Nuclear Power Plant Design and Analysis Codes: Development, Validation, and Application

Title-page_2021_Nuclear-Power-Plant-Design-and-Analysis-Codes
Nuclear Power Plant Design and Analysis Codes
Copyright_2021_Nuclear-Power-Plant-Design-and-Analysis-Codes
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Contents_2021_Nuclear-Power-Plant-Design-and-Analysis-Codes
Contents
List-of-contributors_2021_Nuclear-Power-Plant-Design-and-Analysis-Codes
List of contributors
Chapter-1---Road-map_2021_Nuclear-Power-Plant-Design-and-Analysis-Codes
1 Road map
1.1 Overview
1.2 Fuel codes
1.2.1 FRAPCON/FRAPTRAN
1.2.2 TRANSURANUS
1.2.3 GERMINAL
1.3 System code
1.3.1 ATHLET
1.3.2 RELAP5
1.3.3 RELAP5-3D
1.3.4 CATHARE
1.3.5 TRAC
1.3.6 MELCOR
1.3.7 Modular Accident Analysis Program
1.3.8 System Analysis Module
1.4 Subchannel analysis code
1.4.1 COBRA
1.4.2 HAMBO
1.4.3 FLICA
1.4.4 THINC
1.4.5 MATRA
1.4.6 CTF
1.4.7 VIPRE
1.5 Computational fluid dynamics code
1.5.1 CFX
1.5.2 Fluent
1.5.3 TransAT
1.5.4 STAR-CD
1.6 Coupled multiscale thermal-hydraulics codes
1.6.1 Multiphysics Object-Oriented Simulation Environment
1.6.2 Consortium for the Advanced Simulation of Light Water Reactors
1.6.3 Other coupled codes
1.7 Emerging methods for NPPs
1.7.1 Projection-based particle method
1.7.2 Lattice Boltzmann method
1.7.3 Cybersecurity system
1.7.4 Artificial intelligence and artificial neural network
1.7.5 Temporal data mining
References
Chapter-2---Guidance-of-nuclear-power-pl_2021_Nuclear-Power-Plant-Design-and
2 Guidance of nuclear power plant code development
2.1 Nuclear power plant code
2.1.1 Code classification
2.1.2 Code development process
2.1.3 Code development skills
2.2 Code development examples
2.2.1 Reactor
2.2.1.1 Model development
2.2.1.1.1 Reactor core
2.2.1.1.2 Lower and upper chamber
2.2.1.2 Numerical scheme
2.2.1.3 Verification
2.2.1.3.1 Steady-state results
2.2.1.3.2 Transient results
2.2.2 Pressurizer
2.2.2.1 Model development
2.2.2.1.1 Governing equations
2.2.2.1.2 Empirical correlations
Bulk evaporation and condensation
Condensation on spray
2.2.2.2 Numerical scheme
2.2.2.3 Verification
2.3 Source code
2.3.1 Fourth-order explicit Runge–Kutta method
2.3.2 Gauss–Jordan elimination with partial pivoting
References
Chapter-3---Nuclear-engineering-software_2021_Nuclear-Power-Plant-Design-and
3 Nuclear engineering software quality assurance
3.1 Introduction
3.2 Software development life cycle
3.3 Software quality
3.4 Software quality assurance for nuclear engineering
3.4.1 Implementation framework
3.4.2 Verification and validation
3.5 Summary
References
Chapter-4---Multiphysics-coupling_2021_Nuclear-Power-Plant-Design-and-Analys
4 Multiphysics coupling plan
4.1 Introduction
4.2 Multiphysics coupling methods
4.2.1 Operator splitting methods
4.2.2 Jacobian-free Newton–Krylov methods
4.2.3 Approximate block Newton methods
4.3 Current status of research in multiphysics coupling
4.3.1 Neutronic and thermal-hydraulic code-to-code coupling
4.3.2 NURESAFE European project
4.3.3 Multiphysics Object-Oriented Simulation Environment
4.3.4 Consortium for the Advanced Simulation of Light Water Reactors
4.4 Conclusion
References
Chapter-5---Nuclear-physics-determin_2021_Nuclear-Power-Plant-Design-and-Ana
5 Nuclear physics deterministic code
5.1 Nuclear data processing codes
5.1.1 NJOY
5.2 Cross section generation codes
