IMPACTS OF INTERCONNECTING A NUCLEAR
POWER PLANT ON THE PERFORMANCE OF AN
EXISTING POWER SYSTEM NETWORK.
By
Amr Adel Fathy Mohammed
A thesis submitted to the
Faculty of Engineering at Cairo University
In Partial Fulfillment of the
Requirements for the Degree of
MASTER OF SCIENCE
In
Electrical Power and Machines Engineering
Under supervision of
Prof. Dr. Husain Magdy Zein el-din
Electrical Power and Machines Department
Faculty of Engineering
Cairo University
FACULTY OF ENGINEERING, CAIRO UNIVERSITY
GIZA, EGYPT
OCTOBER 2011
I
ACKNOWLEDGMENTS
First of all, thanks Allah who supported and strengthened me all through
my life.
It is a great pleasure to express my profound gratitude and sincere
appreciation to Prof. Dr. Husain Magdy Zein el-din of the Electrical Power and
Machines department, Faculty of Engineering, Cairo University for his faithful
supervision with a continuous enthusiastic encouragement. Also, I would like to
thank Prof. Dr. Magdy El-Marsfawy for his guidance and help.
Also, I would like to express my deep gratitude to Eng. Shady Abd ElMongy and Eng. Haytham Mahmoud for their support to me. Finally, I would like
to thank my parents and each member of my family for their words of great
inspiration and encouragement.
II
ABSTRACT
The nuclear power plant became one the most optimal solution for
satisfying the continuous and increased demand for electric energy because of its
high capacity, its very low running cost, and assumed to be environmentally clean
source of base load electrical generation. For these reason a lot of developing
countries such as Egypt decides to begin its nuclear program.
In this thesis, the impacts of interconnecting between the Egyptian electric
power system and proposed nuclear power plant (Dabaa) are studied and analyzed
through the dynamic modeling and simulation of both nuclear power plant and
electric transmission grid which are fascinating engineering achievements on their
own. When they are connected together in a highly controlled, dynamic and
distributed network, further complexity is created. This complexity of engineered
systems is a consequence of several factors: the sheer size and interconnectivity of
the electric grid, the nuclear safety requirements imposed on NPP, and the need to
balance electricity supply and consumption throughout the grid at all times.
A study for different type of nuclear reactors is performed to the select the
best type for the first Egyptian nuclear power plant based on technical,
economical, and environmental factors. Also a study for Egyptian unified power
system with some assumptions, related to the detailing of system as whole and the
details for modeling of several elements, where the objective of this study is
determining the optimal way for interconnecting between the proposed NPP
(Dabaa) and the power system. Then, for different disturbances like outage of
main transmission lines, different fault locations and system contingencies, the
behavior of the nuclear unit (pressure, different temperatures, coolant flow rate,
steam flow rate and reactivity) under these disturbances is monitored.
III
TABLE OF CONTENTS
ACKNOWLEDGMENTS ................................................................................................... I
ABSTRACT...................................................................................................................... III
TABLE OF CONTENTS .................................................................................................. IV
LIST OF TABLES ........................................................................................................... VII
LIST OF FIGURES ....................................................................................................... VIII
LIST OF SYMBOLS AND ABBREVIATIONS ............................................................ XII
CHAPTER 1 ....................................................................................................................... 1
INTRODUCTION AND REVIEW OF LITERATURE .................................................... 1
1.1
Thesis motivation and objectives ............................................................................... 1
1.2
Literature review ........................................................................................................ 2
1.3
Thesis layout .............................................................................................................. 4
CHAPTER 2 ....................................................................................................................... 5
CLASSIFICATION OF NUCLEAR POWER STATION ................................................. 5
2.1
Introduction ................................................................................................................ 5
2.2 Basic construction of the reactor................................................................................... 6
2.3 Classification of the traditional nuclear power reactors ............................................... 8
2.3.1 Classification by type of nuclear reaction ................................................................ 10
2.3.1.1 Nuclear fission reactors......................................................................................... 10
2.3.1.2 Nuclear fusion reactors. ........................................................................................ 11
2.3.2 Classification by moderator material ....................................................................... 11
2.3.2.1 Graphite moderated reactors ................................................................................. 11
2.3.2.2 Water moderated reactors ..................................................................................... 13
2.3.2.3 Light element moderated reactors. ........................................................................ 14
2.3.2.4 Organically moderated reactors (OMR) ............................................................... 15
2.3.3 Classification by coolant .......................................................................................... 15
2.3.3.1 Water cooled reactor ............................................................................................. 15
2.3.3.1.1 Pressurized water reactor (PWR) ....................................................................... 15
2.3.3.1.2 Boiling water reactor (BWR) ............................................................................. 16
IV
2.3.4.1 Electric power reactors ......................................................................................... 18
2.3.4.2 Desalination and Heating reactors ........................................................................ 18
2.3.4.3 Nuclear propulsion ................................................................................................ 18
2.3.4.4 Production of elements ........................................................................................ 18
2.6.1 Safety facilities......................................................................................................... 25
2.6.2 Economics ................................................................................................................ 28
2.6.2.1 The construction cost of building the plant........................................................... 28
2.6.2.2 The operating cost of the plant.............................................................................. 29
2.6.2.3 The cost of waste disposal from the plant............................................................. 29
2.6.2.4 The cost of decommissioning the plant ................................................................ 30
2.6.3 Other factors............................................................................................................. 30
2.6.4 Conclusion for selecting the type of reactor in Dabaa NPP..................................... 30
CHAPTER 3 ..................................................................................................................... 32
DETAILED CONSTRUCTION AND DYNAMIC MODELING OF PRESSURIZED
WATER REACTOR ................................................................................................ 32
3.1 Introduction ................................................................................................................. 32
3.2 Basic construction of PWR ......................................................................................... 34
3.2.1 Primary system of PWR........................................................................................... 35
3.2.2 Secondary system of PWR....................................................................................... 37
3.2.3 The auxiliary systems of PWR ................................................................................ 38
1) The chemical and volume control system .................................................................... 38
2) The decay heat removal system .................................................................................... 39
3) The residual heat removal system ................................................................................ 40
4) Emergency core cooling systems ................................................................................. 41
3.3 Mathematical model of proposed NPP ....................................................................... 44
3.3.1 Neutron Dynamics Model ........................................................................................ 44
3.3.2 The Steam Generator Model .................................................................................... 45
3.3.3 Core Fuel and Coolant Temperature Model ............................................................ 45
3.3.4 The Hot Leg and Cold Leg Temperature Model ..................................................... 46
V
3.3.5 The coolant pump model ......................................................................................... 46
3.3.6 The steam turbine model .......................................................................................... 48
3.3.6 The reactor control system model ............................................................................ 49
3.3.7 The reactor protection system .................................................................................. 50
CHAPTER 4 ..................................................................................................................... 51
INTERCONNECTING DABAA TO THE EGYPTIAN UNIFIED POWER SYSTEM . 51
4.1 Introduction ................................................................................................................. 51
4.2 Studying the Egyptian unified power system ............................................................. 52
4.2.1 Overview about the Egyptian unified power system ............................................... 52
4.2.2 Overview about the analysis software package used in this study (PSAT) ............. 53
4.2.3 Size selection for units of Dabaa NPP ..................................................................... 55
4.2.4 Proposed model of the Egyptian unified power system........................................... 57
4.3 Cases of contingency .................................................................................................. 60
4.3.1 Fault at bus of Dabaa-500 accompanied by outage of double circuit line T.L1 .... 60
4.3.2 Fault at bus of Dabaa-500 accompanied by outage of double circuit line T.L2 ...... 66
4.3.3 Fault at bus of Dabaa-500 accompanied by outage of two lines T.L1 and T.L2 ..... 71
4.3.4 Fault at bus of Dabaa-500 accompanied by islanding of NPP................................. 75
4.3.5 Fault at bus of Dabaa-220 accompanied by outage of lines T.L3 and T.L4 ............ 81
CHAPTER 5 ..................................................................................................................... 85
CONCLUSIONS AND FUTURE WORK ....................................................................... 85
5.1
Conclusions .............................................................................................................. 85
5.2
Future work .............................................................................................................. 86
REFERENCES ................................................................................................................. 87
APPENDIX (A) ................................................................................................................ 90
APPENDIX (B) ................................................................................................................ 93
VI
LIST OF TABLES
Title
Page
Table 2.1: Summary of the main thermal reactors …………………………………………...…..20
Table 4.1: Data of new lines which connect Dabaa NPP with the grid………….……………….57
Table A.1: Reactor design data …………………..………………………………………...…..90
Table A.2: Steam generator data …………………..…………………………………………..91
Table A.3: Pressurizer design data …………………..………………………………….………..92
Table (B.1): Electricity for 2009/2010……………………………………………………………93
Table (B.2): General Power Stations Statistics…………………………………………………...95
Table (B.3): Development of Installed Capacities………………………………………………..96
Table (B.4): Performance Statistics for Power Plants…………………………………………....97
VII
LIST OF FIGURES
Title
Page
Figure 2.1: Evolution of nuclear power reactors…………………………………………………...6
Figure 2.2: Basic construction of Nuclear Steam Supply System (NSSS)……………...................8
Figure 2.3: Classification of nuclear power reactors………………………………………………9
Figure 2.4: Basic Gas-Cooled Reactor (MAGNOX)………………………………………….….12
