| Nome |
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An overview of Thorium Utilization in Nuclear Reactors and Fuel Cycle, file e0d6c92b-1223-fcf8-e053-d805fe0aa794
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1.178
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Insights into BEPU, file e0d6c92b-4119-fcf8-e053-d805fe0aa794
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1.014
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Scaling Issues for the Experimental Characterization of Reactor Coolant System in Integral Test Facilities and Role of System Code as Extrapolation Tool, file e0d6c928-8ffb-fcf8-e053-d805fe0aa794
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839
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The role of nuclear thermal-hydraulics in the licensing of Atucha-II: the LBLOCA, file e0d6c92d-bcfc-fcf8-e053-d805fe0aa794
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539
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Comparative study between cold leg and hot leg safety injection during SBLOCA in a 4-loop PWR NPP, file e0d6c929-e0ec-fcf8-e053-d805fe0aa794
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518
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Scaling in System Thermal-Hydraulics Applications to Nuclear Reactor Safety and Design: a State-of-the-Art Report., file e0d6c928-9051-fcf8-e053-d805fe0aa794
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381
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The REPAS approach to the evaluation of passive safety system reliability, file e0d6c92a-7ac7-fcf8-e053-d805fe0aa794
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336
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Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors, file e0d6c928-9082-fcf8-e053-d805fe0aa794
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262
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Instrumenting Full scale Boron Injection Test Facility to support Atucha-2 NPP licensing, file e0d6c928-8fd7-fcf8-e053-d805fe0aa794
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238
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Status report on thermal-hydraulic passive systems design and safety assessment, file e0d6c92b-1220-fcf8-e053-d805fe0aa794
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216
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Panel Session: Reliability of Passive Systems, file e0d6c92b-61b1-fcf8-e053-d805fe0aa794
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198
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Conversion of Small Modular Reactors Fuel to Use Mixed (U-Th)O2 Fuel, file e0d6c92b-6c16-fcf8-e053-d805fe0aa794
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193
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A Procedure for Characterizing the Range of Input Uncertainty Parameters by the Use of FFTBM, file e0d6c92a-7f38-fcf8-e053-d805fe0aa794
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176
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Analysis of Channel Blockage of MNSR Reactor Using the System Thermal-Hydraulic Code RELAP5/MOD3.3, file e0d6c928-9941-fcf8-e053-d805fe0aa794
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175
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Analysis of the OSU-MASLWR 001 and 002 tests by using the TRACE code, file e0d6c928-8fd4-fcf8-e053-d805fe0aa794
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143
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Application of best estimate plus uncertainty in review of research reactor safety analysis, file e0d6c929-fcd5-fcf8-e053-d805fe0aa794
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140
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PREMIUM, a benchmark on the quantification of the uncertainty of the physical models in the system thermal-hydraulic codes: methodologies and data review, file e0d6c92a-0bfa-fcf8-e053-d805fe0aa794
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134
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Prioritization of nuclear thermal-hydraulics researches, file e0d6c92b-16c9-fcf8-e053-d805fe0aa794
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132
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Uncertainty Methods and Approaches in Nuclear System Safety, file e0d6c92a-6507-fcf8-e053-d805fe0aa794
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126
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TRENDS AND PERSPECTIVES IN NUCLEAR THERMAL-HYDRAULICS, file e0d6c92e-8e68-fcf8-e053-d805fe0aa794
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124
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International Standard Problem No 50 – The University of Pisa contribution, file e0d6c92a-750f-fcf8-e053-d805fe0aa794
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117
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Independent assessment for new nuclear reactor safety, file e0d6c929-fb53-fcf8-e053-d805fe0aa794
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115
