Enhancing Nuclear Energy Sustainability Using Advanced Nuclear Reactors

ENHANCING NUCLEAR

ENERGY SUSTAINABILITY

USING ADVANCED NUCLEAR

REACTORS

A thesis submitted to the University of Manchester for the degree of

Doctor of Philosophy

in the Faculty of Engineering and Physical Sciences

School of Mechanical, Aerospace and Civil Engineering

Ayah Elshahat

2015

CONTENTS

List of Figures ...... 5

List of Tables ...... 6

Abstract ...... 7

Declaration ...... 9

Acknowledgements ...... 11

Abbreviations ...... 13

Chapter 1 Introduction ...... 15

Chapter 2 Literature review on sustainability models ...... 26

2.1 DEFINITION OF SUSTAINABLE DEVELOPMENT ...... 26

2.1.1 SUSTAINABILITY INDICATORS ...... 29

2.1.2 QUANTIFICATION MODELS FOR SUSTAINABLE DEVELOPMENT ...... 31

2.2 MODELS OF NUCLEAR ENERGY SUSTAINABILITY ...... 36

Chapter 3 Nuclear Energy Sustainability ...... 40

3.1 INTRODUCTION ...... 40

3.2 SUSTAINABILITY INDICATORS FOR NUCLEAR ENERGY ...... 41

3.2.1 ENVIRONMENTAL INDICATORS ...... 43 3.2.1.1 Reactor Safety and Engineered Safety Features ...... 44 3.2.1.2 Environmental impact for nuclear fuel cycle ...... 46

3.2.1.3 Radioactive waste management ...... 46

3.2.2 ECONOMICAL INDICATORS ...... 47

3.2.3 SOCIOPOLITICAL INDICATORS ...... 51 3.2.3.1 Proliferation risk ...... 51 3.2.3.2 Public opinion ...... 54

3.3 FUZZY LOGIC ...... 57

3.3.1 FUZZY LOGIC METHODOLOGY ...... 57

3.3.2 IF-THEN RULE ...... 59

3.3.3 PRESSURE, STATUS, AND RESPONSE ...... 61

Paper 1: Optimum selection of an energy resource using fuzzy logic ...... 63

Paper 2: Assessment of nuclear energy sustainability index using fuzzy logic ...... 65

Chapter 4 Literature review on advanced nuclear reactors and passive safety systems ...... 68

4.1 DIFFERENT GENERATIONS OF NUCLEAR REACTORS ...... 68

4.2 CHARACTERISTICS OF ADVANCED NUCLEAR REACTORS ...... 74

4.3 CONCEPT OF PASSIVE SAFETY SYSTEMS ...... 75

4.4 ADVANCED NUCLEAR REACTORS ...... 77

4.4.1 AP600 ...... 77

4.4.2 CANDU6 ...... 79

4.4.3 AP1000 ...... 80

4.4.4 ABWR ...... 81

4.4.5 VHTR...... 83

Paper 3: Assessment the safety performance of nuclear power plants using Global Safety Index (GSI) ...... 86

Block diagram for GSI model ...... 89

Paper 4: Simulation of the Westinghouse AP1000 response to SBLOCA using RELAP/SCDAPSIM ...... 90

Chapter 5 Results and conclusions ...... 94

5.1 RESULTS OF ASSESSING NUCLEAR ENERGY SUSTAINABILITY ...... 96

5.2 RESULTS OF GLOBAL SAFETY INDEX ...... 99

5.3 RESULTS OF AP1000 MODELLING ...... 101

5.4 RECOMMENDATIONS FOR FUTURE WORK ...... 104

REFERENCES ...... 106

List of Figures

Figure ‎3.1 Sustainability indicators for nuclear energy AP600 Layout . . . . . 41

Figure ‎4.1 AP600 Layout ...... 74

Figure ‎4.2 CANDU6 ...... 75

Figure ‎4.3 Westinghouse AP1000 plant (section)AP600 Layout ...... 76

Figure ‎4.4 ABWR ...... 78

Figure ‎4.5 VHTR ...... 80

List of Tables

Table 4.1 Progression of nuclear power ...... 68

Abstract

The safety performance of nuclear power reactors is a very important factor in evaluating nuclear energy sustainability. Improving the safety performance of nuclear reactors can enhance nuclear energy sustainability as it will improve the environmental indicator used to evaluate the overall sustainability of nuclear energy.

Great interest is given now to advanced nuclear reactors especially those using passive safety components.

Investigation of the improvement in nuclear safety using advanced reactors was done by comparing the safety performance of a conventional reactor which uses active safety systems, such as Pressurized Water Reactor (PWR), with an advanced reactor which uses passive safety systems, such as AP1000, during a design basis accident, such as Loss of Coolant Accident (LOCA), using the PCTran as a simulation code.

To assess the safety performance of PWR and AP1000, the “Global Safety Index” GSI model was developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence, and the consequences of the accident. Only the second indicator was considered in this work.

A more detailed model for studying the performance of passive safety systems in AP1000 was developed. That was done using SCDAPSIM/RELAP5 code as it is capable of modelling design basis accidents (DBAs) in advanced nuclear reactors

Declaration

Candidate name: Ayah Elshahat

Faculty: Engineering and Physical Sciences

Thesis Title: ENHANCING NUCLEAR ENERGY SUSTAINABILITY USING ADVANCED NUCLEAR REACTORS.

I declare that no portion of the work referred to in the thesis has been submitted in support of an application for another degree or qualification of this or any other university or other institute of learning.

Signed: Ayah Elshahat

The Author of this thesis (including any appendices and/or schedules to this thesis) owns any copyright in it (the “Copyright”) and he has given The University of Manchester the right to use such Copyright for any administrative, promotional, educational and/or teaching purposes.

Copies of this thesis, either in full or in extracts, may be made only in accordance with the regulations of the John Ryland’s University Library of Manchester. Details of these regulations may be obtained from the Librarian. This page must form part of any such copies made.

The ownership of any patents, designs, trademarks and any and all other intellectual property rights except for the Copyright (the “Intellectual Property Rights”) and any re-productions of copyright works, for example graphs and tables (“Reproductions”), which may be described in this thesis, may not be owned by the author and may be owned by third parties. Such Intellectual Property Rights and Reproductions cannot and must not be made available for use without the prior written permission of the owner(s) of the relevant Intellectual Property Rights and/or Reproductions.

Further information on the conditions under which disclosure, publication and exploitation of this thesis, the Copyright and any Intellectual Property Rights and/or Reproductions described in it may take place is available from the Head of School of Mechanical, Aerospace and Civil Engineering.

Acknowledgements

First of all, all thanks are due to ALLAH for giving me this opportunity to invest his grants serving all people around the world. This work is only for the sake of Allah. O’Allah, please accept it from me.

I would like to express my deep gratitude and appreciation to

Professor Timothy Abram, my research supervisor, for giving me the opportunity to study at the University of Manchester.

Also for his patient guidance and enthusiastic encouragement along the whole study.

My special thanks are extended to Professor Gianfranco Saiu, and Dr. Monica Linda Frogheri at ANSALDO Energia for their sincere help and advice in making the nodalization of the model.

I would like to express my deep appreciation to Dr. Chris

Allison and the staff in Innovative System Software (ISS) for their great help and guidance in using SCDAPSIM/RELAP.

I would also like to thank departed Professor Naguib H. Aly, for his advice and assistance in keeping my progress on schedule.

During the long journey of my PhD, starting from 2006, many people really loved me and encouraged me to complete this work, for all of them; I gift this work as I will never forget them.

I would like to offer my special thanks to my sincere friends who pushed me and encouraged me to finish my work.

Difficulties made me filter real friends and know people who really concern and love me.

I wish to thank my parents and brothers for their love, support and encouragement throughout my study.

All my love and passion are to my husband Maged and my kids, Yusuf, Khadijah, and Ahmad.

Abbreviations

ACC: accumulator ADS: Automatic Depressurization System CMT: Core Make-up Tank CVS: Chemical and Volume Control System CDF: Core Damage Frequency DVI: Direct Vessel Injection IRWST: In-containment Refueling Water Storage Tank FWS: Start-up Feedwater System LOCA: Loss of Coolant Accident PMS: Protection and Safety Monitoring System PRHR: Passive Residual Heat Removal System PRHR-HX: Passive Residual Heat Removal System Heat Exchanger PRZ: pressurizer RCP: reactor coolant pump RCS: Reactor Coolant System RNS: Normal Residual Heat Removal System RPV: Reactor Pressure Vessel SG: steam generator Sv: Sievert IAEA: International Atomic Energy Agency NRC: Nuclear Regulatory Commission ISS: Innovative System Software KAERI: Korea Atomic Energy Research Institute LPZ: Low Population Zone ABWR: Advanced Boiling Water Reactor

PWR: Pressurized Water Reactor BWR: Boiling Water Reactor AP1000: Advanced Passive 1000 GSI: Global Safety Index OSI: Occupational Safety Index DBA: design basis accidents CSI: Criticality Safety Index

Chapter 1 Introduction

One of the greatest challenges facing humanity during the twenty-first century is to provide safe, clean and sustainable energy supplies. Sustainability can be considered a relationship or balancing act between many factors (social, environmental and economic) that may constantly change. Sustainability is not only limited to the availability of an energy resource but also must consider the needs of future generations.

Nuclear energy is increasingly considered as an attractive energy source that can deliver an answer to increasing worldwide energy demands. Nuclear energy currently provides nearly 7% of our primary energy requirements. It is based on harnessing the very large quantities of energy that are released when the nuclei of certain atoms, such as uranium-235, are induced to split or fission. Estimates suggest that there is sufficient fuel for many decades or even centuries depending on use, but there are major concerns regarding nuclear safety and the disposal of nuclear waste products.

Nuclear energy is attractive as it has limited greenhouse gas emissions and its waste is more compact than any

other energy source. However, it requires investment on a large scale to construct the plant.

The safety of using nuclear energy has great arguments between acceptance and rejection. One of the important applications of nuclear technology is nuclear power plants. A major advantage of nuclear power plants, in contrast with fossil fuelled plants, is that they do not emit greenhouse gases.

Although nuclear power is considered safe, general public perception is often concealing.

The safety performance of nuclear power reactors is a very important factor in evaluating nuclear energy sustainability, as it can improve the environmental indicator used to evaluate the overall sustainability of nuclear energy.

Great interest is given now to advanced nuclear reactors especially those using passive safety components. Advanced reactors have some distinguishing features, such as a standardised design to expedite licensing, reduced capital cost and shortened construction time, a simplified and more rugged design, higher availability and longer operating life, further reduced possibility of core melt accidents, higher burn-up reducing the need to refuel and the volume of waste produced, and many passive or inherent safety features were added which require no active

controls or operational intervention to avoid accidents in the event of malfunction. Therefore, it is important to evaluate the safety performance of nuclear reactors.

In the field of nuclear safety, a safety measurement tool has been trialed to assess the safety in different areas in nuclear applications. The criticality safety index (CSI) was developed by Nuclear Regulatory Commission (NRC) as a number assigned to a package type or freight container containing fissile material, to provide control over the accumulation of these containers. [1]

The occupational safety index (OSI) was developed by the process industry, so that managers and workers can monitor their safety performance and give a warning signal when the probability of errors or deviations at the workspace starts to increase. [2]

A safety index was developed by Korea Atomic Energy Research Institute (KAERI) to quantify plant status [3]. It is only concerned with reducing information loss resulting when simplifying the safety state of the plant into a binary state. But none of these trials consider the safety of nuclear power plants specifically.

The purpose of this study is to investigate whether nuclear energy can be considered a sustainable energy resource by assessing an index for nuclear energy sustainability. This su

index was evaluated through the assessment of three individual indices: Economical, Environmental, and Sociopolitical. The OSI was then evaluated.

The Overall Sustainability Index was evaluated using a Fuzzy Logic tool box (Matlab), as proposed by Lotfi A. Zadeh [4], which has the ability to deal with complex concepts, which are not amenable to a straightforward quantification, such as sustainability. Another important aspect of fuzzy logic is that it uses linguistic variables, thus performing computation with words.

Improvement to the environmental sustainability index was chosen, by improving the safety performance of nuclear reactors through the use of passive safety system instead of active safety systems. Hence, the safety performance of a conventional reactor (which uses active safety systems), such as a current PWR was compared with an advanced reactor (which uses passive safety systems), such as the AP1000, during a design basis accident, such as Loss of Coolant Accident (LOCA). This was conducted using the PCTran code [5] as a simulation code. Assessment of the safety performance of both reactors was done by developing an index which we have called “Global Safety Index”, to investigate whether using advanced nuclear reactors will enhance the nuclear energy sustainability.

The thesis is divided into three parts:

The first part of this work consists of paper 1 and paper 2. Paper 1 discusses the nuclear energy in a general view, considering the nuclear energy option as a solution for some developing countries and taking Egypt (my home country) as a case study. The paper investigates whether the nuclear option will be suitable for Egypt or not by using Fuzzy Multi-Attribute Utility Theory (FMAUT). Considering energy resources available in Egypt as alternatives (nuclear, hydroelectric, gas/oil, and solar) and factors upon which the proper decision will be taken as attributes (economics, availability, environmental impact, and proliferation).

The paper shows that although Egypt depends most on natural gas as a major electricity production source, however, nuclear energy has the highest utility result, then solar energy, gas/oil, and finally hydroelectric. Thus, nuclear energy is considered as an optimum choice of energy sources for Egypt.

The paper concludes that Egypt’s need to use nuclear power becomes inevitable due to the diminishing petroleum and gas reserves and the increasing consumption rates.

In Paper 2, nuclear sustainability is discussed in more details. A sustainability index of nuclear energy is

evaluated. As sustainability is difficult to be defined or measured because it is a complex concept; so in such cases, fuzzy logic offers an effective technical tool to assess sustainability. Fuzzy logic has a systematic approach to handling quantitative situations where traditional mathematics is ineffective.

Nuclear energy sustainability is evaluated through the assessment of Economical, Environmental, and Sociopolitical sustainability indices. The Overall Sustainability Index (OSUS) is then evaluated using two Fuzzy Inference Systems (FIS) [6].

First, each indicator, environmental, socioplolitical, and economics is assessed using its own FIS. We obtain the environmental sustainability index “ENVIOS” to be high, the economic sustainability index “ECOS” to be low, and the socio-political sustainability index “SOCIOS” to be high.

Second, the sustainability index for nuclear energy is then obtained from the second fuzzy inference system using ECOS, ENVIOS, and SOCIOS as inputs and then OSUS is found to be 67%.

Nuclear energy sustainability can be enhanced if one of these indicators can be improved.

As the main concern of public and stake holders is the environmental impact of any energy resource. Therefore, I chose to improve the environmental sustainability indicator to enhance the sustainability index of nuclear energy.

The environmental sustainability indicator also has an effect on the sociopolitical indicator, as it participates in building the public opinion which is a main indicator in assessing the sociopolitical indicator.

The second part consists of paper 3. Having defined and evaluated the environmental sustainability index, the next stage of the work investigated methods of improving the index.

As the safety of nuclear power reactors has a great effect on the environmental sustainability index, therefore, improving environmental sustainability index can be achieved by improving the safety performance of nuclear reactors through the use of passive safety systems instead of active safety systems.

a design basis accident, such as Loss of Coolant Accident (LOCA), using the PCTran as a simulation code.

To assess the safety performance of a PWR and AP1000, the “Global Safety Index” GSI model was developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence, and the consequences of the accident. Only the second indicator was considered in this work. As there are many researches done for studying the probability of accidents occurrence and some models are developed for assessing the consequences of the accident, but not much researches on evaluating the response of safety systems during an accident. That is why response of safety systems during an accident was chosen to be studied

The GSI was evaluated through the determination of:

. Sense time, St: time during which the safety signal sent to the safety system, or time during which the safety system sensed with the accident,

. Response time, Rt: time during which the safety system responds to the accident and starts to mitigate its consequences,

. Recovery time, Rct: time during which the accident seems to be recovered and the reactor reaches the safe conditions, and

. Core Damage Frequency (CDF): which is the probability of core damage.

The GSI was obtained using the Fuzzy Inference System

(FIS), by creating a membership function of the inputs; St,

Rt, Rct, and CDF, so the output of the FIS is the GSI.

GSI = ƒ (St, Rt, Rct, CDF)

After I applied the GSI model using PCTran and Fuzzy logic toolbox in Matlab, it was found that the GSI of the advanced reactor, employing passive safety features, is better than that of the conventional one. This, indeed, will improve the environmental sustainability index and hence the Overall Sustainability Index of nuclear energy.

Some advantages of the proposed model of GSI are:

 It can be easily used for different types of nuclear power reactors.  It can compare the performance of passive and active Safety systems.  It can evaluate the innovative and advanced designs of nuclear power reactors from the safety performance point of view.  It can enhance the design of future safety systems

The third part consists of paper 4. In this paper, a more detailed model for studying the performance of passive

safety systems in AP1000 is developed. That was done using SCDAPSIM/RELAP5 code [7] as it is capable of modelling design basis accidents (DBAs) in advanced nuclear reactors. It is used in the process of licensing nuclear reactors of varied types. The AP1000 was chosen as it received Final Design Approval from the United States Nuclear Regulatory Commission in September 2004 and because AP1000 and its predecessor AP600 are the only PWR designs using passive safety technology currently licensed anywhere in the world.

The model was used to track the response of each of the passive safety systems in the AP1000 during a Loss Of Coolant Accident (LOCA) and the results were analysed. RELAP/SCDAPSIM is found to be a very good modelling tool for simulating different passive components of the AP1000 and to model a small break LOCA. The model is capable of modelling a LOCA in the AP1000 and enabled the investigation of each safety system component response separately during the accident. The model was also capable of simulating natural circulation and other different phenomena.

Validation of the developed model is also performed using results of the NOTRUMP code [8], which is used by Westinghouse, for the same LOCA scenario in licensing the AP1000. The results are found to have the same trends but

some differences in the initiation of safety systems are found.

Chapter 2 Literature review on sustainability models

The concept of ‘sustainability’ is gaining increased interest as a potential solution for the global, regional, and local problems facing society in the late twentieth century. Sustainability offers a new way of looking at problems on both large and small scales, seeking to ensure that the needs of humanity are met in the present without restricting future human requirements. Sustainability can be considered a relationship between many factors (social, environmental and economic realities and constraints) that are constantly changing [9-11]. The United Nations World Commission on Environment and Development coined a definition of sustainable development, in the “Report of the World Commission on Environment and Development :Our Common Future”, which is probably the most well-known in all of the sustainability literature: “development that meets the needs of the present without compromising the ability of future generations to meet their own needs.” [12]

2.1 Definition of sustainable development

“Sustainability” is the capacity to create, test, and maintain adaptive capability. “Development” is the process of

creating, testing, and maintaining opportunity. The phrase “sustainable development” thus refers to the goal of fostering adaptive capabilities and creating opportunities [10]. Sustainability can be defined more specifically to mean: a. Using methods, systems, and materials that will not deplete resources or harm natural cycles. b. A concept and attitude in development that looks at a site's natural land, water, and energy resources as integral aspects of the developments. c. Integrating natural systems with human patterns and celebrating continuity, uniqueness and place making. Environmental context is an important variable to most working definitions of sustainability. This emphasis is expressed in the following composite definition [10]: Sustainable developments are those which fulfill present and future need while using and not harming renewable resources and unique human-environmental systems of a site: air, water, land, energy, and human ecology and/or those of other sustainable systems. Many researchers have tried to develop a definition for sustainability [13, 14]. Johnson, Hays, Center, and Daley [13] developed a definition of sustainability and an associated planning model for sustaining innovations (pertinent to both infrastructure and interventions) within organizational, community, and state systems. The model

assumes a five-step process (i.e. assessment, development, implementation, evaluation, and reassessment/modification) and addresses factors known to inhibit efforts to sustain an innovation. One set of factors concerns the capacity of prevention systems to support sustainable innovations. Another set pertains to the extent to which a particular innovation is sustainable. A sustainability action strategy was presented that includes goals with corresponding sets of objectives, actions, and results that determine the extent of readiness to sustain an innovation. Pearce [14] described the basic scientific foundations upon which the concept of sustainability is built to be thermodynamic foundations and human component. Thermodynamic by identifying two objectives of sustainability: ecosystem impact; by minimizing negative impacts on natural ecosystem, and resource consumption; by minimizing the gain in entropy. The human component is described by identifying three objectives: the motivation for initiators; to maintain standard of living at least as high as currently exist, intergenerational equity; to leave the earth in a condition at least as good as it presently exists, and intragenerational equity; to bring everyone to at least a decent standard of living.

2.1.1 Sustainability indicators

Indicators of sustainable energy development will help in evaluating the role of different energy sources in sustainable development and in making necessary modifications and improvements in systems. A number of studies have found that indicators for sustainability can be divided into three main groups: 1. Environmental pressure indicators that indicate human activities that will directly influence the state of the environmental (e.g. emission rates of toxic substances), and indicators of the state of the environment or environmental quality indicators that indicate the state of the environment. 2. Economic indicators that indicate market competitiveness and natural resource availability. 3. Social activity indicators that describe activities occurring within society: the use of extracted minerals, the productions of toxic chemicals and recycling of materials. In 2005 [15], the Southeast False Creek Steering Committee (SEFC) listed SEFC Sustainability Indicators and Targets to be environmental (energy, water, storm water, solid waste and recycling, urban agriculture, transportation, and SEFC green buildings), social (basic needs, enhancing human capacity, and enhancing social

capacity), and economic (economic security, local self- reliance, ecological economy, and economic advantage).

Recently, Stamford [16] identified sustainability criteria developed in the context of UK electricity generation, comprised 43 techno-economic, environmental and social indicators which were assessed on a ‘life cycle’ basis; taking into account all relevant techno-economic, environmental and social sustainability issues. The indicators and their relevance to each life cycle stage of energy generation were indicated as follows.

 Environmental indicators, Material recyclability; Water eco-toxicity; Global warming potential (GWP); Ozone layer depletion potential; Acidification potential (AP); Eutrophication potential (EP); Photochemical smog creation potential; and Land use and quality.

 Techno-economic indicators, Operability; Technological lock-in; Immediacy; levelised cost of generation; cost variability; and financial incentives.

 Social indicators, Provision of employment; Human health impacts; large accident risk; Local community impacts; Human rights and corruption; Energy security; nuclear proliferation; and Intergenerational equity.