Figure 2.5: Advanced Gas-Cooled Reactor (AGR)……………………………………….……...13
Figure 2.6: CANDU Reactor (HWR)…………………………………………………………….14
Figure 2.7: Pressurized water Reactor (PWR)……………………………………………………16
Figure 2.8: Boiling water Reactor (BWR)………………………………………………………..17
Figure 2.9: Advanced pressurized water reactor (AP-1000)……………………………………...23
Figure 2.10: Advanced boiling water reactor (ABWR)…………………………………………..24
Figure 2.11: Very high temperature reactor (VHTR)…………………………………………….24
Figure 2.12: Sodium-cooled fast reactor (SFR)………………………………………………..…25
Figure 2.13: Safety facilities……………………………….……………………………………..27
Figure 3.1: Modeled components in PWR power plant…………………………………………..33
Figure 3.2: Basic components of PWR………………………………………………………..….34
Figure 3.3: Primary system of PWR……………………………………..……………………….35
Figure 3.4: Schematic diagram of the secondary system……………………………………...… 37
Figure 3.5: Chemical and volume control system………………………………………………..38
Figure 3.6: Decay heat removal system…………………………………………………………..40
Figure 3.7: Residual heat removal system……………………………………………………..…41
Figure 3.8: Emergency core cooling system…………………………………………………...…43
VIII
Figure 3.9: Mathematical model of PWR power plant……………………………………….…..44
Figure 3.10: Typical model of the reactor coolant pump…………………………………………47
Figure 3.11: Typical model of steam turbine used in NPP……………………………………….48
Figure 3.12: Typical control system of reactor………………………………………………...…49
Figure 3.13: Typical protection system of reactor………………………………………………..50
Figure 4.1: The distribution of annual electric generated energy in percent………………..……53
Figure 4.2: The installed capacity growth of Dabaa NPP…………………………………….…..57
Figure 4.3: The interconnection between Dabaa and unified Egyptian grid……………………..59
Figure 4.4: Delta of Dabaa gen. unit for fault at Dabaa 500 and line (L1) outage……………….61
Figure 4.5: Frequency of Dabaa gen. unit for fault at Dabaa-500 and line (L1) outage………….62
Figure 4.6: Voltages buses for fault at Dabaa 500 and line (L1) outage…………………...…….62
Figure 4.7: Change of Steam pressure of Dabaa PWR for fault at Dabaa 500 and line (L1)
outage…………………………………………………………………..…………………………63
Figure 4.8: Change of Primary temperature of Dabaa PWR for fault at Dabaa 500 and L1
outage………………………………………………………………………………..……………63
Figure 4.9: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L1
outage………………………………………………………………………………..……………64
Figure 4.10: Change of fuel temperature of Dabaa PWR for fault at Dabaa 500 and L1
outage……………………………………………………………………………………………..64
Figure 4.11: Change of hot-leg temperature of Dabaa PWR for fault at Dabaa 500 and L1
outage………………………………………………………………………………….………….65
Figure 4.12: Change of cold-leg temperature of Dabaa PWR for fault at Dabaa 500 and L1
outage………………………………………………………………………………………..……65
Figure 4.13: Delta of Dabaa gen. unit for fault at Dabaa 500 and line (L2) outage………...……67
IX
Figure 4.14: Frequency of Dabaa gen. unit for fault at Dabaa-500 and line (L2) outage………...67
Figure 4.15: Voltage buses for fault at Dabaa 500 and line (L2) outage……………………..…..68
Figure 4.16: Change of Steam pressure of Dabaa PWR for fault at Dabaa 500 and line (L2)
outage……………………………………………………………………………………….…….68
Figure 4.17: Change of Primary temperature of Dabaa PWR for fault at Dabaa 500 and L2
outage……………………………………………………………………………………….…….69
Figure 4.18: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L2
outage……………………………………………………………………………………..………69
Figure 4.19: Change of fuel temperature of Dabaa PWR for fault at Dabaa 500 and L2
outage……………………………………………………………………………………….…….70
Figure 4.20: Change of Hot-leg temperature of Dabaa PWR for fault at Dabaa 500 and L2
outage…………………………………………………………………………….……………….70
Figure 4.21: Change of Cold-leg temperature of Dabaa PWR for fault at Dabaa 500 and L2
outage…………………………………………………………………………………….……….71
Figure 4.22: Delta of Dabaa gen. unit for fault at Dabaa 500 and lines (L1 & L2) outage….…...72
Figure 4.23: Frequency of Dabaa unit for fault at Dabaa-500 and lines (L1 & L2) outage….......72
Figure 4.24A: Voltage buses for fault at Dabaa 500 and lines (L1 & L2) outage………………..73
Figure 4.24B: Voltage buse of Dabaa unit for fault at Dabaa 500 and lines (L1 & L2)
outage……………………………………………………………………………………………..73
Figure 4.25: Change of Steam pressure of Dabaa PWR for fault at Dabaa 500 and L2&L1
outage……………………………………………………………………………..………………74
Figure 4.26: Change of Primary temperature of Dabaa PWR for fault at Dabaa 500 and L2&L1
outage……………………………………………………………………………………………74
X
Figure 4.27: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L2&L1
outage…………………………………………………………….……………………………….75
Figure 4.28: Delta of Dabaa gen. unit for fault at Dabaa 500 and Islanding of NPP……...……..77
Figure 4.29: Frequency of Dabaa unit for fault at Dabaa-500 and Islanding of NPP………….....77
Figure 4.30: Voltage of Dabaa unit for fault at Dabaa-500 and Islanding of NPP………….……78
Figure 4.31: Voltage buses of 500KV system for fault at Dabaa-500 and Islanding of
NPP……………………………………………………………………………………………….78
Figure 4.32: Voltage buses of 220 KV system for fault at Dabaa-500 and Islanding of
NPP……………………………………………………………………………………………….79
Figure 4.33: Change of steam pressure of Dabaa PWR for fault at Dabaa 500 and L2 Islanding of
NPP…………………………………………………………………………………...…………..79
Figure 4.34: Change of primary temperature of Dabaa PWR for fault at Dabaa 500 and Islanding
of NPP……………………………………………………………………………….……………80
Figure 4.35: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and
Islanding of NPP………………………………………………………………………...………..80
Figure 4.36: Delta of Dabaa gen. unit for fault at Dabaa-220 and line (L4 & L3) outage…...…..81
Figure 4.37: Frequency of Dabaa unit for fault at Dabaa-220 and lines (L4 & L3) outage……...82
Figure 4.38: Bus voltages for fault at Dabaa-220 and line (L4 & L3) outage……………...…….82
Figure 4.39: Change of steam pressure of Dabaa PWR for fault at Dabaa 220 and L3&L4
outage……………………………………………………………………………………………..83
Figure 4.40: Change of primary temperature of Dabaa PWR for fault at Dabaa 220 and L3&L4
outage……………………………………………………………………………………………..83
Figure 4.41: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L3&L4
outage……………………………………………………………………………………………..84
XI
LIST OF SYMBOLS AND ABBREVIATIONS
• Symbols
A
: The heat transfer area,
C (t )
: Precursor density, p.u.
C
C
PC
PF
m
2
.
: The specific heat of coolant, kJ/kg. c˚.
: The specific heat of fuel, kJ/kg. c˚.
F
: The frequency of plant electricity bus, p.u.
h
: The heat transfer coefficient from fuel to coolant, W / m . c .
2
: the mass of coolant, kg;
m
m'
m '
m
M
M
e
: The electrical moment, p.u.
m
: The mechanical moment, p.u.
N (t )
: Neutron flux density, p.u.
P
∆P
∆T
∆T
∆T
∆T
∆T
∆T
∆T
: The initial power level, Mw
C
: the mass flow rate in core, kg/s;
C
: The coolant mass flow rate at normal speed.
Cn
F
O
: The mass of fuel, kg.
S
: The steam pressure deviation, p.u.
C1
: The coolant node 1 temperature deviation, c˚.
C2
: The coolant node 2 temperature deviation, c˚.
CL
: The cold leg temperature deviation, c˚.
HL
: The hot leg temperature deviation, c˚.
F
: The fuel temperature deviation, c˚.
m
: The tube metal lump temperature deviation, c˚.
P
: The primary loop coolant temperature deviation, c˚.
XII
°
V
: the voltage of plant electricity bus, p.u.
ρ (t )
: Reactor core reactivity.
∆ρx
: Reactivity change induced by control rod.
λ
: The equivalent decay constant, S −1 .
Λ
: Effective prompt neutron lifetime, S.
β
: Delayed neutrons total quota.
σ
σ
ω
ω
τ
τ
τ
τ
τ
F
C
P
Pn
HL
CL
P
m
Ps
: The fuel coefficient of reactivity, 1/ c˚.
: The coolant coefficient of reactivity, 1/ c˚.
: The coolant speed, p.u.
: The rated coolant speed, p.u.
: The hot leg heat transfer time constant, s.
: The cold leg heat transfer time constant, s.
: The primary coolant heat transfer time constant, s.
: The tube metal heat transfer time constant, s.
: The steam pressure heat transfer time constant, s.
∆µ
: Valve opening.
Γ
: The fraction of the total power produced in the fuel.
XIII
• Abbreviations
AGR
: Advanced Gas Cooled Reactor
BWR
: Boiling Water Reactor
CCW
: Component Cooling Water
CVCS
: Chemical and Volume Control System
ECCS
: Emergency Core Cooling Systems
EPS
: Electric Power System
FWP
: Feed Water Pump
GUT
: Graphical User Interface
HWR
: Heavy Water Reactor
LWR
: Light Water Reactor
MSR
: Molten Salt Reactors
MS/R
: Moisture Separator/Reheaters
NPP
: Nuclear Power Plant
NSSS
: Nuclear Steam Supply System
OMR
: Organically Moderated Reactors
PSAT
: Power System Analysis Toolbox
PSASP
: Power System Analysis Software Package
PWR
: Pressurized Water Reactor
RCP
: Reactor Coolant Pump
RHR
: Residual Heat Removal
RWST
: Refueling Water Storage Tank
SFR
: Sodium cooled Fast Reactor
VHTR
: Very High Temperature Reactor
XIV
CHAPTER 1
INTRODUCTION AND REVIEW OF LITERATURE
1.1 Thesis motivation and objectives
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. This review presents the
recent advances in the field of nuclear power and addresses the aspects of nuclear,
safety, nuclear reactor design and spent fuel processing and waste management.
For a country that does not yet use nuclear power, the introduction and
development of nuclear power is a major undertaking. It requires the country to
build the necessary infrastructure so it can construct and operate a nuclear power
plant (NPP) profitably in a safe, secure and technically sound manner. A major
part of the necessary infrastructure is the electric grid to which the NPP will
connect. While most countries already have an electric grid system, it may require
significant development to be suitable for the connection of an NPP. The efficient,
safe, secure and reliable operation of the NPP requires that the grid to which it
connects is also efficient, safe, secure and reliable.
NPPs are unique and powerful generators compared to other electricity generating
plants. Moreover, they are both electricity generators and customers. They thus
maintain a symbiotic relationship with the electric grid at all times. NPPs supply
large amounts of energy to the grid as well as relying on it to receive power for
crucial safety operations, especially during emergency conditions. The safe
startup, operation and shutdown of NPPs require a reliable and stable power
supply from the electric grid, referred to generally as ‘off-site power’.
1
The objective of this thesis is to study the impacts of interconnecting
“Dabaa’’ as a nuclear power plant to the Egyptian electric grid and determine the
main recommendations for the size of nuclear reactors, the optimized way for
interconnecting the nuclear power plant to the existing network, and monitoring
the behavior of nuclear reactor for several contingency cases in power system to
determine the best procedure of operation of NPP during normal and contingency
cases.
1.2
Literature review
In previous studies, some detailed models for nuclear power plant have
been proposed for power system stability analysis. But some factors that have a
close relationship with the power system are not taken into consideration. For
example, the frequency deviation by the grid disturbance is supposed to be small,
also the coolant pump, rapid closing mechanism, bypass valve control, and
protection system are ignored. However, models in these studies cannot simulate
the dynamic behavior of power system with nuclear power plant under large
disturbance, such as load rejection. In recent studies some of pervious factors are
taken into account, so the model becomes able to simulate the dynamic behavior
of power system with nuclear power plant under large disturbance.
Mohamed M. Megahed; [1] proposed a study to provide the decision-makers
with the necessary information regarding the technical and economical feasibility
and viability of the nuclear option for electricity generation and seawater
desalination. In particular, the study provided an analysis of the Egyptian
economic situation and estimation of future needs for electricity and water, a
technical and economic evaluation of nuclear power generation and desalination
systems, the necessary conditions for launching a successful nuclear power and
2
desalination project, and finally the recommendations on the use of nuclear energy
for electricity generation and seawater desalination.
Aly Karam el-din, Samer Mekhemar; [2] showed a study up to the year of
2025 which considered the population growth, rise in standard of living, and
development plans. A medium-sized nuclear desalination reactor was found to be
a viable choice. This study was made to determine a suitable site for a plant in the
water-scarce coastal governates of Matrouh, Red Sea, North Sinai, and South
Sinai. Considering the effect of extreme climatic and geomorphologic conditions
on the plant and the effect of the plant on the area, it was found that the coastal
strips from El-Arish to Rafah and from El-Dabaa to Saloum are the most suitable
areas for construction of a nuclear power plant.
Huimin. Gao , Chao. Wang, and Wulue; [3] showed the detailed pressurized
water reactor (PWR) nuclear unit model for medium-term and long-term power
system transient stability is proposed. The model is implemented by a user-defined
program in PSS/E through PSS/E Matlab Simulink Interface.
Li Xiong, Dichen Liu; [4] proposed an approach for a new and detailed
pressurized water reactor plant model is presented and developed. The model is
based on the basic physical theorem and implemented by a user-defined program
in power system analysis software package (PSASP). The dynamic response of the
nuclear power plant is studied. Obviously, it can provide not only enormous
reference value for those who operate power systems but also further research on
power system analyze and computation.