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Limitations in the Use of the Equivalent Diameter, file e0d6c928-8fcb-fcf8-e053-d805fe0aa794
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113
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Independent Assessment for new nuclear reactor safety, file e0d6c929-c737-fcf8-e053-d805fe0aa794
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102
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Nuclear Energy for Sustainable Economic Development, file e0d6c928-8ed3-fcf8-e053-d805fe0aa794
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101
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Introduction of an additional safety barrier for nuclear power reactors, file e0d6c92b-42dd-fcf8-e053-d805fe0aa794
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97
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Introduction of an additional safety barrier for nuclear power reactors, file e0d6c92b-42de-fcf8-e053-d805fe0aa794
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96
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Conjugating ALARA, BEPU, Safety Margins and Independent Assessment in Nuclear Reactor Safety, file e0d6c929-c146-fcf8-e053-d805fe0aa794
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95
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Post-BEMUSE Reflood Model input uncertainty methods (PREMIUM) Benchmark Phase II: identification of influential Parameters, file e0d6c928-8acb-fcf8-e053-d805fe0aa794
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86
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A view on identification of thermal-hydraulic phenomena for validation of best-estimate computer codes, file e0d6c929-e75b-fcf8-e053-d805fe0aa794
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83
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Innovation needs in nuclear reactor safety and risk, file da9444e8-20ab-424f-a6f9-cfd3c5ab2786
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82
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Introduction of a Safety Barrier for Nuclear Power Reactors, file e0d6c92b-008b-fcf8-e053-d805fe0aa794
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78
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Nuclear Fission: from E. Fermi to Adm. Rickover, to industrial exploitation, to nowadays challenges, file e0d6c92e-861d-fcf8-e053-d805fe0aa794
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78
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THE UTILIZATION OF THORIUM-232 IN ADVANCED PWR – FROM SMALL TO BIG REACTORS, file e0d6c92b-325c-fcf8-e053-d805fe0aa794
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77
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The Findings from the OECD/NEA/CSNI UMS (Uncertainty Method Study), file e0d6c92a-1c12-fcf8-e053-d805fe0aa794
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68
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CONUSAF PRESENTATION & FOUNDING MOTIVATIONS, file e0d6c92b-344b-fcf8-e053-d805fe0aa794
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68
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Editorial of J NED Special Issue devoted to SWINTH, file e0d6c92b-60f1-fcf8-e053-d805fe0aa794
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68
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Using of BEPU methodology in a Final Safety Analysis Report, file e0d6c928-8e37-fcf8-e053-d805fe0aa794
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61
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FONESYS and SILENCE Networks: Looking to the Future of T-H Code Development and Experimentation, file e0d6c928-9945-fcf8-e053-d805fe0aa794
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61
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Investigation of Coolant Mixing in Reactor VVER-1000, file e0d6c92a-1c14-fcf8-e053-d805fe0aa794
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55
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NEW SAFETY BARRIER FOR CURRENT AND FUTURE NUCLEAR REACTORS, file e0d6c92b-17ef-fcf8-e053-d805fe0aa794
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55
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The Best-Estimate Plus Uncertainty (BEPU) Challenge in the Licensing of Nuclear Power Plants (NPP), file e0d6c92a-3cf9-fcf8-e053-d805fe0aa794
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48
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BEPU and Safety Margins in Nuclear Reactor Safety, file e0d6c929-c149-fcf8-e053-d805fe0aa794
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46
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Validation of CATHARE TH-SYS Code Against Experimental Reflood Tests, file e0d6c928-9950-fcf8-e053-d805fe0aa794
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44
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[116] THERMAL-HYDRAULIC PHENOMENA, file e0d6c92b-38aa-fcf8-e053-d805fe0aa794
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42
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The need of adding a safety barrier to water cooled nuclear reactors, file e0d6c92d-c43e-fcf8-e053-d805fe0aa794
