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CHAPTER 1

INTRODUCTION

1.1 Background

In 2008, total worldwide energy consumption was 80 to 90 percent derived from the

combustion of fossil fuels. This is equals to an average power consumption rate of 15

terawatts. In the International Energy Outlook 2009 (IEO2009) by U.S. Energy

Information Administration (EIA) states that world energy consumption increases from

472 quadrillion Btu (1 Btu = 1.06 kilojoules) in 2006 to 552 quadrillion Btu in 2015 and

678 quadrillion Btu in 2030 by prediction. The recession of current economic slow down

the world demand for energy in the near term as manufacturing and consumer demand for

goods and services are slow.

The use of all energy sources increases over the time. IEO2009 states that the

world oil prices will remain relatively high through most of the projection period. Liquid

fuels and other petroleum are the world’s slowest growing source of energy. The liquids

fuel consumption increases at an average annual rate of 0.9 percent from 2006 to 2030.

Renewable energy source are the fastest growing source of the world energy, with

consumption increasing by 3.0 percent per year. The increasing of oil prices, as well as

rising concern about the environmental impact of using fossil fuel and strong government

incentives for increasing renewable energy sources in most countries around the world

improves the prospects for renewable energy sources worldwide.

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FIGURE 1.1 : World Marketed Energy Consumption 1980-2030

From the figure 1.1 above, the usage of liquids fuel such as petroleum is huge

throughout the years. It follows by coal, natural gases and renewable energies. Renewable

energies increase actively after year 2005. Nuclear power remains the lowest from year

1980 towards year 2030.

Although liquid fuels are predicted to be the largest source of energy, but the

liquids share of the world marketed energy consumption declines from 36 percent in 2006

to 32 percent in 2030. IEO2009 states that this declining of consumption is due to the

world oil prices and lead to many energy users especially in the industrial and electric

power sectors to switch from liquid fuels to another energy sources. From 2006 to 2030,

the liquid consumption rate is declining in residential, commercial and electric power

sectors throughout the whole world.

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Natural gas remains an important fuels for electricity generation worldwide after

liquid fuels, because it is more efficient and emit less carbon than other fossil fuels.

IEO2009 stated that total natural gas consumption increases by 1.6 percent per year on

average and its use in the electric power sector increases by 2.1 percent per year. With

world oil prices increasing, consumers are expected to choose less expensive natural gas

to meet their energy needs especially in the industrial sector.

World coal consumption increases by 1.7 percent per year on average from 2006

to 2030. There is no policies or legislation that would limit the growth of coal use,

therefore the United States, China and India are expected to turn to coal in place of more

expensive fuels. These three nations account for 88 percent of the projected net increase

in coal consumption from 2006 to 2030 reported by IEO2009. The only decreases in coal

consumption is in Europe and Japan where the populations are either growing slowly or

declining, therefore the electricity demand growth is slow and renewable energy sources,

natural gas and nuclear power are likely to be used to replace coal for electricity

generation.

Electricity generation from nuclear power worldwide increases from 2.7

trillionkilowatthours in 2005 to 3.0 trillion kilowatthours in 2015 and 3.8 trillion

kilowatthours in 2030 in the IEO2009 reference. The increasing fossil fuel prices, energy

security, and greenhouse gas emissions lead to the development of new nuclear

generating capacity.However, there are still some issues that slow the development of

new nuclear power plant such as plant safety, radioactive waste disposal, and he

proliferation of nuclear weapons. High capital and maintenance costs may keep some

countries from expanding their nuclear power programs.

As the world’s population increases and there is likely to be demand for more

electrical power. Energy sources available in the world including coal, nuclear,

hydroelectric, gas, wind, solar, and biomass. In addition, fusion had been originally

proposed as the long-term source.

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TABLE 1.1 : Advantages And Disadvantages Of All Form Of Energy

Source Advantages Disadvantages

Coal In expensive

Easy to recover (in US

and Russia)

Requires expensive air pollution

controls

Significant contributor to acid

rain and global warming

Requires extensive

transportation system

Hydroelectri

c

Very inexpensive once

dam is built

Government has

invested heavily in

building dams,

particularly in the

Western U.S.

Very limited source since

depends on water elevation

Many dams available are

currently exist

Dam collapse usually leads to

loss of life

Dams have affected aquatics

Environmental damage for ares

flooded and downstream

Gas/Oil Good distribution

system for current use

levels

Easy to obtain

Better as space heating

energy source

Very limited availability as

shown by shortage

Could be a major contributor to

global warming

Very expensive for energy

generation

Large price swings with supply

and demand

Liquefied Natural Gas storage

facilities and gas transmission

system have met opposition

from enviromentalists

Wind Wind is free if available Need 3X the amount of installed

5

Good source for

periodic water pumping

demands of farms as

used earlier in 1990s

Generation and

maintenance costs have

decreased significantly

Well suited to rural

areas

generation to meet demand

Limited to windy areas

Limited to small generator size

Highly climate dependent

May affect endangered birds,

however tower design can

reduce impact.

