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Accelerator driven subcritical reactors / H. Nifenecker, O. Meplan and S. David
| Accelerator driven subcritical reactors / H. Nifenecker, O. Meplan and S. David |
| Autore | Nifenecker, H. |
| Pubbl/distr/stampa | Bristol ; Philadelphia : IOP Publishing, 2003 |
| Descrizione fisica | IX, 316 p. : ill. ; 24 cm. |
| Disciplina | 621.4834 |
| Collana | Series in fundamental and applied nuclear physics |
| Soggetto topico | Reattori nucleari |
| ISBN | 0-7503-0743-9 |
| Formato | Materiale a stampa  |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNIBAS-000028099 |
Nifenecker, H.
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| Bristol ; Philadelphia : IOP Publishing, 2003 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. della Basilicata | ||
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Éléments de sûreté nucléaire - Les réacteurs à eau sous pression / / Jean Couturier
| Éléments de sûreté nucléaire - Les réacteurs à eau sous pression / / Jean Couturier |
| Autore | Couturier Jean |
| Pubbl/distr/stampa | [Place of publication not identified] : , : EDP SCIENCES, , 2021 |
| Descrizione fisica | 1 online resource (1248 pages) |
| Disciplina | 621.4834 |
| Soggetto topico | Pressurized water reactors |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | fre |
| Record Nr. | UNINA-9910688589903321 |
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Couturier Jean
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| [Place of publication not identified] : , : EDP SCIENCES, , 2021 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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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) | HassanY. 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 |
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D'Auria Francesco
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| San Diego : , : Elsevier Science & Technology, , 2024 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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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) | HassanY. A |
| Collana | Woodhead Publishing Series in Energy Series |
| Soggetto topico |
Thermal hydraulics
Nuclear reactors |
| ISBN | 0-323-85611-X |
| 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 |
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D'Auria Francesco
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| San Diego : , : Elsevier Science & Technology, , 2024 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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Handbook on thermal hydraulics in water-cooled nuclear reactors . Volume 1 Foundations and principles / / edited by Francesco d'Auria, Yassin A. Hassan
| Handbook on thermal hydraulics in water-cooled nuclear reactors . Volume 1 Foundations and principles / / edited by Francesco d'Auria, Yassin A. Hassan |
| Edizione | [Second edition] |
| Pubbl/distr/stampa | Cambridge, MA : , : Woodhead Publishing is an imprint of Elsevier, , [2024] |
| Descrizione fisica | 1 online resource (932 pages) |
| Disciplina | 621.4834 |
| Collana | Woodhead Publishing Series in Energy |
| Soggetto topico |
Thermal hydraulics
Nuclear reactors Reactors nuclears Hidràulica tèrmica |
| ISBN | 0-323-85607-1 |
| 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 1: Foundations and Principles -- Dedication -- Copyright -- 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) -- 1 - Introduction -- Foreword -- 1.1 Introduction -- 1.1.1 Scope and framework -- 1.1.1.1 Origin of nuclear thermal hydraulics -- 1.1.1.2 Single- and two-phase flows -- 1.1.1.3 Prominent scientists, origins, and textbooks -- 1.1.1.4 Journals, conferences, the web, and selected international institutions -- 1.1.1.5 Education and training -- 1.1.1.6 Target nuclear power plant and research reactor types and structures, systems, and components of nuclear installations -- 1.1.1.7 Experiments and instrumentation -- 1.1.1.8 Numerical methods and computer science -- 1.1.1.9 Nuclear safety, licensing process, and design basis accident (moving) boundaries -- 1.1.1.10 Severe accidents -- 1.1.1.11 Containment and reactor coolant system -- 1.1.1.12 Passive systems, reliability, and stability issues -- 1.1.1.13 Accident phenomenology -- 1.1.1.14 Role of startup and shutdown phenomena -- 1.1.1.15 Accident management and its procedures -- 1.1.1.16 Generation IV and small modular reactor thermal hydraulics -- 1.1.1.17 Connections to neutron physics, probabilistic safety assessment, radioprotection, chemistry, mechanics, nuclear fuel, and e ... -- 1.1.1.18 Summary of scope and framework of the textbook -- 1.1.2 Objectives, innovations, and target audience -- 1.1.2.1 Reformulation of textbook objective -- 1.1.2.2 To whom this book is addressed -- 1.1.3 Structure and content -- 1.1.3.1 Chapter 1: Introduction.