5.2.1 Winfrith improved multigroup scheme
5.2.2 CASMO-4
5.2.3 HELIOS
5.2.4 APOLLO2
5.2.5 Standard Reactor Analysis Code
5.2.6 DRAGON4
5.2.7 AEGIS
5.2.8 SCALE
5.2.9 PARAGON
5.2.10 Bamboo-Lattice
5.2.11 MC2-3
5.2.12 ECCO
5.3 Whole-core computational codes
5.3.1 SIMULATE-3
5.3.2 ANC9
5.3.3 PANTHER
5.3.4 Purdue Advanced Reactor Core Simulator
5.3.5 DONJON4
5.3.6 DYN3D
5.3.7 Bamboo-Core
5.3.8 SCOPE2
5.3.9 DIF3D/VARIANT/REBUS
5.3.10 European Reactor Analysis Optimized calculation System
References
Chapter-6---Nuclear-physics-probabilit_2021_Nuclear-Power-Plant-Design-and-A
6 Nuclear physics probability code: OpenMC
6.1 Introduction to Monte Carlo method
6.2 Recently developed Monte Carlo codes
6.3 Typical methodologies in OpenMC
6.3.1 Random number generator
6.3.2 Computational geometry
6.3.3 Random walk
6.3.4 Tallies and statistics
6.3.5 Eigenvalue calculation
6.3.6 Fixed-source calculation
6.4 Usage of OpenMC
6.4.1 Preparation
6.4.2 Data library
6.4.3 Geometry definition
6.4.4 Material definition
6.4.5 Tally definition
6.4.6 Settings definition
6.4.7 Plots definition
6.4.8 Simulation in serial and parallel
6.4.9 Output description
6.4.10 Python API
6.5 Verification and validation
6.6 Summary
References
Chapter-7---FRAPCON-and-FRAPTRAN-codes--Fuel-rod-p_2021_Nuclear-Power-Plant-
7 FRAPCON and FRAPTRAN codes: Fuel rod performance analysis codes under normal and accident conditions
7.1 Objectives, relations, and limitations
7.1.1 Objectives of the two codes
7.1.2 Relations
7.1.3 Limitations
7.2 Code structures and physical models
7.2.1 Thermohydraulic models
7.2.1.1 Coolant conditions
7.2.1.2 Cladding temperature and heat generation
7.2.1.3 Gap heat conductance
7.2.1.4 Pellets temperature distributions
7.2.1.5 Plenum gas temperature model and energy storage model
7.2.2 Mechanical models
7.2.3 Internal gas response and fission gas release
7.3 Assessments
7.3.1 FRAPCON
7.3.2 FRAPTRAN
7.3.2.1 RIA condition
7.3.3 Loss of Coolant Accident condition
7.4 Summary
References
Chapter-8---The-TRANSURANUS-fuel-perf_2021_Nuclear-Power-Plant-Design-and-An
8 The TRANSURANUS fuel performance code
8.1 Introduction: General overview of the TRANSURANUS code
8.2 TRANSURANUS code structure
8.2.1 Thermal analysis
8.2.2 Mechanical analysis
8.2.3 Fission gas behavior modeling
8.2.4 The TRANSURANUS burn-up module
8.2.5 Material conservation equations
8.3 Application to water reactor conditions
8.3.1 Assessment against the pressurized water reactor Super-Ramp irradiation experiment
8.3.2 Application to the safety analysis of the Atucha-II Nuclear Power Plant
8.3.2.1 Overview of the Atucha-II Nuclear Power Plant
8.3.2.2 Atucha-II large break – loss-of-coolant accident transient
8.3.3 Application to water–water energetic reactor conditions
8.4 Preliminary assessment against fast reactor conditions
8.4.1 Development of TRANSURANUS for fast reactor conditions
8.4.1.1 Plutonium redistribution model
8.4.1.2 Formation and closure of the fuel central void
8.4.2 Assessment against the HEDL P-19 irradiation experiment
8.4.3 Application to the safety assessment of the ALFRED reactor
8.5 Conclusions and future code developments
Nomenclature
References
Chapter-9---Two-fuel-performance-codes-of-the-_2021_Nuclear-Power-Plant-Desi
9 Two fuel performance codes of the PLEIADES platform: ALCYONE and GERMINAL
9.1 General overview of the PLEIADES fuel software environment
