628 results on '"D'Auria, FRANCESCO SAVERIO"'
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552. Scaling Capabilities of Thermalhydraulic System Codes
553. A verification or Relap5/mod2 Code using OECD CSNI ISP 26 Standard Problem Calculation (in Croatian)
554. PIPER-ONE Research: Overview of the Experiments carried out
555. Application of the Cathare-2 Code to Integral and Separate Effect Tests
556. Survey of activities for assessing thermalhydraulic aspects of new reactors
557. Application of Relap5/mod2 to PWR International Standard Problems
558. A mechanistic model for counter-current flow at the Upper Tie Plate
559. 'Thermal-hydraulic Analysis of intrinsically safe reactors - Final Report' (in Italian)
560. Characterization of Instabilities during Two-Phase Natural Circulation in PWR Typical Conditions
561. Effect of Steam Generator Heat Transfer upon Core Reflood in a PWR
562. Analysis of Hannover experiments on counter-current flow in the fuel element top nozzle area
563. Flowrate and Density Oscillations During Two-Phase Natural Circulation in PWR Typical Conditions
564. Assessment of RELAP5/MOD2 Code on the Basis of Experiments Performed in LOBI Facility
565. Evaluation of the accuracy of code calculations
566. User and Models Interactions in System Codes Predictions
567. Experience gained in the Operation of PIPER-ONE Apparatus
568. Review of the Activities in the Thermalhydraulic Field carried out at Dipartimento di Costruzioni Meccaniche e Nucleari of the University of Pisa
569. Comparison report of the OECD/CSNI International Standard Problem 21 (PIPER-ONE experiment SB-7)
570. OECD NEA Sample Problem Analysis with RELAP5/MOD1-019 Computer Code
571. Improvement of RELAP5/MOD1 and Application of the New Code Version to ISP 15 Post-Test Analysis
572. Non-Stationary Steam Water Jets
573. State of the Art in CCFL Modelling
574. Assessment of RELAP5/MOD2 Code on the Basis of Experiments performed in LOBI Facility
575. Accuracy in the Prediction of Two-Phase Flow Regimes
576. Application of RELAP5/MOD2 to Design Calculations of More Passive Safety Reactors
577. Conceptual Design of a PWR Experimental Simulator
578. Simulazione dello scambio di calore per conduzione nelle strutture degli impianti nucleari ad acqua leggera in seguito ad incidente [Simulation of Conduction Heat Transfer in the Structures of a Nuclear Plant]
579. Advancements in Evaluating Accuracy of Thermalhydraulic Codes Calculations
580. Evaluation of the Two-Phase Mixture Entropy during Blowdown
581. Blowdown from a Pressure Vessel Containing Steam Water Mixtures
582. Comparison between some constitutive models adopted in CATHARE V1.3 and RELAP5/MOD2 system codes
583. Analysis of Well-Posed Thermal-Hydraulic Models through Numerical Methods Based on the Characteristic Form of the Balance Equations
584. Two-Phase Critical Flow Models
585. Similarity Study of BWR SB-LOCA Counterpart Test
586. Analysis of Experiments performed at the University of Hannover with RELAP5/MOD2 and CATHARE Codes on Fluiddynamic Effects in the Fuel Element Top Nozzle Area during Refilling and Reflooding - Final report. CEC Report
587. Capabilities of System Codes in Simulating Small Break LOCAs in PWRs
588. Mutual Influence of Conduction and Convection Heat Transfer in Nuclear Rod Simulators
589. BLOWAES: a Method to Evaluate Blowdown Two-Phase Flowrate from Pressure and Thrust Measurements
590. Present Capabilities of CCFL Models
591. Analysis of ISP 18 by Relap4/mod6, Relap5/mod1, Relap5/mod1-EUR, Relap5/Mod2 and Cathare 1 v1.3 Codes
592. 'Thermo-mechanical design of the core of PEC Fast Reactor' (in Italian) - CNEN-DRV Report VT.CC.00065, Bologna (I), Apr. 1978
593. CSNI Code Validation Matrix of Thermo-Hydraulic Codes for LWR LOCA and Transients
594. Pressure Drops of Two-Phases Critical Flows
595. Critical Flow Modelling in Nuclear Safety
596. A Distributed Model to Perform a Linear Analysis of Density-Wave Instabilities in a Natural Circulation Loop
597. Evaluation of Fluiddynamic Loads on RPV Internals during Blowdown
598. Loops available in Italy for performing Computer Codes Assessment. CSNI Report
599. Measurement of Two-Phase Critical Flow from a Small Break
600. Advancements in the Assessment of System Codes used for Safety Analyses of Nuclear Power Plants
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