1.1.3.2 Chapter 2: Historical remarks -- 1.1.3.3 Chapter 3: Definitions -- 1.1.3.4 Chapter 4: Needs -- 1.1.3.5 Chapter 5: Balance equations -- 1.1.3.6 Chapter 6: Phenomena -- 1.1.3.7 Chapter 7: Heat transfer -- 1.1.3.8 Chapter 8: Pressure drops -- 1.1.3.9 Chapter 9: Constitutive equations -- 1.1.3.10 Chapter 10: Special models -- 1.1.3.11 Chapter 11: System thermal-hydraulic codes -- 1.1.3.12 Chapter 12: Computational fluid dynamics codes -- 1.1.3.13 Chapter 13: Verification, validation, scaling, and uncertainty -- 1.1.3.14 Chapter 14: Best estimate plus uncertainty approach -- 1.1.3.15 Chapter 15: Design basis accident/condition calculations -- 1.1.3.16 Chapter 16: Thermal hydraulics of nuclear power plant accident occurrences -- 1.1.3.17 Chapter 17: Instrumentation and (basic) experiments -- 1.1.3.18 Chapter 18: Subchannel thermal hydraulics -- 1.1.3.19 Chapter 19: Containment thermal hydraulics -- 1.1.3.20 Chapter 20: Numerics in nuclear thermal hydraulics -- 1.1.3.21 Chapter 21: Scaling insights -- 1.1.3.22 Chapter 22: New perspectives for verification and validation -- 1.1.3.23 Chapter 23: Thermal-hydraulic design of water-cooled nuclear reactors -- 1.1.3.24 Chapter 24: Controversial issues and perspectives -- 1.1.3.25 Chapter interconnections -- Glossary -- Exercises and questions -- Acknowledgments -- 2 - A historical perspective of nuclear thermal hydraulics -- Foreword -- 2.1 Introduction -- 2.1.1 Key actors and stakeholders in nuclear thermal hydraulics -- 2.1.2 Objective -- 2.2 System thermal hydraulics history and trends -- 2.2.1 Role of nuclear thermal hydraulics in nuclear reactor safety -- 2.2.1.1 Roles of probabilistic and deterministic safety assessment -- 2.2.1.2 Role of nuclear thermal hydraulics -- 2.2.2 Definitions -- 2.2.3 History -- 2.2.3.1 Regulatory history in the United States -- 2.2.3.2 Before 1960 -- 2.2.3.3 During 1960-70. 2.2.3.4 During 1970-80 -- 2.2.3.5 During 1980-90 -- 2.2.3.6 During 1990-2000 -- 2.2.3.7 During 2000-10 -- 2.2.3.8 During 2010-20 -- 2.2.3.9 Historical list of topics -- 2.2.3.10 Summary history of nuclear thermal hydraulics -- 2.2.4 An interpretation of current trends -- 2.3 Perspectives for system thermal hydraulics -- 2.3.1 "Local form loss" coefficients (also reported as "K-factors") -- 2.3.2 Multidimensional heat transfer coefficient surface -- 2.3.3 Energy and entropy balance following reactor coolant system blowdown and containment pressurization -- 2.3.4 Precision targets -- 2.3.5 Applying computational fluid dynamics-like approaches to nuclear power plant design and nuclear reactor safety technologies -- 2.3.6 Thermal hydraulics of passive systems -- 2.3.7 Scaling issue and experiments -- 2.3.8 Verification and validation of system thermal hydraulics codes -- 2.3.9 Uncertainty analysis -- 2.3.10 Coupling system thermal hydraulics -- 2.3.11 Modeling and structuring of computational tools -- 2.3.12 Licensing needs -- 2.3.13 Probabilistic safety assessment and system thermal hydraulics -- 2.3.14 Severe accidents and system thermal hydraulics -- 2.3.15 User effects and training -- 2.3.16 Best estimate plus uncertainty approach -- 2.3.17 Summary remarks -- 2.4 Conclusions -- Exercises and questions -- Acknowledgments -- 3 - Parameters and concepts in nuclear thermal hydraulics -- Foreword -- 3.1 General remarks -- 3.1.1 General remarks on parameters and concepts -- 3.1.2 Importance of two-phase flow -- 3.1.3 Multiscale and multiphysics -- 3.1.4 Turbulence and two-phase flow -- 3.1.5 Empirical database and instrumentation -- 3.2 Concepts involved in two-phase flow development -- 3.2.1 General aspects -- 3.2.2 Parameters and concepts -- 3.2.2.1 Void fraction/quality -- Interfacial area/interfacial area concentration -- 3.2.2.2 Mass velocity. 