Dynamic Simulation of Sodium Cooled Fast Reactors

This book provides the basis of simulating a nuclear plant, in understanding the knowledge of how such simulations help in assuring the safety of the plants, thereby protecting the public from accidents. It provides the reader with an in-depth knowledge about modeling the thermal and flow processes in a fast reactor and gives an idea about the different numerical solution methods. The text highlights the application of the simulation to typical sodium-cooled fast reactor. The book • Discusses mathematical modeling of the heat transfer process in a fast reactor cooled by sodium. • Compares different numerical techniques and brings out the best one for the solution of the models. • Provides a methodology of validation based on experiments. • Examines modeling and simulation aspects necessary for the safe design of a fast reactor. • Emphasizes plant dynamics aspects, which is important for relating the interaction between the components in the heat transport systems. • Discusses the application of the models to the design of a sodium-cooled fast reactor It will serve as an ideal reference text for senior undergraduate, graduate students, and academic researchers in the fields of nuclear engineering, mechanical engineering, and power cycle engineering.

Chapter 1 Introduction1.1 General1.2 Basics of Breeding1.3 Uranium Utilization1.4 Components of Fast Reactors 1.5 Overview of Fast Reactor Programs1.6 Need for Dynamic Simulation1.7 Design Basis 1.8 Plant Protection System1.9 Sensors and Response Time1.10 Scope of Dynamic Studies1.11 Modelling Development References Assignment Chapter 2 Description of Fast Reactors 2.1 Introduction 2.2 Fast Breeder Test Reactor (FBTR)2.2.1. Reactor Core2.2.2 Reactor Assembly2.2.3. Sodium Systems2.2.4 Decay Heat Removal2.2.5 Generating Plant2.2.6 Instrumentation and Control2.2.7 Safety2.3 Prototype Fast Breeder Reactor2.3.1 Reactor Core2.3.2 Reactor Assembly2.3.3 Main Heat Transport System2.3.4 Steam Water System2.3.5 Instrumentation and control2.3.6 Safety2.4 Neutronic Characteristics of FNRs2.5 Thermal-Hydraulic Characteristics of FNR References Assignment Chapter 3 Reactor Heat Transfer 3.1 Introduction3.2 Reactor Core3.2.1 Core Description3.2.2 Fuel Pin 3.2.3 Subassembly3.3 Coolant Selection 3.4 Control Material Selection3.5 Structural Material Selection3.6 Heat Generation3.7 Reactivity Feedback3.7.1 Doppler Effect3.7.2 Sodium Density and Void Effects 3.7.3Fuel Axial Expansion Effect 3.7.4Structural Expansion 3.7.5Bowing3.8 Decay Heat3.9 Solution Methods3.9.1 Prompt Jump Approximation3.9.2 Runge Kutta Method3.9.3 Kaganove Method3.9.4 Comparison of different Methods3.9.5 Solution Methodology3.10 Heat Transfer in Primary System3.10.1 Core Thermal Model3.10.2 Fuel Restructuring 3.10.3 Gap Conductance3.10.4 Fuel Thermal Model3.10.5 Solution Technique3.11 Determination of Peak Temperatures: Hot Spot Analysis3.12 Core Thermal Model validation in FBTR and SUPER PHENIX3.13 Mixing of Coolant Streams in Upper Plenum3.13.1Solution Technique3.14 Lower Plenum/Cold Pool3.15 Grid Plate 3.16 Heat Transfer Correlations for Fuel Rod Bundle References Assignment Chapter 4 IHX Thermal Model4.1 Introduction4.2 Experience in PHENIX4.3 Thermal Model4.4 Solution Techniques4.4.1Nodal Heat Balance Scheme4.4.2Finite Differencing Scheme4.5 Choice of Numerical Scheme4.5.1Nodal Heat Balance for Unbalanced Flows4.5.2Modified Nodal Heat Balance Scheme (MNHB)4.6 Heat Transfer Correlations4.7 ValidationReferencesAssignment Chapter 5 Thermal Model of Piping5.1 Introduction5.2 Thermal Model5.3 Solution Methods5.4 Comparison of Piping ModelsReferencesAssignment Chapter 6 Sodium Pump6.1 Introduction6.2 Electromagnetic Pumps6.3 Centrifugal Pump6.3.1 Pump Hydraulic Model6.3.2 Pump Dynamic Model6.3.3 Pump Thermal ModelReferencesAssignment Chapter 7 Transient Hydraulics Simulation7.1Introduction7.2Momentum Equations 7.3Free Level Equations7.4Core Coolant Flow Distribution7.5IHX Pressure Drop Correlations7.5.1 Resistance Coefficient for Cross Flow7.5.2. Resistance Coefficient for Axial Flow7.6 Pump Characteristics7.7 Computational Model7.8 Validation Studies7.9 Secondary Circuit Hydraulics7.9.1 Secondary Hydraulics Model7.9.2 Natural Convection Flow in Sodium-Validation StudiesReferencesAssignment Chapter 8 Steam Generator8.1 Introduction8.2 Heat Transfer Mechanisms8.3 Steam Generator Designs 8.3.1. Conventional Fossil Fuelled Boilers8.3.1.1 Drum Type8.3.1.2 Once Through Steam Generators 8.3.2 Sodium Heated Steam Generators8.4 Thermodynamic Models 8.5 Mathematical Model 8.6 Heat Transfer Correlations 8.6.1 Single Phase Liquid Region 8.6.2 Nucleate Boiling 8.6.3 Dry-Out 8.6.4 Post Dry-Out 8.6.5 Superheated Region 8.6.6 Sodium Side Heat Transfer 8.7 Pressure Drop8.8 Computational Model 8.8.1Solution of Water /Steam Side Equations 8.8.2 Solution of Sodium, Shell, And Tube Wall Equations 8.9 Steam Generator Model ValidationReferencesAssignment Chapter 9 Computer Code Development9.1 Introduction9.2 Organization of DYNAM9.3 Axisymmetric Code STITH-2D 9.4 Comparison of Predictions of DYANA-P And DYANA-HMReferences Chapter 10 Specifying Sodium Pumps Coast-Down Time10.1 Introduction10.2 Impact of Coast Down Time in Loop Type FNR10.3 Impact of Coast Down Time in Pool Type FNR10.4 Considerations for Deciding Flow Coast Down Time10.5 Scram Threshold Vs Coast Down Time 10.5.1. FHT Effect on Maximum Temperatures 10.5.2 FHT to Avoid Scram for Short Power Failure10.6 Secondary pump FHT10.7 Primary FHT for Unprotected Loss of FlowReferencesAssignment Chapter 11 Plant Protection System11.1 Introduction11.2 Limiting Safety System Settings for FBTR 11.2.1 Safety Signals and Settings 11.2.2 Limiting Safety System Adequacy for FBTR11.3 Limiting Safety System Settings for PFBR 11.3.1 Design Basis Events 11.3.2 Core Design Safety Limits 11.3.3 Selection of Scram Parameters11.4 Shutdown System11.5 Event AnalysisReferencesAssignment Chapter 12 Decay Heat Removal System12.1 Introduction12.2 Natural Convection Basics12.3 DHR System Options in FNR 12.3.1 DHR in Primary Sodium 12.3.2 DHR in Secondary Sodium 12.3.3 Steam Generator Auxiliary Cooling System 12.3.4 DHR Through Steam-Water System 12.3.5 Reactor Vessel Auxiliary Cooling System12.4 DHR in FBTR 12.4.1 Heat Removal By Air In SG Casing 12.4.2 Loss of Offsite and Onsite Power with SG Air Cooling 12.4.3 Loss Of Offsite And Onsite Power Without Reactor Trip12.5 DHR in PFBR 12.5.1 Thermal Model 12.5.2 Decay Heat Exchanger (DHX) Model 12.5.3 Hot Pool Model 12.5.4 Air Heat Exchanger Model (AHX) Model 12.5.5 Piping 12.5.6 Expansion Tank 12.5.7 Air Stack/Chimney 12.5.8 Hydraulic Model 12.5.9 DHDYN Validation on SADHANA Loop12.6 Role of Inter Wrapper Flow12.7 Role of Secondary thermal capacityReferencesAssignment Chapter 13 Modelling of Large Sodium-Water Reaction13.1 Introduction13.2 Leak Rate 13.2.1 Water Leak Rate model 13.2.2 Steam leak rate model 13.3 Reaction site dynamics 13.3.1 Spherical bubble model 13.3.2 Columnar bubble model 13.3.3 Solution technique 13.3.4 Validation of Reaction site model13.4 Sodium Side System Transient13.5 Discharge Circuit System Transient13.6 Analysis of Pressure Transients for PFBR13.7 Failure of a greater number of tubes than design basis leakReferencesAssignment