3
1.3 Thesis layout
Chapter one “Introduction and review of literature” begins with the
motivation and objective of the thesis then gives a literature review of the main
publications considering the modeling of nuclear power plant and its performance.
Chapter two “Types of nuclear power station” explains the different types of
nuclear power reactors, also the characteristics of each reactor. Finally,
determining the suitable phenomena to select the best type of the reactor to be
connected to the Egyptian electrical grid.
Chapter three “Detailed Construction and Dynamic Modeling of Pressurized
Water Reactor” covers the dynamic model of the several component of the
nuclear reactor.
Chapter four “Interconnecting Dabaa to the Egyptian Unified Power System”
covers the dynamic the model of the Egyptian unified power system to which the
NPP will connect. The dynamic model of the NPP will be interfaced with the
electric power system to investigate the dynamic and steady state response of both
systems after interconnection under several contingency cases. The chapter
includes a study for the Egyptian electrical grid to ensure that it is efficient, safe,
secure and reliable for interconnection with a NPP.
Chapter five “Conclusions and future work” gives the overall conclusions of
these thesis work and suggestion for future work.
4
CHAPTER 2
CLASSIFICATION OF NUCLEAR POWER STATION
2.1 Introduction
Nuclear power reactors used in nuclear plants power for the generation of
electricity are designed, built and operated to produce heat through the fission
chain reaction of 235U and 239Pu, the chain reaction which takes place in the
reactor core being under control and automatically adjusted to supply the desired
power output.
The reactor system provides the heat source which replaces the furnace in a
fossil-fuelled electrical generating station. The remaining conventional part of the
plant consists of a steam/water circuit feeding steam to a turbine driving an
electrical generator. Heat from the reactor is thus transferred by the coolant and
used to generate steam. The steam/water circuit is adjusted to the steam conditions
achievable with a nuclear heat source. The control of the reactor and of the
conventional circuit is carried out from a control room.
Today there are some 439 operable nuclear power reactors in 29 countries,
with a combined capacity of 375,876 MW. In September 2011 these provided
2630 billion kWh, over 13.8% of the world's electricity. Also for reactors under
construction, there are some 63 under construction nuclear power reactors in 14
countries, with a combined capacity of 64,724 MW. [5]
The Evolution of Nuclear Power can be illustrated by figure 2.1 which
shows several reactor “generations.” Generation I corresponds to the research and
prototype reactors of the 1950s and early 1960s. Generation II gathers most of the
existing commercial nuclear power plants, built between the late 1960s and the
early 1990s. Generation III refers to the advanced designs that are now being
5
constructed or about to be licensed, with more passive safety systems. A
distinction is sometimes drawn between Generation III (current designs) and
Generation III+ (designs that should be available in the near-term, within 20
years). Finally, current nuclear research aims to develop a new generation of
reactors, Generation IV, mostly composed of breeder reactors. This technology is
still at the research stage.
Figure 2.1: Evolution of nuclear power reactors [6]
2.2 Basic construction of the reactor
A nuclear reactor produces and controls the release of energy from splitting
the atoms of certain elements. In a nuclear power reactor, the energy released is
used as heat to make steam to generate electricity.
The principles for using nuclear power to produce electricity are the same
for most types of reactor. The energy released from continuous fission of the
atoms of the fuel is harnessed as heat in either a gas or water, and is used to
6
produce steam. The steam is used to drive the turbines which produce electricity
(as in most fossil fuel plants).
In addition to the cooling system servicing the heat transfer, other main
elements of the reactor core are the moderator and the neutron-absorbing materials
(control rods). The moderator has the function of slowing down the neutrons
emitted in the fission process to the thermal energy range at which they are more
effective in producing further fissions to maintain the chain reaction. In the case of
fast reactor types, no moderator is required.
The function of the neutron-absorbing materials, which are either in the
form of movable rods inside the core, or chemical compounds dissolved in the
coolant, is to regulate the fission chain reaction and control the power level of the
reactor. A reflector surrounds the core, to prevent neutrons from escaping from the
core and hence reduce fissile material requirements and improve the power
distribution within the core.
The main elements of the reactor core are assembled according to the
design of the reactor system, inside a tank or a pressure vessel. For safety reasons,
the whole reactor circuit and reactor vessel are enclosed inside a leak tight
containment building which provides a safety barrier against radioactive products
release to the environment. The containment building may also provide protection
to the nuclear reactor against outside hazards.
7
Figure 2.2: Basic construction of Nuclear Steam Supply System (NSSS)
2.3 Classification of the traditional nuclear power reactors
Nuclear Reactors are classified by several methods; a brief outline of these
classification schemes is provided.
8
Figure 2.3:
2 Classification of nuclear power reactors
9
2.3.1 Classification by type of nuclear reaction
According to nuclear reaction the reactor can be classified to:
2.3.1.1 Nuclear fission reactors.
Most reactors, and all commercial ones, are based on nuclear fission. They
generally use uranium and its product plutonium as nuclear fuel, though a thorium
fuel cycle is also possible. Fission reactors can be divided roughly into two
classes, depending on the energy of the neutrons that sustain the fission chain
reaction:
The first class is the thermal reactor which uses slowed or thermal neutrons.
Almost all current reactors are of this type. These contain neutron moderator
materials that slow neutrons until their neutron temperature is thermalized until
their kinetic energy approaches the average kinetic energy of the surrounding
particles. Thermal neutrons have a far higher probability of fissioning the fissile
nuclei uranium-235, plutonium-239, and plutonium-241, and a relatively lower
probability of neutron capture by uranium-238 (U-238) compared to the faster
neutrons that originally result from fission, allowing use of low-enriched uranium
or even natural uranium fuel. The moderator is often also the coolant, usually
water under high pressure to increase the boiling point. These are surrounded by
reactor vessel, instrumentation to monitor and control the reactor, radiation
shielding, and a containment building.
The second class is the fast neutron reactor which uses fast neutrons to
cause fission in their fuel. These reactors do not have a neutron moderator, and use
less-moderating coolants. Maintaining a chain reaction requires the fuel to be more
highly enriched in fissile material (about 20% or more) due to the relatively lower
probability of fission versus capture by U-238. Fast reactors have the potential to
produce less waste because all actinides are fissionable with fast neutrons, but they
10
are more difficult to build and more expensive to operate. Overall, fast reactors are
less common than thermal reactors in most applications. Some early power
stations were fast reactors, as some Russian naval propulsion units.
2.3.1.2 Nuclear fusion reactors.
Fusion power is an experimental technology, generally with hydrogen as
fuel. While not suitable for power production, Farnsworth-Hirsch fusors are used
to produce neutron radiation.
Fusion fuels include tritium (H3) and deuterium (H2) as well as helium-3 (H
e3). Many other elements can be fused together, but the larger electrical charge of
their nuclei means that much higher temperatures are required. Only the fusion of
the lightest elements is seriously considered as a future energy source. Although
the energy density of fusion fuel is even higher than fission fuel, and fusion
reactions sustained for a few minutes have been achieved, utilizing fusion fuel as a
net energy source remains a theoretical possibility.
2.3.2 Classification by moderator material
According to moderator materials the reactors can be classified into four
categories of thermal reactors:
2.3.2.1 Graphite moderated reactors
Both Magnox and Advanced Gas Cooled Reactor (AGR) owe much to the
very earliest reactor designs in that they are graphite moderated and gas cooled.
Magnox reactors were built in the UK from 1956 to 1971 but have now been
superseded. The Magnox reactor is named after the magnesium alloy used to
encase the fuel, which is natural uranium metal. Fuel elements consisting of fuel
rods encased in Magnox cans are loaded into vertical channels in a core
constructed of graphite blocks. Further vertical channels contain control rods
11
(strong neutron absorbers) which can be inserted or withdrawn from the core to
adjust the rate of the fission process and, therefore, the heat output. The whole
assembly is cooled by blowing carbon dioxide gas past the fuel cans, which are
specially designed to enhance heat transfer. The hot gas then converts water to
steam in a steam generator. Early designs used a steel pressure vessel, which was
surrounded by a thick concrete radiation shield.
Figure 2.4: Basic Gas-Cooled Reactor (MAGNOX)
In later designs, a dual-purpose concrete pressure vessel and radiation
shield was used. In order to improve the cost effectiveness of this type of reactor,
it was necessary to go to higher temperatures to achieve higher thermal
efficiencies and higher power densities to reduce capital costs. This entailed
increases in cooling gas pressure and changing from Magnox to stainless steel
cladding and from uranium metal to uranium dioxide fuel. This in turn led to the
need for an increase in the proportion of U235 in the fuel. The resulting design,
known as the AGR, still uses graphite as the moderator and, as in the later Magnox
12
designs, the steam generators and gas circulators are placed within a combined
concrete pressure-vessel/radiation-shield.
Figure 2.5: Advanced Gas-Cooled Reactor (AGR)
2.3.2.2 Water moderated reactors
There two main categories for reactors which use water as a moderator:
• Heavy water reactors.
• Light water reactors.
Heavy water moderated reactors
The only design of heavy water moderated reactor (HWR), in commercial
use is the CANDU, designed in Canada and subsequently exported to several
countries. In the CANDU reactor, unenriched uranium dioxide is held in
zirconium alloy cans loaded into horizontal zirconium alloy tubes. The fuel is
cooled by pumping heavy water through the tubes (under high pressure to prevent
boiling) and then to a steam generator to raise steam from ordinary water (also
known as natural or light water) in the normal way. The necessary additional
moderation is achieved by immersing the zirconium alloy tubes in an
13
unpressurised container (called a callandria) containing more heavy water. Control
is affected by inserting or withdrawing cadmium rods from the callandria. The
whole assembly is contained inside the concrete shield and containment vessel.
Figure 2.6: CANDU Reactor (HWR)
Light water moderated reactors (LWRs)
Light water reactors use ordinary water to moderate and cool the reactors.
When at operating temperature, if the temperature of the water increases, its
density drops, and fewer neutrons passing through it are slowed enough to trigger
further reactions. That negative feedback stabilizes the reaction rate. Graphite and
heavy water reactors tend to be more thoroughly thermalized than light water
reactors. Due to the extra thermalization, these types can use natural uranium
unenriched fuel.
2.3.2.3 Light element moderated reactors.
These reactors are moderated by lithium or beryllium.
14
Molten salt reactors (MSRs) are moderated by a light elements such as
lithium or beryllium, which are constituents of the coolant/fuel matrix salts LiF
and BeF2.
Liquid metal cooled reactors, such as one whose coolant is a mixture of
Lead and Bismuth, may use BeO as a moderator.
2.3.2.4 Organically moderated reactors (OMR)
Use biphenyl and terphenyl as moderator and coolant.
2.3.3 Classification by coolant
According to coolant materials the reactors can be classified into four
categories of thermal reactors:
2.3.3.1 Water cooled reactor
In thermal nuclear reactors, the coolant acts as a moderator that must slow
down the neutrons before they can be efficiently absorbed by the fuel.
2.3.3.1.1 Pressurized water reactor (PWR)
A primary characteristic of PWRs is a pressurizer, a specialized pressure
vessel. Most commercial PWRs and naval reactors use the pressurizer. During
normal operation, a pressurizer is partially filled with water, and a steam bubble is
maintained above it by heating the water with submerged heaters. During normal
operation, the pressurizer is connected to the primary reactor pressure vessel
(RPV) and the pressurizer "bubble" provides an expansion space for changes in
water volume in the reactor. This arrangement also provides a means of pressure
control for the reactor by increasing or decreasing the steam pressure in the
pressurizer using the pressurizer heaters.
15
Figure 2.7: Pressurized water Reactor (PWR)
2.3.3.1.2 Boiling water reactor (BWR)
BWRs are characterized by boiling water around the fuel rods in the lower
portion of a primary reactor pressure vessel. A boiling water reactor uses 235U,
enriched as uranium dioxide, as its fuel. The fuel is assembled into rods that are
submerged in water and housed in a steel vessel. The nuclear fission causes the
water to boil, generating steam. This steam is pumped through pipes into turbines.