9.1.1 Architecture and generic tools for fuel performance codes
9.1.2 Multiphysics computational scheme for fuel rod type geometries
9.1.2.1 Algorithm
9.1.2.2 Global scale
9.1.2.3 Local scale
9.1.2.4 Software implementation
9.1.3 Verification process and quality control for the fuel performance codes of the PLEIADES platform
9.1.3.1 Software quality control
9.1.3.2 Unit nonregression tests
9.1.3.3 Integral nonregression tests
9.2 ALCYONE fuel performance code for GEN II and III
9.2.1 General presentation
9.2.2 Physical models
9.2.2.1 Isotopic vector evolution and nuclear reactions products
9.2.2.2 Fission gas behavior and helium release
9.2.2.3 Thermochemical analysis and corrosive fission products release
9.2.2.4 Crack extension modeling in fuel
9.2.2.4.1 Macroscopic scale
9.2.2.4.2 Microscopic scale
9.2.2.5 Heterogeneous mechanical behavior
9.2.2.6 Multidimensional and multiscale analysis
9.2.2.6.1 Thermal model
9.2.2.6.2 Mechanical model
9.2.2.6.3 Pellet-to-cladding gap model
9.2.2.6.4 Multiscale analysis
9.2.3 3D simulation results and integral validation of the ALCYONE code
9.2.3.1 Base irradiation
9.2.3.2 Power ramp test
9.2.4 International benchmarks
9.3 GERMINAL fuel performance code for GEN IV
9.3.1 General presentation
9.3.2 Physical models
9.3.2.1 Fuel restructuring
9.3.2.2 Gap closure and relocation model
9.3.2.3 Joint Oxyde Gaine formation and interaction with thermomechanical behavior
9.3.3 Validation and application for fuel design
9.3.3.1 Validation
9.3.3.2 Fuel design for the ASTRID project
9.3.4 International benchmarks
9.3.4.1 NEA expert group on innovative fuels
9.3.4.2 ESFR-SMART
9.3.4.3 CEA-JAEA collaboration
9.3.4.4 INSPYRE European project
9.4 Conclusion and prospects
References
Chapter-10---Subchannel-codes--CTF-and-VIPRE-01-----Note--This-ch_2021_Nucle
10 Subchannel codes: CTF and VIPRE-01*
10.1 Introduction and CTF code overview
10.2 CTF assessment: Motivation and work scope
10.3 VIPRE-01 code overview
10.4 CTF assessment: Flow mixing
10.4.1 Two-channel single-phase flow split problem
10.4.2 General Electric 3×3 benchmark
10.4.2.1 Single-phase flow benchmark
10.4.2.2 Two-phase flow benchmark
10.5 CTF assessment: Pressure drop
10.5.1 Boiling-water reactor full-size fine-mesh bundle tests pressure drop benchmark
10.5.1.1 Single-phase flow benchmark
10.5.1.2 Two-phase flow benchmark
10.5.2 Risø round tube benchmark
10.6 CTF assessment: Void fraction
10.6.1 Boiling-water reactor full-size fine-mesh bundle tests void benchmark
10.6.2 Pressurized-water reactor subchannel and bundle tests void benchmark
10.6.2.1 Single subchannel benchmark
10.6.2.2 Rod bundle benchmark
10.7 Summary
References
Chapter-11---System-level-code-T_2021_Nuclear-Power-Plant-Design-and-Analysi
11 System-level code TRACE
11.1 Features of TRACE
11.2 Constitutive equations of TRACE
11.2.1 Conservation equations
11.2.2 Closure equations
11.2.2.1 Interfacial mass-transfer rate
11.2.2.2 Interface to gas heat transfer
11.2.2.3 Interface to liquid heat transfer
11.2.2.4 Wall-to-gas heat transfer
11.2.2.5 Wall-to-liquid heat transfer
11.2.2.6 Shear at the phase interface
11.2.2.7 Wall shear force to gas
11.2.2.8 Wall shear force to liquid
11.2.3 Heat conduction equations
11.2.4 Other models
11.2.4.1 Critical flow
11.2.4.2 Countercurrent flow
11.2.4.3 Form loss
11.2.4.4 Pump models
11.2.4.5 Pressurizer model
11.2.4.6 Steam separators model