Dr G.Vaidyanathan, B.E., MBA, PhD served the Department of Atomic Energy, India over a period of 38 years until 2010. As a group director of the Fast Reactor Technology group at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, which is devoted to the development of Fast Reactors in India. He is a specialist in numerical and experimental thermal hydraulics and safety analysis. He was one of the key persons involved in the design and development of the experimental Fast Breeder Test Reactor (FBTR) and the Prototype Fast Breeder Reactor (PFBR) in India. He has contributed to the inhouse development of thermal hydraulic computer codes and during commissioning tests in FBTR the predictions of these codes were substantiated. After 2010 Dr Vaidyanathan has been teaching nuclear energy and alternative systems in the Indian universities. He has brought out 4 books on nuclear energy. These books have been welcomed by many professors teaching nuclear subjects in in Indian and Foreign universities. He has developed a video module comprising of 30 lectures on nuclear reactors and safety under the NPTEL program of IIT & IISc and these have been extensively used. Dr Vaidyanathan is a life fellow of the Institution of Engineers India, life member f the Indian nuclear society and Indian society of heat & mass transfer. He has 37 journal publications to his credit. He continues to contribute to the department as a member of the sodium safety panel (SSP) at IGCAR and the Advisory Committees for Safety review of various Projects (ACPSR) at the Atomic Energy Regulatory Board, India.