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42
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Main Results of the OECD BEMUSE Programme, file e0d6c92a-76d4-fcf8-e053-d805fe0aa794
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41
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Analysis of the Multi-Application Small Light-Water Reactor (MASLWR) Design Natural Circulation Phenomena, file e0d6c92a-7972-fcf8-e053-d805fe0aa794
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40
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A vision for nuclear reactor safety, file e0d6c928-8432-fcf8-e053-d805fe0aa794
|
39
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Scaling and Scaling role within the BEPU approach – The μλ-I³TF, file e0d6c928-9053-fcf8-e053-d805fe0aa794
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39
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CIAU Method for Uncertainty Evaluation for System Thermal-Hydraulic Code Calculations, file e0d6c92a-4f69-fcf8-e053-d805fe0aa794
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38
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International Training Program in Support of Safety Analysis: 3C S.UN.COP – Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars, file e0d6c92a-7969-fcf8-e053-d805fe0aa794
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38
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Analysis of break size effects on SBLOCA scenario in a 4 loop PWR using Relap5/mod 3.3, file e0d6c92d-b5a1-fcf8-e053-d805fe0aa794
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37
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BEPU-FSAR: A new paradigm in Nuclear Reactor Safety, file e0d6c929-daab-fcf8-e053-d805fe0aa794
|
36
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ADDING A SAFETY BARRIER FOR EXISTING AND NEW NUCLEAR POWER PLANTS, file e0d6c92b-5fdd-fcf8-e053-d805fe0aa794
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31
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Evaluation of the Simulation of Severe Accidents in Angra 2 Using Melcor1.8.5 Code Compared with RELAP5/MOD3, file e0d6c92a-438c-fcf8-e053-d805fe0aa794
|
30
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Enhanced Nuclear Engineering Simulators (ENES), file e0d6c929-ec09-fcf8-e053-d805fe0aa794
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29
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Assessment of NEPTUNE_CFD Code Capabilities to Simulate Two-Phase Flow in the OECD/NRC PSBT Subchannel Experiments, file e0d6c928-9618-fcf8-e053-d805fe0aa794
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26
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Comments, questions and answer to the passive system status report (> 1000 comments addressed), file e0d6c92e-2793-fcf8-e053-d805fe0aa794
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26
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A ‘new’ Vision for Nuclear Reactor Safety, file e0d6c928-8e3c-fcf8-e053-d805fe0aa794
|
25
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Capabilities of TRANSURANUS Code in Simulating BWR Super-Ramp Project, file e0d6c92a-1cfd-fcf8-e053-d805fe0aa794
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25
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V&V and more in Nuclear Thermal-Hydraulics, file e0d6c92b-2d50-fcf8-e053-d805fe0aa794
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24
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Benchmark Analysis of EBR-II Protected Loss-of-Flow Transient, file e0d6c928-8d12-fcf8-e053-d805fe0aa794
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23
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Perspectives for improving the safety of NPP, file e0d6c928-8ffe-fcf8-e053-d805fe0aa794
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22
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Application of Best Estimate Plus Uncertainty Methodology in a Final Safety Analysis Report of a Generic Plant, file e0d6c928-8bfa-fcf8-e053-d805fe0aa794
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21
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CFD Simulations of a Normally-Impinging Jet from a Circular Nozzle, file e0d6c92a-6b3c-fcf8-e053-d805fe0aa794
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21
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Physical Phenomena in Nuclear Thermal Hydraulics and Current Status, file e0d6c931-91da-fcf8-e053-d805fe0aa794
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21
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The OECD/NEA/CSNI SOAR on scaling (the S-SOAR), file e0d6c928-8fda-fcf8-e053-d805fe0aa794
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19
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Enhanced Nuclear Engineering Simulators, file e0d6c929-e75d-fcf8-e053-d805fe0aa794
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19
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Core Exit Temperature Effectiveness in Accident Management of Nuclear Power Reactor, file e0d6c92a-1fd7-fcf8-e053-d805fe0aa794