11.3 Validation of TRACE
References
Chapter-12---Nuclear-thermal-hydraulics-with_2021_Nuclear-Power-Plant-Design
12 Nuclear thermal hydraulics with the AC² system code package
12.1 The system thermal-hydraulic code ATHLET
12.1.1 Introduction
12.1.2 History of Analysis of Thermal Hydraulics of LEaks and Transients development
12.1.3 Modeling basis
12.1.3.1 Thermal-hydraulic field equations
12.1.3.1.1 The six-equation model
12.1.3.1.2 The five-equation model
12.1.3.1.3 3D model
12.1.3.1.4 Noncondensable gases
12.1.3.2 Constitutive equations
12.1.3.2.1 Working fluid properties
12.1.3.2.2 Interfacial shear
12.1.3.2.3 Wall friction and form losses
12.1.3.2.4 Interfacial heat and mass transfer
12.1.3.3 Heat conduction and heat transfer
12.1.3.4 Neutron kinetics
12.1.3.5 General control and simulation
12.1.3.6 Numerical approach and new Numerical Toolkit
12.1.3.7 Steady-state calculation
12.1.4 Specific models for certain reactor designs
12.1.4.1 Critical discharge models
12.1.4.2 Boron tracking model
12.1.4.3 Gas-cooled reactor models
12.1.5 Validation
12.1.6 Code coupling
12.1.7 Scope of application and limits
12.2 COntainment COde SYStem thermal-hydraulic module THY
12.2.1 History of THY development
12.2.2 COntainment COde SYStem THY modeling basis
12.2.2.1 Basic thermal-hydraulic equations
12.2.2.2 Junction models
12.2.2.3 Fluid properties
12.2.2.4 Heat conduction, heat transfer, and interfacial heat transfer
12.2.2.5 Additional models
12.2.2.6 Numerical approach
12.2.3 COntainment COde SYStem validation
12.2.4 Scope of application and limits
12.3 Quality assurance measures
12.4 Outlook and summary
Nomenclature
Subscripts, superscripts
Abbreviations
References
Chapter-13---Development-and-application-of-Sys_2021_Nuclear-Power-Plant-Des
13 Development and application of System Analysis Module from the user’s view
13.1 Introduction
13.2 Software development
13.2.1 Structure
13.2.2 Models
13.2.2.1 Fluid dynamics
13.2.2.2 Heat transfer
13.2.2.3 Closure models
13.2.2.4 Mass transport and reactor kinetics
13.2.2.5 Numerical schemes
13.2.3 Current capabilities
13.3 Verification and demonstration
13.3.1 Spatial and temporal discretization
13.3.2 Three-dimensional finite element flow model
13.3.3 Pseudo-three-dimensional full-core conjugate heat transfer
13.3.4 EBR-II benchmark
13.3.5 Compact Integral Effects Test experiments
13.3.6 Heat pipe modeling
13.3.7 High-temperature gas reactor primary loop modeling
13.4 Integration and coupling
13.4.1 Implementation in the system code
13.4.2 Implementation in STARCCM+ code
13.4.3 Implementation in other codes
References
Chapter-14---Mechanism-based-codes-for-se_2021_Nuclear-Power-Plant-Design-an
14 Mechanism-based codes for severe accident analysis
14.1 COPRA code
14.1.1 Governing equations
14.1.2 Turbulence model
14.1.3 Crust model
14.1.4 Conduction model
14.1.5 Radiation model
14.1.6 Code validation
14.2 IVRASA code
14.2.1 Heat transfer model
14.2.2 Heat transfer relationships
14.2.3 Benchmark and in-vessel retention analysis
14.3 Thermal EXplosion Analysis Simulation code
14.3.1 Basic assumption
14.3.2 Code validation
14.4 MOCO code
14.4.1 Chemical reaction model
14.4.2 Concrete ablation model
14.4.3 Corium cooling model
14.4.4 Code validation
14.5 DETAC code
14.5.1 Mathematical model
14.5.2 Code validation
References
Chapter-15---Severe-accident-analysi_2021_Nuclear-Power-Plant-Design-and-Ana
