Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors : Volume 3: Procedures and Applications

• Autore: D'Auria Francesco
• Pubblicazione: San Diego : , : Elsevier Science & Technology, , 2024; ©2024
• Edizione: 2nd ed.
• Descrizione fisica: 1 online resource (818 pages)
• Disciplina: 621.4834
• Soggetto topico: Nuclear reactors - Design and construction; Nuclear engineering
• Altri autori: HassanY. A
• 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.
• Sommario/riassunto: This handbook provides a comprehensive exploration of thermal hydraulics in water-cooled nuclear reactors, offering detailed insights into various procedures and applications. Edited by Francesco D’Auria and Yassin A. Hassan, the book is part of the Woodhead Publishing series in energy and serves as a crucial resource for understanding the complexities of nuclear reactor safety and efficiency. It covers topics such as subchannel modeling, containment thermal hydraulics, numerical methods, scaling techniques, and thermal hydraulic design of reactors. Aimed at researchers, practitioners, and students in the field of nuclear engineering, the handbook draws from contributions of thousands of researchers to enhance the understanding of thermal hydraulics. The book emphasizes the importance of safety, validation, and the evolution of methodologies within the field.
• Titolo autorizzato: Handbook on Thermal Hydraulics in Water-Cooled Nuclear Reactors
• ISBN: 9780323856096; 0323856098
• Formato: Materiale a stampa
• Livello bibliografico: Monografia
• Lingua di pubblicazione: Inglese
• Record Nr.: 9911045226403321
• Lo trovi qui: Univ. Federico II
• Serie: Woodhead Publishing Series in Energy Series