Solar Sunlight is free when

available

Cost are dropping

Limited to sunny areas

throughout the world

Does not require special

materials for mirrors/panels that

can affect environment

Current technology requires

large amounts of land for small

amounts of energy generation

Biomass Industry in its infancy

Could create jobs

because smaller plants

would be used

Inefficient if small plants are

used

Could be significant contributor

to global warming because fuel

has low heat content

Hydrogen Combines easily with

oxygen to produce

water and energy

Very costly to produce

Takes more energy to produce

hydrogen then energy that could

be recovered

Fusion Hydrogen and tritium

could be used as fuel

source

Higher energy output

per unit mass than

Breakeven point has not been

reached after 40 years of

expensive research and

commercially available plants

not expected for at least 35

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fission

Low radiation levels

associated with process

than fission-based

reactors

years

Nuclear Fuel is inexpensive

Energy generation is the

most concentrated

source

Waste is more compact

than any source

Extensive scientific

basis for the cycle

Easy to transport as new

fuel

No greenhouse or acid

rain effects

Does not emit carbon

Requires larger capital cost

because of emergency,

containment, radioactive waste

and storage systems

Requires resolution of the long-

term high level waste storage

issue in most countries

Potential nuclear proliferation

issue

Throughout the world, we need energy source that are cheap in production, less

bad effect to the environment and produces massive energy. As one can see from the table

1.1 above, all energy sources have both advantages and disadvantages. Nuclear has a

number of advantages that warrant its use as one of the many method of supplying an

energy-demanding world. Energy demand will continue to increase with time. Therefore,

the world need to choose wisely which energy source can stay long and do not depleted in

the future. Nuclear method is one of the examples that can last long in the sense of energy

supplying.

Several major reasons that people working in the field still remain optimistic about

nuclear power are :

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The energy produced per amount of material consumed is highest available

Costs are competitive with coal, the major source used in the world

Uranium, the source material is abundant

Plutonium, a by-product of commercial nuclear plant operation, can also be used

as a fuel

The amount of waste produced is the least of any major energy production process

Nuclear energy provides benefits other than electricity generation

Uranium-235 is the isotope of uranium that is used in nuclear reactor. Uranium-235

can produce 3.7 million times as much energy as the same amount of coal. The fuel

assemblies remain in the reactor for 3 to 5 years. The waste, in the form of the radioactive

fission products, remains inside the fuel. 2000 kg of uranium are converted to waste after

1.5 years of operation. Currently, the fuel assemblies remain in the pools for about 10 to

15 years. After that time, they are being transferred to special storage where air can be

used for cooling.

1.2 Objectives

Till today, the nuclear power industry has been developing and improving reactor

technology for more than five decades of improvement and new generations of nuclear

reactors are in research and development to meet the rapidly increasing energy demand in

the world market. Therefore, the overall objective of this study is to do comparative study

on nuclear reactor technologies and understanding each reactor’s technology and their

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requirements. The specific objectives of this study is to perform research study on nuclear

energy in nuclear power industry and its concepts, study on current nuclear reactor

technologies, comparing on several types of nuclear reactor that are currently being used

in the world market, review on technical, economical, commercial assessment and safety

of nuclear power technology, and finally, the comparison and conclusions included in this

project are intended to provide an overall picture of the current status of reactor

technology.

1.3 Methodology

The following approaches are adapted in achieving the objectives. Firstly, is literature

studies. Literature studies done on nuclear reactor history, development by region,

statistic, nuclear reactor generations, nuclear reactor types and latest nuclear reactor.

Secondly is information collection. Next will be information analysis and data

interpretation for collected information. Analyze and summarize collected information

and comparison will be make for nuclear reactor comparison studies. Follow by that will

be the model making process. Model of AP1000 and APR1400 reactor coolant system is

made. And finally is documentation. All data collected after summarized and analyze will

be documented in a systematic manner. Documentation will be done throughout the

whole project process so that follow up of information shall be easy to trace at the end of

the project and thesis will be the last documentation work for this project.

CHAPTER 2

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LITERATURE REVIEW

2.1 Introduction To Nuclear Power

2.1.1 What Is Nuclear Power

Many elements exist in nature with a variety of isotopes. Chemically identical, the

various isotopes only differ in the number of neutrons in their nuclei. The majority of the

isotopes found on earth are stable but several, including uranium 238 (238U) shown in

figure 2.3, are not and these are termed radioactive elements. These can spontaneously

naturally decay to form other elements by three processes which are the , , and

decay. During -decay, a helium nucleus is emitted, with -decay a high energy electron

is formed and -decay results in the formation of a high energy photon.

Conversely to the above natural decay processes, a nucleus can be transformed

through fission. This usually occurs in highly unstable nuclei, for example, if a U235

nucleus absorbs an extra neutron, it undergoes nuclear fission and splits into two or more

fragments, which form atoms of other elements along with some more neutrons. The

atoms remaining are termed fission products and examples including strontium and

xenon. The neutrons produced in the fission process can be absorbed by other U235 nuclei

and the process can continue in a self-sustaining chain reaction if the concentration of

U235 in the material is sufficiently high, beside other elements is form, this process emit

high energy to surroundings in the form of heat. Figure 2.1 shows a clear picture of

nuclear fission.

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FIGURE 2.1 : Nuclear Fission Process

Using neutron moderators to change the portion of neutrons that will go on to

cause more fission controls these nuclear chain reactions. Nuclear bomb is an

uncontrolled nuclear fission chain reaction. All these nuclear activity happens in the

nuclear reactors. A nuclear reactor is a device which nuclear chain reaction are initiated,

controlled and sustained at a steady state to convert nuclear energy into extremely high

heat.

Beside nuclear fission, there is another process which can produce massive energy

which are the nuclear fusion. In figure 2.2, nuclear fusion is the process by which charged

atomic nuclei join together to form a heavier nucleus. When the nuclear fusion is an

uncontrolled chain reaction, it can result in thermonuclear explosion similar to hydrogen

bomb. Research into controlled nuclear fusion to produce fusion power for production of

electricity has been conducted over the past 50 years. There are some technological and

scientific difficulties with the nuclear fusion process while doing the research and

therefore it is still in progress till today.

