
Mumbai 104
Budapest 102
Curitiba 100
Dhaka 99
Bremen 98
Gif-sur-yvette 98
The Dalles 97
Brisbane 96
San Francisco 96
Nuremberg 95
Columbus 94
Jinan 93
Kocaeli 92
Chengdu 90
Pisa 90
Totale 114.436
Salva come PNG Salva come JPEG
Nome #
An overview of Thorium Utilization in Nuclear Reactors and Fuel Cycle 2.393
Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation. IAEA Safety Report Series 1.000
Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power Plants 713
Scaling in System Thermal-Hydraulics Applications to Nuclear Reactor Safety and Design: a State-of-the-Art Report. 626
OECD/NRC PSBT Benchmark: Investigating the CATHARE2 Capability to Predict Void Fraction in PWR Fuel Bundle 531
Insights into BEPU 422
Pressurized Heavy Water Reactors – Atucha II 417
Deterministic Safety Analysis for Nuclear Power Plants. IAEA Specific Safety Guide 407
A general method of predicting critical heat flux in Advanced Water Cooled Reactors 365
"Blowdown experiments from a pressure vessel with internal structures" (in Italian), University of Pisa Report, IIN - RL 317(78), Pisa (I), 355
Atucha II Nuclear Reactor: Design Safety and Licensing 332
PIPER-ONE loop: simulation of heat transfer between fluid and structures during a SBLOCA in a BWR" (in Italian) 330
"Proposal of planning hydrogen related experiences in PIPER-ONE facility", University of Pisa Report, DCMN - NT 001(82), Pisa (I) 324
Methodology for the evaluation of the reliability of passive systems 324
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), 321
"Thermo-mechanical design of the core of PEC Fast Reactor" (in Italian) - CNEN-DRV Report VT.CC.00065, Bologna (I), Apr. 1978 318
Problems in the evaluation of thrusts during a Loss of Coolant Accident in a Light Water Reactor" (in Italian) 312
OECD CSNI ISP 15: pre-test analysis of a split break transient in FIX-II loop performed at Pisa University by RELAP4/MOD6 code 312
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) 309
OECD/NEA/CSNI/WGAMA PERSEO benchmark: main outcomes and conclusions 308
Post-BEMUSE Reflood Model input uncertainty methods (PREMIUM) Benchmark Phase II: identification of influential Parameters 305
"Determination of electrical power to be supplied to LOBI heaters", University of Pisa Report, DCMN - NT 011(80), Pisa (I), June 1980 298
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) 296
ATLAS Program & Physical Models of SPACE, Critical evaluation of current status 296
A proposed Methodology for the Analysis of a Phenomenon in Separate Effects and Integral Test Facilities 290
PREMIUM, a benchmark on the quantification of the uncertainty of the physical models in the system thermal-hydraulic codes: methodologies and data review 288
Natural Circulation Phenomena and Modeling for Advanced Water Cooled Reactors 284
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) 284
Post-test analysis of LOBI test SD-SL-03 performed by RELAP4/MOD6 code 282
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 281
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 280
Density Measurements for Two-Phase Critical Flows" (in Italian), 38th Conf. Italian Thermo-technic Association (ATI), Bari (I), Sept. 28-30 1983 278
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 277
Design of PIPER-ONE loop -Final Version- " (in Italian) 271
Analysis of the influence of the upper downcomer nodalization on the LOBI test SD-SL-03 using RELAP4/MOD6 270
Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP3: Achievements and Recommendations Report 267
Blowdown two-phase flowrate evaluation from pressure and thrust measurements 266
Numerical Simulation of Free Surface Flows With Heat and Mass Transfer 265
A methodology for the qualification of thermalhydraulic codes Nodalizations 248
V & V in System Thermal-Hydraulics 247
Thermal-hydraulic design of PIPER-ONE loop", University of Pisa Report, DCMN - RL 003(82), Pisa (I) 247
Nuclear Energy and its History: Past Consequences, Present Inadequacies and a Perspective for Success 247
Methodology for the reliability evaluation of a passive system and its integration into a Probabilistic Safety Assessment 240
Conversion of Small Modular Reactors Fuel to Use Mixed (U-Th)O2 Fuel 239
Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP1: Data Requirements and Databases Needed for Transient Simulations and Qualification 234
Safety Analysis for Research Reactors 231
V&V&C in nuclear reactor thermal-hydraulics 229
Assessment of RELAP5/MOD2 Code on the Basis of Experiments Performed in LOBI Facility 227
A procedure to optimize the timing of operator actions of accident management procedures 225