2.1.2 Quantification models for sustainable development

In 2004, Matt Williams [17], presented two information visualization (infovis) systems to explore a 294- dimensional dataset. He developed the environmental dataset from expert knowledge on ecological, economical, and social systems which were used to model future scenarios consisting of 294 measures of environmental sustainability such as urban population, water supply levels, or tonnes of waste. To overcome the complexity of these systems with the large dataset, he developed QuestVis, a tool that applies infovis theories and techniques to improve comprehensibility. In the same context, S. Rashad in [18] dealt with how to choose a set of sustainable development indicators on a national scale, and took Egypt as an example, to explore the efficiency, sufficiency, equity, and quality of life. In [19], an energy sustainability index was developed to evaluate the American energy policy. In [20], Pearce has examined four dimensions that define and categorize the major issues of sustainability as it pertains to engineering and design. Technology, Ecology, Economics, and Ethics served to summarize these. Pearce concluded with a examination of two philosophies of design woven into the GE curriculum to teach engineers

approaches for managing complexity and evaluating the tradeoffs inherent in sustainable engineering. Many models and codes were developed to measure sustainability as discussed in the following [21-25].  Phillis and Andriantiatsaholiniaina [21] developed a model called Sustainability Assessment by Fuzzy Evaluation (SAFE), which provides a mechanism for measuring development sustainability. Ecological (land, water, air, and biodiversity) and human (economical, social, educational, and political) inputs are treated individually and then combined with the aid of fuzzy logic to provide an overall measure for sustainability.  Bryte Energy [22] developed software that can be used to model and test the viability of any number of potential options for energy sustainability, Future Energy Scenario Assessment (FESA). Using environmental, technical and financial parameters FESA can derive the most cost effectivity technology mix to serve all energy sectors while meeting a range of user-definable targets [22].  Evans et. al. [23], used indicators to assess each technology which were price of generated electricity, greenhouse gas emissions during full life cycle of the technology, the availability of renewable sources, efficiency of energy conversion, land requirements,

water consumption and social impacts. Each indicator was assumed to have equal importance to sustainable development and used to rank the renewable energy technologies against their impacts. The ranking revealed that wind power is the most sustainable, followed by hydropower, photovoltaics and then geothermal. Wind power was identified with the lowest relative greenhouse gas emissions, the least water consumption demands and with the most favorable social impacts comparing to other technologies, but requires larger land and has high relative capital costs.  Recently, Doukasa et.al. [24] assessed energy sustainability of rural communities using Principal Component Analysis method. The Principal Component Analysis method is a method frequently used among multivariate techniques to construct Sustainable Development Indicators. It can be used when variables are highly correlated; it reduces the number of observed variables to a smaller number of Principal Components that account for most of the variation of the observed variables and is a large sample procedure. Energy Sustainability Index is calculated as a weighted sum of the normalized version of a few but concise indicators for rural communities. The indicators and the performances are dependent on the communities’ type and characteristics as well as their

development needs and perspectives. Therefore, this method can be used for the calculation of an Energy Sustainability Index for rural communities, supporting the design efforts towards the development and the implementation and monitoring of such communities sustainable energy action plans (SEAPs). Carrera et.al. [25] developed an expert-based set of social indicators which was verified by the European stake holders for the assessment of societal effects of energy systems, by choosing four sample countries France, Germany, Italy and Switzerland. The indicator set covers the four main criteria: security and reliability of energy provision, political stability and legitimacy, social and individual risks, and quality of life. The sampled experts seemed to evaluate large-scale energy technologies such as nuclear and coal power rather critically both in terms of associated costs and risks as well as in terms of acceptance through the public. One of the most pronounced country-specific differences was the French experts evaluation of nuclear power, which was far more positive on all dimensions than that of other experts. German experts seemed to consider public participation far less important than did their colleagues from other countries. Overall it appeared that especially small-scale technologies such as photovoltaics and fuel cells were evaluated more positively by experts in regard to ecological

and societal compatibility as well as future innovation potential. Yigitcanlar et.al. [26] introduced a new urban sustainability assessment model: the Sustainable Infrastructure, Land- use, Environment and Transport model (SILENT). The SILENT Model is an advanced geographic information system and indicator-based comparative urban sustainability indexing model, which aims to assist planners and policy makers in their daily tasks in sustainable urban planning and development by providing an integrated sustainability assessment framework. The model uses an indicator-based assessment system. From many indicators, 30 of the most relevant indicators are selected by using Delphi method to form the indicator system. (The Delphi method is a scientific process, which aims at gathering opinions of different persons (e.g. experts and stakeholder) in order to inform decision-making processes). The model uses the Likert scale. In the Likert scale, values are placed into distribution-free scale, thus potentially bringing researchers, practitioners and public perceptions together for the normalization procedure. In order to convert all values into standard ordinal scale (i.e., Low, Medium-low, Medium, Medium-high, and High), the last step is visualization of the composite index values in an advanced (Geographic Information System) GIS environment.

2.2 Models of nuclear energy sustainability

In [27], Greening and Schneider discussed the sustainability of nuclear energy to meet the electric energy demand. They concluded that nuclear energy hold tremendous potential as an option for long-term future energy. However, Potential risks of proliferation, accidents and other negative consequences of nuclear energy need to be fully examined and included in any public decision and that a strategy for dealing with spent nuclear fuel should probably be included in any plan to expand nuclear capacity in the US. nuclear capacity. Durpel, Yacout, and Wadem[28] developed an integrated system dynamics model of nuclear energy systems performing material flow accounting, environmental impact, economic competitiveness and socio-political analysis; they used the DANESS-code. DANESS, "Dynamic Analysis of Nuclear Energy System Strategies," Version 1.0, is a software code that permits the integrated process modeling of nuclear energy systems for parameter studies, economic analysis, and for variations in the fuel cycle. The software can currently describe 10 nuclear reactor types and 10 fuel types with a cross-flow of fissile material. The reactor and fuel types are stored in a library that can be updated as necessary. The nuclear reactor and fuel type history can be traced as an operating facility to determine the cost of

energy generation per reactor and for the total nuclear system.

Jeong et. Al. [29] developed a ‘must-go path’ scenario, which is the projection of expectations to a sustainable development of the modern civilization for future sustainability, is proposed by extrapolating human history data for the last three decades. Minimizing the use of non- renewable energy resources without further pollution is a key factor to realize the ‘must-go path’. One of the most important parameters in order to realize the ‘must-go path’ scenario is the sustainability of energy production without further pollution. The effect of nuclear reactor technology on sustainability as an option for near-term energy source was discussed. It is concluded that it is important to minimize energy consumption by increasing the efficiency of energy use.

Piera in [30] presented some issues for nuclear energy sustainability development that should be taken into account. These issues are: the risk of nuclear weapons proliferation, the negative consequences of potential nuclear accidents, and the peculiarities of nuclear waste. It was shown that sustainability technical criteria in nuclear energy can be established by: enhanced safety in nuclear reactors and nuclear fuel facilities, high-level exploitation of natural nuclear materials, minimization of the radioactive

inventory in the waste, and development of proliferation resistant technologies.

C. Forsberg [31] introduced two sustainable energy challenges: the need to avoid climate change and the need to replace traditional crude oil as the basis of our transport system. It considers energy systems that tightly couple nuclear, fossil, and renewable energy sources to create nuclear– fossil systems and nuclear–renewable systems. In the longer term, nuclear energy is potentially the enabling technology for the large-scale use of renewable electricity because nuclear energy may be able to provide peak electricity when the sun does not shine or the wind does not blow.

In [32] Adisa discussed driver led decision-making and the changing policy in UK Government on nuclear power. It includes barriers which have prevented new nuclear power construction in the past, and barriers which are still faced by both government and industry in the UK today. Three main drivers were discussed: security of energy supply; diminishing energy generation capacity; and climate change. Barriers were identified through examination of public perception, and policy. It was shown that the reason of delay in nuclear construction that it was not economy attractive, alongside the issue of public perception of radioactive waste. They concluded that the public’s concern

toward nuclear energy appears to be lessening, as the need for low-carbon energy is recognised

From the above literature, one can deduce that:

The most popular indicators used in assessing sustainability are, environmental, economical, and socio-political indicators.

In many researches, sustainability indicators were assessed by judgements of experts and stakeholders.

Some codes used for sustainability assessment, i.e., DANEDSS and DYMOND, are not publicly available.

Fuzzy logic seems a good tool for dealing with sustainability and indicators of sustainability.

The public opinion of nuclear power is moving towards support as a good solution to energy problems in current society.

Fuzzy logic has the ability to deal with complex concepts, such as sustainability assessment, which is not amenable to a straightforward quantification.

Chapter 3 Nuclear Energy Sustainability

3.1 Introduction

All energy systems can be considered as governed by the dynamics of three subsystems: environmental, economic, and sociopolitical. Each indicator (system) can be determined separately according to different subsystems. For instance, for nuclear energy, the environmental indicator should be determined through descriptions of radioactive waste management, safety features of nuclear reactors, and the environmental impact of energy supply. Economic indicators can be determined through costs of mining and milling, building and construction, operation and maintenance, etc. Socio-political indicator can be determined through public opinion, proliferation risk, transportation of nuclear materials, and risk of accidents. After determining each indicator, the overall sustainability index of nuclear energy can be determined as a function of the three indicators [28].

Not all these parameters can be expressed quantitatively. However, some methods can deal with qualitative parameters such as the codes DANESS and DYMOND developed by ANL durpel [33]. These codes provide a good assessment of sustainability; but they are not publicly

available. Fuzzy logic is considered a good approach dealing with qualitative assessment, as it has the ability to run calculations in words (unification) then convert them into numbers (identification.)

The overall sustainability is then a function of the individual sustainability indices.

3.2 Sustainability indicators for nuclear energy

The sustainability indicators for nuclear energy can be divided into three categories; environmental, economics, and socio-political. In this section, each sustainability indicator will be discussed in detail. Figure (3.1) summarizes components of nuclear energy sustainability indicators, under each category based on my model.

Sustainability indicators

For nuclear energy

A Environmental B Economics C Socio-political

Cost of:  Reactor safety  Proliferation and engineered risk(resistance)  Main and auxiliary safety features  Public opinion buildings, major  Environmental components, impact of energy piping, and supply instrumentation  Radioactive  Mining and milling waste  Enrichment of the management uranium and fuel fabrication

 Licensing  Operation/mainten ance/testing  Conditioning the extremely radioactive spent fuel elements plus decontamination  Decontamination of depleted uranium left behind the enrichment  Waste disposal  Energy supply  Decommissioning

Figure 3.1 Sustainability indicators for nuclear energy

3.2.1 Environmental indicators

The Council on Environmental Quality in the USA [34] has defined an environmental indicator as an environmental parameter, a theoretical concept, or an aggregation of data that provides a surrogate representation of some aspects of environmental quality or condition. When dealing with nuclear energy, the safety concerns, radioactive waste management, and environmental impacts are of great importance due to their effects on the public and the environment. This can strongly affect the other two indicators (economic and socio-political) through minimizing accident probability and reprocessing of radioactive wastes. Therefore the environmental indicator is of great importance.

Many factors affect the environmental sustainability indicator of nuclear energy. These include: reactor safety and engineered safety features, the environmental impact of energy supply, and radioactive waste management. Details of each of these parameters, and the way in which the affect the environmental sustainability of nuclear energy, are discussed below.

3.2.1.1 Reactor Safety and Engineered Safety Features

The main objective of nuclear safety concepts is to prevent or mitigate radioactive releases from the nuclear facility to the environment.

The Reactor Site Criteria requires an “exclusion area" surrounding the reactor in which the reactor licensee has the authority to determine all activities, including exclusion or removal of personnel and property, and a "low population zone" (LPZ) which directly follows the exclusion area. It is required by the International Atomic Energy Agency (IAEA) that at any point within the exclusion area boundary and on the outer boundary of the LPZ the exposure of an individual to a postulated release of fission products (as a consequence of an accident) be less than 0.25 Sv total effective dose equivalent [35].

The meteorological factors at a site should be studied to determine dispersion isopleths for radioactive materials released during a postulated accident to reduce the radiation exposures of individuals within the exclusion area and LPZ boundaries. The geologic and seismic characteristics of a site, such as surface faulting, ground motion, and foundation conditions (including liquefaction, subsidence, and landslide potential), may affect the safety of a nuclear power station.

Locating reactors away from densely populated centers is part of the defense-in-depth philosophy (which concerns the protection of both the public and workers and the safety of nuclear installations) and facilitates emergency planning and preparedness as well as reducing potential doses and property damage in the event of a severe accident. The nearest distance to the boundary of a densely populated center containing more than about 25,000 residents must be at least one and one-third times the distance from the reactor to the outer boundary of the LPZ.

To ensure that adequate protective measures can be taken to protect members of the public in the event of an emergency, the characteristics of the site should not preclude development of emergency plans [35]. To prevent plant damage and possible radiological consequences to the public as a result of acts of sabotage, the characteristics of the site should not preclude development of adequate security plans.

Any nuclear reactor has three protective barriers. The first protection barrier, the ceramic fuel and its cladding, retains the radioactive products of the nuclear fission process. The second, the strong metallic primary circuit consisting of the reactor vessel and connecting pipes, retains radioactive material released in the event of fuel damage. The final and ultimate barrier, typified by a large cylindrical containment structure of pre-stressed concrete enclosing the reactor

primary system retains radioactive material that could be released from a primary circuit failure [36].

3.2.1.2 Environmental impact for nuclear fuel cycle

Operation of the nuclear fuel cycle imposes several environmental concerns. The major burdens are: radioactive emissions, risk of accidents, and air emissions of energy used for some stages, such as enrichment and reprocessing [37].

3.2.1.3 Radioactive waste management

One can predominantly distinguish between two types of nuclear waste: spent fuel (in a solid state) and radioactive emissions (in a liquid or gaseous state). Both are produced by nuclear power plants in normal operation. These two forms of waste are dealt with in different ways. The attitude to the former is that of ‘concentration and protection': radioactive contamination of the external environment from spent fuel storage is minimized through several layers of physical containment. To the latter mostly the principle of dilution and dispersion' is typically applied: the emissions of the nuclear industry may therefore lead to increases in ambient radiation levels. The emissions into the atmosphere or surrounding waters from nuclear power plants are typically much lower than those of reprocessing plants. For the latter, after dilution, the additional radiation

doses generated can generally be neglected in comparison to natural levels of radioactivity [38].

3.2.2 Economical indicators

The most important energy costs of a power plant itself, up to and beyond the end of its useful lifetime, are:

* Main and auxiliary buildings, major components, piping, and instrumentation.

* Mining and refining the uranium in the ores,

* Enrichment of the uranium and fabrication of the fuel elements,

* Operation and maintenance.

If fuel is imported, the second and third items may be replaced with fuel price.

Energy debts are energy costs which will have to be paid many generations after nuclear power stations have stopped producing electricity. They are: [40]

 Conditioning the extremely radioactive spent fuel elements plus decontamination or reprocessing.

 Decontamination of depleted uranium left behind the enrichment, and decommissioning.

The overwhelmingly important element in the total cost of nuclear power is the cost of construction (capital cost), the temporal component of which is a significant determining factor. This typically accounts for some 60% to 75% of the generating cost of nuclear power. Levelized energy generation costs per reactor and per region, in case of multi-regional analysis, are important indicators for any economic sustainability assessment [39]. The decomposition of these levelized costs into capital, operation & maintenance and fuel cycle costs is important information as input to macro-economic energy market penetration models such as MARKAL [40], MESSAGE [41], ENPEP [42] and others.

While levelized energy generation costs are useful, a more detailed cash-flow analysis is needed especially for fuel cycle facilities, in order to analyse the investment attractiveness for each facility and thus determine the advantages and disadvantages in deploying certain fuel cycle options which may seriously impact nuclear energy system deployment potential. Uyterlinde et al in [38] developed a generic model tailored by the input data to represent the evolution over a period of usually 40 to 50 years of a specific energy system at the national, regional, state or province, or community level.

Within this system, the basic components in a model are specific types of energy or emission control technology.

Each is represented quantitatively by a set of performance and cost characteristics. A menu of both existing and future technologies can be input to the model, and both the supply and demand sides are integrated, so that one side responds automatically to changes in the other.

The model is able to select that combination of technologies that minimizes total energy system cost [40] which can be used for varied purposes such as:

* to identify least-cost energy systems

* to identify cost-effective responses to restrictions on emissions

* to perform prospective analysis of long-term energy balances under different scenarios

* to evaluate new technologies and priorities for R&D

* to evaluate the effects of regulations, taxes, and subsidies

* to project inventories of greenhouse gas emissions

* to estimate the value of regional cooperation

The model for Energy Supply Strategy Alternatives and their General Environmental Impact (MESSAGE) is a systems engineering optimization model used for medium- to long-term energy system planning, energy policy

analysis, and scenario development. MESSAGE provides a framework for representing an energy system with all its interdependencies including resource extraction, imports and exports, conversion, transport, and distribution, and the provision of energy end-use services such as light, space conditioning, industrial production processes and transportation.

A typical MESSAGE application is constructed by specifying performance characteristics of a set of technologies and defining a Reference Energy System (RES) to be included in a given study/analysis that includes all the possible energy chains that the model can make use of. In the course of a model run, MESSAGE then determines how much of the available technologies and resources are actually used to satisfy a particular end-use demand, subject to various constraints, while minimizing total discounted energy system costs. [41]

The Energy and Power Evaluation Program (ENPEP- BALANCE) is the premier energy systems analysis software in use in many countries.

ENPEP-BALANCE is a nonlinear equilibrium model that matches the demand for energy with available resources and technologies. Its market- based simulation approach allows ENPEP-BALANCE to determine the response of

various segments of the energy system to changes in energy prices and demand levels [42].

The Generation IV International Forum (GIF) Code of Accounts (COA) is a numeric system designed to provide cost information for any component of a project, from design, layout, and procurement of equipment, to final installation [43].

The next generations of nuclear power plants will play an important role in enhancing the economics of nuclear energy. As adopted by the Generation IV International Forum (GIF), the economic goals of Generation IV nuclear energy systems are:

. To have a life cycle cost advantage over other energy sources (i.e., to have a lower levelized unit cost of energy on average over their lifetime).

. To have a level of financial risk comparable to other energy projects (i.e., to involve similar total capital investment and capital at risk) [43].

3.2.3 Sociopolitical indicators

3.2.3.1 Proliferation risk

The Non-Proliferation Treaty (NPT) defines proliferation as the manufacture or acquisition of nuclear weapons or other nuclear explosive devices by countries which do not

currently possess them. Conventionally, the actual detonation of a device has determined the transition from non- nuclear weapons to nuclear weapons status. Recently, this approach has been questioned on the grounds that there are many stages in the acquisition of a nuclear weapons capability. A nation can make all the preparations for the construction of a weapon or the testing of a device without actually “proliferating.” If it is possible to come within hours of a bomb and still not violate the NPT, the traditional definition conceals more than it reveals [44].

In the recent report of nuclear proliferation and safeguards [44], the definition of proliferation has been broadened to encompass any country that has acquired the capability to very rapidly produce a nuclear explosive device, i.e., a nation that has all the components of an explosive on hand ready for assembly. A country which has decided to acquire the components of a nuclear weapon, and has done so, is a nuclear weapons state even if the mechanics of assembling, arming, and detonating the devices remain to be completed.

The International Nuclear Fuel Cycle Evaluation (INFCE) reviewed nuclear proliferation resistance technologies and divided them into three types which are [45]:

 Nuclear Proliferation Resistance focusing on the compositions and forms of nuclear materials

Concerning nuclear materials handled in the fuel cycle, reviews were made on the compositions and forms of materials in which it would be generally difficult to produce metallic plutonium and highly enriched uranium (HEU) and use them for the components of nuclear explosive devices. The following technologies were reviewed:

 Technologies using irradiation before use, spiking and partial decontamination & reprocessing to build up radiation barriers inaccessible to human beings around nuclear materials; and

 Technologies using for mixing & extraction, mixing & conversion and MOX to prevent simple plutonium and uranium substances from being handled

Nuclear Proliferation Resistance focusing on the structures of facilities. Reviews were made on the safety of facility structures and their inaccessibility to thieves; the facilities were collectively constructed on one site to eliminate the need for transporting nuclear materials between facilities and reduce the possibility of theft. Methods for preventing the removal of nuclear materials from facilities in order to use them for the production of nuclear explosive devises were also reviewed. Nuclear Proliferation Resistance focusing on management methods Preventive systems

against the use of nuclear materials for nuclear explosive devises include the safeguards system established by IAEA.

Apart from this system, reviews were made on the proposals that only specified countries should be permitted to use reprocessing and enrichment technology in order to prevent the spread of this technology to other countries. It was also noted that many countries should jointly manage the produced nuclear materials [46].

3.2.3.2 Public opinion

The peaceful use of nuclear energy serves the purpose of environmental protection. The radioactive emissions induced by nuclear energy are in the range of the natural radioactivity and therefore do not present an environmental burden. Presently, nuclear energy is the only large-scale system which produces large amounts of electricity without releasing substances known to adversely affect the global climate.

There is a continuing public concern that the use of nuclear power is inherently associated with a further spread of nuclear weapons and a risk of terrorism [47]. The US Atoms for Peace policy announced in 1953 promoted a policy of international nuclear co-operation based on the condition that nuclear technology transfer would be used exclusively for peaceful purposes and bilateral safeguards arrangements were introduced Several years after its

creation in 1957, the IAEA initiated on-site inspections at nuclear facilities under binding safeguards agreements. The 1970 Treaty on the Non-Proliferation of Nuclear Weapons (NPT) now commits more than 180 countries to refrain from acquiring nuclear weapons and to accept comprehensive IAEA safeguards on all their nuclear activities. The Treaty was extended indefinitely in 1995 [48].

Reactors can be designed that are less susceptible to proliferation attempts. Practical potential for the development and fabrication of such reactors is available. Despite this all nuclear reactors, however newly designed and incorporating whatever progressive proliferation- beneficent techniques, will always involve some proliferation risks. It would be erroneous to assume that totally proliferation-resistant reactors can ever be built.

One of the most effective factors in public opinion is reactor accidents and their risk.

* Reactor accidents

An intrinsic risk of nuclear energy is the occurrence of reactor incidents or accidents. The present generation of nuclear reactors has had a good safety record when one takes the ratio of incident occurrence and operation-years achieved as reference.

* Accident risk

Nuclear power stations are designed so that safety equipment is duplicated and segregated. This results in a robust approach of the engineered system to abnormal operation and fault conditions. The current safety assessment principles state that safety equipment should be operated automatically, and that no human action should be necessary for at least 30 minutes following an accident initiation [49].

Sustainability Assessment

Recall the problem of assessing sustainability, as what is sustainable from an environmental point of view may not be sustainable from the economic and sociological point of view, therefore, each indicator should be evaluated separately through different parameters. Some of these parameters can be assessed either quantitatively (can be expressed in numbers) by specialized codes or qualitatively (can be expressed in according to stake holders' and experts' judgements and reviews).

For the economic sustainability indicator, most of the parameters included, such as the cost of construction, operation and maintenance, energy supply, etc, can be calculated and evaluated. However, for the environmental sustainability indicator, parameters such as safety features of nuclear plants and waste management cannot be

calculated but can be evaluated qualitatively. For the sociological indicator, public opinion and proliferation cannot be assessed quantitatively and for these parameters one should implement qualitative assessment.

As described in Chapter 2, system simulation codes for assessing sustainability, fuzzy logic, and experts' and stake holders' reviews may be used to evaluate sustainability. Most of these codes used to assess sustainability are not publicly available. Fuzzy logic (as a method for evaluating complex concepts that cannot be expressed in values) is an appropriate method for evaluating sustainability indicators to perform calculations in words then convert it into numbers.

3.3 Fuzzy Logic

3.3.1 Fuzzy logic methodology

Fuzzy logic was first proposed by Lotfi A. Zadeh of the University of California at Berkeley in a 1965 paper [4]. He elaborated on his ideas in a 1973 paper that introduced the concept of "linguistic variables", which in this article equates to a variable defined as a ‘fuzzy set’ [50].

Fuzzy Logic is a departure from classical two-valued sets and logic, that uses " soft" linguistic (e.g. large, hot, tall) system variables and a continuous range of truth values in the interval [O,I], rather than strict binary (True or False)

decisions and assignments. Formally, fuzzy logic is a structured, model-free estimator that approximates a function through linguistic input/output associations.

An important aspect of fuzzy logic is that it uses linguistic variables, thus performing computation with words. If a traditional mathematical approach towards sustainability assessment were adopted, such as cost-benefit analysis or algebraic formulas, then certain factors, which are impossible to quantify, would be left out. There are, however aspects of sustainability that cannot be quantified and yet are very important such as, values and opinions. In this area of human thought fuzzy logic performs very well [50, 51].

The following two basic features justify the use of the fuzzy logic reasoning:

a. Fuzzy logic has the ability to deal with complex and polymorphous concepts, which are not amenable to a straightforward quantification and contain ambiguities. In addition, reasoning with such ambiguous concepts may not be clear and obvious, but rather fuzzy. b. Fuzzy logic provides the mathematical tools to handle ambiguous concepts and reasoning, and gives concrete answers (‘crisp' as they are called) to problems fraught with subjectivity.

Sustainability is, indeed, quite subjective. What appears unsustainable for an environmentalist may be sustainable for an economist and the ingredients signifying sustainability may differ for these specialists [52]. To assess sustainability in a fuzzy logic' manner, the following have to be defined:

- Linguistic variables, which best represent the sustainability of the whole system.

- Linguistic rule bases and fuzzy logical operators which express qualitatively the knowledge and the key features of the overall system.

- An identification method to convert fuzzy statements into a single crisp value of overall sustainability.

Linguistic variables

Briefly, a linguistic variable is defined by four items, (a) The name of the variable (e.g. money); (b) Its linguistic values (e.g. ‘much’ and ‘little’); (c) The membership functions of the linguistic values; and (d) The physical domain over which the variable takes its quantitative values.

3.3.2 IF-THEN rule

Fuzzy logic is a scientific tool that permits simulation of the dynamics of a system without a detailed mathematical

description. Knowledge is represented by IF–THEN linguistic rules, which describe the logical evolution of the system according to the linguistic values of its principal characters, that we call linguistic variables. Real values are transformed into linguistic values by an operation called ‘fuzzification’, and then fuzzy reasoning is applied in the form of IF–THEN rules.