The turbines are driven by the steam, and this process generates electricity. During
normal operation, pressure control is accomplished by controlling the amount of
steam flowing from the reactor pressure vessel to the turbine.
16
Figure 2.8: Boiling water Reactor (BWR)
2.3.3.2 Liquid metal cooled reactor
Since water is a moderator, it cannot be used as a coolant in a fast reactor.
Liquid metal coolants have included sodium (Sodium-cooled fast reactor), lead
(Lead-cooled fast reactor), and in early reactors, the mercury was used as coolant.
2.3.3.3 Gas cooled reactors
These reactors are cooled by a circulating inert gas, often helium in hightemperature designs, while carbon dioxide has been used in past British and
French nuclear power plants. Nitrogen has also been used.[citation needed]
Utilization of the heat varies, depending on the reactor. Some reactors run hot
enough that the gas can directly power a gas turbine. Older designs usually run the
gas through a heat exchanger to make steam for a steam turbine.
2.3.3.4 Molten Salt Reactors (MSRs)
These reactors are cooled by circulating a molten salt, typically a eutectic
mixture of fluoride salts. In a typical MSR, the coolant is also used a matrix in
which the fissile material is dissolved.
17
2.3.4 Classification by usage
2.3.4.1 Electric power reactors
Almost all current reactors are used for electric power generation.
2.3.4.2 Desalination and Heating reactors
Where the nuclear reactors may be used for domestic and industrial heating.
In addition to desalination of water.
2.3.4.3 Nuclear propulsion
Where the first generation of prototype reactors were used for marine
propulsion then for rocket propulsion.
2.3.4.4 Production of elements
Fast breeder reactors are capable of producing more fissile materials than
they consume during the fission chain reaction (by converting fertile U-238 to Pu239) which allows an operational fast reactor to generate more fissile material than
it consumes. In addition to Creating various radioactive isotopes, such as a
americium for use in smoke detectors, and cobalt-60, and molybdenum-99 used
for imaging and medical treatment.
2.3.5 Classification by fuel
According to the fuel classification we have two main categories used in
nuclear power reactors may be found in different forms such as:
2.3.5.1 Phase of used fuel
The fuel used in nuclear power reactors may be found in different forms
such as:
18
■ Solid fueled
Where the natural uranium fuel is first formed into ceramic pellets and then
sealed into metal tubes. The tubes are assembled into fuel bundles weighing about
22 kilograms each. One bundle produces the same amount of heat as 400 tonnes of
coal. It is the common fuel for major reactors in service.
■ Fluid fueled
This may be used in aqueous homogeneous reactor and molten salt reactor.
2.3.5.2 Type of refueling system
The reactors may be classified to reactors can be refueled with off-load
condition such as light water reactors and other type must be refueled with on-load
condition such as heavy water reactors.
2.3.5.2.1 Reactors with off-load refueling system
Light-water reactors use enriched uranium and have enough built-in excess
reactivity to enable full-capacity operation for a period varying from 12 to 18
months without refueling. This means that the reactor needs to be shut down at
such time intervals for refueling for a period varying between 4 and 8 weeks
which is named as the refueling outage. Preventive maintenance and in-service
inspection are also done during the refueling outage.
This requires the power system to have enough reserves for backing up the
system during the scheduled NPP down-time period every 12-18 months. After
refueling, the newly added fuel needs 'conditioning', which means gradual increase
of power, at a low rate and under conditions as dictated by the fuel manufacturers.
Reactors with off-load refueling systems are more suitable for quick load
changes over a wide range of power levels.
19
2.3.5.2.2 Reactors with on-load refueling system
Heavy-water reactors use natural uranium and do not have enough built-in
excess reactivity and cannot operate for prolonged periods of time without
refueling, so an on-power fueling system is provided. Fueling is done almost daily
and the unit is not required to be shut down. However, maintenance and inspection
outages are still necessary, but the time of shut-down can be chosen at the best
convenience of the operator and it is generally chosen at periods of low grid
demand.
These reactors must be restarted soon after a trip and brought to nominal
power within one hour. However, if they cannot be restarted within 30-45 minutes
after a reactor trip and quickly loaded to 70% of the pre-trip power level. The
load-follow capability of these reactors is limited to 30-40% of nominal power. [7]
20
2.4 Summary of the main thermal nuclear reactors
Table 2.1: Summary of the main thermal reactors [8]
21
2.5 Future of nuclear reactors
Many different reactor systems have been proposed to be developed to the
so-called advanced reactors where simplicity, safety, economy, and reliability are
taken into consideration in these advanced designs. The various prototype designs
will now be described.
2.5.1 Advanced PWR (AP-1000)
It is an advanced model of standard PWR design with less piping, fewer
valves, less control cabling and reduced seismic building volumes. Also the
construction schedule becomes shorter (About 36 months from first concrete to
fuel loading.) as it depends on the modular manufacturing techniques.
AP 1000 uses the passive safety technique which using only natural forces
such as gravity, natural circulation and compressed gas. Fans, pumps, diesels and
chillers are not required for safety. The passive cooling systems include core
cooling, providing residual heat removal, reactor coolant make-up and safetyinjection, and containment cooling which provides the safety related ultimate heat
sink for the plant. Its operating lifetime is of 60 years with a design plant
availability of more than 90% of its capacity.
22
Figure 2.9: Advanced pressurized water reactor (AP-1000) [9]
2.5.2 Advanced BWR (ABWR)
It is an advanced protype of standard BWR designs with improvements in
efficiency, safety, reliability, and cost effectiveness over previous BWR designs
where there is a 20% reduction in capital cost versus previous BWR designs. Also
the construction schedule becomes shorter (About 39 months from first concrete to
fuel loading.).
ABWR is available with a capacity ranged from 1350 to 1460 MW, with
24-month refueling cycle, and with a Plant availability factor of 87% or greater.
23
Figure 2.10: Advanced boiling water reactor (ABWR) [10]
2.5.3 Very high temperature reactor (VHTR)
It is an advanced reactor which uses a thermal neutron spectrum and a
once-through uranium cycle. The VHTR system has a coolant outlet temperature
above 1000˚C and it uses the helium as a gas coolant for its core. Commonly this
reactor is used for Hydrogen production.
Figure 2.11: Very high temperature reactor (VHTR)
24
2.5.4 Sodium-cooled fast reactor (SFR)
It is an advanced reactor which features a fast neutron spectrum and a
closed fuel cycle. The SFR system has a coolant outlet temperature about 550˚C
and is available with a capacity ranged from 150 to 1500 MW.
Figure 2.12: Sodium-cooled fast reactor (SFR)
2.6 Considerations for selecting the type of nuclear reactor
There are several factors affect the decision criteria for selecting the type of
reactor, these factors will be mentioned according to its importance.
2.6.1 Safety facilities
Safety is the most important criterion in evaluating the reactor type. All
reactors must meet a minimum safety level, but those that are more inherently safe
are more likely to win the technical and public support.
Even in case of a certain trouble in a nuclear power plant, in order to
prevent it from developing to a big accident, safety facilities to safely shutdown
25
the reactor, cool the reactor core, contain radioactive materials and the equipment
that supplies electricity and cooling water to those safety facilities .
(a) Facilities to control and shutdown a reactor: A reactor shall not go out
of control (an excessive power by an abnormal-rate nuclear fission reaction)
during operation. Moreover, in case of a trouble, it is required to immediately stop
nuclear fission reaction. Therefore, reactor control systems to maintain the nuclear
fission reaction rate at a constant level, and if necessary, facilities to shutdown the
reactor by immediately inserting control rods are provided.
(b) Facilities to cool a reactor: In a reactor core, heat is generated even after
shutdown of the reactor in the decay process for radioactive materials to become
more stable nuclei. This is called decay heat. Therefore, it is necessary to continue
cooling of the reactor core not only during power operation but also after
shutdown. Especially, if a break of reactor cooling system piping, etc. causing loss
of cooling water occurred, the reactor core would become dry-out and reach a high
temperature, probably resulting in core damage. In preparation for such a case,
facilities to immediately inject cooling water into the reactor core in an emergency
are provided.
(c) Facilities to contain radioactive materials Nuclear power plants are
provided with the barriers such as a containment etc. in preparation for an accident
releasing radioactive materials from their reactor cores. These barriers will close
immediately in an abnormal event, and constitute an airtight container.
(d) Facilities to support safety facilities: As facilities necessary to support
safety facilities of a nuclear power plant, instrumentation and control, power
distribution system and equipment cooling system are provided.
26
Figure 2.13: Safety facilities [11]
The fundamental for ensuring safety of nuclear power plants is the concept
of "defense in depth." This concept is to prevent a trouble from escalating to a
significant accident by providing physical and functional protective barriers at
each stage. It is the basics for safety design. Nuclear safety must be ensured with
five protective barriers.
(a) The first protective barrier is to prevent occurrence of an anomaly or
failure by providing a nuclear power plant with high quality and reliability and
verified facilities.
(b) The second protective barrier is to provide facilities. to early detect an
occurring anomaly or failure and shutdown the reactor in order to prevent it from
escalating to an accident.
(c) The third protective barrier is to provide facilities such as emergency
core cooling systems and containment, etc. to mitigate the consequence of an
accident in case of a trouble escalation.
(d) The fourth protective barrier is establishment of the severe accident
management such as installation of a containment to contain radioactivity,
preparation of alternative methods with diversities in response to a multiple-failure
27
event, preparation of the operation procedure, and training of operators. An
accident escalated due to multiple failures.
(e) The fifth protective barrier is off-site emergency measures for protection
of residents in the vicinity from radiation exposure when the fourth protective
barrier is broken.
2.6.2 Economics
The final economic situation of the plant cannot be determined until the
detailed design is complete. Only a rudimentary analysis is available prior to a
complete design. The cost of the NPP can divided into:
•
•
•
•
The construction cost of building the plant.
The operating cost of the plant.
The cost of waste disposal from the plant.
The cost of decommissioning the plant.
2.6.2.1 The construction cost of building the plant
Construction costs are very difficult to quantify for nuclear power plant.
The main difficulty is that third generation power plants now proposed are
claimed, with respect to economy-of-volume, to be both substantially cheaper and
faster to construct than the second generation power plants now in operation
throughout the world.
Westinghouse claims its Advanced PWR reactor, the AP1000, will cost
USD $1400 per KW for the first reactor and fall to USD $1000 per KW for
subsequent reactors. They also claim these will be ready for electricity production
3 years after first pouring concrete. [12]
Proponents of the CANDU ACR and Gas Cooled pebble bed reactors make
similar or stronger claims. However the first wave of new plants in the USA is
expected to cost over $3500 per KW of capacity. [13]
28
The General Electric ABWR was the first third generation power plant
approved. The first two ABWR's were commissioned in Japan in 1996 and 1997.