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FIGURE 2.2 :Nuclear Fusion

FIGURE 2.3 : Uranium Ore

2.1.2 How Nuclear Power Works

The purpose of a nuclear power plant is not to produce or release nuclear power. The

purpose of a nuclear power plant is to produce electricity. Nuclear power plants have

many similarities to other electrical generating facilities. It should also be obvious that

nuclear power plants have some significant differences from other plants. There are

several known methods to produce electricity. The most practical for large scale

production and distribution involves the use of an electrical generator. In the electrical

generator, a magnet or we called it as rotor revolves inside a coil of wire named as stator,

creating a flow of electrons inside the wire. This flow of electrons is known as electricity.

Some mechanical device such as wind turbine, water turbine, steam turbine and diesel

engine must be available to provide the motive force for the rotor. When a turbine is

attached to the electrical generator, the kinetic energy of the wind, falling water, or steam

pushes against the fan-type blades of the turbine causing the turbine to turn and therefore

the attached rotor of the electrical generator to spin and produce electricity.

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FIGURE 2.4 : Hydroelectric Plant

For example, as figure 2.4, in a hydroelectric power plant, water flowing from a higher

level flow to a lower level and travel through the metal blades of a water turbine causing

the rotor of the electrical generator to spin and produce electricity.

FIGURE 2.5 : Fossil Fuel

Steam Plant

In a fossil-fueled power plant as figure 2.5 above, heat generated from the burning of

coal, oil, or natural gases. This will converts the heat and boils the water into steam which

piped to the turbine. In the turbine, the steam passes through the blades which spins the

electrical generator and produce electricity. After the stream leave the turbine, it will

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condensed back into water in the condenser and pumped back to the boiler to reheat into

steam.

FIGURE 2.6 : Nuclear Fuel Steam Plant

In a nuclear power plant as shown in figure 2.6, many of the components are similar to

those in a fossil-fueled plant such as the stream turbine and generator. The only different

is the steam boiler that is replaced by a Nuclear Steam Supply System (NSSS). The NSSS

consist of a nuclear reactor, a device for nuclear fission to happens and transfer heat

emitted from the nuclear reaction to boil the water into high pressure steam. The high

pressure steam created will turn the blades of the steam turbine for electrical generator.

Besides from this, nuclear power plant consist of advance cooling system that used to

cooled or as a safety system to control the heat emitted from the nuclear reaction.

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2.1.3 Nuclear Fuel Cycle

The fuel cycle for a nuclear power station is much more complicated than for a traditional

fossil fuel power plant. For example, a coal power station, the fuel is extracted,

transported to the plant where it is used to burnt and any ash is either sold to the

construction industry or disposed off. The fuel cycle for a nuclear plant can include all of

the steps shown in figure below. When fuel reaches the end of its usable life, it is

removed from the nuclear reactor and can be reprocessed to re-extract the unused

uranium or plutonium. This process is sometimes called “closing the back end of the fuel

cycle” and reduces the amount of fresh uranium that has to be purchased, thus reducing

the cost of the fuel for electricity generation. Nuclear fuel cycle are summarized in the

figure 2.7 below.

FIGURE 2.7 : Nuclear Fuel Cycle

2.1.3.1 Mining

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Most of the uranium comes from Australia, Canada or United States. The impurities are

removed at the side in order to save on transportation costs and the uranium ore

concentrate also known as the “yellow cake” as shown in figure 2.8 below is taken to be

processed into uranium metal or enriched UO2 pallets. The level of radiation is still very

low because it is still stable.

FIGURE 2.8 : Uranium (Yellow Cake)

2.1.3.2 Processing and Enrichment

Once purification is complete, the yellow cake is concerted to uranium hexafluoride

(UF6). Uranium hexafluoride is gaseous, and is spun in a very high speed to separate the

lighter U235 from the heavier U235. If the purification had not been carried out, other light

gases would exist in the centrifuge and contaminate the enriched product. The enrichment

process yields large amount of uranium in which the level of U235 is reduced to about

0.2% to 0.25%, this is termed as depleted uranium. This material is currently stored but it

may be used in the future fast reactors as a fertile fuel.

2.1.3.3 Fabrication

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The next step in the process is to convert the enriched UF6 into uranium dioxide for used

in the reactors, commonly is Pressurize Water Reactor (PWR). Unenriched uranium ore is

converted into uranium metal rods for the use in Magnox reactors.

The production of oxide fuel from the enriched UF6 can be performed via a

method named as the dry route. In dry conversion the UF6 is decomposed by steam to

produce UO2F2 which is a solid. This is then reduced to UO2 using fluidized bed

technology, a two step process using a rotary kiln (a furnace or oven for drying). The UO2

powder produced at the end of this process is then pressed by compression into pellets.

The shape and size of the pellets differs for different reactors. Solid pellets are used in

most PWRs and annular pellets are used in Advanced Gas-Cooled Reactor (AGR).

Annular pellets have a cylindrical hole running through the centre of the pellet and

thus require a retractable pin in the press. The purpose of this hole is to accommodate

distortions in the fuel and fission gasses formed in the reactor. Once the pellets have been

pressed, they are sintered at 1750 oC in a reducing atmosphere of hydrogen or a mixture

of hydrogen and nitrogen to prevent oxidation and the formation of U3O8. This process

increases the density of the pellets and gives them the physical properties they requires to

withstand the high temperature conditions in the reactor.

The next process is to assemble the pellets into fuel pins or elements. These differ

wildly among the different reactors but a similar process is used for both AGR and PWR

fuel elements. In both of these, the fuel pellets are stacked and weighed and then inserted

into the cladding. Once this is complete, the cladding is then filled with helium gas and

the ends are sealed, welded and tested.

The fuel is then transported and installed into the reactor where it is used until the

build-up of neutron absorbing fission products and other detrimental effects such as fuel

swelling, require the fuel to be removed and either disposed off or reprocess again. This

completes the front end of the fuel cycle.