3.2.2.3 Equilibrium/subcooling/superheating -- 3.2.2.4 Pressure drop -- 3.2.2.5 Friction -- 3.2.2.6 Vaporization/evaporation/boiling -- 3.2.2.7 Condensation -- 3.2.2.8 Phase separation/mixture levels -- 3.2.2.9 Parameters of balance equations -- 3.2.2.10 Flow-regime definition -- 3.3 Concepts involved in heat transfer developments -- 3.3.1 General aspects -- 3.3.1.1 Importance of heat transfer developments -- 3.3.1.2 Scope -- 3.3.2 Parameters and concepts -- 3.3.2.1 Related to power generation -- 3.3.2.2 Related to heat conduction -- 3.3.2.3 Related to heat convection-general aspects -- 3.3.2.4 Related to heat convection-pre-critical heat flux convection modes -- 3.3.2.5 Related to heat convection-critical heat flux/boiling crisis/post-critical heat flux convection modes -- 3.3.2.6 Related to heat radiation -- 3.4 Concepts involved in target phenomena -- 3.4.1 General aspects -- 3.4.2 Selected phenomena -- 3.4.2.1 Natural circulation -- 3.4.2.2 Critical and choked flows -- 3.4.2.3 Blowdown/reflood/rewet -- 3.4.2.4 Reflux condenser mode -- 3.4.2.5 Loop seal clearing -- 3.4.2.6 Steam binding -- 3.4.2.7 Boron dilution accidents -- 3.4.2.8 Boron dilution/deboration -- 3.4.2.9 Countercurrent flow limitation -- 3.5 Analytical tools -- 3.5.1 General aspects -- 3.5.2 Models and approximations -- 3.5.2.1 Homogeneous model -- 3.5.2.2 Separated flow model -- 3.5.2.3 Drift-flux model -- 3.5.2.4 Lumped parameter models -- 3.5.2.5 One-dimensional/three-dimensional -- 3.5.2.6 Two-fluid model -- 3.5.3 Concepts related to system codes -- 3.5.3.1 General aspects -- 3.5.3.2 Closure laws or constitutive equations -- 3.5.3.3 Heat transfer correlations/lookup tables -- 3.5.3.4 Special models -- 3.5.3.5 Special components -- 3.5.3.6 Nodalization/nodalization diagram -- 3.5.3.7 User effect/good practices -- 3.5.4 Additional concepts involved in analytical thermal hydraulics. 3.5.4.1 Hydraulic diameter/Reynolds and Froude numbers -- 3.5.4.2 Prandtl and Nüsselt numbers -- 3.5.4.3 Nuclear factors -- 3.5.4.4 Engineering factors -- 3.6 Verification and validation -- 3.6.1 Definitions -- 3.6.2 Concepts involved in experimental thermal hydraulics -- 3.6.2.1 Separate effects tests -- 3.6.2.2 Integral effects tests/integral test facilities -- 3.6.2.3 Analytical support to experimental thermal hydraulics -- 3.6.2.4 Databases of experimental thermal hydraulics -- 3.6.2.5 International standard problems -- 3.6.3 Concepts involved with scaling -- 3.6.3.1 Scaling/scaling issue/addressing the scaling issue -- 3.6.3.2 Power-to-volume scaling -- 3.6.3.3 Hierarchical two-tiered scaling -- 3.6.3.4 Ishii three-level scaling -- 3.6.3.5 Fractional scaling analysis -- 3.6.3.6 Scaling distortion -- 3.6.3.7 Counterpart tests/similar/special counterparts -- 3.6.3.8 Using system codes in scaling analysis -- 3.6.3.9 Kv-scaled calculations -- 3.7 Concepts connected to design and licensing bases -- 3.7.1 Structures, systems, and components -- 3.7.2 Acceptance criteria/design criteria -- 3.7.3 Design basis -- 3.7.4 Licensing basis -- 3.7.5 Quality attributes/availability/reliability -- 3.7.6 Design basis accidents -- 3.7.7 Design basis accident phenomenology -- 3.7.8 Beyond design basis accidents/severe accidents -- 3.8 General safety concepts -- 3.8.1 Safety objectives/safety culture -- 3.8.2 Accident management/emergency operating procedures/severe accident management guidelines -- 3.8.3 Heat extraction safety relevance -- 3.8.4 Energy sources -- 3.8.5 Damage -- 3.8.6 Safety functions -- 3.8.7 Defense in depth -- 3.8.8 Defense-in-depth levels -- 3.8.9 Safety barriers -- 3.8.10 Safety systems/safety-related systems -- 3.8.11 Inherent safety/passive safety/active safety -- 3.8.12 International institutions. 3.9 Concepts related to deterministic safety assessment. |
| Record Nr. | UNINA-9911045227203321 |
| Cambridge, MA : , : Woodhead Publishing is an imprint of Elsevier, , [2024] | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Ninth Pulsed Power Conference, 1993
| Ninth Pulsed Power Conference, 1993 |
| Pubbl/distr/stampa | [Place of publication not identified], : IEEE, 1994 |
| Descrizione fisica | 1 online resource |
| Disciplina | 621.4834 |
| Soggetto topico | Pulsed reactors |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910872973003321 |
