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Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors : Volume 3: Procedures and Applications
| Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors : Volume 3: Procedures and Applications |
| Autore | D'Auria Francesco |
| Edizione | [2nd ed.] |
| Pubbl/distr/stampa | San Diego : , : Elsevier Science & Technology, , 2024 |
| Descrizione fisica | 1 online resource (818 pages) |
| Disciplina | 621.4834 |
| Altri autori (Persone) | HassanYassin A |
| Collana | Woodhead Publishing Series in Energy Series |
| Soggetto topico |
Nuclear reactors - Design and construction
Nuclear engineering |
| ISBN |
9780323856096
0323856098 |
| Formato | Materiale a stampa  |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto |
Front Cover -- Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors -- Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors: Volume 3: Procedures and Applications -- 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) -- 18 - Subchannel modeling and codes -- Foreword -- 18.1 Introduction -- 18.1.1 A historical perspective -- 18.2 The framework for subchannel analyses -- 18.2.1 Key approaches for modeling -- 18.2.2 The integration domain -- 18.3 The balance equations -- 18.4 The constitutive models -- 18.4.1 Flow regime map -- 18.4.2 Pressure drop -- 18.4.2.1 Two-phase flow pressure drop -- 18.4.3 Heat transfer models -- 18.4.3.1 CHF models -- 18.4.4 Inter-subchannel exchange mechanisms, decoupling, and modeling -- 18.4.4.1 Decoupling of single-phase flow inter-subchannel exchange terms -- Correlations having general applicability -- Phenomenology and correlations in single-phase flow considering the presence of spacer grids -- 18.4.4.2 Decoupling of two-phase flow inter-subchannel exchange terms -- Void drift -- Two-phase flow turbulent mixing -- Spacer forced two-phase cross-flow -- 18.5 The codes -- 18.5.1 Focus on LMR codes -- 18.6 The validation -- 18.6.1 Experimental challenges in subchannel analysis code validations -- 18.6.2 Specific validation cases and needs -- 18.6.2.1 Modeling and validation needs -- Scaling needs -- 18.7 Applications and achievements -- 18.7.1 The role of CFD modeling and codes -- 18.7.2 The role of system codes modeling -- 18.7.3 Application of subchannel analysis codes to the whole core -- 18.7.4 The ocean motion -- 18.8 Conclusions.
18.8.1 Chapter summary remarks: Subchannel analysis codes limitations -- Exercises and questions -- Acknowledgments -- 19 - Containment thermal hydraulics -- Foreword -- 19.1 Introduction (evolution and role of containment) -- 19.2 Containment in existing water-cooled nuclear reactors -- 19.2.1 PWR containment -- 19.2.2 Containment for BWR -- 19.2.3 Containment in VVER-1000, CANDU, and evolutionary PWR -- 19.2.4 Containment/confinement in VVER-440 -- 19.2.5 Containment/confinement in RBMK -- 19.3 Containment for advanced reactors (AP-1000 and ESBWR) -- 19.3.1 AP-1000 -- 19.3.2 ESBWR -- 19.4 Containment in SMR (NuScale, SMR160, CAREM, SMART, etc.) -- 19.5 Phenomena in the containment during transients -- 19.5.1 Hydrogen behavior in containment -- 19.6 Computer codes for simulation of containment -- 19.7 Scaling of containment phenomena -- 19.8 Test facilities for experimental investigation of containment phenomena -- 19.9 Summary and conclusions -- Exercises and questions -- Acknowledgment -- 20 - Numerical methods in nuclear thermal hydraulics -- Foreword -- 20.1 An introduction to numerical methods: basic concepts on the discretization of partial differential equations -- 20.1.1 Formulation of exact, discrete approximations (DA) -- 20.1.2 Truncation of exact difference approximations (DA) and the equations really solved, local truncation error (TE), and consis ... -- 20.1.3 The introduction of artificial viscosity -- 20.1.4 Phase error in the solution of DA -- 20.1.5 