2.1.3.4 Reprocessing and Recycling or Disposal

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After the fuel has been removed from the reactor, it enters the back end of the fuel cycle.

At this point, the fuel is highly radioactive due to the presence of fission products and

need to handle it by care. This fuel must be stored and cooled until the level of radioactive

is low enough to allow transport to the reprocessing site or the interim storage facility.

The fuel is normally stored in the ponds at the reactor site. These ponds are sealed

reinforced concrete structures filled with water. This acts as an effective radiation shield

and also provides cooling to the fuel which may otherwise heat to the point where the fuel

or the cladding becomes damaged and release contaminated material into the local

environment.

Once the material has cooled sufficiently, it is either taken to a storage site or to a

reprocessing plant. Reprocessing has several advantages over storage for later disposal

and these are listed here:

Security of Supply : Security of supply is a concern for some countries where

there are no natural uranium deposits. This mean reprocessed uranium is a

valuable resource that should not be wasted. Some countries choose to store the

spend fuel and leave the reprocessing for later but corrosion of the fuel and

cladding materials can be a problem if it is to be stored for a long periods of time.

Waste Management : The recovery of useful material means that the volume of

high level waste that must be disposed of is reduced by a factor of 9. The

radioactive content is also reduced as the alternative, direct disposal, adds

approximately 250kg of plutonium per year to the fuel awaiting burial. Since the

medium to long term radioactivity is dominated by plutonium isotopes, the

radioactivity over 10,000 years can be reduced by over 30%.

Improved Proliferation Resistance : Proliferation is the unlawful diversion of

fissile material. In particular, plutonium is potentially attractive to terrorist

organizations. While considerable effort is made to keep this material safe,

converting this material into mixed oxide (MOX) fuel makes it much less

attractive as the organization would not only have to move the bulky material but

to chemically separate it before it could be used.

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Reprocessing fuel involves separating out the uranium and the plutonium from the

rest of the fuel. These two elements can consist of 97% to 99% of the spent fuel with the

remaining being high level waste including fission products and some of the minor

actinides, including neptunium, americium and californium. Once reprocessing is

complete, the uranium and plutonium are stored waiting to re-enter the fuel cycle at the

enrichment or fuel fabrication step. Figure 2.9 below are to summarized and give an

overview idea of nuclear fuel cycle.

FIGURE 2.9 : Summarize of Nuclear Fuel Cycle

2.1.4 Nuclear Waste

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Any kind of industry develops waste materials along side the desired products and the

nuclear industry is no different. The source of waste in the nuclear industry mainly come

from the following operations:

Reprocessing spent fuel

Final decommissioning when a plant reaches the end of its lifespan

Military waste

Surplus materials

Primary waste from fuel reprocessing includes the fission products, minor actinides

and the remains of the cladding. Secondary wastes that are formed during reprocessing

can include solvents that are no longer recoverable, worn out equipment or clothing and

other domestic waste that may have been contaminated with radioactive material. It’s the

aim of the industry to minimize the amount of secondary waste generated and to convert

as much of the radioactive material into a form that is both suited to long-term storage

and final disposal while taking up as small volume as possible. While doing this, the

environmental impact should be kept as low as reasonably attainable. Store pond shown

in figure 2.10 are used to store nuclear waste.

FIGURE 2.10 : Store Pond For Nuclear Waste

2.1.5 Early Years of Nuclear Power

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In the early years of nuclear power industry, the Union of Soviet Socialist Republics

(USSR)’s Obninsk Nuclear Power Plant became the world’s first nuclear power plant to

generate electricity for a power grid and produced approximately 5 megawatts of electric

power. In 1954, Lewis Strauss, the chairman of the United States Atomic Energy

Commission spoke of electricity in the future being “too cheap to meter” and he refers to

hydrogen fusion but Strauss’s statement was interpreted as a promise of very cheap

energy from nuclear fission.

In 1956, the world’s first commercial nuclear power station, Calder Hall in Sellafield,

England shown in figure 2.11 was opened and generated initial capacity of 50MW and

later generation of 200MW. The first commercial nuclear generator to become

operational in the United States was the Shippingport Reactor.

U.S. Navy is the first organizations to develop nuclear power in the early stage for the

purpose of propelling submarines and aircraft carriers. It has a good nuclear safety record.

The U.S. Navy has operated more nuclear reactors than any other entity, including the

Soviet Navy, with no publicly known major incidents.

In December 1954, the first nuclear-powered submarine, USS Nautilus (SSN-571) was

put to sea. The U.S. Arm also launches a nuclear power program, beginning in the early

of 1954. The SM-1 Nuclear Power Plant at Ft. Belvoir, Va was the first power reactor in

the U.S. to supply electrical energy to a commercial grid in April 1957, before

Shippingport shown in figure 2.12.

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FIGURE 2.11 : Calder Hall, England

There is two major nuclear disaster happened in the nuclear power history. The

Chernobyl disaster and the Three Mile Island accident. The Chernobyl disaster shown in

figure 2.13 was a nuclear accident that happened on 26 April 1986 at Chernobyl Nuclear

Power Plant, Ukraine. It is the worst nuclear power plant disaster ever happened in

nuclear power history. On 26 April 1986, reactor number four at the Chernobyl plant

exploded. This explosion result in massive fire and highly radioactive fallout into the

atmosphere and over an extensive surrounding area, including the nearby town of Pripyat.

It was four hundred times more fallout was released than had been by the atomic bombing

of Hiroshima in world war two. The 2005 report prepared by the Chernobyl Forum, led

by the International Atomic Energy Agency (IAEA) and World Health Organization

(WHO), attributed 56 direct deaths which include 7 accident workers and nine children

with thyroid cancer. Estimated of 4000 extra cancer deaths among the approximately

600,000 most highly exposed people. This accident raised concerns about the safety of

the nuclear power industry as well as nuclear power in general. Ever since this accident

happened, the development of new nuclear power plant in the whole word is slowing

down.

