CINECA IRIS Institutional Research Information System

D'AURIA, FRANCESCO SAVERIO
 Distribuzione geografica
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Continente #
NA - Nord America 66.032
AS - Asia 44.953
EU - Europa 32.352
SA - Sud America 7.349
AF - Africa 955
OC - Oceania 260
Continente sconosciuto - Info sul continente non disponibili 53
Totale 151.954
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Nazione #
US - Stati Uniti d'America 63.063
CN - Cina 13.609
SG - Singapore 11.262
IT - Italia 11.184
HK - Hong Kong 9.109
BR - Brasile 6.039
SE - Svezia 4.461
BG - Bulgaria 3.311
VN - Vietnam 3.053
CA - Canada 2.645
KR - Corea 2.525
UA - Ucraina 2.325
DE - Germania 2.180
RU - Federazione Russa 1.945
TR - Turchia 1.861
GB - Regno Unito 1.688
CH - Svizzera 1.127
FI - Finlandia 1.051
FR - Francia 1.018
IN - India 593
AR - Argentina 580
IR - Iran 577
JP - Giappone 432
ID - Indonesia 368
NL - Olanda 327
BD - Bangladesh 311
AT - Austria 309
ES - Italia 279
AU - Australia 233
EC - Ecuador 219
HU - Ungheria 215
CI - Costa d'Avorio 206
BE - Belgio 204
MX - Messico 173
PK - Pakistan 172
ZA - Sudafrica 163
CO - Colombia 141
CZ - Repubblica Ceca 141
IQ - Iraq 139
SA - Arabia Saudita 115
UZ - Uzbekistan 113
EG - Egitto 102
RO - Romania 102
PY - Paraguay 89
BJ - Benin 85
VE - Venezuela 84
PL - Polonia 81
IL - Israele 78
MY - Malesia 75
TW - Taiwan 75
MA - Marocco 73
AE - Emirati Arabi Uniti 71
PE - Perù 70
DZ - Algeria 62
NO - Norvegia 62
CL - Cile 61
KE - Kenya 61
KZ - Kazakistan 61
JO - Giordania 57
PH - Filippine 57
GR - Grecia 48
IE - Irlanda 44
UY - Uruguay 42
EU - Europa 40
SN - Senegal 40
TH - Thailandia 38
LT - Lituania 34
PT - Portogallo 32
SK - Slovacchia (Repubblica Slovacca) 32
DK - Danimarca 30
TN - Tunisia 28
NP - Nepal 27
NZ - Nuova Zelanda 25
JM - Giamaica 21
AZ - Azerbaigian 20
ET - Etiopia 20
RS - Serbia 20
CR - Costa Rica 19
BO - Bolivia 18
GH - Ghana 18
LB - Libano 18
OM - Oman 18
SI - Slovenia 18
TT - Trinidad e Tobago 18
NG - Nigeria 17
BY - Bielorussia 16
DO - Repubblica Dominicana 16
HN - Honduras 15
HR - Croazia 15
SY - Repubblica araba siriana 15
AL - Albania 14
GE - Georgia 13
BB - Barbados 12
AO - Angola 11
GA - Gabon 11
BH - Bahrain 10
LK - Sri Lanka 10
PA - Panama 10
SD - Sudan 10
XK - ???statistics.table.value.countryCode.XK??? 10
Totale 151.745
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Città #
Woodbridge 10.133
Hong Kong 8.755
Ann Arbor 6.546
Singapore 6.093
Ashburn 5.851
Houston 4.967
Milan 4.748
Sofia 3.291
Jacksonville 3.097
Chandler 3.052
Fairfield 2.598
Hefei 2.572
Dallas 2.411
Shanghai 2.292
Beijing 2.283
Ottawa 2.270
Seoul 2.028
New York 1.928
Serra 1.664
Boardman 1.527
Princeton 1.505
Wilmington 1.264
Lawrence 1.243
Santa Clara 1.229
Nanjing 1.110
Seattle 1.081
Bern 1.030
Izmir 879
Cambridge 868
Los Angeles 756
Des Moines 732
Medford 671
Ho Chi Minh City 650
Dong Ket 632
Istanbul 626
Chicago 614
Dearborn 614
Jüchen 548
Buffalo 535
São Paulo 475
Rome 464
Redwood City 439
Nanchang 424
Redondo Beach 411
Boulder 376
Hanoi 354
Council Bluffs 315
Guangzhou 299
London 296
Changsha 251
Kunming 229
Vienna 216
Shenyang 207
Abidjan 206
Rio de Janeiro 200
San Diego 190
Hebei 172
Düsseldorf 169
Tokyo 164
Wuhan 163
Tianjin 161
San Jose 153
Belo Horizonte 146
Brussels 144
Paris 142
Buenos Aires 140
Washington 137
Norwalk 135
Ogden 134
Jiaxing 121
Munich 120
Brasília 118
Hangzhou 105
Porto Alegre 101
Bremen 98
Gif-sur-yvette 98
Frankfurt am Main 97
Curitiba 96
Kocaeli 92
Jinan 91
Dhaka 90
Chengdu 88
Pisa 87
Cotonou 85
Columbus 82
Moscow 82
Tashkent 81
Haiphong 80
Falls Church 79
Zhengzhou 79
Helsinki 78
Budapest 76
Fremont 76
Orange 76
The Dalles 76
Brisbane 75
Nuremberg 75
Guayaquil 72
Indiana 69
Montecatini Terme 69
Totale 104.717
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Nome #
An overview of Thorium Utilization in Nuclear Reactors and Fuel Cycle 2.112
Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation. IAEA Safety Report Series 892
Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power Plants 638
Scaling in System Thermal-Hydraulics Applications to Nuclear Reactor Safety and Design: a State-of-the-Art Report. 540
Insights into BEPU 386
Deterministic Safety Analysis for Nuclear Power Plants. IAEA Specific Safety Guide 374