The meaning and control of numerical, non-physical solution oscillations -- 20.2 The solution of parabolic PDE -- 20.2.1 The approximation of the solution of time-dependent problems, step-by-step splitting -- 20.2.2 Explicit and implicit approximations in one and multiple space dimensions: alternating direction implicit (ADI) methods -- 20.3 The solution of elliptic PDE. 20.3.1 Characteristics of the linear system -- 20.3.2 Memory and computational time requirements for the solution of the linear system -- 20.3.3 Basic concepts on iterative methods -- 20.3.4 Stationary iterative methods -- 20.3.5 Krylov space-based iterative methods -- 20.3.5.1 The conjugate gradient method -- 20.3.5.2 Preconditioning -- 20.3.5.3 Matrix-free implementation -- 20.3.5.4 Non-SPD matrices: CG over normal equations -- 20.3.5.5 GMRES (Generalized Minimal Residual [method]) -- 20.3.5.6 Other methods for non-SPD matrices -- 20.3.5.7 Pure three-diagonal systems -- 20.3.5.8 Network three-diagonal systems -- 20.3.5.9 Solution of elliptic equations using ADI methods -- 20.3.6 Parallel implementation of direct and iterative methods -- 20.4 The solution of hyperbolic PDE -- 20.4.1 First-order equations, scalar transport -- 20.4.2 The method of characteristics -- 20.4.3 Numerical approximations to the solution of hyperbolic PDE -- 20.5 The validity of computer codes solutions -- 20.6 Automatic computation of sensitivities to parameters in TH codes -- Exercises and questions -- Acknowledgment -- 21 - Scaling in nuclear thermal hydraulics -- Foreword -- Part 1: Scaling background -- 21.1 Introduction -- 21.1.1 The regulatory role of scaling analyses -- 21.1.2 Scaling objectives and general design framework -- 21.1.3 The executive summary from S-SOAR4 -- 21.1.3.1 Scaling distortion -- 21.1.3.2 Scaling analysis for the safety review process -- 21.1.3.3 Scaling methods -- 21.1.3.4 Role of experiments in scaling -- 21.1.3.5 Counterpart test (CT) and similar test (ST) -- 21.1.3.6 Role and characteristics of the system code -- 21.1.3.7 Scaling in uncertainty methods -- 21.1.3.8 Scaling roadmaps -- 21.1.3.9 Role of CFD tools for multi-dimensional and multi-scale phenomena -- Part 2: Scaling techniques (approaches and methods) -- Outline placeholder. 21.2 Scaling techniques -- 21.2.1 Scaling approaches -- 21.2.2 Scaling methods -- 21.2.2.1 Scaling methods used to investigate system phenomena -- 21.2.3 H2TS, FSA, and DSS scaling methods -- 21.2.3.1 Theory -- 21.2.3.2 Hierarchical two-tiered scaling (H2TS) -- 21.2.3.3 Fractional scaling analysis (FSA) -- 21.2.3.4 Dynamical system scaling (DSS) -- Part 3: Scaling database -- 21.3 Scaling database of experiments -- 21.3.1 Roles and requirements for experiments in scaling -- 21.3.2 Scaling distortion -- 21.3.3 Introduction to SETF -- 21.3.4 Examples of SETF -- 21.3.5 Introduction to IETF -- 21.3.6 Examples of IETF -- 21.3.6.1 Current PWR-related facilities -- 21.3.6.2 Current BWR-related facilities -- 21.3.6.3 Current VVER-related facilities -- 21.3.6.4 Current designs related IETF scaling considerations -- Time scaling -- Height scaling -- Volumetric scaling -- Pressure scaling -- Nuclear core simulator scaling -- Number of loop scaling and main coolant lines scaling -- Fluid scaling -- Recirculation and jet-pump scaling -- 21.3.6.5 Advanced-design-related IETF scaling considerations -- 21.3.7 SETF and IETF for phenomena in containment -- 21.3.7.1 Scaling considerations related to the PCV-IETF PWR -- Time scaling -- Volumetric scaling -- Height scaling -- Material scaling -- Compartment subdivision and interconnection among compartments -- Compartment shape scaling -- Energy-release scaling into PCV -- 21.3.7.2 Advanced reactor design considerations -- Part 4: Scaling achievements -- 21.4 Scaling extrapolation methods -- 21.4.1 General remarks -- 21.4.2 Introduction -- 21.4.2.1 Scaling and integral test facilities -- 21.4.2.2 The scaling issue -- 21.4.2.3 The concept of Kv scaling -- 21.4.2.4 Goals and limitations of Kv scaling -- 21.4.2.5 A literature review of applications of Kv scaling -- 21.4.3 The Kv-scaled SCUP methodology. 