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FIGURE 2.12 : Shippingport, U.S.

FIGURE 2.13 : Chernobyl Power Plant After The Explosion

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The Three Mile Island nuclear accident shown in figure 2.14 happened in Three Mile

Island Nuclear Generating Station in Dauphin County, Pennsylvania, United States on the

28 March 1979. The Three Mile Island nuclear accident was a partial core meltdown in

Unit 2 which are the pressurized water reactor manufactured by Babcock & Wilcox. It

was the most significant accident happened in the history of American commercial

nuclear power generating industry. It result in release of massive radioactive gases to the

surrounding atmosphere. This accident happened at 4 a.m. on the Wednesday, 28 March

1979 with failures in the non-nuclear secondary system, followed by stuck-open pilot-

operated relief valve in the primary system which leak large amounts of reactor coolant.

This mechanical failures is due to inadequate training and human factors because the

plant operators failed to recognize the situation as a loss of coolant accident. Since the

coolant is loss, the reactor will eventually heat up and result in partial core meltdown.

There are no deaths in this accident. The accident was followed by a slow development of

new nuclear plant construction in the United States.

FIGURE 2.14 : Three Mile Island Nuclear Generating Station

2.1.6 World Nuclear Statistics

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By year 2007, 14% of the world’s electricity came from nuclear power. Table below

shows that the percentage of nuclear power supply produced for a particular country. This

data is gained from International Energy Agency (IEA) year 2007.

TABLE 2.1 : Nuclear Power Percentage of Total Primary Energy Supply

France 42.6Sweden 36.2

Lithuania 31.9Armenia 27.7Slovakia 24.8Bulgaria 24.3

Switzerland 22.5Belgium 21.9Slovenia 21

Korea 17.9Finland 17.3Ukraine 16.1Japan 15

Czech Republic 14.3Hungary 13Germany 12.3

Spain 10.3United kingdom 9.1

United states 9Canada 8.8

Russian Federation 6.1Romania 3.8Argentina 2.8

South Africa 2.3Mexico 1.6

Netherlands 1.3Brazil 1.2China 0.8India 0.8

Pakistan 0.8From the table 2.1 above, the highest nuclear power generating as the primary

energy supply is France is year 2007 with 42.6% of its total generation from other source.

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TABLE 2.2 : Nuclear Electricity Generation 2008 and Reactor Operable

Country Nuclear Electricity Generation 2008

Reactor Operable1 Sept 2009

Billion kWh % e No. MWeArgentina 6.8 6.2 2 935Armenia 2.3 39.4 1 376

Bangladesh 0 0 0 0Belarus 0 0 0 0Belgium 43.4 53.8 7 5728Brazil 14.0 3.1 2 1901

Bulgaria 14.7 32.9 2 1906Canada 88.6 14.8 18 12652China 65.3 2.2 11 8587

Czech Republic 25 32.5 6 3686Egypt 0 0 0 0

Finland 22.0 29.7 4 2696France 418.3 76.2 59 63473

Germany 140.9 28.3 17 20339Hungary 14.0 37.2 4 1826

India 13.2 2.0 17 3779Indonesia 0 0 0 0

Iran 0 0 0 0Israel 0 0 0 0Italy 0 0 0 0Japan 240.5 24.9 53 46236

Kazakhstan 0 0 0 0North Korea 0 0 0 0South Korea 144.3 35.6 20 17726

Lithuania 9.1 72.9 1 1185Mexico 9.4 4.0 2 1310

Netherlands 3.9 3.8 1 485Pakistan 1.7 1.9 2 400Poland 0 0 0 0

Romania 7.1 17.5 2 1310Russia 152.1 16.9 31 21743

Slovakia 15.5 56.4 4 1688Slovenia 6.0 41.7 1 696

South Africa 12.7 5.3 2 1842Spain 56.4 18.3 8 7448

Sweden 61.3 42.0 10 9104Switzerland 26.3 39.2 5 3237

Thailand 0 0 0 0Turkey 0 0 0 0

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Ukraine 84.3 47.4 15 13168UAE 0 0 0 0

United Kingdom 52.5 13.5 19 11035USA 809.0 19.7 104 101119

Vietnam 0 0 0 0WORLD 2601 15 436 372,553

Table 2.2 above are showing that the nuclear electricity generation by year 2008

on each country around the world. Total of 2601 billion kWh been generate from these 44

countries. The total of 15% of the world generated electricity is from nuclear power.

There are 436 reactor operating worldwide by year 2009 which produce 372,533 MWe.

TABLE 2.3 : Reactors Under Construction and Reactor Planned

Country Reactors Under Construction1 Sept 2009

Reactors PlannedSept 2009

No MWe No. MWeArgentina 1 692 1 740Armenia 0 0 0 0

Bangladesh 0 0 0 0Belarus 0 0 2 2000Belgium 0 0 0 0Brazil 0 0 1 1245

Bulgaria 0 0 2 1900Canada 2 1500 4 4400China 16 16440 35 37480

Czech Republic 0 0 0 0Egypt 0 0 1 1000

Finland 1 1600 0 0France 1 1630 1 1630

Germany 0 0 0 0Hungary 0 0 0 0

India 6 2976 23 2976Indonesia 0 0 2 0

Iran 1 915 2 915Israel 0 0 0 0Italy 0 0 0 0Japan 2 2285 13 2285

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Kazakhstan 0 0 2 0North Korea 0 0 1 0South Korea 5 5350 7 5350

