Urgences médicales : revue européenne d'oxyologie

• Record Nr.: UNINA9910146574103321
• Pubbl/distr/stampa: Paris, : Elsevier, ©1989-
• ISSN: 1872-9290
• Descrizione fisica: 1 online resource
• Soggetti: Emergency medical services; Disaster medicine; Critical Care; Emergency Medical Services; First Aid; Wounds and Injuries; Periodical; Periodicals.
• Lingua di pubblicazione: Francese
• Formato: Materiale a stampa
• Livello bibliografico: Periodico
• Note generali: Title from cover.; Refereed/Peer-reviewed

Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors : Volume 2: Modelling

• Record Nr.: UNINA9911045227503321
• Autore: D'Auria Francesco
• Pubbl/distr/stampa: San Diego : , : Elsevier Science & Technology, , 2024; ©2024
• ISBN: 0-323-85611-X
• Edizione: [2nd ed.]
• Descrizione fisica: 1 online resource (1012 pages)
• Collana: Woodhead Publishing Series in Energy Series
• Altri autori (Persone): HassanY. A
• Disciplina: 621.4834
• Soggetti: Thermal hydraulics; Nuclear reactors
• Lingua di pubblicazione: Inglese
• Formato: Materiale a stampa
• Livello bibliografico: Monografia
• Nota di contenuto: Front Cover -- Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors -- Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors: Volume 2: Modelling -- Copyright -- Dedication -- Contents -- List of contributors -- Contributors for volumes 1, 2 and 3 -- Foreword -- Glossary -- Preface to the first edition of the book -- Preface to the second edition of the book -- Acknowledgments (for the past) and wishes (for the future) -- 10 - Special models for nuclear thermal hydraulics -- Foreword -- 10.1 Introduction -- 10.1.1 Streamlining the content of the chapter -- 10.2 Special process models -- 10.2.1 Two-phase critical flow -- 10.2.1.1 Models in system codes -- 10.2.2 Countercurrent flow limitation -- 10.2.2.1 Correlations for the description of the countercurrent flow limitation phenomenon -- Vertical countercurrent flow -- Horizontal countercurrent flow -- 10.2.2.2 Vertical heterogeneous countercurrent steam-water flow in the downcomer -- 10.2.2.3 Vertical heterogeneous countercurrent steam-water flow through the upper tie plate -- 10.2.2.4 Vertical subscale countercurrent flow through the upper tie plate -- 10.2.2.5 Conclusions regarding CCFL -- 10.2.3 Entrainment and deentrainment -- 10.2.4 Reflood -- 10.2.5 Break flow in branch -- 10.2.5.1 Vapor pull-through -- 10.2.5.2 Upward-oriented break -- 10.2.5.3 Side break -- 10.2.5.4 Models in system codes -- 10.3 Special components -- 10.3.1 Pumps -- 10.3.1.1 Model in system codes -- 10.3.2 ; Separators and dryers -- 10.3.2.1 Model in system codes -- 10.3.3 Accumulators -- 10.3.4 Valves, safety valves, control valves, check valves, and flow limiters -- 10.4 Summary remarks -- Exercises and questions -- 11 - The structure of system thermal-hydraulic code for nuclear reactor applications -- Foreword -- 11.1 Introduction to system codes -- 11.2 The requirements and the domain of simulation.; 11.3 Key features of system codes -- 11.3.1 Best estimate code -- 11.3.2 Safety code -- 11.3.3 Industrial code -- 11.3.4 Some features of the first BE codes -- 11.3.4.1 The RELAP5 code -- 11.3.4.2 TRAC and TRACE codes -- 11.3.4.3 The CATHARE code -- 11.3.4.4 The ATHLET code -- 11.4 The "nodalization" concept: Modeling of systems, components, selected phenomena, and aspects -- 11.4.1 The 1D modules in system codes -- 11.4.2 The tee junctions -- 11.4.3 0D modules in system codes -- 11.4.4 Critical flow -- 11.4.5 Singular pressure losses -- 11.4.6 Countercurrent flow limitation -- 11.4.7 Separators -- 11.4.8 Dryers -- 11.4.9 Pumps -- 11.4.10 Turbines -- 11.4.11 ECC injections -- 11.4.12 Accumulators -- 11.4.13 Valves, safety valves, control valves, check valves, and flow limiters -- 