A final crisp value is obtained by defuzzification, which does the opposite of fuzzification. A simple example of IF–THEN fuzzy approximate reasoning is the assessment of human happiness based on the popular feeling about the importance of health. Choosing money and health as the principal factors of happiness, the fuzzy rules might be.

_ IF one has ‘much’ money AND ‘good’ health, THEN he is ‘very’ happy.

_ IF one has ‘much’ money AND ‘bad’ health, THEN he is ‘insufficiently’ happy.

_ IF one has ‘little’ money AND ‘good’ health, THEN he is ‘satisfactorily’ happy.

_ IF one has ‘little’ money AND ‘bad’ health, THEN he is ‘insufficiently’ happy.

‘Much’ and ‘little’ are linguistic values of the linguistic variable money; they correspond to the fuzzification of a fixed amount of money. (Good, bad), and (very,

satisfactorily, insufficiently) are, respectively, linguistic values of the state of health and happiness. Defuzzification of the linguistic values ‘very’, ‘insufficiently’, and ‘satisfactorily’ provides a crisp measurement of happiness.

The overall sustainability is then a function of the individual subsystem’s integrity, which will be devised logically via fuzzy logic. This function consists of combinations of IF– THEN rules operating on rule bases derived from expert knowledge. By their nature, such functions are highly non- linear. The term integrity is defined as the degree to which each sustainability variable fulfils the criteria and principles of sustainability.

Criteria and principles of sustainability are therefore recommended critical or target states that the system should satisfy in order to be considered sustainable.

3.3.3 Pressure, status, and response

Each subsystem is evaluated by means of three types of indicator, pressure; status; and response indicator. Status is the present state of a component such as the size of forested land. Pressure is a force tending to change status such as the deforestation rate. Finally, response is the reaction taken to bring pressure to a level that will guarantee a better status as, for example, protecting a given area. The combination of each group of indicators by

means of fuzzy logic provides a measurement of sustainability for each subsystem. To assess each secondary variable, we use three tertiary linguistic variables, pressure integrity (PRESSUREi); status integrity (STATUSi); and response integrity (RESPONSEi). Finally, these tertiary variables are assessed by means of fuzzy inference IF–THEN rules applied to indicators of sustainability, which are the inputs of the system.

Paper 1: Optimum selection of an energy resource using fuzzy logic

Fuzzy Multi Attribute Utility Theory (FMAUT) was applied to compare between different energy resources and for the optimization of an energy resource to be used in Egypt (as an example of a developing country). The energy resources (nuclear, oil/gas, hydroelectric, solar) act as the alternatives and some attributes were chosen to compare between them, (economics, environmental impact, availability, and proliferation). FMAUT was applied by giving an importance to each attribute ,and a linguistic appraisal (based on literature and experts’ knowledge and opinions) to each alternative , then multiply each importance by attribute and add the results of each alternative and finally to compare the centre values to find out the optimum choice. - FMAUT model is a simple and easy model that can be applied for different alternatives using different attributes. -The model allows comparing between different energy resources even renewable energy. -The model has simple calculations and qualitative evaluations of some attributes.

Paper 2: Assessment of nuclear energy sustainability index using fuzzy logic

Nuclear energy sustainability was assessed by the Overall Sustainability Index (OSI) model. That was done using two fuzzy inference systems (FIS). The first FIS was to assess the primary indicators of sustainability (Environmental, Economical, and Sociopolitical) using three secondary indicators (pressure, status, and response) using Fuzzy Logic Toolbox in Matlab. After getting the outputs from the first FIS, they were taken as inputs for the second FIS to evaluate the OSUS. The obtained value of sustainability index is an approximation value not an exact value. Its importance is to give an indication of how good or bad the nuclear energy sustainability is. Fuzzy logic is capable of evaluating such parameters those could not be quantified by using simple computations with words. If any algebraic method or cost-benefit method were used, such parameters would be ignored. This method takes into consideration not only the present situation but also the future plans. The model can be more comprehensive to include many details on each indicator. The OSUS can be improved by improving any of these indicators. Environmental: using passive safety systems and safety by design.

Economical: new reactors with simplified design and modular and standardized design to reduce the capital cost.

Linguistic variables Fuzzy inference system Linguistic variables

Pressure If-then rules ENVIOS Response Defuzzification ECOS Status SOCIOS

Overall Sustainability Fuzzy inference system Index (OSUS) If-then rules Defuzzification

Fuzzy inference systems for assessing OSI

Chapter 4 Literature review on advanced nuclear reactors and passive safety systems

The development of modern nuclear energy can be considered to have occurred in three broad phases:

1. The exploration of the atomic structure and radiation during the late 19th and early 20th centuries, 2. The military developments which drove nuclear science during and after the Second World War, and 3. Eisenhower’s “atom for peace” era and the development of civil nuclear power.

This leads to the first nuclear reactor to produce electricity which was the Small Experimental Breeder reactor EBR-I in Idaho in The United States of America, which started up in December 1951.

4.1 Different generations of nuclear reactors

There are four different generations of nuclear power reactors that can be summarized as follows [53].

• Generation I reactors were the first to be developed. Gen I refers to the prototype and power reactors that launched civil nuclear power, such as Shippingport

(1957–1982) in Pennsylvania, Dresden-1 (1960–1978) in Illinois, and Calder Hall-1 (1956–2003) in the United Kingdom. This kind of reactor typically ran at power levels that were “proof-of-concept”.

• Generation II refers to a class of commercial reactors designed to be economical and reliable. These reactors were intended for a typical operational lifetime of 40 years. Gen II reactors include Pressurized Water Reactors (PWR), CANada Deuterium Uranium reactors (CANDU), Boiling Water Reactors (BWR), Advanced Gas-cooled Reactors (AGR), and Vodo-Vodyanoi Energetichesky Reactors(VVER). Gen II systems began operation in the late 1960s and comprise both PWRs and BWRs. These reactors, typically referred to as Light Water Reactors (LWRs), use traditional active safety features involving electrical or mechanical operations that may be initiated automatically or by the operator action.

Some engineered systems still operate passively (for example, using pressure relief valves) and function without operator control or loss of auxiliary power.

Gen II reactors will remain the predominant type in operation up to 2020 and beyond. With extensions of their operating life to 60 years, many of these reactors are expected to operate past 2035.

• Generation III nuclear reactors are essentially Gen II reactors with evolutionary design improvements.

These reactors have some distinguishing features such as: a standardised design for each type to expedite licensing, reduced capital cost and reduced construction time, a simpler and more rugged design (making them easier to operate) higher availability and longer operating life (typically 60 years) further reduced possibility of core melt accidents (compared to Gen II), resistance to serious damage that would allow radiological release from an aircraft impact, higher burn-up to reduce fuel use and the amount of waste, and burnable absorbers ("poisons") to extend fuel life.

The greatest departure from second-generation designs is that many passive or inherent safety features were added which require no active controls or operational intervention to avoid accidents in the event of malfunction[53].

The Westinghouse 600 MW advanced PWR (AP-600) was one of the first Gen III reactor designs. On a parallel track, GE Nuclear Energy designed the Advanced Boiling Water Reactor (ABWR) and obtained a design certification from the NRC. The first of these units went online in Japan in 1996.Other Gen III reactor designs

include the Enhanced CANDU 6, which was developed by Atomic Energy of Canada Limited (AECL); and System 80+, a Combustion Engineering design.

• Generation III+ reactor designs are an evolutionary development of Gen III reactors, offering significant improvements in safety over Gen III reactor designs certified by the NRC in the 1990s. Examples of Gen III+ designs include:

VVER-1200/392M Reactor of the AES-2006 type

Advanced CANDU Reactor (ACR-1000)

AP1000: based on the AP600, with increased power output.

European Pressurized Reactor (EPR): evolutionary descendant of the Framatome N4 and Siemens Power Generation Division KONVOI reactors

Economic Simplified Boiling Water Reactor (ESBWR): based on the ABWR

APR-1400: an advanced PWR design evolved from the U.S. System80+, originally known as the Korean Next Generation Reactor (KNGR)

EU-ABWR: based on the ABWR, with increased power output and compliance with EU safety standards

Table 4.1 Progression of nuclear power.

Category Description Examples Time frame

Generation Early Shippingport, 1950 I prototypes Dresden, Magnox to 1960+

Generation Commercial PWRs, BWRs, 1960+ II power CANDU to 1980–

Generation Advanced CANDU 6, System 1980– III LWRs 80+, AP600 to 2010–

Generation Evolutionary ABWR , ACR1000, 2010– III+ Designs AP1000, APWR , to EPR, ESBWR 2020+

Generation Revolutionary 2030 + IV designs

Source: U.S. Department of Energy. Years are approximate. PWR, pressurized water reactor; BWR, boiling water rector; LWR, light-water reactor; ABWR, advanced boiling water reactor; APWR, advanced pressured water reactor; EPR, European pressurized reactor; ESBWR, economic simplified boiling water reactor.

Advanced PWR (APWR): designed by Mitsubishi Heavy Industries (MHI).

ATMEA I: a 1,000–1,160 MW PWR, the result of collaboration between MHI and AREVA.

 Generation IV reactors are usually referred to as advanced reactors. Gen IV reactors have all of the features of Gen III+ units, as well as the ability, when operating at high temperature, to support economical hydrogen production, thermal energy off-taking, and water desalination. In addition, these designs include advanced actinide management [54].

Gen IV reactors include:

High temperature water-, gas-, and liquid salt–based pebble bed thermal and epithermal reactors. One such design is the Power Reactor Innovative Small Module (PRISM), a compact modular pool-type reactor developed by GE-Hitachi with passive cooling for decay heat removal.

Traveling wave reactors that convert fertile material into fissile fuel as they operate, using the process of nuclear transmutation being developed by TerraPower. This type of reactor is also based on a liquid metal primary cooling system. It is also being designed with passive safety

features for decay heat removal, and has as a major design goal the minimization of life cycle fuel costs by both substantially increasing the burn up percentage and internally breeding depleted uranium.

Gen IV reactors are two-to-four decades away, although some designs could be available within a decade.

Table (4-1) summarizes the progression of nuclear power reactors.

4.2 Characteristics of Advanced Nuclear Reactors

There are some features of innovative designs which contribute to enhanced performance relative to the current generation of reactors. These features can be summarized as:  Safety features that contribute to enhanced safety, including those that reduce the probability and severity of both core damage and radioactive release following core damage;  Features that contribute to improved economic competitiveness relative to other methods of generating power, including those that reduce construction cost, construction time, operating and maintenance costs, or fuel-cycle costs, and those that improve the reliability or capacity factor;

 Features that contribute to enhanced proliferation resistance, including reduced generation or transport of fissile materials, particularly plutonium; increased technical difficulty of extracting weapons-grade fissile materials from spent fuel; and increased ease of implementing international safeguards;  Features that contribute to enhanced performance with respect to waste management, including those that reduce the generation of some or all categories of radioactive waste or ease the management or disposal of such waste;  Features that contribute to the efficient use of nuclear fuel through means such as higher burn up, recycling, increased plant efficiency or use of fuels other than uranium;  Features that facilitate flexible use of the reactor, for electricity generation, cogeneration applications and process-heat applications [55].

4.3 Concept of passive safety systems

The IAEA [56] defined passive safety systems as components and systems (but not structures) having safety functions, so that there must be at least two states corresponding to the normal function and to the safety function. Then, to change from the normal to the safety state:

- There must be "intelligence" such as a signal or parametric change to initiate action; - There must be power and potential difference or motive force to change states; and - There must be the means to continue to operate in the second state. A component or system can be called passive when all three of these considerations are satisfied in a self- contained manner. Conversely, it is considered active if external inputs are needed. There are, however, other considerations that must be taken into account include: - Reliability and availability in the short term, the long term and under adverse conditions; - Longevity; the equivalent of shelf life, against corrosion or deformation by creep etc; - The requirements for testing or demonstration; and - Simplification and man-machine interaction. Due to the simplicity of passive system design, the cost is expected to be a fraction of the cost of a comparable active system. However, special aspects like lack of data on some phenomena, missing operating experience over wide range of conditions expected, and driving forces must be taken into account [57]. Active and passive safety systems may have comparable reliability; due to limited experience in passive safety

systems compared to utilisation of active safety systems in almost current nuclear reactors [58].

4.4 Advanced nuclear reactors

In this section, some advanced reactors from different generations and of different types will be described and discussed briefly. From Generation III: AP600 and CANDU6, from Generation III+: AP1000 and ABWR, and from Generation IV: Very High Temperature Reactor (VHTR).

4.4.1 AP600

AP600 is a two-loop 600 MWe pressurised water reactor, designed to combine the simplicity and enhanced safety associated with passive safety systems with high levels of reliability. AP600 safety systems are almost all passive, reliant upon gravity, natural circulation, natural convection, evaporation or condensation in place of other power systems. There are three important safety systems, including passive residual heat removal (PRHR), passive safety injection and passive containment cooling. The PRHR system includes a HX which protects the plant against transients. Heat is removed from the Reactor Coolant System (RCS) by natural circulation [59].

Figure 4.1 AP600 Layout

The passive Safety Injection System (SIS) utilises three sources of water including core make-up tanks (CMTs), accumulators and the In-Refuelling Water Storage Tank (IRWST). The CMTs accommodate small leaks at any RCS pressure using only gravity.

The core melt frequency is estimated at 1.7x10-7 per year compared with the ALWR target of 1.0x10-5 per year. There is an automatic depressurisation system (ADS) which eliminates the threat of high-pressure core melt injection following a core melt.

4.4.2 CANDU6

AECL CANDU 6 nuclear power plants have been operating successfully since the early 1980s and produce around 700 MWe per plant.

Figure 4.2 CANDU6

The CANDU 6 design includes an advanced control centre to improve the human machine interface.

An advanced fuel design based on the CANFLEX fuel bundle has been introduced to allow flexibility for different fuels. For the CANFLEX fuel, an advanced 43 element fuel bundle enables the same overall bundle power at 20% lower pin rating compared with the earlier 378 elements fuel design.

Incremental improvements have been made to the safety concept including increased redundancy of components,

simplified containment design and improved fuel thermal margin.

There have been improvements to the power system design, including improved materials for pipes (e.g. higher chromium content) to achieve a 60-year operating life [60].

4.4.3 AP1000

The AP1000 has been developed by Westinghouse as an extension of the AP600 design, as many of the AP600 design and safety studies proved that a two-loop configuration of the passive technology could produce over 1000 MWe with minimal changes.

Figure 4.3 Westinghouse AP1000 plant (section)

The main changes relate to component changes to accommodate the increased rating while retaining safety

margins. These include an increase in the steam generator transfer area and increased coolant pump size. The larger pumps ensure increased inertia of the flywheel and hence an improved safety margin for pump trip transients. The containment is also taller (but not wider) in view of the increased mass and energy present in the reactor system [1].

The target for AP1000 is a construction schedule of 36 months from first concrete to the load of fuel. This can be achieved through design simplifications, which result in fewer components requiring install. Another feature is the modular design employed. Many modules can be built in parallel with other site activities.

4.4.4 ABWR

The Advanced Boiling Water Reactor (ABWR) is a 1315 MWe BWR. The ABWR received design certification from the USNRC in 1997 [61].

Figure 4.4 ABWR

New features of ABWR include a reactor internal pump for the recirculation system to replace the conventional external loop system. A fine motion control rod drive replaced the conventional locking piston control drive.

A reinforced concrete containment vessel is adopted in place of the thick steel containment vessel in past designs. In the new design, there is a thin steel liner to prevent leakages, while concrete provides pressure containment functionality [61].

High-level development goals included core damage frequency less than 10-7, a construction period of less than

48 months, a first refuelling outage of less than 55 days and occupational radiation exposures less than 0.36 man- Sv per year.

Unique feature of the ABWR is its simplified active safety system. The ABWR has three completely independent and redundant divisions of safety systems. These are mechanically separated and have no cross connections as seen in earlier BWRs. They are electronically separated so that each division has access to redundant sources of AC power and for added safety, each has its own dedicated emergency diesel generator.

4.4.5 VHTR

The VHTR is a graphite-moderated, helium-cooled reactor with a thermal neutron spectrum. It is designed to be a high-efficiency system that can supply electricity and process heat to a broad range of high-temperature and energy-intensive processes. The reactor core can be a prismatic block core or a pebble-bed core. The reactor supplies heat with core outlet temperatures up to 1000°C, which enables such applications as hydrogen production or process heat for the petro- chemical industry. Thus, the VHTR offers high-efficiency electricity production and a broad range of process heat applications, while retaining the desirable safety characteristics in normal as well as off- normal events [62].

Figure 4.5 VHTR

It is expected that the VHTR will be constructed in the future as either electricity producing plant with a direct cycle gas turbine or a hydrogen producing plant. The plant size, reactor thermal power, and core configuration is designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle is a once-through very high burnup low-enriched uranium fuel cycle [63].

The graphitic core structure, helium coolant, and coated fuel particles allow the VHTR to withstand accident temperatures without structural damage or fission product

release. This provides a significant amount of inherent safety which eliminates the need for active, and expensive, safety systems such as those in current LWRs. TRISO coated fuel particles represent another intrinsic safety feature of HTGRs. Each fuel particle is essentially its own pressure vessel able to retain fission products. This results in very little radioactive release and plate-out during operation, as has been shown by past HTGR prototype and demonstration plants. For waste storage, TRISO fuel may require less overpacking than traditional LWR fuel, reducing the total amount of repository space required. Further intrinsic safety feature of HTGRs is the helium coolant, which precludes many safety complications due to irradiation of the coolant or corrosion of component materials. Helium does not undergo a phase change at or above reactor operating temperatures which simplifies the mechanical design and operation of the reactor, thereby improving safety. However, helium does not have the same biological shielding effect as water which results in higher radiation exposure in and around the core than traditional LWRs.

Paper 3: Assessment the safety performance of nuclear power plants using Global Safety Index (GSI)

In searching for some ways to enhance nuclear energy sustainability, improving the environmental sustainability indicator was chosen. As the main concern of the public and stake holders are the environmental impacts of energy resource. As in most sustainability definitions, the main goal is not to harm the future needs.

As the safety of nuclear power plants affecting environmental sustainability, a model was developed to assess the safety performance of nuclear power plants, Global Safety Index (GSI) model. The model was applied to both conventional PWR and advanced PWR, AP1000 (uses passive safety systems) to investigate if the advanced reactors and passive safety systems will improve the safety performance of NPPs.

The GSI model was developed by introducing:

1. Probability of accident occurrence,

2. Safety systems’ response in case of accident,

3. Consequences of accident

Only the second indicator will be discussed as both the first and third indicators had been studied and modelled before in many researche.

The GSI was assessed by tracking the safety performance of the reactor during a Design Basis

Accident (DBA) such as LOCA using PCTran as a simulation code. The accident was introduced in an event-time scale to deduce some events:

Sense time: time during which the safety signal sent to safety system (safety system sensed with the accident).

Response time: time during which the safety system responds to the accident and activated to mitigate the accident consequences.

Recovery time: time during which accident recovered and the reactor reaches the safe conditions.

CDF: probability of core damage.

The GSI was evaluated for both conventional and advanced PWR for the same LOCA conditions.

It was found that the AP1000, in most cases, has better response time, recovery time, and CDF than those of the conventional PWR but not a better sense time. Generally, the GSI of the AP1000 is better than that of conventional PWR. This will improve the environmental sustainability indicator of nuclear energy and hence the OSUS.

 GSI model can be easily applied to any reactor type.

 GSI model can compare active and passive safety systems’ performance.

 GSI model can evaluate different types of nuclear reactors (innovative designs)

 GSI model can evaluate the safety design of future nuclear reactors.

 The model has a wide range of applications.

Simulation Design manuals

Sense time Response time CDF

Recovery time

Fuzzy Inference System

Global Safety Index

Block diagram for GSI model

Paper 4: Simulation of the Westinghouse AP1000 response to SBLOCA using RELAP/SCDAPSIM

It was needed to study the safety performance of the AP1000 in a more detailed and more precise model.

A model of AP1000 using RELAP/SCDAPSIM was developed by adopting a model of the Westinghouse Surry reactor pressure vessel and some other components and by making the necessary changes and model some passive safety systems.

A 2-in break LOCA was modelled by opening the LOCA valve in the broken cold leg.

Time sequence of events is as follows:

Reactor SCRAM was generated at low pressurizer pressure of 12.4 MPa.

At reactor low low pressurizer pressure of 11.7 MPa, “S” signal was initiated, CMTs and PRHR were activated and initiated to inject cold borated water to reactor coolant system and to absorb decay heat.

ADS first stage after CMTs drained and CMTs liquid level reached 67.5% of its initial level.

ADS second stage after first stage by 70 seconds.

ADS third stage after first stage by 120 seconds.

Accumulators were activated at pressurizer pressure of 4.83 MPa.

ADS forth stage after CMTs liquid level reached 20% of its initial value.

IRWST initiated at reactor coolant system pressure less than the atmospheric pressure.

The results of the model were validated by comparing to the NOTRUMP code results. Both results were consistent and in a good agreement with small differences in injection initiated times and end of injection time and mass flow rates of injection.

These may be due to many approximation were made in deducing some known design parameters of AP1000 due to lack of available design data in the open literature.

 The AP1000 shows good improvement in safety performance of PWRS and in mitigating LOCA consequences.

 RELAP/SCDAPSIM is a good modelling tool in simulating advanced reactors and passive safety systems and in modelling the LOCA.

 Good model in simulating each passive safety system component of AP1000 in details.

 Good results obtained when comparing with NOTRUMP code although all difficulties in obtaining, deducing, and approximated some design data of the AP1000.

 Short time in building the model.

Chapter 5 Results and conclusions

In this thesis, nuclear energy sustainability has been assessed by developing a model of the “Overall Sustainability Index”. That was done by evaluating three indicators for nuclear energy sustainability which are environmental, economical, and socio-political sustainability indices.

It has also been shown that nuclear energy sustainability can be improved by improving the environmental sustainability index by enhancing the safety performance of nuclear reactors using passive safety systems which are applied in advanced nuclear reactors.

In order to evaluate the safety performance of advanced nuclear reactors, a GSI “Global Safety Index” model was developed by introducing sense time, response time, recovery time, and Core Damage Frequency (CDF). These values were obtained by modelling different cases of LOCA in both an advanced type PWR and a conventional PWR using PCTran program. The AP1000 was taken as an example of advanced PWR, because of its extensive use of passive safety systems. Then a comparison between the two types of PWR was made and it was found that the performance of the advanced type PWR during a LOCA is better than that of a conventional type in both response

time, recovery time, and CDF, but that the conventional type is better in sense time.

A detailed model of the AP1000 was required to study the performance of different passive safety systems’ components in more details. Although PCTran is a good simulation code, but it is not capable of changing any parameter in the design to investigate the effect of that on the LOCA scenario. Also it was needed to track the performance of each of the safety systems separately and how they were activated.

The developed model for AP1000 was used to evaluate the safety performance of AP1000 during a small break LOCA using RELAP5/SCDAPSIM simulation code. Passive safety components in AP1000 showed clear improvement in accident mitigation. RELAP5/SCDAPSIM was found to be a very good modelling tool for simulating different passive components of AP1000 and model a small break LOCA. When comparing the model results with those of NOTRUMP code used by Westinghouse, it was found to be in a good agreement. Differences between two models were caused by assumptions and approximations of some of design parameters of AP1000 due to limitations in the published AP1000 design parameters.

The passive safety systems were found to be capable of depressurizing the reactor coolant system while maintaining

acceptable core conditions and establishing stable delivery of cooling water from the IRWST. The AP1000 was found to have a good performance in mitigating SBLOCA consequences. These findings have important implications for enhancing the safety performance of nuclear reactors. This, in turn, will improve the nuclear energy environmental sustainability index and the overall sustainability index of nuclear energy. However, more research on the economic and socio-political sustainability indicators of nuclear energy needs to be undertaken.

5.1 Results of Assessing Nuclear Energy Sustainability

In the first part (paper 1), Fuzzy Multi Attribute Utility Theory (FMAUT) was applied to optimize an energy resource to be used in Egypt.

It was used to compare between available energy resources in Egypt (presented as alternatives) and the comparison was done using some attributes (such as availability, economics, environmental impacts, and proliferation).

The comparison was done based on specific values for quantitative indicators published in the IAEA, energy association and some based on experts’ judgments and knowledge (qualitative indicators such as environmental impacts and proliferation).

It was found that nuclear energy has the highest utility result, then solar energy, gas/oil, and finally hydroelectric. Thus, nuclear energy is considered as an optimum choice of energy sources for Egypt.