Lithuania 0 0 0 0Mexico 0 0 0 0

Netherlands 0 0 0 0Pakistan 1 300 2 300Poland 0 0 0 0

Romania 0 0 2 0Russia 9 7130 7 7130

Slovakia 2 840 0 840Slovenia 0 0 0 0

South Africa 0 0 3 3565Spain 0 0 0 0

Sweden 0 0 0 0Switzerland 0 0 0 0

Thailand 0 0 2 2000Turkey 0 0 2 2400Ukraine 0 0 2 1900

UAE 0 0 3 4500United Kingdom 0 0 4 6400

USA 1 1180 11 13800Vietnam 0 0 2 2000WORLD 50 45,438 137 151,185

Table 2.3 above showing the statistic of reactors building and ordered or planned

reactors in each countries. There are total of 50 reactor in the building process worldwide

which produce 45,438 MWe. 137 nuclear reactor ordered or planned to built which can

supply 151,185 MWe

.

TABLE 2.4 : Reactor Proposed and Uranium Required

Country Reactors ProposedSept 2009

Uranium Required 2009

No MWe

28

Argentina 1 740 122Armenia 1 1000 51

Bangladesh 2 2000 0Belarus 2 2000 0Belgium 0 0 1002Brazil 4 4000 308

Bulgaria 0 0 260Canada 3 3800 1670China 90 79000 2010

Czech Republic 2 3400 610Egypt 1 1000 0

Finland 1 1000 446France 1 1630 10569

Germany 0 0 3398Hungary 2 2000 274

India 15 20000 961Indonesia 4 4000 0

Iran 1 300 143Israel 1 1200 0Italy 10 17000 0Japan 1 1300 8388

Kazakhstan 2 600 0North Korea 0 0 0South Korea 0 0 3444

Lithuania 2 3400 0Mexico 2 2000 242

Netherlands 0 0 97Pakistan 2 2000 65Poland 5 10000 0

Romania 1 655 174Russia 37 36680 3537

Slovakia 1 1200 251Slovenia 1 1000 137

South Africa 24 4000 303Spain 0 0 1383

Sweden 0 0 1395Switzerland 3 4000 531

Thailand 4 4000 0Turkey 1 1200 0Ukraine 20 27000 1977

UAE 11 15500 0United Kingdom 4 6000 2059

USA 19 25000 18867Vietnam 8 8000 0

29

WORLD 295 303,405 65,405

Table 2.4 above stating that 295 reactor proposed to be built and the total uranium

required for the whole world for their nuclear power plant is 65,405 tonnes.

FIGURE 2.15 : Nuclear Capacity in Current and Future Nuclear Power Countries

Figure 2.15 above gained from Nuclear Century Outlook by World Nuclear

Association (WNA) stating that the current nuclear power countries in low boundary is

around 84% which are 1725 GW and the future nuclear power countries are 16% which

are 325 GW. In the high boundary, current nuclear power countries are 83% which are

9150 GW and 17% which are 1900 GW come from those future nuclear power countries.

Low boundary represent minimum global nuclear capability expected while high

boundary represent maximum nuclear commitment in most nations. From the Nuclear

Century Outlook, expert support that there are few combination of factors that contribute

to the increasing in nuclear usage by many countries in the world. Factors such as new

ore discoveries, advanced mining techniques, more reprocessing, introduction of the

thorium fuel cycle and, ultimately, employment of breeder reactors which will ensure

30

affordable and continuous of nuclear fuel supplies to produce cheap electricity into the

future.

FIGURE 2.16 : Global Clean Energy Need and Supply

The global population are increasing from 6.6 billion towards 9 billion by year

2050, therefore the demand for electricity by year 2050 will greatly increase to meet

human needs. With the conventional method of producing electricity, it will be

insufficient to provide for future demand and the greenhouse gases that contribute to

global warming must be reduce by 70% in year 2050. Therefore clean energy is

introduced. Clean energy is energy produced without emitting greenhouse gases that will

result in global warming. From the figure 2.16 above, to achieve effective clean energy,

we needs 8000 GW more nuclear power production. Hydropower growth stops at mid-

century while fossil fuel power contribute during the 21st century but does not grow

indefinitely. New renewables energy grow steadily and robustly and in the mean time,

nuclear power grow within the range defined by the WNA outlook boundaries.

31

2.1.7 Nuclear Reactor Technology

Just as many conventional thermal power stations generate electricity by harnessing the

thermal energy released from burning fossil fuels, nuclear power plants converts the

energy released from the nucleus of an atom, typically via nuclear fission. A cooling

system removes heat from the reactor core and transports it to another area of the plant,

where the thermal energy can be harnessed to produce electricity or to do other useful

work. Typically the hot coolant will be used as a heat source for a boiler, and the

pressurized steam from that boiler will power one or more steam turbine driven electrical

generators. Reactors are the place where nuclear fission happens and produce massive

heat to transfer to the surrounding coolant.

Most nuclear electricity is generated using just two kinds of reactors that were

developed in the 1950s and improved since.There are many different reactor designs,

utilizing different fuels and coolants and incorporating different control schemes. Some of

these designs have been engineered to meet a specific need. Enriched uranium commonly

used as a fuel as this fuel choice increases the reactor’s power density and extends the

usable life of the nuclear fuel load, but is more expensive and a greater risk to nuclear

proliferation than some of the other nuclear fuels.

A nuclear reactor is a device in which nuclear chain reactions are initiated,

controlled, and sustained at a steady rate. The most significant use of nuclear reactors is

as an energy source for the generation of electrical power and for the power in some

ships. Early stage of nuclear reactors are use in the naval ship to power up the engine in

military usage in the United States.