A general method of predicting critical heat flux in Advanced Water Cooled Reactors 331
"Blowdown experiments from a pressure vessel with internal structures" (in Italian), University of Pisa Report, IIN - RL 317(78), Pisa (I), 313
Methodology for the evaluation of the reliability of passive systems 303
PIPER-ONE loop: simulation of heat transfer between fluid and structures during a SBLOCA in a BWR" (in Italian) 301
OECD/NEA/CSNI/WGAMA PERSEO benchmark: main outcomes and conclusions 296
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), 291
"Proposal of planning hydrogen related experiences in PIPER-ONE facility", University of Pisa Report, DCMN - NT 001(82), Pisa (I) 287
"Thermo-mechanical design of the core of PEC Fast Reactor" (in Italian) - CNEN-DRV Report VT.CC.00065, Bologna (I), Apr. 1978 285
Problems in the evaluation of thrusts during a Loss of Coolant Accident in a Light Water Reactor" (in Italian) 284
OECD CSNI ISP 15: pre-test analysis of a split break transient in FIX-II loop performed at Pisa University by RELAP4/MOD6 code 278
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) 278
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) 270
A proposed Methodology for the Analysis of a Phenomenon in Separate Effects and Integral Test Facilities 267
"Determination of electrical power to be supplied to LOBI heaters", University of Pisa Report, DCMN - NT 011(80), Pisa (I), June 1980 266
Post-BEMUSE Reflood Model input uncertainty methods (PREMIUM) Benchmark Phase II: identification of influential Parameters 265
Natural Circulation Phenomena and Modeling for Advanced Water Cooled Reactors 259
PREMIUM, a benchmark on the quantification of the uncertainty of the physical models in the system thermal-hydraulic codes: methodologies and data review 257
ATLAS Program & Physical Models of SPACE, Critical evaluation of current status 256
Post-test analysis of LOBI test SD-SL-03 performed by RELAP4/MOD6 code 255
Density Measurements for Two-Phase Critical Flows" (in Italian), 38th Conf. Italian Thermo-technic Association (ATI), Bari (I), Sept. 28-30 1983 255
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 254
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 254
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) 254
Numerical Simulation of Free Surface Flows With Heat and Mass Transfer 251
Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP3: Achievements and Recommendations Report 247
Analysis of the influence of the upper downcomer nodalization on the LOBI test SD-SL-03 using RELAP4/MOD6 246
Blowdown two-phase flowrate evaluation from pressure and thrust measurements 245
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 242
Nuclear Energy and its History: Past Consequences, Present Inadequacies and a Perspective for Success 240
Pressurized Heavy Water Reactors – Atucha II 233
V & V in System Thermal-Hydraulics 231
Design of PIPER-ONE loop -Final Version- " (in Italian) 231
Methodology for the reliability evaluation of a passive system and its integration into a Probabilistic Safety Assessment 229
Thermal-hydraulic design of PIPER-ONE loop", University of Pisa Report, DCMN - RL 003(82), Pisa (I) 224
Neutronics/Thermal-hydraulics Coupling in LWR Technology – CRISSUE-S WP1: Data Requirements and Databases Needed for Transient Simulations and Qualification 220
Conversion of Small Modular Reactors Fuel to Use Mixed (U-Th)O2 Fuel 220
A methodology for the qualification of thermalhydraulic codes Nodalizations 219
Assessment of RELAP5/MOD2 Code on the Basis of Experiments Performed in LOBI Facility 217
A model for the analysis of pump start-up transients in Tehran Research Reactor 217
V&V&C in nuclear reactor thermal-hydraulics 214
A procedure to optimize the timing of operator actions of accident management procedures 213
PERSEO Benchmark - UNIPI results 210
Safety Analysis for Research Reactors 206
Scaling of Natural Circulation in PWR Systems 203
Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled Thermal-Hydraulic 3D Kinetics code 203
Critical Flow Modelling in Nuclear Safety 202
Assessment of RELAP5/MOD2 Code on the Basis of Experiments performed in LOBI Facility 202
Use of the Natural Circulation Flow Map for Natural Circulation Systems Evaluation 202
Deterministic Safety Technology for RBMK Reactors 200