The presidents of Russia and Egypt announced that the two countries are to build Egypt's first nuclear power plant together and they signed an agreement on the plant's construction. The plant would be built at the existing nuclear site in Dabaa, to the west of the city Alexandria, where a research reactor has stood.[80]

But, from my point of view, I do not think that the instable situation of Egypt encourages letting such huge project to come true.

In the second part (paper 2), the Sustainability Assessment by Fuzzy Evaluation (SAFE) model, developed by Phillis and Andriantiatsaholiniaina [21] to assess development sustainability of eight countries, was applied to assess nuclear energy sustainability index.

The model was generalized to include global indicators for sustainability such as Economics, Environmental, and Sociopolitical indicators.

After assessing nuclear energy sustainability indicators, environmental, socio-political, and economics, nuclear

energy sustainability index is roughly estimated to be 76%. This value is a rough value as the model developed has many judgemental evaluations of sustainability indicators. Also, fuzzy logic does not give an exact value for each indicator. For example, when a certain linguistic variable, used in assessing the three primary indicators, is given a linguistic value of “M” (medium), it does not have an exact value. As the linguistic value of “M” has a wide range of values between 0.5 and 0.65.

Therefore, in evaluating every linguistic variable, there are some approximations and judgemental values.

It is very difficult to quantify certain values for both public opinion and proliferation resistance. Fuzzy logic is very appropriate for such cases. However, a linguistic value should be firstly given before used as input in the fuzzy inference system.

For all previous reasons, this model to assess the sustainability index of nuclear energy is considered as a rough, simplified, and judgemental model.

The simplification of the model does not affect the results obtained. When comparing the value obtained for nuclear energy sustainability to other values for different energy sources, obtained in a previous work [paper 1], it is found to be logical and acceptable.

5.2 Results of Global Safety Index

I developed a model for assessing the safety performance of nuclear reactors. I assumed the model to be based on three primary indicators:

1. Probability of accident occurrence,

2. Response of safety systems in case of an accident,

3. Consequences of the accident.

As the first and third indicators were studied and investigated in some researches before, only the second indicator was considered in the study.

I defined four indicators in the model participate in assessing the GSI: sense time, response time, recovery time, and Core Damage Frequency (CDF).

A simulation code, PCTran, was used to model a DBA such as LOCA, to investigate the performance of different components of the safety system during the accident. After defining the four indicators obtained from the simulation code, and CDF from design manuals, they were used as inputs for the FIS to assess the GSI.

As I have some indicators different in units, fuzzy logic is found to be a good tool in dealing with such cases.

The model was applied for both conventional PWR and advanced PWR to investigate the safety performance of both passive and active safety systems, and to check whether using advanced nuclear reactors and passive safety system will improve the GSI or not.

The model is simplified to be only dependent on the performance of safety systems during an accident.

PCTran is a simple simulation code that has many assumptions and approximations in modelling different types of nuclear reactors. However, PCTran was able to demonstrate the response of each safety systems during the different types of LOCAs.

It was clear and easy to determine every indicator (sense time, response time, and recovery time) those contribute to evaluate the GSI by using PCTran.

Definitely, the accuracy of the simulation code cannot be the same as that of the real case.

After the determination of sense time, response time, recovery time, and Core Damage Frequency (CDF), fuzzy logic was used again to evaluate the GSI due to its capability of dealing with these cases where we have many parameters affecting in the same way the required index.

The safety performance of AP1000 is found to be better than that of current PWR in both the response time and the

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recovery time of accident and also in CDF, but the conventional type PWR is better only in sense time than the AP1000. It takes longer time for AP1000 to respond to accident than conventional PWR.

It was found that using a suitable simulation code and tracking the performance of the safety systems during an accident, determining the four parameters (sense time, response time, recovery time, and CDF), and use the fuzzy logic tool box to evaluate the GSI, we can assess the safety performance of any type of nuclear power plants.

PCTran has a limited capability in changing any value of the design parameters of the reactor to investigate the effect of this change on the accident conditions and the response of safety systems. For this reason, a detailed model of AP1000 using a more accurate and flexible simulation code, such as RELAP/SCDAPSIM, was needed.

5.3 Results of AP1000 Modelling

It was needed from assessing GSI for both conventional and advanced PWR to investigate the performance of passive safety system components in the AP1000 in more details.

It was decided to use a more precise and sophisticated simulation code to model the AP1000 and its passive safety system components.

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Many simulation codes can model passive safety systems and advanced nuclear reactors but they are not publicly available.

I chose to use TRACE code in the beginning but I faced some problems in getting thew license to be used within University of Manchester.

I tried after that to get a license to use RELAP/SCDAPSIM, and I succeeded in having a license for a year.

To build my model, I adopted the model of Westinghouse Surry reactor pressure vessel and some other components, and made necessary changes. I modeled some passive safety systems in AP1000.

The main problem in building the model is the lack of AP1000 design data available in the open literature.

To overcome this problem, I used the design parameters found in Westinghouse published work and from other published researches on AP1000. The other design data, some of them could be deduced, some were approximated, and the rest were assumed to be the same as in the Surry reactor.

To validate my model, I compared RELAP/SCDAPSIM model results to the available results of the NOTRUMP code used by Westinghouse for the same accident conditions.

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The results showed almost the same trend with small differences in the initiations times of injection, end times of injection, and flow rates

The RELAP/SCDAPSIM model of Surry reactor (2-loop PWR) was adopted. Some modifications were done to model passive safety components in AP1000 and some changes in some parameters were made. Many assumptions and approximations were made in the model due to the lack of many design parameters of AP1000.

RELAP/SCDAPSIM was found to be a very good modelling tool for simulating different passive components of the AP1000 and to model a small break LOCA. The model was capable of modelling a LOCA in the AP1000 and enabled the investigation of each safety system component response separately during the accident. The model was also capable of simulating natural circulation and other different phenomena.

When comparing the results with that of the NOTRUMP code used by Westinghouse, they were found to have the same trends but some differences in the initiation of safety systems were found. These differences are caused by the assumptions and approximations made by both codes and also because of the many assumed values for the design data used in the RELAP model due to the lack of published values for the AP1000 design parameters. However, the

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results show the same trend in the response of each safety system during the LOCA, i.e. the increasing flow rate, the decreasing pressure, etc.

5.4 Recommendations for future work

As a future work, some research points can be investigated as the following:

 A study of the economics and socio-political indicators of advanced reactors can be made to check if the economic sustainability index and socio-political sustainability index will be improved by using advanced reactors. The overall sustainability index of nuclear energy can then be re-assessed for advanced reactors to see if it will make any improvement.

 Assessment of the accident probability can be done by studying safety by design of each reactor type.

 A model of assessing the consequences of the accident releases can be developed then The GSI model can be reassessed.

 The Global Safety Index can be generalized by taking into consideration the probability of accidents’ occurrence in addition to the response of the reactor safety systems in case of accident and also the consequences of the accident, i.e. amounts of releases of radioactive materials.

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Further investigation of AP1000 instrumentation and control, responsible of sense time and sending “S” signal to initiate the safety systems, is required to know the reason that AP1000 has longer time to sense accidents than conventional PWR and to find out some ways to enhance the sense time.

Further research should be done to investigate the effect of each component of the passive safety systems separately.

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Nuclear Engineering and Design 239 (2009) 3062–3068

Contents lists available at ScienceDirect

Nuclear Engineering and Design

journal homepage: www.elsevier.com/locate/nucengdes

Optimum selection of an energy resource using fuzzy logic

Ayah E. Abouelnaga ∗, Abdelmohsen Metwally, Mohammad E. Nagy, Saeed Agamy

Nuclear Engineering Department, Faculty of Engineering, Alexandria University, 21544 Alexandria, Egypt article info abstract

Article history: Optimum selection of an energy resource is a vital issue in developed countries. Considering energy Received 14 June 2008 resources as alternatives (nuclear, hydroelectric, gas/oil, and solar) and factors upon which the proper Received in revised form 31 August 2009 decision will be taken as attributes (economics, availability, environmental impact, and proliferation), Accepted 7 September 2009 one can use the multi-attribute utility theory (MAUT) to optimize the selection process. Recently, fuzzy logic is extensively applied to the MAUT as it expresses the linguistic appraisal for all attributes in wide and reliable manners.

The rise in oil prices and the increased concern about environmental protection from CO2 emissions have promoted the attention to the use of nuclear power as a viable energy source for power generation. For Egypt, as a case study, the nuclear option is found to be an appropriate choice. Following the introduction of innovative designs of nuclear power plants, improvements in the proliferation resistance, environmental impacts, and economics will enhance the selection of the nuclear option. © 2009 Elsevier B.V. All rights reserved.

1. Introduction 2. Available energy resources

One of the greatest challenges facing humanity during the 2.1. Conventional energy sources twenty-first century is to provide safe, clean and sustainable energy supplies. About 80% of the world’s energy is currently supplied by fossil Throughout history, the use of energy has been central to the fuels such as coal, oil and gas. Present estimates suggest that, at cur- functioning and development of human societies. But during the rent consumption rates, there are over 200-years’ worth of coal left, nineteenth and twentieth centuries, humanity learned how to har- 60-years’of gas, and 40-years’ of oil. Fossil fuels are hydrocarbons, ness the highly concentrated forms of energy contained within and their combustion releases carbon dioxide into the atmosphere, fossil fuels. These provided the power that drove the industrial one of the main causes of the human-induced greenhouse effect. revolution, bringing unparalleled increases in affluence and pro- The majority of the world’s scientists now believe that ductivity to millions of people throughout the world. As we enter anthropogenic greenhouse gas emissions are causing the earth’s the third millennium, however, there is a growing realization that temperature to increase at a rate unprecedented since the ending the world’s energy systems will need to be changed radically if they of the last ice age. This is very likely to cause significant changes in are to supply our energy needs sustainability on a long-term basis. the world’s climate system, leading to disruption of agriculture and The world’s current energy systems have been built around ecosystems, to sea level rises that could overwhelm some low-lying the many advantages of fossil fuels, and we now depend over- countries, and to accelerated melting of glaciers and polar ice. whelmingly upon them. Concerns that supplies will ‘run out’ in the The combustion of biofuels such as wood or other biomass mate- short-to-medium term have probably been exaggerated. Contin- rial gives us bioenergy. To be sustainable, the forests that provide ued discovery of new reserves and the application of increasingly traditional wood fuel need to be re-planted at the same rate as advanced exploration technologies are being accredited. Neverthe- they are cut down. The incomplete combustion of wood can also less it remains the case that fossil fuel reserves are ultimately finite. release a mixture of greenhouse gases with a greater overall global In the long term they will eventually become depleted and substi- warming effect than can be offset by the CO2 absorbed by growing tutes are urgently needed (Boyle, 2004). replacement trees. Modern bioenergy power plants burn straw and forestry wastes. Hydroelectricity is the power from flowing water, a source which has been used by humanity for many centuries. In 2000, it con- tributed over 17% of world electricity. Its original source is the sun; ∗ water evaporated from oceans falls as rain or snow into rivers, Corresponding author. Tel.: +20 105614004. E-mail address: [email protected] (A.E. Abouelnaga). where its flow can be harnessed using water wheels or turbines.

Larger installations can have adverse environmental effects, but 2.4. Available energy resources as alternatives and the relevant smaller projects may have little, if any, impact. attributes

2.2. Renewable energy sources For applying the optimization techniques, it is useful to treat different energy resources as alternatives (e.g. nuclear energy, Fossil and nuclear fuels are often described as non-renewable hydroelectric energy, gas/oil energy, and solar energy). Param- because supplies are finite and will eventually run out. Renewable eters affecting the optimum selection are referred as attributes fuels are those energy sources that will not run out in the future. (e.g. economics, availability, environmental impact, and prolifer- Most renewable energy sources originate from the sun (solar ation). energy), while tidal energy originates from the gravitational pull of the moon, and geothermal energy results from heat trapped below 3. Fuzzy multi-attribute utility theory the surface of our planet. Renewable energy sources can fall in five categories as explained 3.1. General below. Fuzzy logic was first proposed by Lotfi A. Zadeh (Zadeh, 1965) 2.2.1. Solar energy of the University of California at Berkeley in a 1965 paper. He elab- It can be used directly in two different ways. The heating part orated on his ideas in a 1973 paper that introduced the concept of the suns rays can be used to directly produce hot water (solar of “linguistic variables”, which in this article equates to a variable thermal), while the light energy can directly produce electricity defined as a fuzzy set (Zadeh, 1973). from photovoltaic cells (PV). Fuzzy logic is a departure from classical two-valued sets and There are, however, a number of indirect ways in which solar logic, that uses “soft” linguistic (e.g. large, hot, tall) system vari- energy can be utilized. ables and a continuous range of truth values in the interval [0,1], rather than strict binary (true or false) decisions and assign- 2.2.2. Wind energy ments. It has been utilized for mills and pumps for hundreds of years, Formally, fuzzy logic is a structured, model-free estimator that but has been harnessed to produce electricity during the past sev- approximates a function through linguistic input/output associa- eral decades. Wind ‘farms’ have been set up, and around the world, tions. both inland and offshore. Fuzzy rule-based systems apply these methods to solve many types of “real-world” problems, especially where a system is dif- 2.2.3. Wave energy ficult to model, is controlled by a human operator or expert, or The technology is still in the early development phase but sev- where ambiguity or vagueness is common. A typical fuzzy system eral devices have been tested in pilot projects. The most widely consists of a rule base, membership functions, and an inference used is the Oscillating Water Column which uses waves to push air procedure. through a turbine which generates electricity. 3.2. Fuzzy logic advantages 2.2.4. Tidal energy It’s utilized by allowing the tide to build up a head of water The following two basic features justify the use of the fuzzy logic behind a barrier, then allowing the water to flow through a turbine reasoning: to produce electricity. The effectiveness depends on the position of the moon, being greatest at full and new moons. Tidal currents (a) Fuzzy logic has the ability to deal with complex and polymor- can also be used to generate power, by installing turbines in the phous concepts, which are not amenable to a straightforward flows, and that is seen as a more likely way ahead than building quantification and contain ambiguities. In addition, reasoning large invasive barrages. with such ambiguous concepts may not be clear and obvious, but rather fuzzy. 2.2.5. Geothermal energy (b) Fuzzy logic provides the mathematical tools to handle ambigu- It’s an energy taken from hot regions below the surface. The ous concepts and reasoning, and finally gives concrete available heat varies from place to place. answers (‘crisp’ as they are called) to problems fraught with All of these renewable resources hold out a great promise for subjectivity. the future. At present, they are more expensive than fossil fuels but their sustainability and lesser environmental impacts make them Sustainability is, indeed, quite subjective. What appears unsus- attractive alternatives to fossil fuels. tainable for an environmentalist may be sustainable for an economist and the ingredients signifying sustainability may 2.3. Nuclear energy differ for these specialists (Phillis and Andriantiatsaholiniaina, 2001). Nuclear power has grown in importance since its inception just Another important aspect of fuzzy logic is that it uses linguistic after World War II and now supplies some 7% of world primary variables, thus performing computation with words. If a traditional energy. A major advantage of nuclear power plants, in contrast mathematical approach towards sustainability assessment were with fossil fuelled plants, is that they do not emit greenhouse gases. adopted, such as cost-benefit analysis or algebraic formulas, then Also, supplies of uranium, the principal nuclear fuel, are sufficient certain factors, which are impossible to quantify, would be left for many decades – and possibly centuries – of supply at current out. There are, however, aspects of sustainability, which cannot use rates. However the use of nuclear energy gives rise to problems be quantified and yet are very important as, for example, values arising from the routine emissions of radioactive substances, some and opinions. In this area of human thought fuzzy logic performs difficulties of radioactive waste disposal, and dangers from the pro- successfully (Zimmermann, 1991). liferation of nuclear weapons material. To these must be added the Fuzzy logic is used here in order to compare between many possibility of major nuclear accidents. alternatives, energy sources, according to many attributes. In other 3064 A.E. Abouelnaga et al. / Nuclear Engineering and Design 239 (2009) 3062–3068 words, the fuzzy logic is used to assess the sustainability of each Australian company Worley Parsons to take over the consultation energy source to find the more sustainable one. contract (NTI, 2009).

3.3. The procedure (Nguyen and Walker, 1966) 4.2. Available energy sources: advantages and disadvantages

The procedure used to utilize fuzzy logic in getting the opti- 4.2.1. Nuclear energy mum energy resource can be summarized through the following The rise in oil prices and the increased concern about environ- steps: mental protection from CO2 emissions have promoted the attention to the use of nuclear power as a viable energy source for power generation. Step 1: Prepare a list of linguistic appraisal for the assigned Also Egypt’s need to use nuclear power became inevitable due attributes, if possible, five reasons for each appraisal. to the diminishing petroleum and gas reserves and the increasing Step 2: Assign an importance to each attribute and transform it to consumption rates. fuzzy numbers. Advantages of nuclear energy: Step 3: Multiply importance by attribute. Step 4: Add the results of each alternative. • Nuclear reactions release a million times more energy, as com- Step 5: Compare the central values of alternatives’ results. pared to hydro or wind energy. Hence, a large amount of electricity can be generated. Presently, 12–18% of the world’s 4. Case study electricity is generated through nuclear energy. • The biggest advantage of nuclear energy is that there is no release In this section, Egypt is taken as a case study and some energy of greenhouse gases, which cause global warming and climate resources available in Egypt are taken to compare between them changes (carbon dioxide, methane, ozone, chlorofluorocarbon), in order to get the optimum energy resource. during nuclear reaction. So, there is very little effect on the environment. 4.1. Current status of energy in Egypt • Nuclear reactors make use of uranium as fuel. Fission reaction of a small amount of uranium generates large amount of energy. Cur- Egypt’s fast growing demand for electricity requires signifi- rently, the high reserves of uranium found on Earth, are expected cant investment in generation capacity each year (the increase to last for another 100 years. in demand for electricity in Egypt averaged about 7% during • High amount of energy can be generated from a single nuclear 1997/98–2003/04 and is expected to remain in the 6–7% range power plant. over the next 10 years). Installed capacity of electric power was • Also, nuclear fuel is inexpensive and easier to transport. 18,119 MW in 2003/04, of which 84% comprised thermal power (8% of which is provided by the private sector through 3 Independent Disadvantages of nuclear energy: Power Producers, IPPs). The remaining capacity was attributed to hydropower (15%) and wind (1%). Peak load reached 14,735 MW, • The waste produced after fission reactions contains unstable and about 90% of the thermal power production was based on nat- elements and is highly radioactive. It is very dangerous to the ural gas (Feller, 2008). environment as well as human health, and remains so, for thou- In pursuance of its reform agenda, the Egyptian government has sands of years. It needs professional handling and should be kept set an ambitious renewable energy program to generate 500 MW isolated from the living environments. The radioactivity of these of solar energy, 600 MW of wind power, and 600 MW of hydroelec- elements reduces over a period of time, after decaying. tric power by 2017. Egypt is building a new hybrid power plant • The impact of serious nuclear accident can have a very large – the Integrated Solar Combined Cycle power plant – at Kureimat immediate impact and also a legacy for many years. as a BOOT project, which will have 30 MW of solar capacity out of a total planned capacity of 150 MW. Egypt has also built a wind Egypt is considered one of the first countries that used peace- farm at Zafarana that has been operational since 2004 at an output ful nuclear energy in the region where there were many attempts capacity of 80 MW. A Netherlands-funded project is also building through establishing the Nuclear Energy Committee in 1955 and 60 MW worth of wind power units in the Suez Canal area. Egypt is the Nuclear Energy Institution in 1957. also working with Nuclear power. It has a 22-MW nuclear research The first nuclear reactor in Egypt was operated in 1960 in reactor at Inshas in the Nile Delta, built by INVAP S.A. of Argentina, Anshas for the purpose of making nuclear researches and pro- which began operation in 1997. In March 2008 Egypt also signed an ducing radioactive isotope to serve the industrial and medical agreement with Russia to assist in building Egypt’s first 1000-MW development as well as search for petroleum in wells and gas tubes nuclear plant at al-Daba’a. soldering. Its capacity was 2 MW. The reactor was developed over Egypt plans to expand electricity capacity to 32,000 megawatts the past 15 years to cope with hi-nuclear technology in operation (MW) over the next 5 years. The additional capacity will come prin- and control. cipally from 11 new thermal plants and expansions: Kureimat 2 and A new multi-purpose and peaceful usage research reactor was 3, Talkha, Tabbin, Nuberiya 3, Cairo West, SidiKrier, el-Atf, Abu Qir, established in 1997. However, few amendments were needed to Ain Sokhna and Sharm el-Sheikh. In 2005, nearly 75% of Egypt’s operate safely to the maximum capacity of 22 MW. electric generating capacity was powered by natural gas, some 14% Then, it was reopened in 2003 and worked regularly ever by petroleum products, and the remaining 12% by hydroelectric, since. As for producing electricity from nuclear energy. Egypt has mostly from the Aswan High Dam according to the IEA (Energy launched establishing a nuclear power station to generate electric- Information Administration, 2008). ity and water desalination in 1967. Egyptian Energy Minister Hassan Younus announced that Egypt In 1981, the first electricity generating station from nuclear intends to award a contract by 2010 for the provision of two nuclear energy was allocated in Al-Daba’a region on an area of 12 km2. But power reactors, which will be ready to operate by 2017 or 2018 at the project stopped in the aftermath of Chernople station incident the latest. At the same time, Egypt will begin work on a second set in 1986. The discovery of natural gas had a great effect on closing of two power reactors to be operational by 2022. Egypt chooses the the nuclear file for producing electricity temporarily. A.E. Abouelnaga et al. / Nuclear Engineering and Design 239 (2009) 3062–3068 3065

After the technology development of nuclear reactors, Egypt The High Dam of Aswan in Egypt was started in1960 in President launched to establish electricity generating stations from nuclear Gamal Abdul Naser reign and it was completed in July 21-1970, and energy especially that Egypt has many advantages such as the It now powers twelve generators each rated at 175 MW, producing equipped site in Al-Daba’a and a big group of experts. The usage a hydroelectric output of 2.1 GW. of nuclear energy will make a leap in many industries such as in The High Dam was a great idea as it has not only provided the petroleum by using thermal energy generated from nuclear energy electricity for almost all the villages and some of the areas around it, (Egypt States Information Service, 2006). but it has also provided many other benefits for the government as well as the people. For example the dam reduced the effects of dan- 4.2.2. Solar gerous floods and threatening droughts in the period of 1964–1985, Like nuclear energy, solar energy does not put CO2 into the A new fishing industry has been created around Lake Nasser, and atmosphere, and so needs to be considered if global warming has of course it had a part in the agriculture field as it releases on aver- to be avoided. age 55 billion m3 water per year of which some 46 billion m3 are Although solar energy would be expensive enough to put diverted into the irrigation canals In Nile valley and delta, almost 8 nations that decided to depend on it alone at a serious economic million Fadden benefit from these waters producing on average 1.8 disadvantage compared to nations that were not constrained to rely crop per year. However, as it was a great idea and useful it also has on it. caused some issues such as It flooded much of lower Nubia (A region Advantages of a solar energy system: in south Egypt) and over 90,000 people were displaced, it also has flooded valuable archaeological sites, and the Mediterranean fish- • Solar energy is renewable, plentiful and free. ing declined after the dam was finished because nutrients that was • A solar energy system is environmentally friendly and will not used to flow down the Nile to the Mediterranean were trapped pollute the air. behind the dam (Hydroelectrical Energy of the High Dam in Egypt, • Solar panel maintenance is minimal. 2007).