The reactor core generates heat in a number of ways :

The kinetic energy of fission products is converted to thermal energy when these

nuclei collide with nearby atoms.

Gamma rays produced during fission are absorbed by the reactor in the form of

heat

32

Heat produced by the radioactive decay of fission products and materials that have

been activated by neutron absorption. This decay heat source will remain for some

time even after the reactor is shutdown.

The heat power generated by the nuclear reaction is 1,000,000 times that of the equal

mass of coal. There are several components common to most of the reactors type :

Fuel – usually come in pallets form of uranium oxide (UO2) arranged in tubes to

form fuel rods.

Moderator – this is a material which installed in the core to slows down the

neutrons released from the fission reaction so that they cause more fission.

Usually is water but heavy water or graphite might be used for different reactor

design.

Control rods – these are made with neutron-absorbing material such as cadmium,

hafnium or boron and are to inserted or withdrawn from the core to control the

rate of reaction or to stop it if emergency. Totally insert of control rod will stop

the fission reaction immediately.

Coolant – a liquid or gas circulating the core and to function as a heat transfer

medium. The heat produce in the reactor will be cooled or controlled by the

coolant.

Pressure vessel or pressure tubes – a robust steel vessel containing the reactor core

and moderator but it may be a series of tubes holding the fuel and conveying the

coolant through the moderator

Steam generator – part of the cooling system where the primary coolant transfer

heat from the reactor is used to make pressurized steam for the turbine.

Containment – structure designed around the reactor core to protect it from

outside intrusion and to protect those outside from the effects of radiation in case

of any malfunction inside such as leak of coolant accident that contain high

radioactive coolant. It is a meter thick concrete and steel structure.

TABLE 2.5 : Nuclear Power Plants in Commercial Operation

33

Reactor Type Main Countries No. GWe Fuel Coolant Moderator

Pressurized Water

Reactor (PWR)

US, France,

Japan, Russia,

China

265 251.6 Enriche

d UO2

water Water

Boiling Water

Reactor (BWR)

US, Japan,

Sweden

94 86.4 Enriche

d UO2

water Water

Pressurized Heavy

Water Reactor

CANDU (PHWR)

Canada 44 24.3 Natural

UO2

Heavy

water

Heavy

water

Gas-cooled Reactor

(AGR & Magnox)

UK 18 10.8 Natural

U,

Enriche

d UO2

CO2 Graphite

Light Water

Graphite Reactor

(RBMK)

Russia 12 12.3 Enriche

d UO2

water Graphite

Fast Neutron

Reactor (FBR)

Japan, France,

Russia

4 1.0 PuO2,

UO2

Liquid

sodium

None

Other Russia 4 0.05 Enriche

d UO2

Water Graphite

Total 441 386.5

A cooling source often water but sometimes a liquid metal is circulated past the

reactor core to absorb the heat that it generates. The heat is carried away from the reactor

and is then used to generate steam. Most reactor systems employ a cooling system that is

physically separate from the water that will be boiled to produce pressurized steam for the

turbines, like the pressurized water reactor. But for boiling water reactor, the water for the

steam turbines is boiled directly by the reactor core.

34

In some reactors, the coolant acts as a neutron moderator. A moderator increases

the power of the reactor by causing the fast neutrons that are released from fission to lose

energy and become thermal neutrons. Thermal neutrons are more likely than fast neutrons

to cause fission, so more neutron moderation means more power output from the reactors.

If coolant is a moderator, then temperature changes can affect the density of the

coolant/moderator and therefore change power output. A higher temperature coolant

could be less dense, and therefore a less effective moderator.

There are several types of reactor in current market which are :

Boiling Water Reactor (BWR)

Pressurized Water Reactor (PWR)

CANDU Reactor

Heavy Water Reactor

Light Water Reactor

Many more

Reactors divided into 4 generation :

Generation I reactor (prototype)

Generation II reactor (most current nuclear power plants)

Generation III reactor (evolutionary improvements of existing designs)

Generation IV reactor (technologies still under development)

2.1.8 Inside a Nuclear Power Plant

Nuclear power plants are similar to conventional power plant such coal-fired power plant.

Beside the common steam turbine, there are other building such as :

35

FIGURE 2.17 : Typical Nuclear Power Plant Layout

Figure 2.17 above shows that a typical nuclear power plant layout which are

labeled. Nuclear fission reaction happens in C and the fission rate being controlled by B.

Water from the lake or sea will transfer through I as a coolant and the heat dissipate to the

surrounding atmosphere by J. the heat generated by nuclear fission in C transfer to

primary loop through F and D generate pressurized steam to push H to turn. And the

rotating force of H will turn the G to produce electricity. The steam being condenses at I

and being pump back by F to the D and the process repeat itself.

Containment or Drywell Building

A building shown in figure 2.18 was designed to sustain pressures of about 345kPa.

Normally houses the reactor and the related cooling system that contains highly

radioactive fluids. Building is made of steel construction. Sometimes the building is

surrounded by a concrete structure that is designed for much lower pressures. The area

between the steel and concrete building is called the ‘Annulus’. Designs vary. At one

facility there are 1.37 meter concrete walls reinforced with steel. The dome is 0.762 meter

thick and the base 3.66 meter thick. The containment is the 3rd fission product barrier. In

BWRs, the drywell is located in the reactor building.

36

FIGURE 2.18 : Containment Building

Auxiliary or Reactor Building

A building separate from the containment that houses much of the support equipment that

may contain radioactive liquids and gases. Emergency equipment is also normally located

in this building.

Turbine Building

A building that houses the turbine, generator, condenser, condensate and feedwater

systems shown in figure 2.19.

FIGURE 2.19 : Turbine Building

37

Intake Structure or Screenhouse

A building that houses the circulating water pumps used to pump water from the river,

lake, sea for cooling the condenser. Trash racks and traveling screens also remove debris

to clean the water so that it can pass through the condenser tubes.

