Severe Accidents in Nuclear Reactors

Corium Retention Technologies and Insights

1st Edition - August 2, 2021
Imprint: Woodhead Publishing
Authors: Arun K. Nayak, Parimal Pramod Kulkarni
Language: English
Paperback ISBN:
9 7 8 - 0 - 1 2 - 8 2 2 3 0 4 - 8
eBook ISBN:
9 7 8 - 0 - 1 2 - 8 2 2 3 0 5 - 5

Severe Accidents in Nuclear Reactors: Corium Retention Technologies and Insights presents an authoritative and practical analysis of the latest severe accident management strategie… Read more

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Severe Accidents in Nuclear Reactors: Corium Retention Technologies and Insights presents an authoritative and practical analysis of the latest severe accident management strategies based on previous events and experiments. Written for utilities and industries operating and researching nuclear cooled reactor power plants, this book presents the exponential growth in research since major nuclear accidents and acts as a guide to retaining molten corium, both inside and outside the reactor vessel. Sections cover the physics behind several complex phenomena occurring during corium coolability, providing the reader with an in-depth understanding by presenting the insights obtained from simulated severe accidents.

In addition, the book validates several severe accident codes and provides evidence on the termination of severe accident progressions to help the reader evaluate the safety of existing reactors and design the next generation of nuclear reactors.

Cover Image
Title Page
Copyright
Table of Contents
Foreword
Nuclear Power – an Integral Part of the Energy Solution for the Current Century
Preface
1 Introduction
Chapter Outline
Abstract
1.1 History of light water reactor safety
1.2 Pre-history of nuclear safety and nuclear safety assessment
1.3 Evolution of siting criteria
1.4 Safety in design of nuclear reactors
1.5 Risk of nuclear power
1.6 The nuclear accidents and lessons learnt
1.7 Evolution of safety in design of new reactors in post-TMI accident scenario
1.8 The accident at Chernobyl, 1986
1.9 Initiation of severe accident research
1.10 Concept of core catcher: How cooling is achieved in the Core Catcher?
1.11 Evolution of small modular reactors
1.12 Another blow to nuclear industry – the Fukushima accident
1.13 Closure
References
2 Progression of severe accidents in water cooled reactors
Chapter Outline
Abstract
2.1 Introduction
2.2 Transient evolution of severe accident progression
2.3 Managing core melt accidents
2.4 Managing core melt accidents in PHWRs
2.5 Closure
References
3 Experiments with high temperature melts: challenges and issues
Chapter Outline
Abstract
3.1 Introduction
3.2 Scaling consideration for simulant materials
3.3 High temperature melt generation
3.4 Thermite melting
3.5 Measurement of high temperatures
3.6 Safety issues in conducting high temperature melts
3.7 Summary
References
4 Corium coolability in PHWRs: In-vessel retention
Chapter Outline
Abstract
4.1 Introduction
4.2 Basis for scaling
4.3 In-calandria corium coolability with decay heat simulation for a prolonged duration
4.4 In calandria corium coolability in stored heat dominated regime
4.5 Integrity of calandria vessel weld joints against high temperature load
4.6 Influence of moderator drain pipe in calandria vessel on retention of molten corium
4.7 Critical heat flux on curved calandria vessel vs the imposed heat flux due to molten corium
4.8 Insights
References
5 Numerical modelling of in-vessel retention in PHWRs
Chapter Outline
Abstract
5.1 Introduction
5.2 Heat transfer in calandria vessel
5.3 CFD simulation of melt pool coolability in calandria vessel in prototypic condition
5.4 Simulation of thermal and structural loads on the calandria vessel
5.5 Closure
References
6 Ex-vessel molten corium coolability
Chapter Outline
Abstract
6.1 Introduction
6.2 Issues in exvessel corium coolability
6.3 Corium coolability under top flooding
6.4 Corium coolability with bottom flooding
6.5 Corium coolability in core catcher with external vessel cooling and top flooding
6.6 Closure
References
7 Molten core concrete interaction and ablation of sacrificial material in ex-vessel scenarios
Chapter Outline
Abstract
7.1 Introduction
7.2 Molten core concrete interaction (MCCI)
7.3 Benchmarking of the model
7.4 Thermal decomposition characteristics of different concretes
7.5 Corium coolability during MCCI under top flooded conditions
7.6 Summary
7.7 Ablation behaviour of sacrificial material
7.8 Application to prototypic condition
7.9 Closure
References
8 Fuel coolant interaction
Chapter Outline
Abstract
8.1 Introduction
8.2 Mechanism of steam explosion
8.3 State of the art
8.4 Fuel coolant interaction experiments
8.5 High temperature experiments with prototypic melt
8.6 Discussions
8.7 Further discussions
8.8 Closure
References
9 Debris bed hydrodynamics, convective heat transfer and dryout
Chapter Outline
Abstract
9.1 Introduction
9.2 Formation and characterization of debris beds formed during severe accident
9.3 Issues in coolability of heat generating debris beds
9.4 Hydrodynamics and heat transfer behavior of irregularly shaped particulate debris bed
9.5 Natural convection heat transfer behavior of a radially stratified particulate debris bed
9.6 Natural convection heat transfer behavior of a large multidimensional debrs bed with volumetric heat generation
9.7 Closure
References
10 Conclusions
Chapter Outline
Abstract
10.1 Introduction
10.2 Summary of severe accident phenomena in nuclear power plants
10.3 Severe accident management strategies
10.4 SAMG for new nuclear plants
10.5 Insights from corium cooling studies
10.6 Way forward
References