Thermal-hydraulic performance of primary system of RBMK in case of accidents 199
Flowrate and Density Oscillations During Two-Phase Natural Circulation in PWR Typical Conditions 197
Advancements in Evaluating Accuracy of Thermalhydraulic Codes Calculations 197
MTR benchmark static calculations with MCNP5 code 197
The Individual Channel Monitoring (ICM) proposal to improve the safety performance of RBMK 197
Fluiddynamic Analysis of Steam-Water Flows from a Pressure Vessel 195
Validation of NEPTUNE CFD Module with Data of a Plunging Water Jet Entering a Free Surface 194
Evaluation of the accuracy of code calculations 193
Density Wave Instabilities in Steam Generators 193
Evaluation of uncertainties in system thermal-hydraulic calculations and key applications by the CIAU method 193
Analyses of pressure perturbation events in boiling water reactor 193
Thermal Hydraulics in Water-Cooled Nuclear Reactors 193
Sensitivity analysis of the Peach Bottom Turbine Trip 2 experiment 192
Effect of Steam Generator Heat Transfer upon Core Reflood in a PWR 191
Scaling of complex phenomena in System Thermalhydraulics 191
Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code 190
Overview of accident analysis in nuclear research reactors 190
FONESYS : The Forum & Network of SYStem Thermal-Hydraulic Codes in Nuclear Reactor Thermal-Hydraulics 190
Uncertainty Methods and Approaches in Nuclear System Safety 190
Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code 189
Scaling Issues for the Experimental Characterization of Reactor Coolant System in Integral Test Facilities and Role of System Code as Extrapolation Tool 189
Assessment study of the coupled code Relap5/Parcs against the Peach Bottom BWR turbine trip test 188
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures 188
Uncertainties in Predictions by Thermal-Hydraulic Codes: Approaches and Results 187
Proposed Set of Criteria in Designing Nuclear Power Plants Experimental Simulators 185
A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP 185
Report of the Uncertainty Methods Study for advanced best estimate thermalhydraulic code applications 184
Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results of the BEMUSE programme 184
A Procedure for Characterizing the Range of Input Uncertainty Parameters by the Use of FFTBM 184
Determination of code accuracy in predicting small break LOCA experiment 182
Validation of Neptune CFD Module With Data of a Plunging Water Jet Entering a Free Surface 182
Analysis of Hannover Experiments on Countercurrent Flow in the Fuel Element Top Nozzle Area 181
The utilization of thorium in Small Modular Reactors – Part I: Neutronic assessment 181
Three-Dimensional Thermal-Hydraulics Analysis of ROCOM Mixing Experiment By Relap5-3d© Code 180
BEMUSE Phase V Report Uncertainty and Sensitivity Analysis of a LBLOCA in Zion Nuclear Power Plant. OECD/NEA Report 180
Summary of SWINTH-2016 : a Specialists Workshop on advanced instrumentation and measurement techniques for nuclear reactor thermal-hydraulic experimentation 180
Assessment of 12 CHF Prediction Methods, for an Axially Non-Uniform Heat Flux Distribution, with the RELAP5 Computer Code 179
RBMK Fuel Channel Blockage Reactivity Analysis by MCNP5 and RELAP5-3D© codes 178
Application of an optimized AM procedure following a SBO in a VVER-1000 178
Application of CFX-10 to the investigation of RPV coolant mixing in VVER reactors 177
Use of coupled code technique for Best Estimate safety analysis of nuclear power plants 177
Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis 176
CSNI Code Validation Matrix of Thermo-Hydraulic Codes for LWR LOCA and Transients 175
Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core 175
Deterministic accident analysis for RBMK 175
Totale 25.462
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Categoria #
all - tutte 384.064
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 384.064
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Totale Lug Ago Sett Ott Nov Dic Gen Feb Mar Apr Mag Giu
2020/20215.581 0 0 0 0 0 524 759 650 622 840 447 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
2023/202416.125 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
2025/202624.556 1.906 6.091 4.943 4.867 3.042 3.707 0 0 0 0 0 0
Totale 153.650