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19
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Current possibilities for performing licensing analyses for accidents in NPP, file e0d6c92a-5aa8-fcf8-e053-d805fe0aa794
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19
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Proposal of a BEPU-FSAR, file e0d6c928-91b7-fcf8-e053-d805fe0aa794
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18
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COAUTHOR - a MoU to create a COnsortium of Academics from Universities promoting the use of THORium, file e0d6c92d-b646-fcf8-e053-d805fe0aa794
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17
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Benchmark of Atucha-2 PHWR RELAP5-3D Control Rod Model by Monte Carlo MCNP5 Core Calculation, file e0d6c92a-455e-fcf8-e053-d805fe0aa794
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16
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Best Estimate Plus Uncertainty (BEPU) approach and safety margins in nuclear reactor safety, file e0d6c92a-ff9f-fcf8-e053-d805fe0aa794
|
15
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|
AN ADDITIONAL SAFETY BARRIER FOR EXISTING AND NEW NPP, file e0d6c92b-5d08-fcf8-e053-d805fe0aa794
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15
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A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP, file 0af7fc0d-bdac-40a0-9d1d-905c6f1bd42b
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14
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Data, Data Science, and Data-Driven Thermal- Hydraulics – Panel, file e0d6c928-8b6c-fcf8-e053-d805fe0aa794
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14
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An Integrated Software Platform for Best Estimate Safety Analyses of Nuclear Power Plants, file e0d6c92a-1c16-fcf8-e053-d805fe0aa794
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14
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|
Training on Scaling, Uncertainty, 3-D Coupled Calculations in Nuclear Technology, file e0d6c92a-2f81-fcf8-e053-d805fe0aa794
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14
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Evaluation of uncertainty connected with Transient thermal-Hydraulics Calculations, file e0d6c92a-5389-fcf8-e053-d805fe0aa794
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14
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|
UNIPI STARTING REFERENCE DB, file e0d6c92b-52cc-fcf8-e053-d805fe0aa794
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14
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A MoU to create a COnsortium of Academics from Universities promoting the use of THORrium (COAUTHOR), file d100c14c-1a15-4c4d-936d-dee350c981de
|
13
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Prediction of Void Fraction in PWR Subchannel by CATHARE2 Code, file e0d6c92a-4f67-fcf8-e053-d805fe0aa794
|
13
|
|
Pressurized Heavy Water Reactors – Atucha II, file e0d6c931-ac28-fcf8-e053-d805fe0aa794
|
13
|
|
The BEPU Challenge in Current Licensing of Nuclear Power Reactors, file e0d6c92a-2513-fcf8-e053-d805fe0aa794
|
12
|
|
Identification of Limiting Case Between DBA and Selected BDBA (CL Break Area Sensitivity): A New Model for the Boron Injection System, file e0d6c92a-558e-fcf8-e053-d805fe0aa794
|
10
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|
Acquired experience on organizing 3D S.UN.COP: International Course to support Nuclear Licensing by user training in the areas of Scaling, Uncertainty, and 3D Thermalhydraulics/ Neutron-Kinetics CouPled Codes, file e0d6c92a-770f-fcf8-e053-d805fe0aa794
|
10
|
|
An Enhanced Nuclear Engineering Simulator, file e0d6c92b-0ee6-fcf8-e053-d805fe0aa794
|
10
|
|
Thermal-Hydraulics: from Fundamentals to Applications, file e0d6c928-89d3-fcf8-e053-d805fe0aa794
|
9
|
|
Natural Circulation Limits achievable in a PWR, file e0d6c92a-7008-fcf8-e053-d805fe0aa794
|
9
|
|
Critical flow prediction by system codes – Recent analyses made within the FONESYS network, file e0d6c92f-80ef-fcf8-e053-d805fe0aa794
|
9
|
|
Recent Applications of RELAP5-3D at GRNSPG, file e0d6c928-776d-fcf8-e053-d805fe0aa794
|
8
|
|
SCRED Database to Support BEPU Licensing, file e0d6c928-82f7-fcf8-e053-d805fe0aa794
|
8
|
|
The technological challenge for current generation nuclear reactors, file e0d6c92d-b683-fcf8-e053-d805fe0aa794
|
8
|
|
OECD/NEA/CSNI – MADRID 2020 SM: Towards the planning of the Specialists Meeting, file e0d6c92f-8e89-fcf8-e053-d805fe0aa794
|
8
|
|
Methodology for Pressurized Thermal Shock Analysis in Nuclear Power Plant, file e0d6c928-6fab-fcf8-e053-d805fe0aa794
|
7
|
|
USE AND QUALIFICATION OF THE UMAE UNCERTAINTY METHODOLOGY HAVING AS BASIS THE CATHARE2 CODE, file e0d6c92a-3b9d-fcf8-e053-d805fe0aa794
|
7
|
|
Comparison between RELAP5 Mod3.2 and Mod3.3 in Simulating a Large Break LOCA in Low Power RRs, file e0d6c92d-b5a4-fcf8-e053-d805fe0aa794
|
7
|
|
Passive systems and nuclear thermal-hydraulics, file e0d6c92f-e441-fcf8-e053-d805fe0aa794
|
7
|
| Totale |
10.081 |