| [Place of publication not identified], : IEEE, 1994 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors : [Newport Beach, California, August 1-5, 1999] / / sponsored by the Minerals, Metals and Materials Society, American Nuclear Society, National Association of Corrosion Engineers International ; edited by Steve Bruemmer, Peter Ford, Gary Was
| Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors : [Newport Beach, California, August 1-5, 1999] / / sponsored by the Minerals, Metals and Materials Society, American Nuclear Society, National Association of Corrosion Engineers International ; edited by Steve Bruemmer, Peter Ford, Gary Was |
| Pubbl/distr/stampa | Warrendale, Pennsylvania : , : Minerals, Metals & Materials Society, , [1999] |
| Descrizione fisica | 1 online resource (1252 p.) |
| Disciplina |
621.48
621.4834 |
| Altri autori (Persone) |
BruemmerS. M
FordF. P (F. Peter) WasGary S <1953-> (Gary Steven) |
| Soggetto topico |
Nuclear power plants - Corrosion
Water cooled reactors - Corrosion Nuclear power plants - Materials - Effect of radiation on |
| ISBN |
1-118-78777-3
1-118-78761-7 1-118-78795-1 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto |
Cover; Title Page; Copyright Page; FOREWORD; TABLE OF CONTENTS; PWR Primary-1: Mechanisms; An Overview of Internal Oxidation as a Possible Explanation of Intergranular Stress Corrosion Cracking of Alloy 600 in PWRs; Methodology to Understand the Mechanisms of PWSCC; Hydrogen Effects on PWR SCC Mechanisms in Monocrystalline and Polycrystalline Alloy 600; Insights into Environmental Degradation Mechanisms from Analytical Transmission Electron Microscopy of SCC Cracks; Measurement of the Fundamental Parameters for the Film-Rupture/Oxidation Mechanism-The Effect of Chromium
Comparison of Hydrogen Effects on Alloy 600 and 690Comments on a Proposed Mechanism of Internal Oxidation for Alloy 600 as Applied to Low Potential SCC; Internal Oxidation and Embrittlement of Alloy 600; PWR Primary-2: Chemistry and Failure Analysis; The Effect of Primary Coolant Zinc Additions on the SCC Behaviour of Alloy 600 and 690; PWSCC of Alloy 600: A Parametric Study of Surface Film Effects; Modelling of Stress Corrosion Crack Initiation on Alloy 600 in Primary Water of PWRs; Effect of Water Chemistry on Environmentally Assisted Cracking in Alloy in Simulated PWR Environments Unique Primary Side Initiated Degradation in the Vicinity of the Upper Roll Transition in Once Through Steam Generators from Oconee Unit 1PWR Primary-3: Hydrogen Effects & Microstructure; On the Possibility of Forming Ordered Ni2Cr in Alloy 690; Hydrogen Embrittlement of PH 13-08 Mo Stainless Steel in PWR Environment Effect of Microstructure; The Effect of Special Grain Boundaries on IGSCC of Ni-16Cr-9Fe-xC; Fracture Behavior of Nickel-Based Alloys in Water; Hydrogen-Assisted Failure of Alloys X-750 and 625 under Slow Strain-Rate Conditions An Experimental Study of the Hydrogen Embrittlement of Alloy 718 in PWR Primary WaterA Study of Corrosion Mechanisms and Kinetics of Alloy 718 in PWR Primary Water; Stress Corrosion Crack Propagation Rate of Alloy 600 in the Primary Water of PWR: Influence of a Cold Worked Layer; PWR Primary-4: Crack Growth & Creep; Stress Corrosion Crack Growth Rate Measurements in Alloys 600 and 182 in Primary Water Loops Under Constant Load; Initial Results on the Stress Corrosion Cracking Monitoring of Alloy 600 in High Temperature Water Using Acoustic Emission Stress Corrosion Crack Propagation Rates in Reactor Vessel Head Penetrations in Alloy 600Stress Corrosion Life Assessment of Alloy 600 PWR Components; Influence of Chromium Content and Microstructure on Creep and PWSCC Resistance of Nickel Base Alloys; A Simplified Model for SCC Initiation Susceptibility in Alloy 600, with the Influence of Cold Work Layer and Strength Characteristics; Creep of Nickel Base Alloys in High Temperature Water; An Investigation of Alloy 182 Stress Corrosion Cracking in Simulated PWR Environment; BWR-1: Cracking Response Characteristics of Crack Propagation Through SCC under BWR Conditions in Stainless Steels Stabilized with Titanium or Niobium |