Arun K. Nayak

Arun Kumar Nayak is the Outstanding Scientist and Head of the Thermal Hydraulics Section of Bhabha Atomic Research Centre and a Professor at Hombi Bhabha National Institute in Mumbai, India. He is a graduate in Mechanical Engineering from NIT, Rourkela, India, and joined the Reactor Design and Development Group in BARC in 1990 after completing the Orientation Course in Nuclear Engineering from the BARC Training School in Mumbai, India. He obtained his PhD in Nuclear Engineering from Tokyo Institute of Technology, Japan, and has worked at the Interfaculty Reactor Institute of the Technical University Delft, as well as in the Nuclear Power Safety Division of the Royal Institute of Technology, in Stockholm as a Post-doc Scientist. He has also worked at Institute of Nuclear Energy (IKE) in the University of Stuttgart under the DST-DAAD programme. He has received several awards and is a Fellow of Maharashtra Academy of Science and Indian National Academy of Engineering. He has published more than 350 papers in academic journals and conferences, and is an Associate Editor of the journals “Computational Thermal Sciences” and “Frontiers in Energy Research – Nuclear Energy” and is a member of the editorial board of “Life Cycle Reliability and Safety Engineering."

Affiliations and expertise

Outstanding Scientist and Head, Thermal Hydraulics Section of Bhabha Atomic Research Centre Professor, Hombi Bhabha National Institute, Mumbai, India

Parimal Pramod Kulkarni

Parimal Pramod Kulkarni, a Chemical Engineering Graduate from Institute of Chemical Technology, Mumbai, India, joined Reactor Design and Development Group in BARC in 2005 after completing his masters in Chemical Engineering from Indian Institute of Technology, Kanpur. He obtained his PhD in Engineering Sciences from Homi Bhabha National Institute, Mumbai, India, in year 2015. He has also worked at Institute of Nuclear Energy (IKE), University of Stuttgart during 2008-09 in the field of debris bed coolability. He has received the DAE Young Engineer Award, 2012 and group achievement award in 2015. He has published more than 50 papers in academic Journals and Conferences. Currently he is working as Scientific Officer in Thermal Hydraulics Section, Reactor Engineering Division, BARC. He has done extensive research in the field of severe accident management which has contributed towards safety enhancement of Indian Advanced Heavy Water and Pressurized Heavy Water Reactors.

Affiliations and expertise

Institute of Chemical Technology, Mumbai, India

Life Sciences

Physical Sciences & Engineering

Social Sciences & Humanities

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