These took just over 3 years to construct. Their construction costs were around
$2000 per KW. [14]
For Chinese Nuclear Power Industry, they build new plants of their own
design at capital costs reported to be ranged between $1300 -$1500 per KW at
sites in South-East and North-East China. [15]
2.6.2.2 The operating cost of the plant
These costs are much easier to quantify and are independently verified as
they relate directly to the profitability of the Utilities which operate them. The
operating cost includes a charge of 0.2 cents per KWH to fund the eventual
disposal of waste from the reactor and for decommissioning the reactor. The price
of Uranium ore contributes approximately 0.05 cents per KWH. [16]
2.6.2.3 The cost of waste disposal from the plant.
Nuclear Power operators in USA are charged 0.1 cents per KWH for the
disposal of nuclear waste. In Sweden this cost is 0.13 US cents per KWH. These
Countries have utilized these funds to pursue research into Geologic disposal of
waste and both now have mature proposals for the task. In France the cost of waste
disposal and decommissioning is estimated to be 10% of the construction cost. So
far provisions of 71 billion Euros have been acquired for this from the sale of
electricity. [17],[18]
29
2.6.2.4 The cost of decommissioning the plant
The average cost for decommissioning a nuclear power plant is USD $300
million in US industry. The funds for this activity are accumulated in the operating
cost of the plant. The French and Swedish Nuclear Industries expect
decommissioning costs to be 10 -15 % of the construction costs and budget this
into the price charged for electricity. On the other hand the British
decommissioning costs have been projected to be around 1 Billion pounds per
reactor. [19]
2.6.3 Other factors
There are a lot of factors but with less importance and may depend on
pervious factors such as public acceptance, government support, construction time,
efficiency, refueling time, availability, modularity, level of waste disposal, and
enrichment of fuel.
2.6.4 Conclusion for selecting the type of reactor in Dabaa NPP
Depending on pervious considerations the advanced pressurized water
reactor will be the most suitable reactor for the first nuclear project in Egypt for
the following reasons:
• It is the common reactor used in all over the world (PWR represents
about 61% of all reactors in operation also major of reactors under
construction are of this type).
• The long experience of operation makes it is easier to ensure safety.
•
Its construction cost is assumed to be the lowest (ranged from 1000
to 1300 $USD per KW).
• Modular manufacturing techniques giving a shorter construction
schedule (about 3 years).
30
• Because of the tremendous experience the industry ensures long life
time for this type of reactor (about 60 years with availability of 90
% of its installed capacity).
• Fission products are contained and not circulated which ensure
more safety in normal and emergency cases.
• Having good load-follow characteristics where the reactor can be
controlled through control rods, soluble boron and coolant
temperature changes.
31
CHAPTER 3
DETAILED CONSTRUCTION AND DYNAMIC
MODELING OF PRESSURIZED WATER REACTOR
3.1 Introduction
Power system dynamic study is one of the most important and common
analyzing technique used to evaluate performance of the system against any
disturbance such as faults, large difference between available generation capacity
and required demand, and successive disconnecting of generating units or
transmission lines which may lead to large excursions in the system voltage and
frequency from the rated values. Power system dynamic, after these disturbances,
is greatly affected by the dynamic characteristics of generating units. Therefore the
accurate dynamic model of NNP is so important as this power plant is relatively
the largest generating unit in the system, where its rating is ranged between 600 to
1100 MW, in addition to its large influence on the stable operation of the power
system. [20]
Some approaches for modeling of nuclear plant dynamics for power system
long-term dynamics have been presented in previous studies. These models are
generally simple where the coolant pump, rapid closing mechanism, bypass valve
control, and protection system are not taken into account.
Simplified plant models traditionally used for short-term stability analysis
are inadequate for power system long-term dynamics analysis. And it is generally
based on the assumption that excursions in power system frequency and voltage
caused by small system disturbances. Thus they are not able to be used to simulate
the dynamic behavior of power system with nuclear power plant under large
disturbances.
32
To develop an adequate nuclear plant model which is able to simulate the
plant performance against any disturbance, considering reactions of the plant
control and protection systems, it is important to be aware of nuclear power plant
components and main function. It is necessary to know which control and
protection systems of the plant components must be modeled or neglected to
accurately represent plant dynamics analysis.
PWR plant can be illustrated by figure 3.1 which shows that the NPP is
composed of nuclear reactor, steam generator, condenser, steam turbine generator
unit and other attached equipment. The basic design objective of a PWR is to
transfer the energy generated in the reactor core to a steam generator where it is
converted to steam to drive a turbine generator. Factors that have close
relationship with the power system should be kept, while the others should be
ignored or simplified.
Figure 3.1: Modeled components in PWR power plant [21]
33
3.2 Basic construction of PWR
The first step for modeling the PWR system is knowing its main
components and their function. The PWR is consists of two major systems which
are utilized to convert the heat generated in the fuel into electrical power in
addition to the auxiliary systems. The first main system is the primary system
which transfers the heat from the fuel to the steam generator, and the second is the
secondary system which transfers the steam formed in the steam generator to the
main turbine generator, where it is converted into electricity. After passing
through the low pressure turbine, the steam is routed to the main condenser. The
cooling water which flows through the tubes in the condenser is used to remove
excess heat from the steam, which allows the steam to condense. Then the water is
pumped back to the steam generator for reuse.
Figure 3.2: Basic components of PWR
34
3.2.1 Primary system of PWR
The primary system consists of the reactor vessel, the steam generator, the
reactor coolant pump, the pressurizer, and the connecting piping. The function of
the reactor coolant system is to transfer the heat from the fuel to the steam
generators. A second function is to contain any fission products that escape the
fuel. The function of each component will be shown as follow:
Figure 3.3: Primary system of PWR [22]
1) Reactor vessel
The reactor core, all associated support and alignment devices are housed
within the reactor vessel. The reactor vessel is a cylindrical vessel with a
hemispherical bottom head and a removable hemispherical top head. The top head
35
is removable to allow for the refueling of the reactor. There will be one inlet
nozzle which is called cold leg nozzle and one outlet nozzle which is called hot leg
nozzle for each reactor coolant system loop. The reactor vessel is constructed of
manganese molybdenum steel, and all surfaces that come into contact with reactor
coolant are clad with stainless steel to increase corrosion resistance. The core
barrel slides down inside of the reactor vessel and houses the fuel.
2) Steam generator
The reactor coolant flows from the reactor to the steam generator. Inside of
the steam generator, the hot reactor coolant flows inside of the many tubes. The
secondary coolant flows around the outside of the tubes, where it picks up heat
from the primary coolant. When the secondary coolant absorbs sufficient heat, it
starts to boil and form steam.
3) Reactor Coolant Pump
The purpose of the reactor coolant pump is to provide forced primary
coolant flow to remove the amount of heat being generated by the fission process.
Even without a pump, there would be natural circulation flow through the reactor.
However, this flow is not sufficient to remove the heat being generated when the
reactor is at full-power. The used motor is a large, air cooled, electric motor. Its
horsepower rating will be from 6,000 to 10,000 HP. This large amount of power is
needed in order to provide the necessary flow of coolant for heat removal.
4) Pressurizer
The main function of the pressurizer is providing a means of controlling the
system pressure. Pressure is controlled by the use of electrical heaters, pressurizer
spray, power operated relief valves, and safety valves.
36
3.2.2 Secondary system of PWR
The secondary system of PWR consists of the main steam system and the
condensate/feed-water system. Since the primary and secondary systems are
physically separated from each other, by the steam generator tubes, the secondary
system will contain little or no radioactive material.
Figure 3.4: Schematic diagram of the secondary system [22]
The main steam system starts at the outlet of the steam generator. The steam
is routed to the high pressure main turbine. After passing through the high pressure
turbine, the steam is piped to the moisture separator/reheaters (MS/R). In the
MS/R the steam is dried with moisture separators and reheated using other steam
as a heat source. From the MS/R, the steam goes to the low pressure turbines.
After passing through the low pressure turbines, the steam goes to the main
condenser, which is operated at a vacuum to allow for the greatest removal of
37
energy by the low pressure turbines. The steam is condensed into water by the
flow of circulating water through the condenser tubes.
3.2.3 The auxiliary systems of PWR
1) The chemical and volume control system
The chemical and volume control system (CVCS) is a major support system
for the reactor coolant system. The main functions of the system are:
• Purifying the reactor coolant system using filters and demineralizers,
• Adding and removing boron as necessary, and
• Maintaining the level of the pressurizer at the desired set-point.
Figure 3.5: Chemical and volume control system [22]
38
2) The decay heat removal system
During normal operation, the heat produced by the fission process is
removed by the reactor coolant and transferred to the secondary coolant in the
steam generators. Here, the secondary coolant is boiled into steam and sent to the
main turbine.
Even after the reactor has been shutdown, there is a significant amount of
heat produced by the decay of fission products (decay heat). The amount of heat
produced by decay heat is sufficient to cause fuel damage if not removed.
Therefore, systems must be designed and installed in the plant to remove the decay
from the core and transfer that heat to the environment, even in a shutdown plant
condition. Also, if it is desired to perform maintenance on reactor coolant system
components, the temperature and pressure of the reactor coolant system must be
reduced low enough to allow personnel access to the equipment.
The auxiliary feed-water system and the steam dump system are work
together to allow the operators to remove the decay heat from the reactor. The
auxiliary feed-water system pumps water from the condensate storage tank to the
steam generators. This water is allowed to boil to make steam. The steam can then
be dumped to the main condenser through the steam dump valves. The circulating
water will then condense the steam and take the heat to the environment.
If the steam dump system is not available the steam can be dumped directly
to the atmosphere through the atmospheric relief valves (as the secondary system
of PWR does not contain any radioactive material). By using either method of
steam removal, the heat is being removed from the reactor coolant system, and the
temperature of the reactor coolant system can be reduced to the desired level.
39
Figure 3.6: Decay heat removal system [22]
3) The residual heat removal system
At some point, the decay heat being produced will not be sufficient to
generate enough steam in the steam generators to continue the cooldown. When
the reactor coolant system pressure and temperature have been reduced to within
the operational limits, the residual heat removal system (RHR) will be used to
continue the cooldown by removing heat from the core and transferring it to the
environment. This is accomplished by routing some of the reactor coolant through
the residual heat removal system heat exchanger, which is cooled by the
component cooling water system (CCW). The heat removed by the CCW system
is then transferred to the service water system in the component cooling water heat
exchanger. The heat picked up by the service water system will be transferred
directly to the environment from the service water system.
40
Figure 3.7: Residual heat removal system [22]
4) Emergency core cooling systems
There are two purposes of the emergency core cooling systems (ECCS).
The first is to provide core cooling to minimize fuel damage following a loss of
coolant accident. This is accomplished by the injection of large amounts of cool,
borated water into the reactor coolant system. The second is to provide extra
neutron poisons to ensure the reactor remains shutdown following the cooldown
associated with a main steam line rupture, which is accomplished by the use of the
same borated water source. This water source is called the refueling water storage
tank (RWST).
41
To perform this function of injection of large quantities of borated water,
the emergency core cooling systems consist of four separate systems shown. In
order of highest pressure to lowest pressure, these systems are: the high pressure
injection system, the intermediate pressure injection system, the cold leg
accumulators, and the low pressure injection system (residual heat removal). Even
though the diagram shows only one pump in each system, there are actually two,
each of which is capable of providing sufficient flow. Also, these systems must be
able to operate when the normal supply of power is lost to the plant. For this
reason, these systems are powered from the plant emergency (diesel generators)
power system.
The high pressure injection system uses the pumps in the chemical and
volume control system. Upon receipt of an emergency actuation signal, the system
will automatically realign to take water from the refueling water storage tank and
pump it into the reactor coolant system. The high pressure injection system is
designed to provide water to the core during emergencies in which reactor coolant
system pressure remains relatively high (such as small break in the reactor coolant
system, steam break accidents, and leaks of reactor coolant through a steam
generator tube to the secondary side).
The intermediate pressure injection system is also designed for emergencies
in which the primary pressure stays relatively high, such as small to intermediate
size primary breaks. Upon an emergency start signal, the pumps will take water
from the refueling water storage tank and pump it into the reactor coolant system.
The cold leg accumulators do not require electrical power to operate. These
tanks contain large amounts of borated water with a pressurized nitrogen gas
bubble in the top. If the pressure of the primary system drops below low enough,
the nitrogen will force the borated water out of the tank and into the reactor
coolant system. These tanks are designed to provide water to the reactor coolant
42
system during emergencies in which the pressure of the primary drops very
rapidly, such as large primary breaks.