15 Severe accident analysis with AC²
15.1 The severe accident code ATHLET-CD for in-vessel phenomena
15.1.1 Introduction
15.1.2 History of ATHLET-CD development
15.1.3 ATHLET-CD modules and models
15.1.3.1 ECORE
15.1.3.2 FIPREM
15.1.3.3 FIPISO
15.1.3.4 SAFT
15.1.3.5 AIDA/LHEAD
15.1.4 Numerical approach
15.1.5 Specific models for certain reactor designs
15.1.6 Validation of ATHLET-CD
15.1.6.1 Simulation of Phébus FPT-3
15.1.6.2 AC2 application for a generic PWR accident scenario
15.1.7 Scope of application and limits
15.2 Severe accident analysis for containment phenomena with Containment Code System
15.2.1 Introduction
15.2.2 History of Containment Code System development
15.2.3 The Containment Code System main modules
15.2.3.1 Thermal hydraulic module
15.2.3.2 Aerosol and fission product module
15.2.3.3 Core concrete interaction module and other ex-vessel corium issues
15.2.4 Numerical approach
15.2.4.1 Thermal hydraulic module
15.2.4.2 Aerosol and fission product module
15.2.4.3 Core concrete interaction module
15.2.5 Models for specific reactor designs
15.2.6 Validation of Containment Code System
15.2.7 Scope of application and limits
15.3 Quality assurance measures
15.4 Outlook and summary
Nomenclature
1 Symbols
Subscripts, superscripts
Abbreviations
References
Chapter-16---Integral-severe-accident-co_2021_Nuclear-Power-Plant-Design-and
16 Integral severe accident codes: IMPACT/SAMPSON
16.1 Introduction
16.2 SAMPSON main modules
16.2.1 Fuel rod heat up and molten core relocation
16.2.2 Validation of the fuel rod heat up analysis and molten core relocation analysis modules
16.3 3D containment modules
16.3.1 POOL3D module
16.3.2 Debris spreading analysis module
16.4 Application of the SAMPSON code to the Fukushima Daiichi nuclear power plant accident
16.4.1 Self-controlling behavior of the RCIC turbine in Unit 2
16.4.2 Three pressure peaks period in Unit 2
16.4.3 PCV pressure behavior in Unit 3
16.5 Advantages and disadvantages in the use of SAMPSON
References
Chapter-17---Engineering-level-system-code-fo_2021_Nuclear-Power-Plant-Desig
17 Engineering-level system code for severe accident analysis: MELCOR
17.1 Introduction
17.2 Quality control
17.2.1 Verification and validation study of core heat up and degradation
17.2.2 Bundle design
17.2.3 Results of analysis
17.3 Capabilities and limitations
17.3.1 Advantages
17.3.2 Limitation of MELCOR
17.4 MELCOR version update history
17.5 Demonstration problems: Experiments for validation
Reference
Chapter-18---Moving-Particle-Semi-imp_2021_Nuclear-Power-Plant-Design-and-An
18 Moving Particle Semi-implicit method
18.1 Introduction
18.2 Moving Particle Semi-implicit method
18.2.1 Governing equations
18.2.2 Discretization scheme
18.2.3 Detection of free surface particles
18.2.4 Semi-implicit algorithm and pressure calculation
18.3 Application to nuclear engineering
18.3.1 Bubble dynamics
18.3.2 Vapor explosion
18.3.3 Jet and droplet
18.3.4 Multiphase flow instability
18.3.5 In-vessel phenomena during severe accident
18.3.6 Corium spreading and molten core concrete interaction
18.3.7 Flooding accident
18.4 Conclusion
Reference
Chapter-19---Lattice-Boltzmann-met_2021_Nuclear-Power-Plant-Design-and-Analy
19 Lattice Boltzmann method code
19.1 Introduction
19.2 Lattice Boltzmann multiphase models
19.2.1 Color-gradient model and its application to jet breakup