PERSEO Benchmark - UNIPI results 225
A model for the analysis of pump start-up transients in Tehran Research Reactor 224
Critical Flow Modelling in Nuclear Safety 217
Deterministic Safety Technology for RBMK Reactors 214
Assessment of RELAP5/MOD2 Code on the Basis of Experiments performed in LOBI Facility 213
Sensitivity analysis of the Peach Bottom Turbine Trip 2 experiment 213
Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation 212
Advancements in Evaluating Accuracy of Thermalhydraulic Codes Calculations 211
Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled Thermal-Hydraulic 3D Kinetics code 210
Validation of NEPTUNE CFD Module with Data of a Plunging Water Jet Entering a Free Surface 210
Scaling of Natural Circulation in PWR Systems 208
Uncertainty Methods and Approaches in Nuclear System Safety 208
Thermal Hydraulics in Water-Cooled Nuclear Reactors 208
Flowrate and Density Oscillations During Two-Phase Natural Circulation in PWR Typical Conditions 207
Report of the Uncertainty Methods Study for advanced best estimate thermalhydraulic code applications 207
Scaling Issues for the Experimental Characterization of Reactor Coolant System in Integral Test Facilities and Role of System Code as Extrapolation Tool 206
A Procedure for Characterizing the Range of Input Uncertainty Parameters by the Use of FFTBM 206
Evaluation of the accuracy of code calculations 205
Thermal-hydraulic performance of primary system of RBMK in case of accidents 205
Fluiddynamic Analysis of Steam-Water Flows from a Pressure Vessel 204
Analyses of pressure perturbation events in boiling water reactor 204
The Individual Channel Monitoring (ICM) proposal to improve the safety performance of RBMK 203
Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code 202
MTR benchmark static calculations with MCNP5 code 202
V&V and more in Nuclear Thermal-Hydraulics 201
Capabilities of Transuranus Code in Simulating Power Ramp Tests from the IFPE Database 200
FONESYS : The Forum & Network of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics 200
Evaluation of uncertainties in system thermal-hydraulic calculations and key applications by the CIAU method 199
Effect of Steam Generator Heat Transfer upon Core Reflood in a PWR 198
Density Wave Instabilities in Steam Generators 198
Scaling of complex phenomena in System Thermalhydraulics 198
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures 198
Validation of Neptune CFD Module With Data of a Plunging Water Jet Entering a Free Surface 198
Assessment study of the coupled code Relap5/Parcs against the Peach Bottom BWR turbine trip test 197
RBMK Fuel Channel Blockage Reactivity Analysis by MCNP5 and RELAP5-3D© codes 197
Overview of accident analysis in nuclear research reactors 197
Proposed Set of Criteria in Designing Nuclear Power Plants Experimental Simulators 196
Uncertainties in Predictions by Thermal-Hydraulic Codes: Approaches and Results 195
Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code 195
Analysis of Hannover Experiments on Countercurrent Flow in the Fuel Element Top Nozzle Area 194
Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code 194
Three-Dimensional Thermal-Hydraulics Analysis of ROCOM Mixing Experiment By Relap5-3d© Code 194
BEMUSE Phase V Report Uncertainty and Sensitivity Analysis of a LBLOCA in Zion Nuclear Power Plant. OECD/NEA Report 191
A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP 191
Determination of code accuracy in predicting small break LOCA experiment 190
Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems 190
Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme 190
Summary of SWINTH-2016 : a Specialists Workshop on advanced instrumentation and measurement techniques for nuclear reactor thermal-hydraulic experimentation 190
Improvement of RELAP5/MOD1 and Application of the New Code Version to ISP 15 Post-Test Analysis 189
Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core 189
Best Estimate Analysis and Uncertainty evaluation of the Angra-2 LBLOCA DBA 189
Totale 28.393
Salva come PNG Salva come JPEG
Categoria #
all - tutte 430.972
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 430.972
Salva come PNG Salva come JPEG


Totale Lug Ago Sett Ott Nov Dic Gen Feb Mar Apr Mag Giu
2020/20211.739 0 0 0 0 0 0 0 0 0 0 0 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
2023/202416.125 2.625 1.920 1.945 1.255 2.283 2.310 338 407 291 565 350 1.836
2024/202532.983 498 1.742 605 1.687 1.446 1.751 2.694 2.387 4.231 5.045 4.441 6.456
2025/202639.846 1.906 6.091 4.943 4.867 3.042 3.924 5.190 2.007 2.452 3.188 1.810 426
Totale 168.940