Figure 3.8: Emergency core cooling system [22]
43
3.3 Mathematical model of proposed NPP
In this study, a detailed nuclear machine model is proposed for a PWR in
H.B.Robinson NPP. It includes the reactor neutron dynamics model, steam
generator dynamics model, core fuel and coolant temperature mode, the hot leg
and cold leg temperature model and reactor control system model. The following
figure (3.9) describes the main components of the proposed NPP
Figure 3.9: Mathematical model of PWR power plant [23]
3.3.1 Neutron Dynamics Model
The reactor neutron dynamic behavior is modeled by six groups of
equivalent delayed neutrons. And six groups of delayed neutrons are simplified to
a group of equivalent delayed neutrons. The neutron dynamic equations of reactor
core are as follows
44
d ∆ N (t )
∆ ρ (t ) − β
=
∆ N (t ) + λ ∆ C (t )
Λ
dt
(3.1)
d ∆ C (t )
β
=
∆ N (t ) − λ ∆ C (t )
dt
Λ
(3.2)
∆ ρ ( t ) = β ∗ ∆ ρ ext ( t ) + σ
∆T
f
f
+σ
(∆ T c 1 +
c
∆T
c2
)
2
(3.3)
3.3.2 The Steam Generator Model
The steam generator model is also a lumped parameter model, in which the
heat transfer processes from the primary coolant loop to the U-type metal tube and
then to the secondary loop fluid are mainly considered. The primary loop coolant
temperature, U-type metal tube temperature, and the steam pressure model at the
outlet of secondary loop model are shown as follows respectively.
∆T
∆T
∆P
p
=
m
s
τ
=
=
1
[ ∗
S + 1 k1 ∆T
p
τ
τ
HL
+
k ∗ ∆T
2
m
]
(3.4)
∆T
p
+
k
4
∗ ∆ P s]
(3.5)
m
1
[
∗
S +1 k3
∆T
m
+
k
6
∗
∆µ
(3.6)
ps
1
[
∗
S +1 k5
]
Where K1, K2, K3, K4, K5, and K6 are constants depend on the steam generator.
3.3.3 Core Fuel and Coolant Temperature Model
The energy which nuclear fuel fission produced in the core cause fuel
temperature to increase which transfer to coolant. Coolant temperature increase as
well. Energy balance equations are shown in (3.7), (3.8) and (3.9). The lumped
parameter model is adopted for core fuel and coolant temperature model.
45
d∆ T
dt
d∆T
dt
d∆T
dt
=
f
1
m c
F
mc
C
c2
=
2
( ∆ T c1 + ∆ T c 2 − ∆ T f )]
PC
1
mc
C
PF
hA
(3.7)
′
m
(3.8)
′
m
(3.9)
[(1 − Γ) Po ∆N (t ) + hA(∆ T f − ∆ T c1)] + m c (∆ T CL − ∆ T c1)
1
=
c1
[ Γ P o ∆ N (t ) +
c
[(1 − Γ) Po ∆N (t ) + hA(∆T f − ∆T c1)] + m c (∆T c1 − ∆T c 2)
PC
c
3.3.4 The Hot Leg and Cold Leg Temperature Model
According to heat balance equation, both the hot leg and cold temperature
are expressed by first order inertial element, as is shown
∆T
HL
=
∆T
CL
=
τ
1
S + 1 ∆T c2
HL
(3.10)
τ
1
S + 1 ∆T
CL
(3.11)
P
3.3.5 The coolant pump model
The reactor coolant flow rate is one of the most important variables for
the PWR plant dynamics under the power system disturbances and so that the
model for the reactor coolant pump must take in consideration both frequency
and voltage disturbances. The coolant pump may be modeled as follows
46
Figure 3.10: Typical model of the reactor coolant pump [24]
dωp
dt
M
= Me − Mm
e =
k



e1 



M m =ω p
(3.12)
V
F

1 +


2

1 −


ω
2
F k
e2
F
p





1 −


ω
p
F
(3.13)
2
(3.14)
ωp
.
m cn ω pn
mc








.
=
(3.15)
47
3.3.6 The steam turbine model
The steam turbine which is used in the NPP, is approximately as the steam
turbine used in fossil-fuelled plants but with some differences where commonly
there is not intermediate pressure turbine section in the turbine used in the NPP,
also its typical type is the single reheat tandem-compound as follows:
Figure 3.11: Typical model of steam turbine used in NPP [25]
The fraction of total turbine power generated by intermediate pressure
section equals zero (FIP = 0) and the other typical values of the parameters are as
follow: [24]
48
3.3.6 The reactor control system model
Reactor power can be controlled via two methods: by inserting or
withdrawing control rods and by changing the water flow through the reactor core.
In our study the reactor control system is a temperature regulating system with
average temperature as main regulating value. There are three input signals to
determine the temperature error in temperature regulating system where the
difference between the setting value and the measured value of coolant average
temperature will actuate the power control rod to move at a limited speed until the
difference is less than a dead band. Power compensation channel is included in
regulation system to improve regulating system response speed and system
stability.
Figure 3.12: Typical control system of reactor [26]
The parameters of the reactor control system are as follow:
τ1: average temperature measurement time constant, s;
τ2, τ3: lag time constant of average temperature lead compensation, s;
τ4: leading time of average temperature lead compensation, s;
τ5: mechanical power measurement time constant, s;
49
τ6: lag time constant of average temperature fixed value, s;
τ7: nuclear power measurement time constant, s;
τ8: lag time constant of equivalence based on in equal power temperature, s;
K: the reactivity deviation.
3.3.7 The reactor protection system
The reactor protection system monitors the plant variables closely related
to the integrity of the reactor such as the neutron flux, coolant pressure,
coolant temperature and coolant level. The following major plant variables
which are represented in the plant model.
Figure 3.13: Typical protection system of reactor [27]
50
CHAPTER 4
INTERCONNECTING DABAA TO THE EGYPTIAN
UNIFIED POWER SYSTEM
4.1 Introduction
The performance of the NPP (availability, life time, required maintenance) is
affected by the performance of the electric power system which the NPP will be
connected to, and the frequent grid disturbances and in this context a study of the
dynamic interaction between grid and NPP becomes extremely important.
Safe start-up, running and shut-down of NPPs require adequate
characteristics of the external grid power supply, since it should be possible to
supply the electric power system (EPS) of the NPP with electric power from the
main electric generator and from the transmission grid. In the case of loss of
power from the generator under operational or accident conditions the EPS should
be supplied by the grid. When determining reliability criteria for reactor safety, the
probability of a certain number of grid power failures per year is assumed. If,
however, the number of grid power failures per year is higher, the reliability
criteria will be adversely affected.
Careful evaluations for the electric power system should lead to a concept of
the proper NPP design, the possible addition of equipment and those system
operation strategies which will most economically ensure the safe operation of the
NPP. Sothat this chapter studies the performance of the Egyptian electrical power
system.
Generally in order to include the nuclear power option in a realistic manner, the
country should have an adequate total installed power capacity, reliable
interconnected by transmission system, of over 7–10 times the unit size of the
51
NPP. The size of bulk transmission network, the extent of interconnection and the
need for strengthening the system stability are important considerations in the
assessment.
4.2 Studying the Egyptian unified power system
In this study, according to the available data of the Egyptian unified power
system, the objective will be determining the optimum design of the NPP (i.e.: the
unit size, mode of operation, strategies for safe start-up and shut-down, and etc.),
the method and configuration of transmission system between the NPP and the
grid, and the recommendation for dealing with the disturbances in the power
system that may affect the operation of the NPP.
4.2.1 Overview about the Egyptian unified power system
The first installation of electric generation in Egypt was in 1893 with small
diesel engines which supply some lighting loads in houses in Cairo and Alexandria
with low voltage direct current. Then the steam engine power stations were
introduced in 1928 in several regions to supply the drainage pumps in Delta. At
the end of 1959 the total installed capacity reached 511 MW. In 1960, the hydro
power plant of Aswan Dam was commissioned with an installed capacity of 345
MW then in 1970; the High Dam hydro power plant was completed with a total
installed capacity of 2100 MW. During the period 1970-1980 the development in
the system was relatively limited. During the period 1980-2002 an additional
capacity of 11024 MW was added to the unified grid to reach 16648 MW installed
capacity. For the period from 2002 until now a lot of additional capacities have
been added to meet the continuous demand on electric energy. The total installed
capacity reaches 21000 MW by 2011. [28]
52
The Egyptian unified power system can be divided into five main zones
(upper-Egypt, Suez-canal, Delta, Cairo, and Alex.) with various voltage levels for
transmission (66, 132, 220, 500 KV). For the existing Egyptian grid the
distribution of annual electric generated energy according to type of generation is
as follow:
Figure 4.1: The distribution of annual electric generated energy in percent [28]
4.2.2 Overview about the analysis software package used in this
study (PSAT)
Power System Analysis Toolbox (PSAT) is a MATLAB toolbox for electric
power system analysis and control. PSAT includes power flow, continuation
power flow, optimal power flow, small signal stability analysis, and time domain
simulation. All operations can be assessed by means of graphical user interface
(GUT) and a Simulink-based library provides user friendly tool for network
design. PSAT core is the power flow routine, which also takes care of state
variable initialization. Once the power flow has been solved, further static and/or
dynamic analysis can be performed. These routines are:
53
1. Continuation power flow
2. Optimal power flow
3. Small signal stability analysis
4. Time domain simulations
5. Phasor measurement unit (PMU) placement
In order to perform accurate power system analysis, PSAT supports a variety
of static and dynamic component models, as follows:
•
Power Flow Data: Bus bars, transmission lines and transformers, slack
buses, PV generators, constant power loads, and shunt admittances.
•
CPF and OPF Data: Bids and limits of power supply, generator power
reserves, generator ramping data, and power demand bids and limits.
•
Switching Operations: Transmission line faults and transmission line
breakers.
•
Measurements: Bus frequency and phasor measurement units (PMU).
•
Loads: Voltage dependent loads, frequency dependent loads, ZIP
(impedance, constant current and constant power) loads, exponential
recovery loads, thermostatically controlled loads and mixed loads.
•
Machines: Synchronous machines (dynamic order from 2 to 8) and
induction motors (dynamic order from 1 to 5).
•
Controls: Turbine Governors, Automatic Voltage Regulators, Power
System
Stabilizer,
Over-excitation
limiters,
Secondary
Voltage
Regulation (Central Area Controllers and Cluster Controllers), and a
Supplementary Stabilizing Control Loop for SVCs.
54
•
Regulating Transformers: Load tap changer with voltage or reactive
power regulators and phase shifting transformers.
•
FACTS: Static Var Compensators, Thyristor Controlled Series
Capacitors, Static Synchronous Source Series Compensators, Unified
Power Flow Controllers, and High Voltage DC transmission systems.
•
Wind Turbines: Wind models, Constant speed wind turbine with
squirrel cage induction motor, variable speed wind turbine with doubly
fed induction generator, and variable speed wind turbine with direct drive
synchronous generator.
•
Other Models: Synchronous machine dynamic shaft, sub-synchronous
resonance model, and Solid Oxide Fuel Cell. [29]
4.2.3 Size selection for units of Dabaa NPP
The high safety standards to which NPPs are licensed for operation require
complex engineered safety systems and reliable auxiliary systems which are
unprecedented in units of conventional types. The investments associated with
these systems penalize the capital costs of NPPs as compared with fossil-fuelled
plants of the same capacity. The economy of scale has consequently a major
impact on NPPs and this is the reason why NPPs have developed in a range of
rapidly increasing unit size. At present, commercially available designs of Nuclear
Steam Supply System (NSSS) may still be too large for the electric grid of many
developing countries which may not be able to ensure system stability when the
NPP is not available. At present, the only proven NPP types commercially
available for export include PWRs, PHWRs, and BWRs. The minimum size range,
55
at which these NSSS are manufactured, may be ranged between 600-1000 MW.