21.4.3.1 Scaling of nodalizations -- 21.4.3.2 Validation of the methodology with a counterpart exercise at the PKL and LSTF facilities -- 21.4.4 Applications of the methodology -- 21.4.4.1 Application of the methodology for the qualification of a full NPP model -- 21.4.4.2 The impact of scale on the uncertainties -- 21.4.5 Forthcoming roles of Kv-scaled calculations -- 21.4.5.1 Support to test design using hybrid calculation results -- 21.4.5.2 The impact of scale on the figures of merit -- 21.4.5.3 Perfecting nuclear power plant model qualification -- 21.5 Conclusions and recommendations from S-SOAR6 -- 21.5.1 Key findings -- 21.5.2 Recommendations -- 21.6 Conclusions and achievements -- Exercises and questions -- 22 - Good practices in V& -- V for system thermal-hydraulic codes -- Foreword -- 22.1 Introduction -- 22.1.1 Framework -- 22.2 Scope for the SYS TH code and requirements -- 22.2.1 Domain of simulation -- 22.2.2 Precision objective -- 22.2.3 Attribute for safety analyses -- 22.2.3.1 Scaling requirements -- 22.3 SYS TH code development process -- 22.3.1 Physical models -- 22.3.1.1 Fundamental models for thermal hydraulics -- 22.3.1.2 Special thermal-hydraulics models -- 22.3.1.3 Physical models for non-thermal-hydraulics systems -- 22.3.2 Numerics -- 22.3.3 Code implementation -- 22.3.3.1 Code structure -- 22.3.3.2 Programming -- 22.3.3.3 Software quality engineering (SQE) -- 22.3.4 Code assessment strategy within the development process -- 22.3.4.1 State of the art -- 22.3.5 Code manual -- 22.3.6 Life cycle -- 22.3.6.1 Quality assurance -- 22.4 Verification -- 22.4.1 Numerical algorithm and numerical solution -- 22.4.1.1 Numerical scheme -- 22.4.1.2 Verification matrix for numerical algorithm and solution -- 22.4.1.3 Accuracy definition and numerical error estimation -- 22.4.1.4 Checklist for review and inspection -- 22.4.2 Source code. 22.4.2.1 Tools for verification. |
| Record Nr. | UNINA-9911045226403321 |
D'Auria Francesco
|
||
| San Diego : , : Elsevier Science & Technology, , 2024 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors : Volume 2: Modelling
| Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors : Volume 2: Modelling |
| Autore | D'Auria Francesco |
| Edizione | [2nd ed.] |
| Pubbl/distr/stampa | San Diego : , : Elsevier Science & Technology, , 2024 |
| Descrizione fisica | 1 online resource (1012 pages) |
| Disciplina | 621.4834 |
| Altri autori (Persone) | HassanYassin A |
| Collana | Woodhead Publishing Series in Energy Series |
| Soggetto topico |
Thermal hydraulics
Nuclear reactors |
| ISBN |
9780323856119
032385611X |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| 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& -- 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. |
| Record Nr. | UNINA-9911045227503321 |
|
D'Auria Francesco
|
||
| San Diego : , : Elsevier Science & Technology, , 2024 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Thermal-hydraulics of water cooled nuclear reactors / / Francesco D'Auria
| Thermal-hydraulics of water cooled nuclear reactors / / Francesco D'Auria |
| Autore | D'Auria Francesco |
| Edizione | [1st edition] |
| Pubbl/distr/stampa | London, [England] : , : Academic Press, , 2017 |
| Descrizione fisica | 1 online resource (1,121 pages) : illustrations |
| Disciplina | 539.7213 |
| Soggetto topico |
Water cooled reactors
Thermal hydraulics |
| ISBN |
0-08-100679-9
0-08-100662-4 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910583384303321 |
|
D'Auria Francesco
|
||
| London, [England] : , : Academic Press, , 2017 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||