Disadvantages of a solar energy system: 4.2.4. Oil/gas Fossil fuels include oil formed out of organic matter deposited • High capital cost and high maintenance cost. and decomposed under the earth’s surface for a very long time. • The need to store energy, because of daily, hourly and weekly Advantages of oil: (from clouds) and seasonal availability. • The need to transport the energy long distances. This might put • Easily combustible, and produces high energy. cloudy regions at high latitudes at a severe economic disadvan- • Widely and easily distributed all over the world through rail and tage (McCarthy, 1995). sea tankers. • Inexpensive. Egypt lies in the North African sun belt with flat desert topog- raphy and perennially clear skies favorable to commercial solar Disadvantages of oil: technologies. Annual solar concentration averages 2300 KWh per square meter. While photovoltaic (PV) solar panels are used to • Oil is a non-renewable energy source that takes millions of years power some low-energy applications such as telecom relay tow- to form and therefore once existing and any new reserves are ers and highway billboards, consumer initiatives to encourage the depleted there is no way to obtain more. use of solar water heating have failed to generate widespread sup- • Oil spills as well as evaporation and fumes pollutes the environ- port. Egypt’s first solar steam plant, which went online in 2004, ment. replaced boilers that burned mazout (heavy oil) to produce steam • Used oil is difficult to recycle. for a pharmaceutical plant. Egypt is also among candidates for a • The price of oil is rising, especially if the real cost of its carbon is 550 billion dollar project to build solar collectors in North African included. deserts to supply 15% of Europe’s energy by 2020 (McGrath, 2009). • Burning it produces carbon dioxide a ‘greenhouse gas’, a major cause of global warming. 4.2.3. Hydroelectric • Oil contains sulphur which when burnt forms sulphur dioxide The hydroelectric energy provided about 19% of the world elec- and sulphur trioxide – these compounds combine with atmo- tricity in 2003 and it reached for over 63% in year 2005. The idea of spheric moisture to form sulphuric acid, leading to ‘acid rain’ the hydroelectric system is creating electricity by using the water. (Fossil Fuel, 2009). Advantages of hydroelectric energy:

Advantages of natural gas: • Hydroelectric power has been feted as a cheap, clean and renewable source of energy. • • Does not consume its fuel as the water may be used downstream Is more environment friendly than oil or coal, as natural gas emits for example in irrigation and for drinking water. 30% less carbon dioxide than burning oil, 45% less carbon dioxide than burning coal, and 60–90% less smog-producing pollutants. • Disadvantages of hydroelectric energy: Cheap and can be safely stored and burned. • Most of the natural reserves of natural gas fields are still under- utilized. • Dams are extremely expensive to build and must be built to a very high standard and it must operate for many decades to become profitable. Disadvantages of natural gas: • The building of large dams can cause serious geological damage. • Hydropower dams can damage the surrounding environment and • It is a non-renewable energy resource. Its availability is finite. alter the quality of the water by creating low dissolved oxygen • Is highly volatile (highly flammable) and can be dangerous, if levels, which impacts fish and the surrounding ecosystems. handled carelessly. 3066 A.E. Abouelnaga et al. / Nuclear Engineering and Design 239 (2009) 3062–3068

• Is colorless, odorless and tasteless that makes detection of its leak Table 1b very difﬁcult. Reasoning for assigned linguistic appraisal (hydroelectric energy). • In gas pipelines, a substance (contains carbon monoxide) that has Attribute Linguistic Explanation a strong odor is added to help detect a leak. But, these substances appraisal may be harmful and cause deaths. Economics M Very inexpensive once dam is built • Constructing and managing such pipelines cost a lot. Government has invested heavily in building dams

In 2007, Egypt produced 664,000 barrels of oil per day (bbl/d) Availability H Renewable source of energy continuing its fall from a high of 950,000 bbl/d in 1995. Yet hav- Does not consume its fuel, water may be ing consumed 653,000 b/d in 2007, production was sufficient to used in irrigation and for drinking water prevent Egypt from becoming a net importer of oil as some had Environmental L Dam collapse usually leads to loss of life predicted. Production and consumption of natural gas continue to impact Dams have affected fish environmental rise with a total of 1.9 trillion cubic feet (Tcf) produced and 1.3 Tcf damage for areas flooded (backed up) consumed in 2006, making Egypt a net gas exporter. and downstream Egyptian oil production comes from four main areas: the Gulf of Proliferation H Not applicable Suez (about 50%), the Western Desert, the Eastern Desert, and the Sinai Peninsula. Most Egyptian production is derived from mature, relatively small fields that are connected to larger regional produc- Table 1c tion systems. Reasoning for assigned linguistic appraisal (gas/oil energy). Due to major recent discoveries, natural gas is likely to be the primary growth engine of Egypt’s energy sector for the foreseeable Attribute Linguistic Explanation appraisal future. Egypt’s natural gas sector is expanding rapidly with production having increased over 30% between 1999 and 2007. According Economics H Good distribution system for current use levels to the Oil and Gas Journal, Egypt’s estimated proven gas reserves Better as space heating energy source stand at 58.5 Tcf, or roughly 1% of world reserves. With the contin- Expensive for energy generation ued expansion of the Arab Gas pipeline, which increased its exports Availability L Easy to obtain to roughly 68 bcf during fiscal 2006 from 8 bcf in 2003, Egypt is on Very limited availability as shown by its way to becoming a leading supplier of natural gas throughout shortages during winters several years the Mediterranean region (Egypt Energy Data, 2008). ago

Environmental L Emits greenhouse gas and contains 4.3. Linguistic appraisal impact sulphur Major contributor to global warming and Linguistic appraisal for attributes associated with each alterna- acid rain tive is provided and translated as: Proliferation H Not applicable

VH: very high H: high Table 1d M: medium Reasoning for assigned linguistic appraisal (solar energy). L: low Attribute Linguistic Explanation VL: very low appraisal Economics H Sunlight is free when available High means preferable situation, and low means not preferable Current technology requires large situation. amounts of land for small amount of energy

4.4. Reasoning for assigned linguistic appraisal Availability M Limited to southern areas of U.S. and other sunny areas throughout the world (demand can be highest when least The above discussion is summarized in the following available, e.g. winter solar heating) Tables 1a–1d which provide explanations for the linguistic Environmental M Does require special materials for appraisal for the attributes pertained to nuclear, hydroelectric, impact mirrors/panels that can affect gas/oil, and solar energy, respectively (Nuclear Tourist, 2005). environment Proliferation H Not applicable

Table 1a Reasoning for assigned linguistic appraisal (nuclear energy).

Attribute Linguistic Explanation Table 2 gives a summary appraisal for the four alternatives. appraisal

Economics H Fuel is inexpensive Easy to transport as new fuel Requires larger capital cost because of Table 2 emergency, containment, radioactive Summary appraisal for four alternatives. waste, and storage systems Alternatives Attributes Availability H Market of LWR fuel is currently diverse Economics Availability Environmental impact Proliferation Environmental H No greenhouse or acid rain effects Nuclear H H H L impact Hydroelectric M M L H Waste is more compact than any source Gas/oil H L L H Proliferation L Potential nuclear proliferation issue Solar L H M H A.E. Abouelnaga et al. / Nuclear Engineering and Design 239 (2009) 3062–3068 3067

5. Results and conclusions

Following the procedure steps explained in Section 3.3, the following results are obtained.

It is clear that nuclear energy has the highest utility result, then even a single large power station. Mining and multiple handling solar energy, gas/oil, and finally hydroelectric. Thus, nuclear energy of so much material of any kind involves hazards, and these are is considered as an optimum choice of energy sources for Egypt. reflected in the statistics. It is shown also from the previous discussion that Egypt depends It was found that in the years from 1970 to 1992, there were most on natural gas as a major electricity production source. 1200 injuries from workers and public due to natural gas acci- So, in this part we compare natural gas and nuclear energy to dents which accounts for 85 deaths per TWy electricity and there show the optimum source to be used in Egypt. were only 31 injuries from workers due to nuclear accidents which accounts for 8 deaths per TWy electricity (World Nuclear Association, 2008). • The designs for nuclear plants being developed for implementa- • The investment cost of the nuclear power stations could seem tion in coming decades contain numerous safety improvements relatively high, but they can be redeemed in a reasonable time based on operational experience. The main feature they have in due to their variable costs (especially that of fuel) is reduced and common is passive safety systems, requiring no operator inter- is not vulnerable in front of fluctuations of the market. With these vention in the event of a major malfunction. characteristics, the nuclear power stations are good to produce Many occupational accident statistics have been generated load-base electricity, it means, working the 24 h of every day of over the last 40 years of nuclear reactor operations in the US and the year (Palacios et al., 2004). UK. These can be compared with those from coal-fired power gen- • As in Simsa et al. (2003) new nuclear power at generating costs eration. All show that nuclear is a distinctly safer way to produce between 3.9 and 8.0 c/kWh can be competitive with coal and nat- electricity. A major reason for coal’s unfavorable showing is the ural gas where coal has to be transported over long distances or huge amount which must be mined and transported to supply natural gas pipelines and infrastructures are not in place. 3068 A.E. Abouelnaga et al. / Nuclear Engineering and Design 239 (2009) 3062–3068

Nuclear power is more expensive than coal-fired or gas-fired Comparisons of various energy sources, Nuclear tourist, December 2005, www.nucleartourist.com/basics/reasons1.htm. generation but can be a cheaper option than CO2 capture and Palacios, J.C., Alonso, G., Ramírez, R., Gómez, A., Ortiz, J., Longoria, L.C., storage. 2004. Levelized costs for nuclear, gas and coal for electricity, under This technology could reduce emissions by about 80% for addi- the Mexican scenario, DOE Scientific and Technical Information, tional costs around 3 c/kWh for pulverized coal, 2.5 c/kWh for coal http://www.osti.gov/bridge/servlets/purl/840500-YJxBpR/native/840500.pdf. Phillis, Y.A., Andriantiatsaholiniaina, L.A., 2001. Sustainability: an ill-defined concept IGCC and 1.5 c/kWh for gas CCGT (Audus, 2000). and its assessment using fuzzy logic. Ecological Economics 37, 435–456. Simsa, R.E.H., Rognerb, H.-H., Gregoryc, K., 2003. Carbon emission and mitigation Egypt’s need to use nuclear power became inevitable due to the cost comparisons between fossil fuel, nuclear and renewable energy resources diminishing petroleum and gas reserves and the increasing con- for electricity generation. Energy Policy (31), 1315–1326. World Nuclear Association. Safety of Nuclear Power Reactors, June 2008. sumption rates. Zadeh, L.A., 1965. Fuzzy sets. Information and Control 8, 338–353. Using nuclear energy in Egypt may widen the electricity grid Zadeh, L.A., 1973. Outline of a new approach to the analysis of complex systems and connections of Egypt, which includes five countries, to include more decision processes. IEEE Transactions on Systems, Man, and Cybernetics SMC-3, 28–44. countries and thus enhance the economic situation of Egypt. Zimmermann, H.J., 1991. Fuzzy Set Theory and Its Applications. Kluwer, Boston, p. Nuclear power alone won’t get us to where we need to be, but 399. we won’t get there without it. Ayah Elsayed Abouelnaga is a staff member in Nuclear Engineering Department, Faculty of Engineering, Alexandria University in Egypt. She was graduated in 2001 Acknowledgement from the department and received a MSc in thermal performance of nuclear reactors (calculation of critical heat flux, CHF, in subcooled boiling) in 2005. She is a PhD candidate since 2006 and working in enhancing the sustainability of nuclear energy I would like to thank Dr. Naguib H. Aly and Dr. Amr S. Galal in using the advanced reactors. She is an assistant lecturer in the Department of Nuclear Nuclear Engineering Department, Faculty of Engineering, Alexan- Engineering, Faculty of Engineering, Alexandria University in Egypt since 2005. dria University for their great support and encouragement. Abdelmohsen Metwally was a staff member in Nuclear Engineering Department, Faculty of Engineering, Alexandria University in Egypt. He received a MSc in 1976 References from Nuclear Engineering Department, Alexandria University. He received a PhD in nuclear engineering from Iowa State University in USA in 1982. From 1979 to Audus, H., 2000. Leading options for the capture of CO2 at power stations. In: 1981, he worked as a NRC investigator for the nuclear accident at Three Mile Island. Proceedings of the Fifth International Conference on Greenhouse Gas Control From 1991 to 2003, he worked as a safeguards inspector in International Atomic Technologies, Cairns, Australia, August. Energy Agency (IAEA). He was a professor in the Nuclear Engineering Department Boyle, G., Energy for a sustainable future, Faculty of technology, the Open University, in Alexandria since 2003. He was specialized in artificial intelligence systems, auto- www.open.ac.uk, December 2004. matic control, reactor safety, and simulation for NPPs. Egypt Energy Data, Statistics and Analysis—Oil, Gas, Electricity, Coal, August 2008. Egypt State Information Service—Year Book 2006.htm. Mohammad E. Nagy is a staff member in Nuclear Engineering Department, Faculty Egypt Energy Data, Statistics and Analysis—Oil, Gas, Electricity, Coal, Energy of Engineering, Alexandria University in Egypt. He received his MSc in 1967 from the information administration, August 2008. http://www.eia.doe.gov/emeu/cabs/- Nuclear Engineering Department in Alexandria University in radiation shielding. He Egypt/pdf.pdf. received his PhD in 1969 from the Nuclear Engineering Department in Iowa State Feller, G., 2008. Solar thermal electricity in Egypt, http://www.ecoworld.com/fuels/- University in the reactor dynamics. He is a professor in the Nuclear Engineering solar-thermal-electricity-in-egypt.html. The Disadvantages of Oil, 2009. Fossil Fuel, http://fossil-fuel.co.uk/oil/the- Department in Alexandria since 1970. He is specialized in reactor dynamics, reactor 518disadvantages-of-oil. analysis, radiation shielding and protection. Hydroelectrical Energy of the High Dam in Egypt, 2007. University of Portsmouth, Saed Agamy is a staff member in Nuclear Engineering Department, Faculty of Engi- Department of Electronic and Computer Engineering. neering, Alexandria University in Egypt. He received his MSc in 1973 from the McCarthy, J., Solar energy, November 1995, http://www-formal.stanford.edu/jmc/- progress/solar.html. Nuclear Engineering Department in Alexandria. He received his PhD in 1979 from McGrath, C., EGYPT Plenty of Sun, Nobody Catching any Rays, IPS, 16 August 2009, McMaster University, Canada in backscattering of low energy ions for Silicon. He was http://ipsnews.net/news.asp?idnews=47861. a Post-Doctoral researcher from 1977 to 1978 at University of Toronto in Hydrogen Nguyen, H.T., Walker, E.A., 1966. A First Course in Fuzzy Logic. CRC press. Energy. He worked in University of Missouri in Saint Luis from 1978 to 1979 and NTI, Egypt Profile, Nuclear Chronology, June 2009, http://nuclearthreatinitiative. in UC Santa Barbra from 1982 to 1983. He is a professor in the Nuclear Engineering org/e research/profiles/Egypt/Nuclear/chronology 2008.html. Department in Alexandria since 1984. He is specialized in material science. Nuclear Engineering and Design 240 (2010) 1928–1933

Contents lists available at ScienceDirect

Nuclear Engineering and Design

journal homepage: www.elsevier.com/locate/nucengdes

Assessment of nuclear energy sustainability index using fuzzy logic

Ayah E. Abouelnaga ∗, Abdelmohsen Metwally, Naguib Aly, Mohammad Nagy, Saeed Agamy

Nuclear Engineering Department, Faculty of Engineering, Alexandria University, Alexandria 21544, Egypt article info abstract

Article history: Nuclear energy is increasingly perceived as an attractive mature energy generation technology that can Received 27 October 2009 deliver an answer to the worldwide increasing energy demand while respecting environmental concerns Received in revised form 22 February 2010 as well as contributing to a reduced dependence on fossil fuel. Advancing nuclear energy deployment Accepted 1 March 2010 demands an assessment of nuclear energy with respect to all sustainability dimensions. In this paper, the nuclear energy, whose sustainability will be assessed, is governed by the dynamics of three subsystems: environmental, economic, and sociopolitical. The overall sustainability is then a non-linear function of the individual sustainabilities. Each subsystem is evaluated by means of many components (pressure, status, and response). The combination of each group of indicators by means of fuzzy logic provides a measurement of sustainability for each subsystem. © 2010 Elsevier B.V. All rights reserved.

1. Introduction be quantified and yet are very important as, for example, values and opinions. In this area of human thought fuzzy logic performs There are many definitions or, better, descriptions of sustain- successfully (Zadeh, 1979; Zimmerman, 1991). ability according to subject (ESI, 2005; Scott, 2003; Measuring Sustainability, 2003; Sutter, 2003; Afgan et al., 2006; Duffey and 2. Characteristics and usability of fuzzy logic Miller, 2002; ESS, 2005; Ellis et al., 2004; WCDE, 1987). One such description of development sustainability is “development that To assess sustainability in a ‘fuzzy logic’ manner, the following meets the needs of the present without compromising the ability have to be defined: of future generations to meet their own needs” (WCDE, 1987). Sustainability is difficult to define or measure because it is an • Linguistic variables, which best represent the sustainability of the inherently vague and complex concept. Fuzzy logic, due to its capa- whole system. bility to emulate skilled humans and its systematic approach to • Linguistic rule bases and fuzzy logical operators which express handling vague situations where traditional mathematics is inef- qualitatively the knowledge and the key features of the overall fective, seems to be a natural technical tool to assess sustainability. system. Fuzzy logic is a scientific tool that permits simulation of the • A defuzzification method to convert fuzzy statements into a sin- dynamics of a system without a detailed mathematical descrip- gle crisp value of overall sustainability. tion. Knowledge is represented by IF–THEN linguistic rules, which describe the logical evolution of the system according to the linguis- Linguistic variables: tic values of its principal characters that we call linguistic variables. Briefly, a linguistic variable is defined by four items, Real values are transformed into linguistic values by an operation called fuzzification, and then fuzzy reasoning is applied in the form (a) The name of the variable; of IF–THEN rules. A final crisp value is obtained by defuzzification, (b) its linguistic values; which does the opposite of fuzzification. (c) the membership functions of the linguistic values; and Another important aspect of fuzzy logic is that it uses linguistic (d) the physical domain over which the variable takes its quantita- variables, thus performing computation with words. If a traditional tive values (Phillis and Andriantiatsaholiniaina, 2001). mathematical approach towards sustainability assessment were adopted, such as cost-benefit analysis or algebraic formulas, then 3. Method for assessing sustainability certain factors, which are impossible to quantify, would be left out. There are, however, aspects of sustainability, which cannot In Phillis and Andriantiatsaholiniaina (2001), a model called Sus- tainability Assessment by Fuzzy Evaluation (SAFE) was developed, which provides a mechanism for measuring development sustain- ∗ Corresponding author. Tel.: +20 105614004; fax: +20 35925550. ability. In this paper, we follow the same method but to assess the E-mail address: [email protected] (A.E. Abouelnaga). nuclear energy sustainability index.

Fig. 1. Membership function of tertiary linguistic variables.

In Guimaraes and Lapa (2004), the fuzzy inference system (FIS) 3.1. Input and output fuzzy membership functions is used for evaluating and improving nuclear power plant operating performance, using different indicators such as energy availability The fuzzy toolbox of Matlab works as a simulator (Matlab, 2004). factor (EAF) and the planned (PUF) and unplanned unavailability For each linguistic variables are defined the total range and param- factor (UUF). eters of each current membership function. Figs. 1 and 2 present In this paper, it is assumed that the overall system, whose the final configuration of the Mamdani type FIS variable. sustainability will be assessed, consists of three subsystems, envi- The primary linguistic variables of sustainability take the lin- ronmental, economic, and sociopolitical. guistic values very bad (=VB), bad (=B), intermediate (=I), good The overall sustainability is then a function of the individual sub- (=G), and very good (=VG). For the three tertiary linguistic variables, system’s integrity, which will be devised logically via fuzzy logic. we use the linguistic values weak (=W), medium (=M), and strong This function consists of combinations of IF–THEN rules operating (=S). For the overall indicator of sustainability we use the linguistic on rule bases derived from expert knowledge. By their nature, such values bad (=B), acceptable (=A), and excellent (=E). Examples of functions are highly non-linear. The term integrity is defined as membership functions of the linguistic values are shown graphi- the degree to which each sustainability variable fulfils criteria and cally in Figs. 1 and 2. Triangular functions are used for the inputs principles of sustainability. and secondary variables, while trapezoidal functions are chosen Criteria and principles of sustainability are recommended criti- for the tertiary and primary variables. The horizontal axis of each cal or target states that the system should satisfy to be sustainable. membership function expresses the normalized values of each sus- Each subsystem is evaluated by means of three types of indica- tainability variable and ranges over [0,1], whereas the vertical axis tor, pressure; status; and response indicator. Status is the present expresses membership grades ranging again over [0,1]. Triangu- state of a component such as developing of the nuclear energy lar membership functions are selected because they are simple. program. Pressure is a force tending to change status such as the Trapezoidal membership functions are straightforward extensions environmental impact. Finally, response is the reaction taken to of triangular ones. bring pressure to a level that will guarantee a better status as, for example, design modifications towards improving safety. Indica- 3.2. Quantitative values of sustainability indicators tors of sustainability that have been developed in the literature use similar components (Bell and Morse, 1999; Durpel et al., 2005; Since the overall sustainability index (OSUS) will be obtained Andriantiatsaholiniaina et al., 2004; Penfold et al., 1995; Swift through calculating the environmental, economic, and sociopolit- and Woomer, 1993; UN, 1993). The combination of each group ical sustainability indices, so they should be discussed as shown of indicators by means of fuzzy logic provides a measurement of below. sustainability for each subsystem. To assess each secondary variable, we use three tertiary linguistic variables, pressure; status; and 3.2.1. Environmental indicators response. • System dynamics Finally, these tertiary variables are assessed by means of fuzzy In developing integrated nuclear energy systems models, the inference IF–THEN rules applied to indicators of sustainability, following objectives from a user’s perspective should be kept which are the inputs of the system. in mind;

Fig. 2. Membership function of primary linguistic variables. 1930 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 1928–1933

transparency of the model, i.e. the final user, being it a ◦ More evaluation on operation limits and conditions include: researcher or an energy policy advisor, needs to be assured Safety limits, limiting safety system settings, limits and condi- that the underlying systems model behaves as in the real world tions for normal operation, surveillance requirements, and action and communicates the results of systems simulations in a for- statements for deviations from the operational limits and condi- mat according standardized sustainability criteria. Appropriate tions (OLCS) (IAEA, 2008). level of detail to guarantee correct results while this level of detail surely depends on the specific case being analyzed; 3.2.2. Economical indicators scalability, i.e. the architecture of the model should allow small The most important energy costs of a power plant itself, up to as well a very complex, e.g. multi-regional systems, be analyzed the end of its useful lifetime, are: using a standard methodology; interactivity, i.e. part of the systems simulation aims at allowing • Main and auxiliary buildings, major components, piping, and iterative intervention capability to the user so he can inves- instrumentation, tigate on the possible synergies in deploying nuclear energy • mining and refining the uranium in the ores, systems as well as to show the trade-offs to be made in advanc- • enrichment of the uranium and fabrication of the fuel elements, ing nuclear energy systems deployment; and connectivity, i.e. such dynamic nuclear energy system models • operating and maintenance. are complementary to other models addressing, for instance, macro-economic energy market aspects. Connectivity, prefer- If fuel is imported, second and third items are replaced with fuel ably on-line, with these other codes is very important (Durpel price. et al., 2005). The energy debts incurred by a nuclear power plant have to be • Radioactive waste management paid after the plant has reached the end of its useful life. They are: One can predominantly distinguish between two types of nuclear waste: spent fuel (in solid state) and radioactive emis- • Conditioning the extremely radioactive spent fuel elements plus sions (in liquid or gaseous state). Both produced by nuclear power decontamination, plants in normal operation. These two forms of waste are dealt • decontamination of depleted uranium left behind the enrich- with in two opposite manners. The attitude to the former is that ment, and of ‘concentration and protection’: radioactive contamination of • decommissioning (Nuclear power, 2005). the external environment from spent fuel storage is minimized through several layers of physical containment. To the latter The overwhelmingly important element in the total costs of mostly the principle of ‘dilution and exposure’ is applied: the nuclear power is the cost of construction (capital cost), the tem- emissions of the nuclear industry may therefore lead to increases poral component of which is a significant determining factor. This in ambient radiation levels. The emissions into the atmosphere typically accounts for some 60–75% of the generating cost of nuclear or surrounding waters from nuclear power plants are typically power (The role of nuclear power in a low carbon economy). much lower than those of reprocessing plants, and even for the latter, after dilution. The additional radiation doses generated can 3.2.3. Sociopolitical indicators generally be neglected in comparison to natural levels of radioactivity (The contribution of nuclear energy, 2006). • Proliferation risk • Safety concepts No real consensus exists even on the interpretation of the word The main objective of safety concepts and concepts is to prevent “proliferation”. The Non-Proliferation Treaty (NPT) defines pro- or mitigate radioactive releases from the nuclear facility to the liferation as the manufacture or acquisition of nuclear weapons environment. or other nuclear explosive devices by countries which do not now possess them. Conventionally, the actual detonation of a The concepts include: device has determined the transition from non-nuclear weapons to nuclear weapons status. Recently, this approach has been ◦ Multiple barriers to radioactive releases. questioned on the grounds that there are many stages in the The barriers include fuel, clad, coolant, pressure vessel, con- acquisition of a nuclear weapons capability. A nation can make tainment, site selection, and evacuation plan (relocation is all the preparations for the construction of a weapon or the test- considered). ing of a device without actually “proliferating”. If it is possible ◦ Defense in depth. to come within hours of a bomb and still not violate the NPT, According to the IAEA (2008) newest definition, five levels are the traditional definition conceals more than it reveals (Nuclear defined (Safety of Nuclear Power Plants): proliferation safeguards, 1977). Level 1: Preventions of abnormal operation and failure. • Public opinion Level 2: Control of abnormal operation and detection of failures. The peaceful use of nuclear energy serves the purpose of Level 3: Control of accidents within design basis. environmental protection. The radioactive emissions induced Level 4a: Mitigation of severe plant conditions including pre- by nuclear energy are in the range of the natural radioactiv- vention of accident progression. ity and therefore do not present an environmental burden. The Level 4b: Mitigation of the consequences of severe accidents. ecological use in respect of climate policy is out of question Level 5: Mitigation of radiological consequences of significant due to the avoidance of gases noxious to the climate. Presently, releases of radioactive materials. nuclear energy is the only large-scale system without releas- ◦ Reliability oriented design concepts. ing substances known to adversely affect the global climate The concepts include: redundancy, diversity, fail safe, self reg- (http://www.kernenergie.net/kernenergie/en). ulation, and physical separation. ◦ Passive and active designs of engineered safety features (ESF). There is a continuing public concern that the use of nuclear For early triggering of safety actuation system as well as reli- power is inherently associated with a further spread of nuclear able and effective operation of ESF, innovative designs of reactors weapons and a risk of terrorism (Sustainable development and introduce passive instrumentation and systems. nuclear power). The US Atoms for Peace policy announced in A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 1928–1933 1931

Table 1 Tertiary linguistic variables for sustainability indicators.