[30]
Nuclear power plants have the lowest marginal fuel cost of all types of
power stations other than run-of-river hydro power plants and their continuous
operation at nominal power is therefore the utility's first choice for economic
electricity generation. However, the size of the power system in some developing
countries may be so limited that its off-peak load demand is too low to permit
constant load operation of the NPP. In this case, the need of providing some loadfollow capability will of necessity add to the plant cost because of additional
design and engineering complexity as well as the required instrumentation and
degree of automation in the plant control system.
In conclusion, the size selection should be a factor in all implications
associated with making NPP operation viable in a low-performance system.
Therefore, the following points should be carefully considered:
• Cost of extensive NPP engineering, such as load-follow capabilities,
additional equipment, adequate instrumentation and control system, and
effective protection system to withstand transient conditions from the
grid, to ensure adequate performance and to guarantee the designed
plant life.
• Cost of meeting the increased reliability requirements for the on-site
emergency power supply to ensure the performance of the essential
safety functions of the NPP.
• Cost of maintaining grid stability when the NPP is not available. This is
comprised of costs for providing additional spinning reserve,
56
establishing effective system generation control, enhancing the
performance of the grid protection system, and reinforcing the
transmission system.
With respect to these factors the unit size of Dabaa NPP will be 1000 MW
where the first unit of Dabaa nuclear power station is planned to be in operation
during 2014-2015. After the first unit other two units will be constructed, and the
total power generation of Dabaa plant will reach 3,000 MW at the final of period
of this study (2020) as is shown in figure (4.2)
Installed Power
3500
3000
2500
2000
1500
1000
500
0
2014-2015
2017-2018
2020-2021
Years
Figure 4.2: The installed capacity growth of Dabaa NPP
4.2.4 Proposed model of the Egyptian unified power system
This model takes in consideration the detailed configuration of the 500 KV
of Egyptian grid and in brief the 220 KV system around the suggested location of
the first Egyptian NPP for the period from 2010 to 2020. The model of different
components in the power system study is simple with lowest details (i.e. the model
57
of generator, turbine, and governor system is third or fourth order). The suggested
way for interconnecting between Dabaa NPP and the Egyptian grid is as follow:
For the 500KV lines:
1-from Dabaa to Saloom
&
2-from Dabaa to Sidi-krir
&
2- from Dabaa to Matroah.
For the 220KV lines:
1- from Dabaa to Omeed
This scenario of interconnecting depends on three concepts, the first one is
the reliability where the interconnection will be on two levels 220KV and 500KV
with double circuit on 500KV lines, the second concept is the security of operation
of NPP where this scenario takes into consideration the requirements of off-site
supply that feeds the utilities of NPP in case of any contingency for the NPP or its
on-site supply for safe shutdown of reactor, and the third concept is related to the
economics of these infrastructure with respect to other concepts. For the 220KV
system, there is an existing double circuit line between Omeed and Matroah. So it
is of no use to build new lines, just one of the two circuits will be opened at the
nearest point to Dabaa and connected to Dabaa station.
(kv)
No. of
cts.
Length
(km)
R pu/km
X pu/km
B pu/km
Saloum
500
2
325
1.00E-05
1.23E-04
9.10E-03
Dabaa
KRIR.PS
500
2
122
1.00E-05
1.23E-04
9.10E-03
T.L3
Dabaa
Omid
220
1
71
2.77E-04
2.143E-03
4.798E-03
T.L4
Dabaa
Matroah
220
1
111
1.77E-04
1.367E-03
3.060E-03
Line code
From
TO
T.L1
Dabaa
T.L2
Voltage
Table 4.1: Data of new lines which connect Dabaa NPP with the grid
58
Figure 4.3: The interconnection between Dabaa and unified Egyptian grid
59
4.3 Cases of contingency
Here the power system will be subjected to some disturbances to monitor
and analyze the behavior of the NPP under these disturbances and thus, the impact
of this behavior on the performance of the power system where some disturbances
may cause one or more units in the NPP to be dropped, the following assumptions
are taken into consideration during applying these faults:
1) Faults will be three-phase short circuit which may be applied to different buses
in and around the NPP.
2) The fault duration is selected to be three cycles (0.06sec) and it applied to the
system after the system initialized by one second.
3) The fault clearing method is followed by disconnection of the line connected to
the faulted bus.
4) For double circuit transmission line which is disconnected, one circuit of them
is assumed to be in maintenance.
5) The mode of operation of NPP is assumed to be constant-load operation.
4.3.1 Fault at bus of Dabaa-500 accompanied by outage of
double circuit line T.L1
A three-phase short circuit fault is applied to the bus of Dabaa-500 and
accompanied by outage for the double circuit line (T.L1) between Dabaa and
Saloum. The assumption of this case that there is a time lag between the circuit
breakers of T.L1 and that of T.L2 and sothat there is an outage for T.L1 only. It is
noticed from the following figures (from figure 4.4 to figure 4.12) that the impacts
60
of this fault on the performance of the units of NPP are limited where the
deviation in the steam pressure of the reactor is about 40 psi which represents
about 1.78 % of the nominal primary system pressure which is assumed to be a
small value. Also the change in the primary and tube metal temperatures is very
small. Hence, this case of contingency has a limited impact on performance of
Dabaa units.
1
Delta of Dabaa generation unit
0.9
0.8
0.7
0.6
0.5
0.4
0
5
10
time (s)
15
20
Figure 4.4: Delta of Dabaa gen. unit for fault at Dabaa 500 and line (T.L1) outage
61
Frequency of Dabaagenerationunit (P.U)
1.008
1.006
1.004
1.002
1
0.998
0.996
0.994
0
5
10
time (s)
15
20
Figure 4.5: Frequency of Dabaa gen. unit for fault at Dabaa-500 and line (T.L1) outage
1.4
VDABAA 500
1.2
VDABBA1
VSALOOM500
1
0.8
0.6
0.4
0.2
0
0
5
10
time (s)
15
Figure 4.6: Voltages buses for fault at Dabaa 500 and line (T.L1) outage
62
20
0
Change in steam pressure in (psi)
-5
-10
-15
-20
-25
-30
-35
-40
-45
0
10
20
30
40
50
Time (sec)
60
70
80
90
100
Figure 4.7: Change of Steam pressure of Dabaa PWR for fault at Dabaa 500 and line (L1) outage
0
Change in Primary Temperature in (F°)
-0.5
-1
-1.5
-2
-2.5
-3
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.8: Change of Primary temperature of Dabaa PWR for fault at Dabaa 500 and L1 outage
63
0
Change in Tube Metal Temperature in (F°)
-0.5
-1
-1.5
-2
-2.5
-3
-3.5
-4
-4.5
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.9: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L1 outage
6
x 10
-3
Change in Fuel Temperature in (F°)
5
4
3
2
1
0
-1
-2
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.10: Change of fuel temperature of Dabaa PWR for fault at Dabaa 500 and L1 outage
64
Change in Hot-leg Temperature in (F°)
x 10
-4
3
2
1
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.11: Change of hot-leg temperature of Dabaa PWR for fault at Dabaa 500 and L1 outage
Change in Cold-leg Temperature in (F°)
x 10
-4
3
2
1
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.12: Change of cold-leg temperature of Dabaa PWR for fault at Dabaa 500 and L1 outage
65
4.3.2 Fault at bus of Dabaa-500 accompanied by outage of
double circuit line T.L2
A three-phase short circuit fault is applied to the bus of Dabaa-500 and
accompanied by outage for the double circuit line (T.L2) between Dabaa and Sidikrir. The assumption of this case that there is a time lag between the circuit
breakers of T.L2 and that of T.L1 and sothat there is an outage for T.L2 only. It is
noticed from the following figures (from figure 4.13 to figure 4. 21) that the impacts
of this fault on the performance of the power system is very bad where the change
in the frequency of Dabaa generation unit from the nominal value exceed 40 %
also the drop in the voltage of Dabaa generation unit exceed 45 % ,and these
deviation will result in a decrease in coolant flow through the reactor vessel and
decrease in feed water flow through the steam generator and sothat the deviation
in the steam pressure of the reactor is about 240 psi which represents about 10.67
% of nominal primary system pressure which is assumed to be a large value. Also
this fault is accompanied by an increase in the primary and tube metal
temperatures.
66
1000
Deltaof Dabaagen. unit
800
δSyn 1
600
400
200
0
-200
0
5
10
time (s)
15
20
Figure 4.13: Delta of Dabaa gen. unit for fault at Dabaa 500 and line (T.L2) outage
Frequencyof D
abaagen. unit (P.U
)
1.5
1.4
1.3
1.2
1.1
1
0.9
0
5
10
time (s)
15
20
Figure 4.14: Frequency of Dabaa gen. unit for fault at Dabaa-500 and line (T.L2) outage
67
1.4
VDABAA 500
1.2
VDABBA1
VSIDI-KRIR 500
1
0.8
0.6
0.4
0.2
0
0
5
10
time (s)
15
20
Figure 4.15: Voltage buses for fault at Dabaa 500 and line (T.L2) outage
Change in steam pressure in (psi)
250
200
150
100
50
0
0
10
20
30
40
50
Time (sec)
60
70
80
90
100
Figure 4.16: Change of Steam pressure of Dabaa PWR for fault at Dabaa 500 and line(L2) outage
68
18
16
Change of Primary Temperature in (F°)
14
12
10
8
6
4
2
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.17: Change of Primary temperature of Dabaa PWR for fault at Dabaa 500 and L2 outage
25
Change in Tube Metal Temperature in (F°)
20
15
10
5
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.18: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L2 outage
69
14
x 10
-4
12
Change in Fuel Temperature in (F°)
10
8
6
4
2
0
-2
-4
-6
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.19: Change of fuel temperature of Dabaa PWR for fault at Dabaa 500 and L2 outage
9
x 10
-5
8
Change in Hot-Leg Temperature in (F°)
7
6
5
4
3
2
1
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.20: Change of Hot-leg temperature of Dabaa PWR for fault at Dabaa 500 and L2 outage
70
9
x 10
-5
Change in Cold-Leg Temperature in (F°)
8
7
6
5
4
3
2
1
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.21: Change of Cold-leg temperature of Dabaa PWR for fault at Dabaa 500 and L2 outage
4.3.3 Fault at bus of Dabaa-500 accompanied by outage of two lines
T.L1 and T.L2
A three-phase short circuit fault is applied to the bus of Dabaa-500 and
accompanied by outage for the two lines (T.L1 & T.L2) which connect Dabaa with
Sidi-krir and Saloum. The assumption of this case that there is no time lag
between the circuit breakers of T.L1 and that of T.L2 and sothat there is an outage
for both T.L1 and T.L2. It is noticed from the following figures (from figure 4.22 to
figure 4.27) that the impacts of this fault on the performance of the units of NPP are
limited where the deviation in the steam pressure of the reactor is about 50 psi
which represents about 2.22 % of the nominal primary system pressure which is
71
assumed to be a small value. Also the change in the primary and tube metal
temperatures is very small. Hence, this case of contingency has a limited impact
on
the
3000
performance
of
the
units
of
Dabaa
NPP.
