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Advanced Propulsion System and Thermal Management Technology / / edited by Jian Liu [and three others]
| Advanced Propulsion System and Thermal Management Technology / / edited by Jian Liu [and three others] |
| Pubbl/distr/stampa | Basel, Switzerland : , : MDPI - Multidisciplinary Digital Publishing Institute, , [2023] |
| Descrizione fisica | 1 online resource (126 pages) |
| Disciplina | 621.4022 |
| Soggetto topico |
Heat - Transmission
Thermal hydraulics |
| Formato | Materiale a stampa  |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910734350803321 |
| Basel, Switzerland : , : MDPI - Multidisciplinary Digital Publishing Institute, , [2023] | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Analysis with TRACE code of Rosa test 1.2 : small LOCA in the hot-leg with HPI and accumulator actuation / / prepared by J.L. Munoz-Cobo, A. Romero, S. Chiva
| Analysis with TRACE code of Rosa test 1.2 : small LOCA in the hot-leg with HPI and accumulator actuation / / prepared by J.L. Munoz-Cobo, A. Romero, S. Chiva |
| Autore | Muñoz-Cobo J. L. |
| Pubbl/distr/stampa | Washington, DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, , April 2013 |
| Descrizione fisica | 1 online resource (various pagings) : illustrations |
| Collana | International agreement report |
| Soggetto topico |
Light water reactors - Analysis
Nuclear power plants Nuclear reactors - Thermodynamics Thermal hydraulics |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Altri titoli varianti | Analysis with TRACE code of Rosa test 1.2 |
| Record Nr. | UNINA-9910704660203321 |
Muñoz-Cobo J. L.
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| Washington, DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, , April 2013 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Assessment of TRACE V5.0.Patch 4 code against PWR PACTEL loop seal clearing experiment / / prepared by: Otso-Pekka Kauppinen
| Assessment of TRACE V5.0.Patch 4 code against PWR PACTEL loop seal clearing experiment / / prepared by: Otso-Pekka Kauppinen |
| Autore | Kauppinen Otso-Pekka |
| Pubbl/distr/stampa | Washington., DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, , December 2018 |
| Descrizione fisica | 1 online resource (xiii, 23 pages) : illustrations (some color) |
| Collana | International agreement report |
| Soggetto topico |
Pressurized water reactors - Loss of coolant - Computer programs
Thermal hydraulics |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Altri titoli varianti | Assessment of TRACE V5.0.Patch 4 code against Pressurized Water Reactor PArallel Channel TEst Loop loop seal clearing experiment |
| Record Nr. | UNINA-9910711940303321 |
|
Kauppinen Otso-Pekka
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| Washington., DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, , December 2018 | ||
| 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 |
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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 |
9780323856072
0323856071 |
| 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 | ||
| ||
Investigation of the loop seal clearing phenomena for the ATLAS DVI line and cold leg SBLOCA tests using MARS-KS and RELAP5/MOD3.3 / / prepared by Minjeong Hwang [and three others]
| Investigation of the loop seal clearing phenomena for the ATLAS DVI line and cold leg SBLOCA tests using MARS-KS and RELAP5/MOD3.3 / / prepared by Minjeong Hwang [and three others] |
| Autore | Hwang Minjeong |
| Pubbl/distr/stampa | Washington., DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, , March 2019 |
| Descrizione fisica | 1 online resource (xvolumes 49 pages) : illustrations (some color) |
| Collana | International agreement report |
| Soggetto topico |
Pressurized water reactors - Loss of coolant - Computer programs
Thermal hydraulics |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Altri titoli varianti | Investigation of the loop seal clearing phenomena for the Advanced Thermal-hydraulic test Loop for Accident Simulation Direct Vessel Injection line and cold leg Small Break Loss Of Coolant Accident tests using Multi-dimensional Advanced Reactor Safety program-Korea Standard and Reactor Excursion and Leak Analysis Program 5/MOD3.3 |
| Record Nr. | UNINA-9910712081803321 |
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Hwang Minjeong
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| Washington., DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, , March 2019 | ||
| 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 |
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D'Auria Francesco
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| London, [England] : , : Academic Press, , 2017 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
TRACE VVER-1000/V-320 model validation / / prepared by: S. Iegan [and four others]
| TRACE VVER-1000/V-320 model validation / / prepared by: S. Iegan [and four others] |
| Autore | Iegan S. |
| Pubbl/distr/stampa | Washington, DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, , December 2018 |
| Descrizione fisica | 1 online resource (various pagings) : illustrations (some color) |
| Collana | International agreement report |
| Soggetto topico |
Nuclear power plants - Safety measures - Mathematical models
Thermal hydraulics |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910711949503321 |
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Iegan S.
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| Washington, DC : , : Division of Systems Analysis, Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, , December 2018 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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