| Altri titoli varianti | Environmental degradation of materials in nuclear power systems--water reactors |
| Record Nr. | UNISA-996213874203316 |
| Warrendale, Pennsylvania : , : Minerals, Metals & Materials Society, , [1999] | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. di Salerno | ||
| ||
The Risks of Nuclear Energy Technology : Safety Concepts of Light Water Reactors / / by Günter Kessler, Anke Veser, Franz-Hermann Schlüter, Wolfgang Raskob, Claudia Landman, Jürgen Päsler-Sauer
| The Risks of Nuclear Energy Technology : Safety Concepts of Light Water Reactors / / by Günter Kessler, Anke Veser, Franz-Hermann Schlüter, Wolfgang Raskob, Claudia Landman, Jürgen Päsler-Sauer |
| Autore | Kessler Günter |
| Edizione | [1st ed. 2014.] |
| Pubbl/distr/stampa | Berlin, Heidelberg : , : Springer Berlin Heidelberg : , : Imprint : Springer, , 2014 |
| Descrizione fisica | 1 online resource (365 p.) |
| Disciplina | 621.4834 |
| Collana | Science Policy Reports |
| Soggetto topico |
Nuclear energy
Radiation - Safety measures Radiation—Safety measures Energy systems System safety Nuclear Energy Effects of Radiation/Radiation Protection Energy Systems Security Science and Technology |
| ISBN | 3-642-55116-5 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto | From the Contents: Inherent safety characteristics of Pressurized Water- and Boiling Water Reactors (PWRs and BWR s) -- Safety design concepts of present and future PWRs and BWR s -- Radiation protection and emission of radioactivity during normal operation of PWRs and BWRs -- Accident and risk analysis as well as additional severe accident measures to be initiated after core cooling accidents -- Safety design concepts against external hazards, e. g. earthquakes, chemical explosions, flooding. |
| Record Nr. | UNINA-9910299619903321 |
|
Kessler Günter
|
||
| Berlin, Heidelberg : , : Springer Berlin Heidelberg : , : Imprint : Springer, , 2014 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Safety of sodium-cooled fast reactors : particle-bed-related phenomena in severe accidents / / Songbai Cheng, Ruicong Xu
| Safety of sodium-cooled fast reactors : particle-bed-related phenomena in severe accidents / / Songbai Cheng, Ruicong Xu |
| Autore | Cheng Songbai |
| Pubbl/distr/stampa | Gateway East, Singapore : , : Springer, , [2021] |
| Descrizione fisica | 1 online resource (313 pages) |
| Disciplina | 621.4834 |
| Soggetto topico | Sodium cooled reactors |
| ISBN | 981-16-6116-2 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNISA-996466741903316 |
|
Cheng Songbai
|
||
| Gateway East, Singapore : , : Springer, , [2021] | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. di Salerno | ||
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Safety of Sodium-Cooled Fast Reactors : Particle-Bed-Related Phenomena in Severe Accidents / / by Songbai Cheng, Ruicong Xu
| Safety of Sodium-Cooled Fast Reactors : Particle-Bed-Related Phenomena in Severe Accidents / / by Songbai Cheng, Ruicong Xu |
| Autore | Cheng Songbai |
| Edizione | [1st ed. 2021.] |
| Pubbl/distr/stampa | Singapore : , : Springer Nature Singapore : , : Imprint : Springer, , 2021 |
| Descrizione fisica | 1 online resource (313 pages) |
| Disciplina | 621.4834 |
| Soggetto topico |
Nuclear physics
Thermodynamics Nuclear engineering Security systems Electric power-plants Nuclear Physics Nuclear Energy Security Science and Technology Power Stations |
| ISBN | 981-16-6116-2 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto | Chapter 1 Introduction -- Chapter 2 Molten-pool Mobility.-Chapter 3 Molten-pool Sloshing Motion.-Chapter 4: Debris Bed Formation Behavior -- Chapter 5 Debris Bed Self-Leveling Behavior -- Chapter 6 Conclusion and Future Prospect. |
| Record Nr. | UNINA-9910502620903321 |
|
Cheng Songbai
|
||
| Singapore : , : Springer Nature Singapore : , : Imprint : Springer, , 2021 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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