δSyn 1
D
eltaof D
abaagen. unit (P.U
)
2500
2000
1500
1000
500
0
0
5
10
time (s)
15
20
Figure 4.22: Delta of Dabaa gen. unit for fault at Dabaa 500 and lines (L1 & L2) outage
Frequencyof D
abaagen. unit (P.U
)
2
1.8
1.6
1.4
1.2
1
0.8
0
5
10
time (s)
15
20
Figure 4.23: Frequency of Dabaa unit for fault at Dabaa-500 and lines (L1 & L2) outage
72
1.4
VDABAA 500
VSALOOM 500
VSIDI-KRIR 500
1.2
1
0.8
0.6
0.4
0.2
0
0
5
10
time (s)
15
20
Figure 4.24A: Voltage buses for fault at Dabaa 500 and lines (TL1& TL2) outage
1.1
VDABBA1
1
0.9
0.8
0.7
0.6
0.5
0.4
0
5
10
time (s)
15
20
Figure 4.24B: Voltage buses for fault at Dabaa 500 and lines (TL1& TL2) outage
73
50
Change of steam pressure in (psi)
45
40
35
30
25
20
15
10
5
0
0
10
20
30
40
50
Time (sec)
60
70
80
90
100
Figure 4.25: Change of Steam pressure of Dabaa PWR for fault at Dabaa 500 and L2&L1 outage
4.5
Change in Primary Temperature in (F°)
4
3.5
3
2.5
2
1.5
1
0.5
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.26: Change of Primary temperature of Dabaa PWR for fault at Dabaa 500 and L2&L1 outage
74
6
Change in Tube Metal Temperature in (F)
5
4
3
2
1
0
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.27: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and L2&L1 outage
4.3.4 Fault at bus of Dabaa-500 accompanied by islanding of NPP
A three-phase short circuit fault is applied to the bus of Dabaa-500 and
accompanied by outage for the line (T.L1 & T.L2) which connect Dabaa-500 with
Saloum and Sidi-krir respectively also the protection system of step-down
transformer between two buses Dabaa-500 and Dabaa-220 is tripped. Under
disturbed grid conditions, keeping a station or unit connected to the grid may
result in outage or damage of costly and vital NPP equipment. Thus, a scheme for
isolating the station or unit from the grid is often followed as an ultimate strategy.
After separation, the station auxiliaries continue to be supplied by the unit at good
voltage and frequency, and the NPP output is reduced by the automatic power setback system to predetermined values compatible with the local loads and the
capacity of other generating units in the isolated system. The unit/station thus
75
'islanded' from the grid can be re-connected when the fault is isolated and normal
conditions are restored in the grid, and the unit/station can then be progressively
re-loaded to nominal power.
The limits of frequency and voltage at which the unit may be islanded
should be very carefully chosen. If they are too conservative, there would be
frequent disturbances in the grid because of loss of generation; if they are too
liberal, the NPP may trip before islanding or important equipment may be
damaged. It is noticed from the following figures (from figure 4.28 to figure 4.35)
that the impacts of this fault on the performance of the power system is very bad
where the change in the frequency of Dabaa generation unit from the nominal
value exceed 100 % also the increase in the voltage of Dabaa generation unit
exceed 35 % ,and these deviation will result in a decrease in coolant flow through
the reactor vessel and decrease in feed water flow through the steam generator and
sothat the deviation in the steam pressure of the reactor is about 190 psi which
represents about 8.44 % of nominal primary system pressure which is assumed to
be a large value.
76
3500
δDabaa unit
3000
2000
1500
1000
500
0
0
5
10
time (s)
15
20
Figure 4.28: Delta of Dabaa gen. unit for fault at Dabaa 500 and Islanding of NPP
2.2
2
Frequencyof D
abaaunit (P.U
)
Deltaof Dabaaunit
2500
1.8
1.6
1.4
1.2
1
0.8
0
5
10
time (s)
15
Figure 4.29: Frequency of Dabaa unit for fault at Dabaa-500 and Islanding of NPP
77
20
Bus voltage of Dabaaunit (P.U)
1.4
VDABBA1
1.2
1
0.8
0.6
0.4
0
5
10
time (s)
15
20
Figure 4.30: Voltage of Dabaa unit for fault at Dabaa-500 and Islanding of NPP
1.4
1.2
1
0.8
VDABAA 500
0.6
VSALOOM500
VSIDI-KRIR 500
0.4
0.2
0
0
5
10
time (s)
15
20
Figure 4.31: Voltage buses of 500KV system for fault at Dabaa-500 and Islanding of
NPP
78
1
0.8
0.6
V
DABBA 220
V
0.4
SALOOM 220
V
SIDI- KRIR 220
0.2
0
0
5
10
time (s)
15
20
Figure 4.32: Voltage buses of 220 KV system for fault at Dabaa-500 and
Islanding of NPP
0
-20
Change of steam pressure in (psi)
-40
-60
-80
-100
-120
-140
-160
-180
-200
0
10
20
30
40
50
Time (sec)
60
70
80
90
100
Figure 4.33: Change of steam pressure of Dabaa PWR for fault at Dabaa 500 and L2
Islanding of NPP
79
0
Change in Primary Temperature in (F°)
-2
-4
-6
-8
-10
-12
-14
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.34: Change of primary temperature of Dabaa PWR for fault at Dabaa 500 and Islanding of NPP
0
-2
Change in Tube Metal Temperature in (F°)
-4
-6
-8
-10
-12
-14
-16
-18
-20
0
10
20
30
40
50
Time (sec.)
60
70
80
90
Figure 4.35: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and
Islanding of NPP
80
100
4.3.5 Fault at bus of Dabaa-220 accompanied by outage of lines
T.L3 and T.L4
A three-phase short circuit fault is applied to the bus of Dabaa-220 and
accompanied by outage for the line (T.L4 & T.L3) which connect Dabaa-220 with
Matroah and Omeed respectively. . It is noticed from the following figures (from
figure 4.36 to figure 4.41) that the impact of this fault on the performance of the
power system is limited where the change in the frequency of Dabaa generation
unit from the nominal value and the change in the voltage are very small. Hence,
this case of contingency has a limited impact on the performance of Dabaa units.
0.65
δSyn 1
0.6
Deltaof Dabaaunit
0.55
0.5
0.45
0.4
0.35
0.3
0.25
0.2
0
5
10
time (s)
15
Figure 4.36: Delta of Dabaa gen. unit for fault at Dabaa-220 and line (T.L4 & T.L3)
outage
81
20
1.005
Frequency of Dabaaunit (P.u)
1.004
1.003
1.002
1.001
1
0.999
0.998
0.997
0.996
0
5
10
time (s)
15
20
Figure 4.37: Frequency of Dabaa unit for fault at Dabaa-220 and lines (T.L4 &
T.L3) outage
1.4
1.2
1
0.8
0.6
VDABBA220
VDABBA1
0.4
VMATROAH
0.2
0
0
VOMEED1
5
10
time (s)
15
Figure 4.38: Bus voltages for fault at Dabaa-220 and line (T.L3 & T.L4) outage
82
20
1.0005
Change in Steam Pressure in (P.U)
1
0.9995
0.999
0.9985
0.998
0.9975
0.997
0
10
20
30
40
50
60
70
80
90
100
Figure 4.39: Change of steam pressure of Dabaa PWR for fault at Dabaa 220 and L3&L4 outage
0
-0.05
Change in Primary Temperature in (F°)
-0.1
-0.15
-0.2
-0.25
-0.3
-0.35
-0.4
-0.45
-0.5
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.40: Change of primary temperature of Dabaa PWR for fault at Dabaa 220 and L3&L4
outage
83
0
Change in Tube Metal Temperature in (F°)
-0.1
-0.2
-0.3
-0.4
-0.5
-0.6
-0.7
0
10
20
30
40
50
Time (sec.)
60
70
80
90
100
Figure 4.41: Change of tube metal temperature of Dabaa PWR for fault at Dabaa 500 and
T.L3 & T.L4 outage
84
CHAPTER 5
CONCLUSIONS AND FUTURE WORK
5.1 Conclusions
This thesis can be summarized in the following points:
1. The proposed NPP which is planned to be constructed in Dabaa is the
optimal solution for solving the problem of continuous demand on electric
energy in Egypt and the most suitable reactor for the first nuclear project in
Egypt is PWR.
2. The detailed model of NPP ,which includes the reactor neutron dynamics
model, reactor thermal dynamics model, steam generator dynamics model,
reactor power control system model and coolant pump system model,
should be used for the medium-term and long-term power system stability
analysis.
3. Pressurized water reactor employs the induction motors to drive their
reactor coolant pumps (RCPs) and feed water pumps (FWPs). For large
disturbance the RCP and FWP will be affected, hence the water flows
through the reactor vessel and through the steam generator will be affected
and any reduction in core flow will result in excessive fuel heat up and
possible fuel damage..
4. The limits of frequency and voltage at which the unit may be islanded
should be very carefully chosen. If they are too conservative, there would
be frequent disturbances in the grid because of loss of generation; if they
85
are too liberal, the NPP may trip before islanding or important equipment
may be damaged.
5. Studying the Egyptian unified power system shows that it is an efficient,
safe, secure and reliable power system and its impacts on the performance
of proposed NPP are accepted.
6. The most severe faults which have a bad impact on the performance of NPP
occur at the 500 KV system.
5.2 Future work
The recommended future work may be:
1.
Applying the study of interconnection to more than one NPP in the system.
2.
Studying the impacts of interconnection on the performance of NPP due to
massive disturbance in power system such as voltage collapse or complete
blackout
3.
Studying the impacts of interconnection if the NPP operates in load-follow
mode.
4.
Taking the behavior of off-site and on-site power supply of NPP into
consideration.
5.
Studying the impacts of interconnection if the transmission system between
the grid and NPP containing a HVDC (high voltage direct current) links.
86
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10
April
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89
APPENDIX (A)
H.B ROBINSON NUCLEAR PLANT DATA
Table (A.1): Reactor design data
Core Thermal and Hydraulics Characteristics
Variable
Value
Total primary output, Mw[th]
2200(7508*106
Btu/h)
Normal primary system pressure, psi
2250
Total coolant flow rate, Ib/h
101.5*106
Average coolant velocity along fuel rods, ft/s
14.3
Total mass of coolant in primary loop, Ib
406050
Normal coolant inlet temperature, F
546.2
Normal coolant outlet temperature, F
602.1
Active heat transfer surface area, ft2
42460
Average heat flux, Btu/h.ft2
171600
Fuel to coolant heat transfer coefficient in fuel, Btu/h.ft2 F
176
90
Kinetic Characteristics
Doppler coefficient, (∆k/k)/F
-1.3*10-5
Moderator temperature coefficient, (∆k/k)/F
-2*10-4
Moderator pressure coefficient, (∆k/k)/F
3*10^-3
Prompt neutron lifetime, sec
1.6*10^-5
Delayed neutron fraction
0.0064
Table (A.2): Steam generator data
Steam generator
Variable
Value
Number of U-tubes
3260
U-tubes diameter, in
0.875
Average tube wall thickness, in
0.05
Mass of U-tube metal, Ib
91800
Total heat transfer area, ft2
44430
Steam condition at full load
91
Table (A.2, cont): Steam generator data
Steam flow, Ib/h
3.169*106
Steam temperature, F
516
Steam pressure, psi
770
Primary side coolant
Reactor coolant flow, Ib/h
33.93*106
Reactor coolant water volume, ft3
928
Secondary side fluid
Feed water temperature, F
435
Secondary side water volume, full power,ft3
1526
Secondary side steam volume, full power,ft3
3203
Table (A.3): Pressurizer design data
Water volume, full power ,ft3
780
Steam volume, full power ,ft3
520
Electric heater capacity, KW(total)
1300
92
APPENDIX (B)
Table (B.1): Electricity for 2009/2010
93
94
Table (B.2): General Power Stations Statistics
95
Table (B.3): Development of Installed Capacities
Table (B.4): Performance Statistics for Power Plants
96
97