Environmental Economics Sociopolitical

1- Pressure: M 1- Pressure: S 1- Pressure: S Justification: Justification: Justification: - Radioactive waste management - Availability of funds to cover costs for main - Proliferation risk, (inspection, additional and auxiliary buildings, main components protocol) and instrumentation - System dynamics - Fuel operation and maintenance - Public opinion - Reactor safety and engineered safety features - Fuel licensing - Environmental impact - Fuel waste disposal and decontamination

2- Status: M 2- Status: M 2- Status: S Justification: Justification: Justification: - The status is subjected to move to another level - Financial problems - Better outline to material balance area with optimum application of containment and surveillance measures - Developing on the progress of the nuclear program - Political pressure - Poor safety culture and communication with public - Slow nuclear program progress - Limited nuclear energy projects

3- Response: S 3- Response: M 3- Response: S Justification: Justification: Justification: Programs to improve: - Planning to make new nuclear projects - Plans to reinforce proliferation resistance - Site selection procedure - A start of political support - Plans to improve communication with the public - Radioactive waste management - Providing more money to support for nuclear projects. - Design modification towards improving safety

1953 promoted a policy of international nuclear co-operation This index is an approximate assessment of nuclear energy sus- based on the condition that nuclear technology transfer would tainability. This index can be improved by many ways. be used exclusively for peaceful purposes and bilateral safe- Sensitivity analysis can be conducted to study how the out- guards arrangements were introduced. Several years after its put, OSUS, could be changed if one or more attributes is creation in 1957, the IAEA initiated on-site inspections at nuclear improved/degraded as shown in Table 3. facilities under binding safeguards agreements. The 1970 Treaty By using advanced nuclear power reactors, which use passive on the Non-Proliferation of Nuclear Weapons (NPT) now com- safety system such as AP600 and AP1000, we can improve the mits more than 180 countries to refrain from acquiring nuclear safety and hence the environmental safety index “ENVIOS”. Also weapons and to accept comprehensive IAEA safeguards on all their using nuclear reactors which apply simplicity in design (short con- nuclear activities. The Treaty was extended indeﬁnitely in 1995 struction period) and safety-by-design concept in their design can (http://www.un.org/Depts/dda/WMD/treaty). From the discussion above and the comparison mensioned before in (Abouelnaga et al., 2009), Table 1 can be deduced as shown Table 2 below. Inputs and outputs for the fuzzy inference system.

Inputs Output ECOS, ENVIOS, SOCIOS 3.3. Summary of the fuzzy inference system Pressure Status Response Resulting integrity SSSVH The rules adopted in the Mamdani model to generate the FIS for SMSI the three indicators are presented below. SWSB To determine the overall sustainability, OSUS, the rule base SSMH SMMI 3 needs 5 = 125 rules because we have five linguistic values and SWMB three variables (ECOS, ENVIOS, and SOCIOS). To determine the three SSWI indicators for overall sustainability (secondary variables), we use SMWB 33 = 27 rules since we have three linguistic values (strong, weak, SWWVB MSS VH and medium) and three variables (indicators) (Table 2). MMSI MWSB MSMH 4. Results and conclusions MMMI MWMB The model developed herein serves a dual purpose, it provides a MSWI MMWB flexible framework defining sustainability as a function of a number MWWVB of variables and at the same time it gives the mathematical machin- WSS VH ery to compute numerical values of sustainability. First we obtain WMSI the environmental sustainability index “ENVIOS” to be high, the WWSB WSMH economic sustainability index “ECOS” to be low, and the sociopo- WMMI litical sustainability index “SOCIOS” to be high. The sustainability WWMB index for nuclear energy is then obtained from the second fuzzy WSWI inference system using ECOS, ENVIOS, and SOCIOS as inputs and WMWB then OSUS is found to be 67%. WWWVB 1932 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 1928–1933

Fig. 3. Rules viewer of primary variables.

Table 3 Andriantiatsaholiniaina, L.A., Kouikoglou, V.S., Phillis, Y.A., 2004. Evaluating strate- Sensitivity analysis for the three indicators. gies for sustainable development: fuzzy logic reasoning and sensitivity analysis. Ecological Economics 48, 149–172. Indicators OSUS Indicators OSUS Bell, S., Morse, S., 1999. Sustainability Indicators: Measuring the Immeasurable. (0.4,0.5,0.5) 0.5 (0.2,0.1,0.1) 0.203 Earthscan Publication, London, p. 175. (0.4,0.4,0.5) 0.395 (0.1,0.1,0.1) 0.203 Durpel, L.V., Yacout, A., Wade, D., 2005. Development of integrated system dynamics models for the sustainability assessment of nuclear energy. In: Proceeding of (0.4,0.4,0.4) 0.395 0.1,0.1,0.05) 0.172 GLOBAL. (0.3,0.5,0.5) 0.5 (0.1,0.05,0.05) 0.172 Drs. Romney B. Duffey and Alistair I. Miller, From Option to Solution: The Nuclear (0.3,0.4,0.5) 0.395 (0.05,0.05,0.05) 0.16 Contribution to Global Sustainability Using the Newest Innovations World (0.3,0.4,0.4) 0.395 (0.1,0.5,0.5) 0.355 Nuclear Association Annual Symposium. September 2002–London. (0.3,0.3,0.4) 0.317 (0.1,0.5,0.1) 0.203 Environmental Sustainability Stewardship, 2005. Index Benchmarking National (0.3,0.3,0.3) 0.31 (0.6,0.5,0.5) 0.5 Environmental Stewardship World Economic Forum Geneva, Switzerland Joint (0.2,0.5,0.5) 0.44 (0.65,0.5,0.5) 0.5 Research Centre of the European Commission, Italy. (0.2,0.4,0.5) 0.395 (0.65,0.65,0.5) 0.645 Ellis, C., Baxter, A., Shenoy, A., 2004. Modular Helium reactor fuel cycle concepts and (0.2,0.3,0.5) 0.31 (0.7,0.5,0.5) 0.5 sustainability. In: 2nd International Topical Meeting on HIGH TEMPERATURE (0.2,0.3,0.4) 0.31 (0.7,0.6,0.5) 0.605 REACTOR TECHNOLOGY, Beijing, CHINA, September 22–24. (0.2,0.3,0.3) 0.304 (0.7,0.6,0.6) 0.605 Environmental Sustainability Index (ESI), 2005. Center for International Earth Sci- (0.2,0.2,0.3) 0.244 (0.7,0.7,0.7) 0.69 ence Information Network and the World Economic Forum,Yale University. (0.2,0.2,0.2) 0.244 (0.8,0.7,0.6) 0.683 http://www.yale.edu/esi/. Guimaraes, A., Lapa, C., 2004. Fuzzy inference system for evaluating and improv- (0.2,0.2,0.1) 0.203 (0.8,0.7,0.7) 0.69 ing nuclear power plant operating performance. annals of nuclear energy (31), (0.8,0.8,0.9) 0.757 311–322. IAEA safety standards, No. NS-G-4.4, 2008. http://www.kernenergie.net/kernenergie/en. 2001. improve the economics and safety of the nuclear energy and hence Measuring Sustainability, 2003. A Framework for Private Sector Investments. Inter- national Finance Corporation Pennsylvania. Washington, USA. can improve the overall sustainability index of nuclear energy, such Matlab 7, 2004. Users guide of the fuzzy logic toolbox. as the IRIS reactor, which also extends the refueling cycle which can Nuclear power, 2005. The energy balance, introduction: general principles of sus- make an improvement in the SOCIOS and hence the OSUS (Fig. 3). tainability, summary of the costs of nuclear energy, 4 August. Nuclear proliferation safeguards, June 1977. For the environmental sustainability index, changing the Phillis, Y., Andriantiatsaholiniaina, L., 2001. Sustainability: an ill-deﬁned con- appraisal from intermediate (I) to high (H) yields an increase in cept and its assessment using fuzzy logic. Ecological Economics (37), the overall sustainability index (OSUS) value from 67% to 70%. 435–456. Also, by improving the economics of nuclear energy, changing Penfold, C.M., Miyan, M.S., Reeves, T.G., Grierson, I.T., 1995. Biological farming for sustainable agricultural production. Aust. J. Exp. Agric. 35, 849–856. the appraisal from high (H) to very high (VH) yields an increase in Swift, M.J., Woomer, P., 1993. Organic matter and the sustainability of agricultural the overall sustainability index (OSUS) value from 70% to 73%. system: deﬁnition and measurement. In: Mulongoy, K., Merckx, R. (Eds.), Soil Organic Matter Dynamics and Sustainability of Tropical Agriculture. Wiley, New York, pp. 3–17. References Scott, J., February, 2003. How can farmers and researchers measure sustainability? In: Australian Agronomy Conference. Deakin University. Sutter, C., 2003. Sustainability Check-Up for CDM Projects. How to assess the sustain- Abouelnaga, A.E., Metwally, A., Nagy, M., Agamy, S., December 2009. Optimum selec- ability of international projects under the Kyoto Protocol. Swiss Federal Institute tion of an optimum energy resource using fuzzy logic. Nuclear Engineering and of Technology (ETH) Zurich. Design Vol. 239 (12), 3062–3068. Safety of Nuclear Power Plants: Design–IAEA NS-R-1. Naim H. Afgan Paul Andre Maria G. Carvalho. Sustainability: The Management Sys- Sustainable development and nuclear power. http://www.iaea.org/Publications/ tem Property. Technology Management for the Global Future, 2006. PICMET Booklets/Development/devnine.html. 2006. A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 1928–1933 1933

The contribution of nuclear energy to a sustainable energy system, volume 3 in the 1991 to 2003, he worked as a safeguards inspector in International Atomic Energy CASECADE MINTS project, March 2006. Agency (IAEA). He was a professor in the nuclear engineering department in Alexan- The role of nuclear power in a low carbon economy Paper 4: The economics of dria since 2003. He was specialized in artificial intelligence systems, automatic nuclear power, March 2006. control, reactor safety, and simulation for NPPs. http://www.un.org/Depts/dda/WMD/treaty. UN (United Nations), 1993. The Global Partnership for Environment and Develop- Naguib Aly is a staff member in nuclear engineering department, Faculty of Engi- ment: a Guide to Agenda 21 Post Rio Edition. United Nations, New York, p. neering, Alexandria University in Egypt. He received an Msc in reactor safety (PWR 239. simulation) in 1979. He received a PhD in reactor dynamics (BWR representation WCDE (World Commission on Environment Development), 1987. Our Common for PWR) in 1984 from nuclear engineering department, Faculty of Engineering, Future, 400. Oxford University Press, Oxford, p. 11. Alexandria University and since 1984 has been a lecturer in nuclear engineering Zadeh, L.A., 1979. A theory of approximate reasoning. In: Hayes, J., Michie, D., department, Faculty of Engineering, Alexandria University. He is a professor in the Mikulich, L. (Eds.), Machine Intelligence, 9. Halstead, New York, pp. 149–194. department specialized in artificial intelligence systems, automatic control, reactor Zimmerman, H.J., 1991. Fuzzy Set Theory and its Applications. Kluwer, Boston, MA, safety, and simulation for NPPs. 399 pp. Mohammad Nagy is a staff member in nuclear engineering department, Faculty of Ayah Abouelnaga is a staff member in nuclear engineering department, Faculty of Engineering, Alexandria University in Egypt. He received his MSc in 1967 from the Engineering, Alexandria University in Egypt. She was graduated in 2001 from the nuclear engineering department in Alexandria University in radiation shielding. He department and received an Msc in thermal performance of nuclear reactors (calcu- received his PhD in 1969 from the nuclear engineering department in Iowa state lation of critical heat flux, CHF, in subcooled boiling) in 2005. She is a PhD candidate university in the reactor dynamics. He is a professor in the nuclear engineering since 2006 and working in enhancing the sustainability of nuclear energy using the department in Alexandria since 1970. He is specialized in reactor dynamics, reactor advanced reactors and in assessing the safety performance of advanced reactors analysis, radiation shielding and protection. which use passive safety systems. She is an assistant lecturer in the department of nuclear engineering, Faculty of Engineering, Alexandria University in Egypt since Saeed Agamy is a staff member in nuclear engineering department, Faculty of Engi- 2005. neering, Alexandria University in Egypt. He received his MSc in 1973 from the nuclear engineering department in Alexandria. He received his PhD in 1979 from Abdelmohsen Metwally was a staff member in nuclear engineering department, McMaster University, Canada in backscattering of low energy ions for Silicon. He was Faculty of Engineering, Alexandria University in Egypt. He received an Msc in 1976 a Post-Doctoral researcher from 1977 to 1978 at University of Toronto in Hydrogen from nuclear engineering department, Alexandria University. He received a PhD in energy. He worked in University of Missouri in Saint Luis from 1978 to 1979 and nuclear engineering from Iowa State University in USA in 1982. From 1979 to 1981, in UC Santa Barbra from 1982 to 1983. He is a professor in the nuclear engineering he worked as a NRC investigator for the nuclear accident at Three Mile Island. From department in Alexandria since 1984. He is specialized in material science. Nuclear Engineering and Design 240 (2010) 2820–2830

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Nuclear Engineering and Design

journal homepage: www.elsevier.com/locate/nucengdes

Assessment the safety performance of nuclear power plants using Global Safety Index (GSI)

Ayah E. Abouelnaga ∗, Abdelmohsen Metwally, Naguib Aly, Mohammad Nagy, Saeed Agamy

Alexandria University, Faculty of Engineering, Department of Nuclear and Radiological Engineering, Alexandria, 21544, Egypt article info abstract

Article history: The safety performance of the nuclear power plant is a very important factor enhancing the nuclear Received 28 September 2009 energy option. It is vague to evaluate the nuclear power plant performance but it can be measured Received in revised form 29 April 2010 through measuring the safety performance of the plant. Accepted 2 July 2010 In this work, the safety of nuclear power plants is assessed by developing a “Global Safety Index” (GSI). The GSI is developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence (during an accident), and the consequences of the accident. The GSI is developed by tracking the performance of the safety system during a design basis accident such as loss of coolant accident (LOCA). This is done by using the PCTran simulation code in simulation a PWR LOCA and introducing four indicators: the sensation time, the response time, and the recovery time together with Core Damage Frequency (CDF). Then Fuzzy Inference System is used for obtaining the GSI. The GSI is also evaluated for the advanced types for nuclear power plants, such as AP1000, and a comparison is made between the GSI evaluated for both conventional and advanced types. © 2010 Elsevier B.V. All rights reserved.

1. Introduction tain potential concerning termination of events or accidents that are effectively coped with by a protective system limited by the The world demand for energy is growing rapidly, particularly reliability of the active safety systems or prompt operator actions in developing countries that are trying to raise the standard of to prevent significant fuel failure and fission product release. Since living for billions of people, many of whom do not have access the reliability of active systems cannot be reduced below a thresh- to electricity or clean water. Climate change and the concern for old and that of the operator’s action is debatable, there is growing increased emissions of green house gases have brought into ques- concern about the safety of such plants due to the large uncertainty tion the future primary reliance of fossil fuels. With the projected involved in Probabilistic Safety Analysis (PSA) particularly in ana- worldwide increase in energy demand, concern for the environ- lyzing human faults. In view of this, a desirable goal for the safety mental impact of carbon emissions, and the recent price volatility characteristics of an innovative reactor is that its primary defense of fossil fuels, nuclear energy is undergoing a rapid resurgence. against any serious accidents is achieved through its design features This “nuclear renaissance” is broad based, reaching across Asia, preventing the occurrence of such accidents without depending North America, Europe, as well as selected countries in Africa either on the operator’s action or the active systems (Spiewak and and South America. Many countries have publicly expressed their Weinberg, 1985). That means, the plant can be designed with ade- intentions to pursue the construction of new nuclear energy plants. quate passive and inherent safety features to provide protection Some countries that have previously turned away from commercial for any event that may lead to a serious accident. Such robustness nuclear energy are reconsidering the advisability of this decision. in design contributes to a significant reduction in the conditional This renaissance is facilitated by the availability of more advanced probability of severe accident scenarios arising out of initiating reactor designs than are operating today, with improved safety, events of internal and external origin. The function of confinement economy, and operations. of any radioactivity released in the containment is also made more The conventional reactors or so-called “Traditional ones” have reliable by adopting robust, redundant, and passive design features. seen an extensive use of “active” engineering safety systems for That means such reactors are different from traditional ones, reactor control and protection in the past. These systems have cer- i.e. they are designed on the philosophy of “safety by design”. Such reactors have the potential to restore the reactor to a stable state in any postulated accident condition and the risk to the public ∗ Tel.: +447401403429. must be at least in the same level or even lower than the other E-mail address: [email protected] (A.E. Abouelnaga). industrial plants. The most important safety tasks of the future

0029-5493/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2010.07.004 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830 2821 reactors are not only to prevent excessive radioactive release to All the above mentioned trials consider the safety of nuclear the environment but also to avoid necessity of evacuation of the power plants from a specific point of view. population. Minimum frequency of such events should not exceed The safety performance of the reactor will be measured by the an acceptable level, which is much lower than that of current assessment of a Global Safety Index (GSI). reactors. Generally, this is very ambitious and, often, economi- The GSI is developed by introducing three indicators: probability cally expensive task if the future reactors are built with active of accident occurrence, performance of safety system in case of an engineering safety features. On the other hand, passive safety sys- accident occurrence (during an accident), and the consequences of tems have higher reliability compared with active safety systems the accident. and could help in meeting the above criteria without much eco- Only the second indicator is considered in this work. nomic penalty. Moreover, advanced reactor safety systems can be The GSI is developed by tracking the performance of the safety designed and built with more inherent and passive systems with system during a design basis accident, such as LOCA, and then intro- multiple lines of defense-in-depth which would provide adequate duce the accident in a time event scale. After that we define some protection against any release of radioactivity outside the plant events. containment (Nayak and Sinha, 2007). The safety of the nuclear power plant affects much the envi- 1. Sense time: time during which the safety signal (signal of acci- ronmental sustainability indicator and hence the nuclear energy dent occurrence) sent to the safety system, or time during which sustainability index measured in (Abouelnaga et al., 2010). the safety system sensed with the accident. In the field of nuclear safety, there are some trials to introduce 2. Response time: time during which the safety system responds a safety measurement tool to help assessing the safety in deferent to the accident and starts to mitigate its consequences. areas in nuclear applications. 3. Recovery time: time during which the accident seems to be The criticality safety index (CSI) was developed by NRC as a num- recovered and the reactor reaches the safe conditions. ber assigned to a package, overpack or freight container containing 4. Core Damage Frequency (CDF): which is the probability of core fissile material which is used to provide control over the accumula- damage. tion of packages, overpacks or freight containers containing fissile material (NRC, 1997). The occupational safety index (OSI) was developed as a tool for 2. Model for GSI estimation safety promotion in process industry. By means of the OSI, the managers and workers can monitor their safety performance. The OSI 2.1. General form gives a warning signal when the probability of errors or deviations at the workspace starts to increase (OSI, 2004). As mentioned before, the GSI is developed by tracking the per- A safety index was developed by KAERI to quantify the plant sta- formance of the safety system during a design basis accident, such tus from the perspective of safety. The information loss resulting as LOCA, and then introduce the accident in a time event scale. when simplifying the safety state of the plant into a binary state is After that we define some events: sense time St, response time Rt, reduced through the combination of the probability density func- recovery time Rct, and Core Damage Frequency (CDF). tion arising from the sensor measurement and the membership Sense time is the main characteristic parameter that determines function representing the expectation of the state of the system the reliability of instrumentation system. Nuclear instrumentation (Cho et al., 1996). is integral to the functioning of the nuclear power plant. With- out monitoring equipment, operators would not have a feedback

Fig. 1. Layout of Westinghouse PWR components on PCTran. 2822 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830 loop to ensure that the reactor is operating safely and within specification limits. Nuclear instrumentation enables control room technicians to identify the status of the nuclear reactor power output, as well as alerting the control room immediately if the reactor operates outside of normal conditions (RINEP, 2006). To cope with the accident safely and quickly, the capability of the safety system to respond in a very short time is important. Also the recovery time of the accident, which leads to accident mitigation and help the reactor to reach the safe conditions, is very important parameter required to be short. This is done by increasing passive systems’ share. Passive systems do more than increasing safety, enhancing public acceptance of nuclear power, and easing licensing, they also simplify overall plant systems, equipment, and operation and maintenance. The simplification of plant systems, combined with large plant operating margins, greatly reduces the actions required by the operator in the unlikely event of an accident (Favennec and Bremnes, 2007). Passive systems use only natural forces, such as gravity, natural circulation, and compressed gas-simple physical principles we rely Fig. 2. Break flow. on every day. There are no pumps, fans, diesels, chillers, or other rotating machinery required for the safety systems. This eliminates the need for safety-related AC power sources. A few simple valves align the passive safety systems when they are automatically actu- ated. In most cases, these valves are “fail safe.” They require power to stay in their normal, closed position. Loss of power causes them to open into their safety alignment. In all cases, their movement is made using stored energy from springs, compressed gas or batteries (Favennec and Bremnes, 2007). Core Damage Frequency which is the likelihood of an accident that would cause damage to a nuclear reactor core which can lead to a nuclear meltdown is an important parameter in assessment the safety performance of the nuclear reactor. The GSI is obtained using the Fuzzy Inference System (FIS), by creating a membership function of the inputs; St, Rt, Rct, and CDF, and the output of the FIS will be the GSI.

GSI = f (St, Rt, Rct, CDF)

2.2. Modeling LOCA

2.2.1. LOCA scenario The loss of coolant accident can be classiﬁed according to the Fig. 3. Pressurizer level. break size to: Small LOCA: <2 in break in diameter, <20% cm2 break size. Medium LOCA: 2–6 in break diameter, till 180% cm2 break size. Large LOCA: >6 in break diameter, till 2800 cm2 break size (MST, 2006).

2.2.2. Results (small break LOCA) 2.2.2.1. PWR (1550 MW MAAP3). The simulation system was made using PCTran (MST, 2006) to run at initial condition number (1). Various parameters in initial condition (1) are mentioned in Table 1. At t = 0 s, a break was initiated in the cold leg (15% break) by opening the LOCA valve. Responses for the various

Table 1 Initial conditions of conventional PWR.

Parameters Initial value

Reactor power 100% Pressurizer level 60% Pressurizer pressure 2300 Psia Accumulator pressure 700 Psia Fig. 4. High pressure injection system response. Refueling water storage tank volume 45,000 ft3 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830 2823

Table 2 Initial conditions of AP1000.