11.4.14 Breaks -- 11.4.15 Spray cooling -- 11.4.16 The 3D modeling of core and pressure vessel in system codes -- 11.4.16.1 The various core modeling scales -- 11.4.16.2 Importance of the nodalization -- 11.5 The numeric solution methods -- 11.5.1 State of the art on numeric schemes in current system codes -- 11.6 The relation between SYS TH code and containment -- 11.7 The relation between SYS TH code, component codes, and subchannel codes -- 11.8 Predicting break flow and choked flow -- 11.8.1 Choked flow in single-phase gas flow -- 11.8.2 Choked flow in two-phase steam water flow -- 11.8.3 Sonic velocity in two-phase flow -- 11.8.3.1 The homogeneous equilibrium model -- 11.8.3.2 Attempts to take the slip ratio into account -- 11.8.3.3 Use of the 1D two-fluid model -- 11.8.4 Observations in two-phase choked flow experiments -- 11.8.5 Choked flow prediction by system codes -- 11.8.5.1 Using 0D choked flow model -- 11.8.5.2 Predicting choked flow with a 1D two-fluid modeling -- 11.9 Predicting two-phase flow in horizontal pipes including stratification.; 11.9.1 Phenomena of interest in HLs and CLs of a PWR -- 11.9.2 Horizontal flow modeling with the two-fluid model -- 11.9.3 Properties of the system of equations for stratified flow -- 11.9.4 Predicting stratification -- 11.9.4.1 Stability of bubbly flow regime -- 11.9.4.2 Interfacial friction -- 11.9.5 Conclusion on predictive capabilities of two-fluid model in horizontal pipes -- 11.9.6 Benchmarking of system codes in horizontal flow -- 11.9.7 Further improvements of flow predictions in horizontal flow -- 11.10 The use of flow regime maps -- 11.10.1 Transition criteria -- 11.10.2 The limitations of flow regime maps -- 11.11 Developing and validating closure relations -- 11.12 Predicting CCFL -- 11.13 Modeling of selected phase change occurrences -- 11.13.1 DCC: Steam injection into liquid -- 11.13.1.1 DCC in pool and in pipe without stratification -- 11.13.1.2 DCC in stratified conditions-Pool or vessel at the free surface -- 11.13.1.3 DCC in stratified conditions-Horizontal pipe until CIWH -- 11.13.1.4 Other DCC situations in horizontal pipes and findings -- 11.13.2 DCC: Liquid injection into steam -- 11.13.2.1 Jet cooling -- 11.13.2.2 Spray cooling -- Spray cooling in containment conditions -- 11.13.3 Flashing in case of fast-rapid depressurization -- 11.13.3.1 The bubble growth -- 11.13.3.2 Mechanistic modeling -- Interfacial transfer terms -- Wall nucleation -- Mechanistic modeling: The reviews by Liao and Lucas -- 11.13.3.3 EoS modeling -- 11.13.3.4 Experimentation -- 11.14 Modeling of pressure wave propagation -- 11.14.1 Relevance in nuclear reactor transient-accident scenarios -- 11.14.2 Insights into ; the physical mechanism of fast depressurization -- 11.14.3 Thermal-hydraulic system codes modeling -- 11.15 Modeling of reflooding in system codes -- 11.15.1 Introduction -- 11.15.2 Scenario of a PWR core reflooding.; 11.15.3 Phenomena in a PWR core reflooding -- 11.15.3.1 Classification of phenomena -- 11.15.3.2 Steam binding -- 11.15.3.3 Oscillatory reflooding -- 11.15.3.4 Thermal-hydraulic phenomena in the core -- Flow regimes and heat transfer regimes -- The film sputtering process -- The droplet behavior in the core -- The effects of spacer grids in reflooding -- The CCFL at top of the core -- The BU quenching and the TD quenching -- 3D effects in the core during reflooding -- Thermo-mechanics of the fuel rods -- 11.15.3.5 TH phenomena in CLs at the breaks in the downcomer and the LP -- TH phenomena in the UP HLs and SGs -- 11.15.4 Modeling of reflooding -- 11.15.4.1 Two-fluid modeling and three-field modeling -- 11.15.4.2 1D modeling and 3D modeling of the core during reflooding -- 11.15.4.3 Modeling of core thermal hydraulics during reflooding -- Quenched