19.2.2 Pseudopotential model and its application to boiling
19.3 Summary
References
Chapter-20---Code-for-nuclear-mat_2021_Nuclear-Power-Plant-Design-and-Analys
20 Code for nuclear materials
20.1 Electronic structure calculations in nuclear materials
20.2 Molecular dynamics simulations in nuclear materials
20.3 Mesoscale modeling in nuclear materials field
20.3.1 Kinetic Monte Carlo
20.3.2 Mean field rate theory and cluster dynamics
References
Chapter-21---Nuclear-power-plant-cyb_2021_Nuclear-Power-Plant-Design-and-Ana
21 Nuclear power plant cybersecurity
21.1 Introduction
21.2 Cybersecurity differences of instrumentation and control system and information technology system
21.3 Cyber-incidents in the history of the nuclear industry
21.4 Regulations
21.4.1 International Atomic Energy Agency guidance
21.4.2 International Electrotechnical Commission standard
21.4.3 Nuclear Regulatory Commission guidance
21.4.4 Nuclear Energy Institute guidance
21.5 Cyberattack detection research using machine-learning algorithms
21.5.1 General procedure of building a machine-learning model
21.5.2 Cyberattack detection using cyber data
21.5.2.1 Artificial neural networks
21.5.2.2 Bayesian network
21.5.2.3 Decision trees
21.5.2.4 K-Nearest neighbor
21.5.2.5 Ensemble learning
21.5.2.6 Bagging
21.5.2.7 Random forest
21.5.3 Cyberattack detection using process data
21.5.4 Discussion
21.6 Conclusion
Acknowledgment
References
Chapter-22---Artificial-neural-network_2021_Nuclear-Power-Plant-Design-and-A
22 Artificial neural network introductions
22.1 What is artificial neural network
22.2 Theory of artificial neural network
22.2.1 Mathematical models and structures of artificial neural network
22.2.1.1 Basic artificial neural network model
22.2.1.2 BPN model
22.2.1.2.1 Weight adjustment between the hidden layer and output layer
22.2.1.2.2 Weight adjustment between the input layer and hidden layer
22.2.1.2.3 The implement of BPN by MATLAB
22.2.1.3 Genetic neural network model
22.2.1.4 Wavelet neural network model
22.2.2 Training process of neural network
22.3 Artificial neural network applications in T/H problem
22.3.1 Prediction of critical heat flux
22.3.2 Prediction of nucleate boiling heat transfer coefficient
22.3.3 Prediction of onset of nucleate boiling in vertical narrow annuli
22.3.4 Characteristic points of boiling curve
22.3.4 Prediction of leak before break leak rate
References
Chapter-23---New-direction-of-nuclear-code-de_2021_Nuclear-Power-Plant-Desig
23 New direction of nuclear code development: artificial intelligence
23.1 Introduction
23.2 A brief history of artificial intelligence
23.3 Artificial intelligence research in nuclear power industry in the 1980s
23.4 Recent application of artificial intelligence in nuclear power plant code
References
Chapter-24---Temporal-data-mining-in-nuclear-si_2021_Nuclear-Power-Plant-Des
24 Temporal data mining in nuclear site monitoring and in situ decommissioning
24.1 Introduction
24.2 Theoretical: Frequent episode and discovery algorithms
24.3 Computational: Application of TDMiner
24.3.1 User interfaces of TDMiner
24.3.2 Verification and validation of TDMiner
24.4 Experimental: In situ decommissioning Sensor Network Test Bed and data collection
24.5 Data analysis and discussions
24.6 Conclusion and future work
References
Index_2021_Nuclear-Power-Plant-Design-and-Analysis-Codes
Index