Parameters Initial value

Reactor power 100% Pressurizer level 53.5% Pressurizer pressure 2300 Psia Accumulator pressure 700 Psia Refueling water storage tank volume 78,900 ft3

The pressurizer did not start to reﬁll and hence the accident was not recovered till the 600 s simulation time as shown in Fig. 5.

2.2.2.2. AP1000. The simulation system was made to run at initial condition number (1). Various parameters in initial condition (1) are mentioned in Table 2. At t = 0 s, a break was initiated in the cold leg (15% break) by

Fig. 5. Reactor coolant system volume. opening the LOCA valve. Responses for the various parameters are discussed below and graphically represented (Fig. 6). Break flow: Flow through the break reaches a maximum value parameters are discussed below and graphically represented of 340 kg/s, sharply at 10 s. Later on, it has a decreasing trend as the (Fig. 1). steam depressurizes. As shown in Fig. 7. Break flow: Flow through the break reaches a maximum value Pressurizer level: Pressurizer level is decreased sharply from its (340 kg/s), sharply at 10 s. Later on, it has a decreasing trend as the steady state value at 54% to zero at 82 s and at pressure of 1752 Psia steam depressurizes. As shown in Fig. 2. as the volume inventory of the system decreases. Then it begins to Pressurizer level: Pressurizer level is decreased sharply from its refill at 160 s till reaches a level of 3% at 280 s. This is illustrated in steady state value at 60% to zero at 65 s and at pressure of 1600 Psia Fig. 8. as the volume inventory of the system decreases. This is illustrated SIS response: The scram occurred at 54 s, then the signal reaches in Fig. 3. the SIS, hence it is initiated at 70 s as the core make up tank SIS response: The scram occurred at 38 s, then the signal reaches starts before the pressurizer empties at the pressurizer pressure the SIS, hence it is initiated at 113 s as high pressure safety injection of 1752 Psia with an increasing mass flow, reaching a maximum starts at the pressurizer pressure of 11 MPa with an increasing mass value of 620 kg/s at 80 s as shown in Fig. 9. flow, reaching a maximum value at 600 s as shown in Fig. 4. The accident starts to be recovered as the pressurizer starts The accumulator did not start till 600 s. The low pressure pump to refill at 85 s and the reactor coolant system volume starts to does not start till the 600 s simulation time. increase at 100 s as shown in Fig. 10.

Fig. 6. Layout of Westinghouse AP1000 components on PCTran. 2824 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830

Fig. 7. Break ﬂow. Fig. 10. Reactor coolant system volume.

2.2.3. Results (medium break LOCA) 2.2.3.1. PWR. The simulation system was made using PCTran (MST, 2006) to run at initial condition number (1). Various parameters in initial condition (1) were mentioned previously in Table 1. At t = 0 s, a break was initiated in the cold leg (180% break) by opening the LOCA valve. Responses for the various parameters are discussed below and graphically represented. Break ﬂow: Flow through the break reaches a maximum value (3600 kg/s), sharply at 5 s. Later on, it has a decreasing trend as the steam depressurizes. As shown in Fig. 11. Pressurizer level: Pressurizer level is decreased sharply from its steady state value at 60% to zero at 10 s and at pressure of 1500 Psia as the volume inventory of the system decreases. This is illustrated in Fig. 12. SIS response: The scram occurred at 6.5 s, then the signal reaches the SIS, hence it is initiated at 12.5 s as high pressure safety injection starts at the pressurizer pressure of 11 MPa with an increasing mass ﬂow, reaching a maximum value at 600 s as shown in Fig. 13. The accumulator started at 115 s. The low pressure pump started Fig. 8. Pressurizer level. at 530 s simulation time, as shown in Figs. 14 and 15, respectively.

Fig. 9. Core makeup tank response. Fig. 11. Break ﬂow. A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830 2825

Fig. 12. Pressurizer level.

Fig. 15. Low pressure pump response.

The pressurizer started to reﬁll at 150 s simulation time and the accident started to be recovered as shown in Fig. 16.

2.2.3.2. AP1000. The simulation system was made to run at initial condition number (1). Various parameters in initial condition (1) were mentioned previously in Table 2. At t = 0 s, a break was initiated in the cold leg (180% break) by opening the LOCA valve. Responses for the various parameters are discussed below and graphically represented. Break ﬂow: Flow through the break reaches a maximum value of 3800 kg/s, sharply at 10 s. Later on, it has a decreasing trend as the steam depressurizes. As shown in Fig. 17. Pressurizer level: Pressurizer level is decreased sharply from its Fig. 13. High pressure injection system response. steady state value at 54% to zero at 20 s and at pressure of 1600 Psia as the volume inventory of the system decreases. It did not reﬁll till 300 s. This is illustrated in Fig. 18.

Fig. 14. Accumulator response. Fig. 16. Reactor coolant system volume. 2826 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830

Fig. 19. Core makeup tank response. Fig. 17. Break ﬂow.

SIS response: The scram occurred at 7.5 s, then the signal reaches the SIS, hence it is initiated at 11.5 s as the core make up tank starts before the pressurizer empties at the pressurizer pressure of 1778 Psia with an increasing mass ﬂow, reaching a maximum value of 1250 kg/s at 180 s as shown in Fig. 19. The accident starts to be recovered as the reactor coolant system volume starts to increase at 18 s as shown in Fig. 20.

2.2.4. Results (large break LOCA) 2.2.4.1. PWR. The simulation system was made using PCTran (RINEP, 2006) to run at initial condition number (1). Various parameters in initial condition (1) were mentioned previously in Table 1. At t = 0 s, a break was initiated in the cold leg (2000% break) by opening the LOCA valve. Responses for the various parameters are discussed below and graphically represented. Break ﬂow: Flow through the break reaches a maximum value (1680 kg/s), sharply at 7 s. Later on, it has a decreasing trend as the steam depressurizes. As shown in Fig. 21. Pressurizer level: Pressurizer level is decreased sharply from its Fig. 20. Reactor coolant system volume. steady state value at 60% to zero at 5 s and at pressure of 1500 Psia

Fig. 21. Break ﬂow. Fig. 18. Pressurizer level. A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830 2827

Fig. 22. Pressurizer level. Fig. 24. Accumulator response. as the volume inventory of the system decreases. This is illustrated in Fig. 22. SIS response: The scram occurred at 3 s, then the signal reaches the SIS, hence it is initiated at 3.5 s as high pressure safety injection starts at the pressurizer pressure of 11 MPa with an increasing mass ﬂow, reaching a maximum value at 110 s as shown in Fig. 23. The accumulator started at 15 s and the low pressure pump started at 50 s simulation time, as shown in Figs. 24 and 25, respectively. The pressurizer started to reﬁll at 40 s simulation time and the accident started to be recovered as shown in Fig. 26.

2.2.4.2. AP1000. The simulation system was made to run at initial condition number (1). Various parameters in initial condition (1) were mentioned previously in Table 2. At t = 0 s, a break was initiated in the cold leg (2000% break) by opening the LOCA valve. Responses for the various parameters are discussed below and graphically represented. Break ﬂow: Flow through the break reaches a maximum value of 1880 kg/s, sharply at 10 s. Later on, it has a decreasing trend as the steam depressurizes. As shown in Fig. 27.

Fig. 25. Low pressure pump response.

Pressurizer level: Pressurizer level is decreased sharply from its steady state value at 54% to zero at 10 s and at pressure of 1220 Psia as the volume inventory of the system decreases. It did not reﬁll till 300 s. This is illustrated in Fig. 28. SIS response: The scram occurred at 2.5 s, then the signal reaches the SIS, hence it is initiated at 3 s as the core make up tank starts before the pressurizer empties at the pressurizer pressure of 2000 Psia with an increasing mass ﬂow, reaching a maximum value of 1470 kg/s at 35 s as shown in Fig. 29. The accident starts to be recovered as the reactor coolant system volume starts to increase at 36 s as shown in Fig. 30.

2.3. Estimation of GSI

For obtaining the GSI, the fuzzy toolbox of Matlab works as a Fig. 23. High pressure injection system response. simulator (MFL, 2006). For each linguistic variables are deﬁned the 2828 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830

Fig. 29. Core makeup tank response.

Fig. 26. Reactor coolant system volume.

Fig. 30. Reactor coolant system volume.

Fig. 27. Break flow. total range and parameters of each current membership function. In Figs. 31 and 32 are presented the final configuration of the Mamdani type FIS variable. Each indicator will be expressed as high (H) (means fast for time and good for CDF), medium (M), and low (L) (means slow or bad for CDF). There will be 34 rules, 81 rules to determine whether the GSI is good (G), satisfactory (S), or bad (B). Triangular functions are used for the inputs and the output. The horizontal axis of each membership function expresses the normalized values of each GSI variable and ranges over [0, 1], whereas the vertical axis expresses membership grades ranging again over [0, 1]. The simulation results, sense time, response time, and recovery time can be summarized in Table 3. The CDF is determined for both the conventional PWR and AP1000, from the design manuals, to be 5 × 10−5 and 2.4 × 10−7, respectively (AP1000, 2007). After the simulation of several small, medium, and large break LOCAs’, the results are normalized over the whole range and then converted into linguistic appraisal and summarized in Table 4. The GSI is obtained for both the conventional and the advanced reactors, and it is found to be better for the advanced reactor than Fig. 28. Pressurizer level. that for the conventional one. A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830 2829

Fig. 31. Membership function of the primary variables.

Fig. 32. Fuzzy Inference System for GSI.

It was found to be good (0.75) for AP1000, and it was index and hence the Overall Sustainability Index of nuclear found to be satisfactory (0.5) for conventional PWR. This, energy. of course, will improve the environmental sustainability

Table 3 3. Conclusions and future work Summary of LOCAs’ results. • The safety performance of AP1000 is better than that of current Sense time Response time Recovery time PWR in both the response time and the recovery time of accident PWR AP1000 PWR AP1000 PWR AP1000 and also in CDF, but the conventional type PWR is better only in 15% 38 54 113 70 Did not 100 sense time than the AP1000. It takes longer time for AP1000 to 180% 6.5 7.5 12.5 11 150 18 respond to accident than conventional PWR. 2000% 3 2.5 3.5 3 40 36 • Using a suitable simulation code and tracking the performance of the safety systems during an accident, determining the four Table 4 Linguistic appraisal for the results. parameters (sense time, response time, recovery time, and CDF), and use the fuzzy logic tool box to evaluate the GSI, we can assess Sense time Response time Recovery time CDF the safety performance of any type of nuclear power plants. PWR M M L M • Advantages of the proposed GSI: AP1000 L H H H o Can be easily used for different types of nuclear power reactors. 2830 A.E. Abouelnaga et al. / Nuclear Engineering and Design 240 (2010) 2820–2830

o Can compare the performance of passive and active safety sys- of Nuclear Power Plants (NPPs) by Improved Instrumentation and Control (I&C) tems. Systems” – 621-I2-TM-2928028-30 May 2007, Prague, Czech Republic. Matlab Fuzzy Logic Toolbox, 2006. The Mathworks Inc. o Evaluate the innovative and advanced designs of nuclear power Micro Simulation Technology, 2006. PCTran for PWR and AP1000, Ver. 5.0.4, USA, reactors from the safety performance point of view. 2000, and User Manual 2006. o Enhance the design of future safety systems. Nayak, A.K., Sinha, R.K., 2007. Role of passive systems in advanced reactors. Progress • in Nuclear Energy 49, 486–498. As a future work, the Global Safety Index can be generalized by Nuclear Regulatory Commission (NRC), 1997. 10 CFR 71.59. taking into consideration the probability of accidents’ occurrence Occupational Safety Index, 2004. Janne sinisammal. http://tuta.oulu.fi/tyotiede/ in addition to the response of the reactor safety systems in case of tti230904.pdf. accident and also the consequences of the accident, i.e. amounts Reactor Instrumentation for the New Era of Nuclear Power, General Electric Com- pany, 2006. GEA-14752 (10/06). of releases of radioactive materials. Spiewak, I., Weinberg, A.M., 1985. Inherently safe reactors. Annual Review of Energy, • Also, the sustainability index of nuclear energy can be obtained 431–462. for advanced reactors, by changing both the economic sustain- Ayah E. Abouelnaga is a staff member in Nuclear Engineering Department, Faculty ability index and sociopolitical sustainability index, and then of Engineering, Alexandria University in Egypt. She was graduated in 2001 from investigate if it will make any improvement to the overall sus- the department and received an Msc in thermal performance of nuclear reactors (calculation of critical heat flux, CHF, in subcooled boiling) in 2005. She is a PhD tainability index or not. candidate since 2006 and working in enhancing the sustainability of nuclear energy • Assessment of the accident probability can be done by studying using the advanced reactors. She is an assistant lecturer in the Department of Nuclear safety by design of each reactor type. Engineering, Faculty of Engineering, Alexandria University in Egypt since 2005. • A model of assessing the consequences of the accident releases Abdelmohsen Metwally was a staff member in Nuclear Engineering Department, can be developed. Faculty of Engineering, Alexandria University in Egypt. He received an Msc in 1976 • A study of the economics of advanced reactors can be made to from Nuclear Engineering Department, Alexandria University. He received a PhD in Nuclear Engineering from Iowa State University in USA in 1982. From 1979 to check if the economic sustainability index will be improved by 1981, he worked as a NRC investigator for the nuclear accident at Three Mile Island. using advances reactors. From 1991 to 2003, he worked as a safeguards inspector in International Atomic • The overall sustainability index can then be re-assessed for Energy Agency (IAEA). He was a professor in the Nuclear Engineering Department in Alexandria since 2003. He was specialized in artificial intelligence systems, auto- advanced reactors to see how using the advanced reactors will matic control, reactor safety, and simulation for NPPs. improve the sustainability index of nuclear energy. Naguib Aly is a staff member in Nuclear Engineering Department, Faculty of Engi- neering, Alexandria University in Egypt. He received an Msc in reactor safety (PWR Acknowledgements simulation) in 1979. He received a PhD in reactor dynamics (BWR representation for PWR) in 1984 from nuclear Engineering Department, Faculty of Engineering, Alexandria University and since 1984 has been a lecturer in Nuclear Engineering We would like to thank Dr. Amr S.G. Mohamed in Argonne Department, Faculty of Engineering, Alexandria University. He is a professor in the National Laboratory in USA for his great support and encourage- department specialized in artificial intelligence systems, automatic control, reactor ment. Also, we would like to thank Dr. Hisham Hegazy in Nuclear safety, and simulation for NPPs. Power Plants Authority in Egypt for his fruitful help and great sup- Mohammad Nagy is a staff member in Nuclear Engineering Department, Faculty of port to make this work end. Engineering, Alexandria University in Egypt. He received his MSc in 1967 from the Nuclear Engineering Department in Alexandria University in radiation shielding. He received his PhD in 1969 from the Nuclear Engineering Department in Iowa State References University in the reactor dynamics. He is a professor in the Nuclear Engineering Department in Alexandria since 1970. He is specialized in reactor dynamics, reactor Abouelnaga, A.E., Metwally, A., Aly, N., Nagy, M., Agamy, S., 2010. Assessment of analysis, radiation shielding and protection. nuclear energy sustainability index using fuzzy logic. Nuclear Engineering and Design 240, 1928–1933. Saeed Agamy is a staff member in Nuclear Engineering Department, Faculty of AP1000 Nuclear Power Plant Overview, 2007. Gianfranco Saiu, Monica Linda Engineering, Alexandria University in Egypt. He received his MSc in 1973 from Frogheri, Ansaldo Energi, Italy. www.ansaldonucleare.it/TPap0305/NNPP/ the Nuclear Engineering Department in Alexandria. He received his PhD in 1979 NPP 37.pdf. from McMaster University, Canada in backscattering of low energy ions for Silicon. Cho, J.H., Lee, G.W., Kwon, J.S., Park, S.H., Na, Y.W., 1996. Quantification of plant He was a Post-Doctoral researcher from 1977 to 1978 at University of Toronto in safety status. Journal of the Korean Nuclear Society 28 (October (5)), 431–439. Hydrogen energy. He worked in University of Missouri in Saint Luis from 1978 to Favennec, J.M., Bremnes, O., 2007. Innovative instrumentation for safety and per- 1979 and in UC Santa Barbra from 1982 to 1983. He is a professor in the Nuclear formance of PWRs: primary coolant flow, temperature and level measurement Engineering Department in Alexandria since 1984. He is specialized in material systems. IAEA Technical Meeting: “Increasing Power Output and Performance science. Hindawi Publishing Corporation International Journal of Nuclear Energy Volume 2014, Article ID 410715, 9 pages http://dx.doi.org/10.1155/2014/410715

Research Article Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

Ayah Elshahat,1 Timothy Abram,2 Judith Hohorst,3 and Chris Allison3

1 Nuclear & Radiation Engineering Department, Faculty of Engineering, Alexandria University, Alexandria 21544, Egypt 2Centre for Nuclear Energy Technology, University of Manchester, Manchester, M13 9PL, UK 3Innovative Systems Software (ISS), Ammon, ID 83406, USA

Correspondence should be addressed to Ayah Elshahat; [email protected]

Received 20 August 2014; Revised 23 November 2014; Accepted 23 November 2014; Published 16 December 2014

Academic Editor: Arkady Serikov

Copyright © 2014 Ayah Elshahat et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow,and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

1. Introduction and maintenance of the plant. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 Nuclear energy is increasingly considered as an attractive MWe2-LOOPPWRbasedcloselyontheAP600design.The energy source that can deliver an answer to increasing AP1000 maintains the AP600 design configuration, the use worldwide energy demands. One of the most important of proven components, and the licensing basis by limiting public concerns, when dealing with nuclear energy, is the the changes to the AP600 design to as few as possible [1]. safety of nuclear power reactors. The main safety concern is The AP1000 safety systems use natural driving forces, such the emission of uncontrolled radiation into the environment as pressurized gas, gravity flow, natural circulation flow, and which could cause harm to humans at both the reactor site convection [2]. and off-site. Therefore, it is important to evaluate the safety In a previous work [3], a model for assessing the safety performance of nuclear reactors. performance of nuclear reactors was developed using a Great interest is now given to advanced nuclear reactors Global Safety Index (GSI) model. The model was devel- especially those using passive safety components. Passive oped by introducing three indicators: probability of accident safety systems are used to provide significant improvements occurrence, performance of safety system in case of an to plant simplification, safety, reliability, investment protec- accident occurrence, and the consequences of the accident. tion, and capital costs. Investigation of the improvement in nuclear safety using The Westinghouse Advanced PWR, AP1000, design advanced reactors was done by comparing the safety per- includes advanced passive safety features and extensive plant formance of a conventional reactor which uses active safety simplification to enhance the safety, construction, operation, systems, such as PWR, with an advanced reactor which uses 2 International Journal of Nuclear Energy passive safety systems, such as the AP1000, during a design ETER-experiment-driven SCDAP models). The new SCDAP basis accident, such as loss of coolant accident (LOCA), using modelling options include (a) an improved fuel rod gap PCTran [4] as the simulation code. The safety performance of conductance model, (b) improvements in the electrically AP1000isfoundtobebetterthanthatofaconventionalPWR. heated fuel rod simulator model, (c) improvements in the It was needed to study the AP1000 safety performance in a shroud model, and (d) models to treat the influence of air more detailed model to investigate the passive safety systems ingression. employed. RELAP/SCDAPSIM MOD 3.5 runs a much wider variety In this work, a model of the AP1000 was developed using of transients faster and more reliably than RELAP5/MOD 3.3 RELAP/SCDAPSIM [5]tosimulatethepassivecorecooling or RELAP5-3D. Also it is more accurate for many transient system components and to investigate the performance of the conditions particularly for those involving core temperatures different passive safety systems during a small break loss of in excess of 1000 K [6]. coolant accident (SBLOCA). For the reasons mentioned above, it was decided to use MOD 3.5 to model the AP1000 as it is capable of modelling 2. RELAP/SCDAPSIM Code severe accidents, such as a LOCA, and it is able to model the passivesafetysystems’componentswithhighaccuracy. 2.1. Difference between RELAP/SCDAPSIM and PCTran. RELAP/SCDAPSIM is an advanced best-estimate computer 3. AP1000 RELAP5/SCDAPSIM Model code designed for simulating severe reactor accidents. The code predicts the thermal hydraulic response of the reactor 3.1. AP1000 Plant Description. The reactor vessel is cylin- coolant system (RCS) under normal operating conditions or drical, with a hemispherical bottom head and a removable under design basis or severe accident conditions [6]. hemispherical upper head. The vessel contains the core, The RELAP5 models calculate the overall RCS thermal core support structures, control rods, and other components hydraulic response, control system behaviour, reactor kinet- directly associated with the core. The coolant and moderator ics, and the behaviour of special reactor system components are light water at a normal operating pressure of 15.5 MPa. such as valves and pumps. The SCDAP models calculate the The fuel, internals, and coolant are contained within a heavy behaviour of the core and vessel structures under normal and walled reactor pressure vessel. An AP1000 fuel assembly accident conditions. consists of 264 fuel rods in a 17 × 17 square array [8]. RELAP/SCDAPSIM has complete flexibility that allows The AP1000 steam generator (SG) is a vertical shell and U- modelling individual components, separate subsystems, or tube evaporator with integral moisture separating equipment. entire reactor complex with fully integrated control system Design enhancements include nickel-chromium-iron alloy logic [7]. 690 thermally treated tubes a in triangular pitch, improved On the other hand, PCTran is a simple reactor transient antivibration bars, single-tier separators, enhanced mainte- and accident simulation software program that operates on nancefeatures,andaprimary-sidechannelheaddesignfor a personal computer. It is used mainly for educational and easy access and maintenance by robotic tooling. training purposes; hence, it is not recommended to be used The AP1000 reactor coolant pumps are high-inertia, high- for severe accident analysis. PCTran does not have the capa- reliability, low-maintenance, sealless pumps of either canned bility to investigate the effects of changing some parameters motor or wet winding motor design that circulate the reactor related to the accident severity and mitigation. It does not coolant through the reactor vessel, loop piping, and steam have the ability to remove or add different components. It also generators. does not allow the study and investigation of different passive Reactor coolant system piping is configured with two safety systems individually. identical main coolant loops, each of which employs a single For the reasons mentioned above, RELAP/SCDAPSIM 31-inch inside diameter hot leg pipe to transport reactor is used in this paper to develop a model of the AP1000 coolanttoasteamgenerator.Two22-inchinsidediameter to investigate the effect of using passive safety systems in cold leg pipes in each loop (one per pump) transport reactor accident mitigation. coolant back to the reactor vessel to complete the circuit [9]. The AP1000 pressurizer is a vertical, cylindrical vessel 2.2. Different Versions of RELAP5/SCDAPSIM. There are with hemispherical top and bottom heads, where liquid and many different versions of RELAP/SCDAPSIM; MOD3.2 is vapour are maintained in equilibrium saturated conditions. the oldest version and includes the publicly available RELAP One spray nozzle and two nozzles for connecting the safety and SCDAP models released by the US Nuclear Regulatory and depressurization valve inlet headers are located in the Commission (USNRC). MOD3.4 is the current production top head. Electrical heaters are installed through the bottom version. MOD4.0 is the latest experimental version and head. includes expanded models and user options. It has the most advanced code architecture, coding, numerics, and RELAP5 3.2. AP1000 Safety Systems Description. The passive core modelling options such as advanced water properties, alter- cooling system protects the plant against reactor coolant native fluids, and an integrated uncertainty analysis package system leaks and ruptures of various sizes and locations. The [7]. passive core cooling system provides the safety functions of RELAP/SCDAPSIM MOD 3.5 has the most advanced core residual heat removal, safety injection, and depressur- SCDAP modelling options (the new QUENCH/PARAM- ization as shown in Figure 1. International Journal of Nuclear Energy 3

#1 40 M M ADS 1 3 stages – #2 (1 of 2) 41 M M

#3 42 MM Containment 25

45 26 Refuel cavity Core makeup 1 12 tank (1of 2) Spargers M (1 of 2) Pressurizer 15 PRHR 2 HX 28 11 IRWST IRWST 16 51 screen M (1 of 2) Loop FAI 52 compart. M 18

33 M M FO Recirc. 30 ADS screen 4 stage (1 of 2) Accum. (1 of 2) RNS (1 of 2) 32 pumps 20 50 N2 6 5

CL 21 HL

DVI conn. RNS 35 pumps (1 of 2) 22 Core M

Reactor vessel

Figure 1: AP1000 passive core cooling system [8].