region below the BU QF -- Inverse annular and inverse slug flow downstream of the BU QF -- Dispersed flow film boiling -- Modeling of rod quenching -- Insights into convection HT, the Leidenfrost, and the MFB temperatures -- Numeric issues related to core reflooding -- 11.15.4.4 Physical modeling in other components -- 11.15.4.5 Potential compensating errors in reflooding modeling -- 11.15.5 Validation of reflooding model -- 11.15.5.1 Scaling of reflooding experiments -- 11.15.5.2 Requirements for validation of reflooding -- 11.15.6 Perspectives for future progress in simulation of reflooding -- 11.16 Upscaling capabilities of system codes -- 11.16.1 PIRT -- 11.16.2 Scaling -- 11.16.3 Distortion in IET -- 11.16.4 Code upscaling capability -- 11.17 Predictive capabilities of SYS TH codes -- 11.17.1 Status of current system codes -- 11.17.2 Capabilities of system codes seen from the validation -- 11.18 Drift flux, two fluids, and TIA in system codes -- 11.18.1 The point of view of time-scale analysis.; 11.18.2 Comparing drift flux with two-momentum equations -- 11.18.3 Polydispersion effects -- 11.18.4 The two-fluid model -- 11.18.5 Perspective for using TIA in future system codes -- 11.18.6 Three-field models in system codes -- Exercises and questions -- 12 - An overview of computational fluid dynamics and nuclear thermal hydraulics applications -- Foreword -- Part 1. Computational fluid dynamics for nuclear thermal hydraulics: The current overview -- 12.1 Introduction -- 12.1.1 Computational fluid dynamics at OECD/NEA and IAEA -- 12.1.2 Computational fluid dynamics reviews -- 12.1.3 Scope, objective, and structure -- 12.2 CFD analysis procedure -- 12.2.1 Understand the problem (phenomena) and set the simulation strategy -- 12.2.2 Generate the geometry and the mesh -- 12.2.3 Set the boundary and the initial conditions -- 12.2.4 Postprocess and interpret the results -- 12.2.5 Refine the mesh (rerun) and perform a sensitivity analysis (rerun) -- 12.2.6 Document the analysis -- 12.3 Methodological aspects: Physical models -- 12.3.1 Single-phase modeling -- 12.3.2 Two-phase modeling -- 12.4 Characteristics of the CFD analysis -- 12.4.1 Applications -- 12.4.2 Validations -- 12.4.2.1 Capabilities of steady RANS -- 12.4.2.2 Limitations of steady RANS -- 12.4.2.3 Final remarks from EPRI round robin -- 12.5 Summary remarks -- Part 2. Insights into computational fluid dynamics for nuclear power plant applications -- 12.6 CFD-related elements and definitions -- 12.7 CFD methods -- 12.7.1 CFD procedure -- 12.7.2 Problem definition and identification of CFD role -- 12.7.3 Selection of physical models -- 12.7.4 Turbulence models -- 12.7.5 ; Characterization of turbulent situations -- 12.7.6 Coupling -- 12.7.7 CFD assessment: V&amp -- V and UQ -- 12.7.8 CFD experiments -- 12.7.9 Best practice guidelines (CFD application) -- 12.8 Applications of CFD: Examples.; 12.8.1 PWR rod bundle problem: EPRI CFD round-robin benchmark exercise.
• Sommario/riassunto: The 'Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors' is a comprehensive guide focusing on the modeling and simulation aspects of nuclear reactor thermal hydraulics. Edited by Francesco D’Auria and Yassin A. Hassan, the book consolidates contributions from numerous researchers in the field. It emphasizes the intricacies of system codes, computational fluid dynamics, and the validation and verification processes crucial for nuclear reactor safety and design. The volume addresses the application of these methodologies in accident analysis, highlighting the best estimate plus uncertainty approach. This resource is aimed at professionals and researchers in nuclear engineering, providing detailed insights into the thermal hydraulic phenomena associated with nuclear reactors.