3.2.1. Core Make-Up Tank (CMT). The core make-up tanks 3.2.2. Passive Residual Heat Removal (PRHR). The passive replace the high-pressure safety injection systems in conven- residual heat removal heat exchanger (PRHR HX) is designed tional PWRs. The two CMTs are located at an elevation above tousepassiveprocessessuchasthegravityeffectandnatural the core; they are filled with borated water and provide the circulation in its operation. It is the main component of reactor coolant system (RCS) makeup and boration for the PRHRS and is immersed in the in-containment refueling loss of coolant accident (LOCA) and non-LOCA events when water storage tank (IRWST) which acts as heat sink. The the normal makeup system is unavailable or insufficient. Each PRHRHXconsistsofC-shapedtubebundlethatisemployed CMT consists of a large volume stainless steel tank with an to remove core decay heat from the RCS for certain postulated inlet line that connects one of the cold legs to the top of accident events where a loss of cooling capacity via the steam the CMT and an outlet line that connects the bottom of the generators occurs. The heat exchanger is maintained full of CMT to the direct vessel injection (DVI) line. The DVI line cold RCS coolant at full RCS pressure. The heat exchanger is connected to the reactor vessel downcomer. Each CMT is is connected to the RCS by an inlet line from one RCS hot filledwithcoldboratedwater.TheCMTinletvalveisnor- leg through a tee from one of the fourth stage automatic mally open and hence the CMT is normally at primary system depressurization lines [10]. pressure. The CMT outlet valve is normally closed, preventing natural circulation during normal operation. When the outlet 3.2.3. Automatic Depressurization System (ADS). The auto- valve is open, a natural circulation path is established. Cold matic depressurization system consists of four stages of valves boratedwaterflowsintothereactorvesselandhotprimary that provide for the controlled reduction of primary system fluid flows upward into the top of the9 CMT[ ]. pressure. The first three stages consist of two trains of valves The CMTs can operate in two different modes, depending connected to the top of the pressurizer. The first stage opens on the RCS conditions. If the cold legs are filled with water, when CMT liquid level reaches 67.5% of its initial liquid level. CMTs operate in a water recirculation mode with the driving ADS stages two and three open shortly after a time force based on gravity and on the density difference between delay of 70 sec and 120 sec, respectively. The ADS 1–3 valves thehotreactorcoolantintheCMTbalancelineandthe discharge primary system steam into a Sparger line that vents colder water in the CMT. If the cold legs become voided, as into the IRWST. The steam is condensed by direct contact they do during LOCAs, the CMTs will operate in a steam with the highly subcooled water in the IRWST. displacement injection or steam drain-down mode. In this The fourth stage of the ADS consists of two large valves mode, the driving force is based on gravity and the density attached to ADS lines on each hot leg. The ADS-4 valves difference between steam from the cold legs and water in the open when the CMT liquid level is low (20% of its initial CMTs. level) and effectively bring primary-side pressure down to 4 International Journal of Nuclear Energy containment conditions. The ADS-4 valves vent directly into Table 1: AP1000 initial conditions. the containment building. Parameter Initial value The first three stages of the ADS are represented by three Core thermal power (MW) 3400.00 parallel pipes connected to the pressurizer upper head and 3 each flow path has two motor valves in series whose action Coolant volume flow per loop (m ) 35772.14 RCS pressure (MPa) 15.52 is controlled by an open trip and a close trip in the input Vessel inlet temperature (K) 553.82 deck. The two ADS stage four paths are connected to the Vessel outlet temperature (K) 594.26 two hot legs, respectively, and the associated squib valves are SG secondary pressure (MPa) 5.61 modelled by trip valves [9]. SG feedwater temperature (K) 499.82 SG feedwater temperature (K) 944.35 3.2.4. In-Containment Refueling Water Storage Tank (IRWST). Steam flow per SG (kg/s) 28.32 The in-containment refueling water storage tank is a very large concrete pool filled with cold borated water. It serves as the heat sink for the PRHR heat exchanger and a source of andsomeapproximationshavebeenappliedtothedeveloped water for IRWST injection. The IRWST has two injection lines model. connected to the reactor vessel DVI lines. These flow paths As the AP1000 is a two-loop PWR, each loop consists arenormallyisolatedbytwocheckvalvesinseries.Whenthe of one hot leg, two cold legs, one steam generator, and two primary pressure drops below the head pressure of the water canned motor pumps integrated into the steam generator intheIRWST,aflowpathisestablishedthroughtheDVIinto channel head. The secondary side of the steam generator is the reactor vessel downcomer. The IRWST water is sufficient simulated by a downcomer through which the main feed- to flood the lower containment compartments to a level above water enters, a boiler consisting of the subcooled region and thereactorvesselheadandbelowtheoutletoftheADS-4 the boiling region, a separator where vapour-water separation lines. occurs, and a steam dome connected to the steam header [11]. The IRWST is modelled as two connected fluid nodes (as done in the NOTRUMP code [11]). The lower node is connected to the direct vessel injection line and is the source 3.3.1. AP1000 Model Nodalization and Initial Conditions. In of injection water to the DVI lines driven by a gravity head. this model, a RELAP/SCDAPSIM model of the Surry reactor The upper node acts as a sink for the ADS flow from the pressure vessel (RPV) is adopted and necessary changes were pressurizerandasaheatsinkforthePRHRheatexchanger. made to model the AP1000 pressure vessel. Some passive Thesenodesaremodelledashavinganinitialtemperatureof safety systems were modelled. The loop designated “b” has 322 K and a pressure of 101.35 kPa. the pressurizer and the PRHR system connections, and loop “a” cold legs have the core makeup tank pressure balance line connections as shown in the nodalization diagram, Figure 2. 3.2.5. Accumulators (ACC). The accumulators are similar to The initial conditions are listed in Table 1. those found in conventional PWRs. They are large spherical The core is composed of a downcomer (104) connecting tanks approximately three-quarters filled with cold borated to the lower head (106) and the lower plenum (108). The water and pre-pressurized with nitrogen. The accumulator fuelled parts of the core are composed of five channels (111, outlet line is connected to the DVI line. A pair of check valves 112, 113, 114, and 115) and a bypass region (118). The upper prevents injection flow during normal operating conditions. plenum of the core is a 3-channel model ([151-152-153-154], When system pressure drops below the accumulator pressure [161-162-163-164],and[171-172-173-174]). The control (plus the check valve cracking pressure), the check valves assembly housing is represented in three parts (181, 182, and open allowing coolant injection to the reactor downcomer via 183). The upper head (190) covers the core and 100 and 102 the DVI line. represent upper annulus and inlet annulus, respectively. TheACCismodelledinRELAP5asalumpedparameter Components 140–149 connect the centre core Channel component and therefore a nitrogen-charged accumulator 1 and centre core Channel 2. Components 120–129 connect andsurgelinesystemcanbesimulated[11]. thecentrecoreChannel2andthemiddlecoreChannel3. Components 80–89 connect the middle core Channel 3 and 3.3. AP1000 Model Description. RELAP/SCDAPSIM/MOD3 the middle core Channel 4. Components 130–139 connect the input requirements can be divided into four distinct areas: middle core Channel 4 and the outer core Channel 5. hydrodynamics, heat structures, control systems, and neu- The pressurizer surge line and the hot and cold leg piping tronics. This model is mainly composed of hydrodynamic are constructed of stainless steel. The power operated relief components that basically represent the parts of the reactor valves (PORV) are located at the top of the pressurizer and where coolant passes through and heat structures that repre- can be used to relieve excess pressure in reactor coolant sent solid parts of the reactor. system (RCS). Pressurizer safety relief valves (SRV) are also Owing to the proprietary nature of Westinghouse techni- available to handle pressure excursions in excess of the cal documentation, very little information related to design pressurizer PORV capacity. parameters and small break LOCAs in AP1000 can be The pipe 400 is the hot leg of the loop. The branch con- obtained in the open literature. Therefore, some parameters nects the pressurizer to the circuit. The pipe 443 represents havebeenassumedtobethesameasinaconventionalPWR thesurgeline.Thepipe441representsthetank.Theupper International Journal of Nuclear Energy 5

448 ADS 1, 2, 3 458 AP1000 nodalization 468

484 384 IRWST 780 800 485 487 489 PRHR 385 387 389 444 790 449

770 760 486 488 490 386 388 390 445 775 765

PRZ 775 720 440 765 620 482 382 441 480 380 470 478 378 370 ADS4 1 ADS4 2 472 372 476 915 376 442 905 700 600 CMT2 376 190 376 CMT1 725 408 308 625 174 164 154 374 474 910 900 100 133 131 173 132 163 153 100

406 443 306 710 SG 2 SG 1 102 172 162 152 102 610

171 161 151

404 HL2 HL1 304 402 406 300 302

416 CL3 CL1 314 422 420 418 416 322 320 318 316 104 115 113 114 714 CL4 104 116 112 111 CL2 614 716 722 616

DVI2 DVI1 750 108 650 106

540 530

Acc. 2 Acc. 1

Figure 2: AP1000 nodalization. part of this tank, the pressurizer dome (440), is represented The pump included in this system is identical for each by a branch-type element. Two valves (444, 445) connect the loop.Itispossibletousethebuilt-inpumpmodelsorto tank to a containment represented by a single volume (449). describe the characteristics of head and torque, pump head Their action is controlled by open trips 631 and 604 and close and torque multiplier, and two-phase difference curve. trips 633 and 606, respectively, in the input deck. Two direct vessel injection (DVI) nozzles are located in 3.3.2. Modelling Safety System Components the upper downcomer and safety injection water from CMTs, accumulators, IRWST, and containment sump is supplied (1) Core Makeup Tank (CMT). Two CMTs, 600 and 700, are through the nozzles. These nozzles are modelled as valves, 650 modelled as pipes consisting of 4 volumes each. CMT-1 (600) and 750. is connected through the discharge line, 610, to the direct Each steam generator model includes a separator (378 and vessel injection (DVI), 650, which is modelled as a single 478resp.),motorvalves387,487,389,and489;pipes374,474, volume. From the top, CMT-1 (600) is connected through the 376,476,308,and408;branches380,480,372,472,306,406, balance line, 630, to the cold leg, 618, modelled as branch. The 310, and 410; time dependent junctions 385, 485, 367,and 467; inlet and outlet valves are modelled as trip valves 630 and 610. single volumes 370, 470, 304, and 404; and time dependent CMT-2 (700) is connected through the discharge line, 710, volumes366,466,390,490,388,488,386,and486.Theaction to the direct vessel injection (DVI), 750, which is modelled of safety and PORV valves is controlled by open trips 610 asasinglevolume.FromthetopCMT-2(700)isconnected and 619 and close trips 612 and 621, respectively, in the input throughthebalanceline,730,tothecoldleg,320,modelledas deck. branch. The inlet and outlet valves are modelled as trip valves There are many heat structures in the core, since all 730 and 710. The two valves, 610 and 710, are adjusted to open walls have conduction properties that cannot be neglected. after the RCS pressure reaches a value of 11.7 MPa, using trip All vessel walls, 100, thermal shields, 104 (in addition to card625.Valvesfromthepressurebalancelineandthecold thevesselwalls,106,107,and108),differentstructuresand legs, 320 and 618, are adjusted to open just after the CMTs plates present in the core to maintain the assemblies and the start to inject water, using trip card 642. integrity of the vessel, the core barrel, 201, and the core baffle, 151, were modelled with heat structures. Also the cold and hot (2) Automatic Depressurization System (ADS).Thefirstthree leg piping, steam generator tubes and shells, pump suction stagesoftheADSaremodelledasvalves,448,458,and468, piping, and PRHRHX were represented by heat structures. and are connected to the pressurizer upper head, 440, and to 6 International Journal of Nuclear Energy

Table 2: Set points for SBLOCA. Event Set point Reactor trip on low pressurizer pressure 12.41 MPa Reactor trip on low-low pressurizer pressure 11.72 MPa “S” signal on low-low pressurizer pressure After reactor trip signal PRHR isolation valve starts to open After “S” signal CMTs injection starts After “S” signal ACC injection starts on low RCS pressure 4.83 MPa ADS-1 control valve trip signal 20 s after 67.5% liquid volume fraction in CMT ADS-2 control valve trip signal 70 s after ADS-1 actuation ADS-3 control valve trip signal 120 s after ADS-2 actuation ADS-4A starts to open 20.0% liquid volume fraction in CMT and 120 s after ADS-3 actuation ADS-4B starts to open 60 s after ADS-4A actuation IRWST injection starts RCS pressure RCS pressure less than 0.18 MPa

the discharge line to the upper IRWST, 780. Their action is 60 controlled by open trip 629 and close trip 627 in the input 50 deck. The fourth stage of ADS is modelled as two valves, 905 40 and915.Onesideofvalve905isconnectedtohotleg400 30 in loop b, and the other side of the valve to containment 449. Onesideofvalve915isconnectedtohotleg300inloopb,and 20 the other side of the valve to containment 449. Their action is 10 controlled by open trip 630 and close trip 627. 0 Flow rate (kg/s) rate Flow −10 (3) In-Refueling Water Storage Tank (IRWST).TheIRWST − is modelled as two connected nodes: a source, 790, and a 20 sink,780.Thesource790isconnected(discharges)tothe −30 0 500 1000 1500 2000 2500 3000 3500 4000 two DVIs, 650 and 750, through two valves, 770 and 760. Their action is controlled by open trip 628. The sink, 780, is Time (s) connected to the first three stages of ADS. Figure 3: CMT injection flow rate during SBLOCA. (4) Accumulator (ACC). There are two accumulators, 530 and 540, driven by compressed nitrogen. Accumulator 1, 530, in loop (a) is connected to DVI1, 650. Accumulator 2, 540, in loop (b) is connected to DVI2, 750. decay heat level with the effect of shutdown rods. The RCS continues depressurization till it reaches the low-low pressure 3.4. Description of the SBLOCA Scenario. For the safety set point (11.72 MPa), which is the “S signal” and safety system analyses of the AP1000, a small break LOCA is defined as actuation point. Then, both CMTs and PRHRHX are then 2 activated. arupturewithatotalcross-sectionalarealessthan1.0ft of The two CMTs are activated just after the “S” signal at the reactor coolant pressure boundary. This is considered a about 𝑡=10sec providing relatively high flow of borated Condition III event (infrequent fault) that may occur during water for a longer duration and with an increasing flow rate thelifeoftheplant[8]. reaching a maximum value of 55 kg/sec at 𝑡=50sec. After A 2 in. cold leg small break LOCA is simulated by that the flow continues with a decreasing rate till both CMTs adding a trip valve, 550, to the broken cold leg, 320, in the empty at 𝑡 = 3200 sec as shown in Figure 3.Thetimesequence nonpressurizer loop. The set points for SBLOCA analysis are of events is listed in Table 3. listed in Table 2. Once the RCS pressure reaches 4.83 MPa, the accumulators begin to provide a high flow of borated water in a short 4. Results and Discussions timethroughtheDVIlines.Theaccumulatorsareactivated at 𝑡 = 900 sec with a flow rate starting at 81 kg/sec which Alossofcoolantaccidentwith2-inchdiameterbreakatthe decreases until both accumulators empty at 𝑡 = 1800 sec as bottom of cold leg is simulated. The scenario of the accident shown in Figure 4. can be analysed as follows. At 𝑡=0s, a break was initiated After the water level in both CMTs reaches 67.5% of their in the cold leg by opening the LOCA valve. The primary original value, the first three stages of automatic depressur- system starts depressurization until the RCS pressure reaches ization system (ADS) start depressurizing the RCS effectively. the low value set point (12.41 MPa), a reactor scram signal The first stage begins at 𝑡 = 1600 sec and the other two stages is generated, and the core thermal power falls rapidly to the start with a delay of 70 sec and 120 sec, respectively. International Journal of Nuclear Energy 7

Table 3: Time sequence of events. 120

Event Time 100

Break opens 0.0 80 “S” signal 10 CMT discharge valves open 10 60 Accumulators injection starts 900 40 Accumulators emptied 1800 Flow rate (kg/s) rate Flow 20 ADS1 starts 1600 ADS2 starts 1670 0 ADS3 starts 1720 −20 ADS4 2100 0 500 1000 1500 2000 2500 3000 3500 4000 IRWST injection starts 2800 Time (s) Figure 5: ADS4 injection flow rate during SBLOCA. 90 80 70 60 NOTRUMPandareinagoodagreementinbothinitiation 50 40 and end of injection times, but there is a difference in the 30 injection flow rates. 20 Flow rate (kg/s) rate Flow 10 0 5.3. ADS-4. Figure 8 shows the ADS4 response in −10 NOTRUMP and RELAP/SCDAPSIM. The two models 0 500 1000 1500 2000 2500 3000 3500 4000 have the same average flow rates, 100 kg/sec, but there is Time (s) a difference in the initiation of injection time. This canbe Figure 4: Accumulators’ injection flow rates during SBLOCA. caused by the assumptions made for some parameters when modelling ADS4 due to lack of design data.

After the water level in both CMTs reaches 20% of their 5.4. IRWST. Figure 9 shows the IRWST response in original level, the fourth stage of ADS is activated at 𝑡= NOTRUMP and RELAP/SCDAPSIM. There is a difference 2100 sec with a flow rate starting at 110 kg/sec as shown in in the initiation time, but in both models the average flow Figure 5. rate is 50 kg/sec. After the primary system pressure drops to near the The calculated results obtained by RELAP/SCDAPSIM containment pressure, injection from the IRWST initiates. are consistent with those obtained using NOTRUMP. The The IRWST is activated at 𝑡 = 2800 sec and provides low main difference between them is the actuation times of flow of borated water at an average flow rate of 50 kg/sec passive systems which are partly due to the difference in the and continues injection for longer time till the end of model break discharge flow rate and also due to lack of available scenario. information in the literature on the AP1000 passive safety system design and the difficulty of deducing all these parameters. However, the results show, in most cases, the same trend in 5. Model Validation the response of each safety system during the SBLOCA, that is, the increasing flow rate. The results of the developed model will be compared in this section with the results of a 2 in. LOCA modelled by the NOTRUMP code [12] which is used by Westinghouse in [13]. 6. Conclusions Comparisons between the responses of different passive safety systems are illustrated in Figures 6–9. The RELAP/SCDAPSIM model of Surry reactor (2-loop PWR) was adopted. Some modifications to the Surry model were made to model the passive safety components in the 5.1. CMT. Figure 6 shows the CMT response in NOTRUMP AP1000 and some changes to several parameters were also and in RELAP/SCDAPSIM. The graph shows the same made. Many assumptions and approximations were made to injection rates in both models, but there is a difference themodelduetothelackoftheavailabilityofmanydesign in the end of injection time. This can be caused by the parameters for the AP1000. approximations and assumptions done in modelling the CMT by the RELAP/SCDAPSIM code. (i)ThedevelopedmodelfortheAP1000wasusedtoeval- uate the safety performance of an AP1000 during a 5.2. Accumulators. Figure 7 shows the accumulator response small break LOCA. The passive safety components in in NOTRUMP and RELAP/SCDAPSIM. The results obtained the AP1000 showed a clear improvement in accident from RELAP/SCDAPSIM model have the same trend of the mitigation. 8 International Journal of Nuclear Energy

140 60 60 120 50 40 100 40 80 20 30 60 0

40 20 (kg/s) rate Flow Mass flow rate (kg/s) rate flow Mass Mass flow rate (lbm/s) rate flow Mass −20 20 10

0 0 −40 0 1000 2000 3000 4000 5000 0 1000 2000 3000 4000 Time (s) Time (s) (a) NOTRUMP (b) RELAP/SCDAPSIM

Figure 6: CMT (1) injection flow rate during SBLOCA.

300 100 120 250 80 100 200 60 80 150 40 60 100 20

40 (kg/s) rate Flow Mass flow rate (kg/s) rate flow Mass Mass flow rate (lbm/s) rate flow Mass 50 20 0

0 0 −20 0 1000 2000 3000 4000 5000 0 1000 2000 3000 4000 Time (s) Time (s) (a) NOTRUMP (b) RELAP/SCDAPSIM

Figure 7: Accumulator (1) injection flow rate during SBLOCA.

400 150

150 300 100

100 200 50

50 (kg/s) rate Flow 100 0 Mass flow rate (kg/s) rate flow Mass Mass flow rate (lbm/s) rate flow Mass

0 0 −50 0 1000 2000 3000 4000 5000 0 1000 2000 3000 4000 Time (s) Time (s) (a) NOTRUMP (b) RELAP/SCDAPSIM

Figure 8: ADS4 response during SBLOCA.

(ii)Thepassivesafetysystemswerefoundtobecapable have important implications for enhancing the safety of depressurizing the reactor coolant system while performance of nuclear reactors. maintaining acceptable core conditions and establishing a stable delivery of cooling water from the IRWST. (iii) RELAP/SCDAPSIM was found to be a very good The AP1000 was found to have a good performance modelling tool for simulating different passive com- in mitigating SBLOCA consequences. These findings ponents of the AP1000 and to model a small break International Journal of Nuclear Energy 9

120 80 50 100 60 40 80 40 30 60 20

20 40 (kg/s) rate Flow 0 Mass flow rate (kg/s) rate flow Mass Mass flow rate (lbm/s) rate flow Mass

20 10 −20

0 0 −40 0 1000 2000 3000 4000 5000 0 2000 4000 6000 Time (s) Time (s) (a) NOTRUMP (b) RELAP/SCDAPSIM

Figure 9: IRWST injection flow rate during SBLOCA.

LOCA. The model was capable of modelling a LOCA [2] G. Saiu and M. L. Frogheri, AP1000 Nuclear Power Plant intheAP1000andenabledtheinvestigationof Overview,ANSALDOEnergia—NuclearDivision,Genoa,Italy. each safety system component response separately [3]A.E.Abouelnaga,A.Metwally,N.Aly,M.Nagy,andS.Agamy, during the accident. The model was also capable “Assessment the safety performance of nuclear power plants of simulating natural circulation and other different using Global Safety Index (GSI),” Nuclear Engineering and phenomena. Design,vol.240,no.10,pp.2820–2830,2010. [4] “PCTran for PWR and AP1000,” Micro Simulation Technology, (iv)Whencomparingtheresultswiththatofthe 2006. NOTRUMP code used by Westinghouse, they were [5] http://www.relap.com. found to have the same trends, but some differ- [6]C.M.AllisonandJ.K.Hohorst,Role of RELAP/SCDAPSIM in ences in the initiation of safety systems were found. Nuclear Safety, Science and Technology of Nuclear Installations, These differences are caused by the assumptions and 2010. approximations made by both codes and also because [7] L. J. Siefken, C. M. Allison, and J. K. Hohorst, of the many assumed values for the design data used “RELAP/SCDAPSIM/MOD3.5—improvements resulting from in the RELAP model due to the lack of published QUENCH and PARAMETER bundle heating and quenching values for the AP1000 design parameters. However, experiments,” in Proceedings of the 8th International Conference theresultsshowthesametrendintheresponseof onNuclearOptioninCountrieswithSmallandMedium each safety system during the LOCA, that is, the Electricity Grids, Dubrovnik, Croatia, May 2010. increasing flow rate, the decreasing pressure, and so [8] Westinghouse AP1000 Design Control Document,Westinghouse forth. Electric Company, 2004. [9] T. L. Schulz, “Westinghouse AP1000 advanced passive plant,” Nuclear Engineering and Design,vol.236,no.14–16,pp.1547– Conflict of Interests 1557, 2006. [10] “Westinghouse AP1000 Design Control Document, ML11171- The authors declare that there is no conflict of interests A500, Rev. 19, Chapter 15: Accident Analyses,” 2011. regarding the publication of this paper. [11] R. C. D. Team, “RELAP5/MOD3.3 Code Manual: NUREG/CR- 5535”. Acknowledgments [12] “NOTRUMP, a nodal transient small-break and general network code,” Tech. Rep. WCAP-10079-P-A, 1985. The authors would like to thank Professor Gianfranco Saiu [13] Chapter 21. Testing and Computer Code Evaluation, “Final and Dr. Monica Linda Frogheri in Ansaldo Energia for their Safety Evaluation Report Related to Certification of the AP1000 sincere help and advice in making the nodalization of the Standard Design,” Initial Report NUREG-1793. AP1000.

References

[1] AP1000 Nuclear Power Plant BAT Assessment,Westinghouse Electric Company, 2011. Journal of International Journal of Journal of Rotating Wind Energy Machinery Energy

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