Fuel Building

A building separate from the containment that is used to spent fuel assemblies in steel

racks in a large 12.2 meter deep storage pool shown in figure 2.20. Casks for shipping or

onsite dry storage of spent fuel assemblies will be loaded or unloaded in this pool. A new

fuel storage area is provided for receipt of new assemblies and storage prior to going into

the containment and subsequently into the reactor during a refueling.

38

FIGURE 2.20 : Fuel Storage Pool

Diesel Generator Building

A building used to house the diesel generators and supporting systems (air, water, radiator

fans, fuel oil, lubricating oil, air conditioning, and ventilation). In some cases, related

electrical switchgear for distributing electrical power produced by the diesel generator.

The Diesel generators that provide backup electrical power to safety and non-safety

systems.

In some plants separate buildings or areas within the buildings mentioned above may

house the following:

Water treatment systems used to purify water so that it can be used in the power

plant.

Radioactive waste treatment systems used to purify and store radioactive liquids

and gases.

Cooling tower pumps used to pump water to cooling towers. Cooling towers are

often used for power plants located on rivers and small lakes so that impact of

temperature of discharged water on fish is minimized. 

Control Room , related electrical cabling, and ventilation systems (sometimes

called the Control Building) shown in figure 2.21

Administration Building

Security

39

FIGURE 2.21 : Nuclear Power Plant Control Room

2.1.9 Pros and Cons of Nuclear Power

Nuclear power boasts a number of advantages, as well as its share of downright

depressing negatives. As far as positives go, nuclear power's biggest advantages are tied

to the simple fact that it doesn't depend on fossil fuels. Coal and natural gas power plants

emit carbon dioxide into the atmosphere, contributing to climate change. With nuclear

power plants, CO2 emissions are minimal.

According to the Nuclear Energy Institute, the power produced by the world's

nuclear plants would normally produce 2 billon metric tons of CO2 per year if they

depended on fossil fuels. In fact, a properly functioning nuclear power plant actually

releases less radioactivity into the atmosphere than a coal-fired power plant. By not

depending on fossil fuels, the cost of nuclear power also isn't affected by fluctuations in

oil and gas prices.

As for negatives, nuclear fuel may not produce CO2, but it does provide its share

of problems. Historically, mining and purifying uranium hasn't been a very clean process.

Even transporting nuclear fuel to and from plants poses a contamination risk. And once

the fuel is spent, you can't just throw it in the city dump. It's still radioactive and

potentially deadly.

On average, a nuclear power plant annually generates 20 metric tons of used

nuclear fuel, classified as high-level radioactive waste. When you take into account every

nuclear plant on Earth, the combined total climbs to roughly 2,000 metric tons yearly. All

of this waste emits radiation and heat, meaning that it will eventually corrode any

container and can prove lethal to nearby life forms. As if this weren't bad enough, nuclear

power plants produce a great deal of low-level radioactive waste in the form of radiated

parts and equipment.

40

Nuclear waste can pose a problem, and it's the result of properly functioning

nuclear power plants. When something goes wrong, the situation can turn catastrophic.

The Chernobyl disaster is a good recent example. In 1986, the Ukrainian nuclear reactor

exploded, spewing 50 tons of radioactive material into the surrounding area,

contaminating millions of acres of forest. The disaster forced the evacuation of at least

30,000 people, and eventually caused thousands to die from cancer and other illnesses.

Chernobyl was poorly designed and improperly operated. While the plant required

constant human attention to keep the reactor from malfunctioning, modern plants require

constant supervision to keep from shutting down. Still, Chernobyl is a black eye for the

nuclear power industry, often overshadowing some of the environmental advantages the

technology has to offer.

2.2 Nuclear Reactor Development

2.2.1 Nuclear Reactor

Many different designs for power reactors have been proposed and many different

prototypes built. Most of the countries that have developed nuclear power started with

graphite or heavy-water moderated system, since only these moderators allow criticality

with natural uranium. However, most of the power reactors now use slightly enriched

uranium. With such enrichments, other moderators especially light water can be used as

shown in figure 2.22

41

FIGURE 2.22 : Nuclear Reactor

Pressurized Water Reactor (PWR)

Pressurized water reactor (PWR) shown in figure 2.23 is the most widely used type of

power reactors, employ two water loops. The water in the primary loop is pumped

through the reactor to remove the thermal energy produced by the core. The primary

water is held at sufficiently high pressure to prevent the water from boiling. This hot

pressurized water is then passed through a steam generator where the secondary loop

water is converted into high temperature and high pressure steam that turns the turbo-

generator unit. The use of a two-loop system ensures that any radioactivity produced in

the primary coolant does not pass through the turbine for safety purposes. Most of the

today’s advanced nuclear reactor design are based on the PWR basic design philosophy.

FIGURE 2.23 : Pressurize Water Reactor (PWR)

42

Boiling Water Reactor(BWR)

In a boiling water reactor (BWR) shown in figure 2.24, cooling water is allowed to boil

while passing through the core. The steam then passes directly to the turbine. The low

pressure steam leaving the turbine is then condensed and pumped back to the reactor for

the same process. By having a single loop, the need for steam generators and other

expensive equipment in a PWR can be avoided. By having only single loop on coolant in

the BWR might expose radioactive radiation to the surroundings if there is a leak of

coolant accident occur.

FIGURE 2.24 : Boiling Water Reactor (BWR)

43

Gas Cooled Reactor

In a gas cooled reactor (GCR) carbon dioxide or helium gas is used as the core’s coolant

by pumping it through channels in the solid graphite moderator. The fuel rods are placed

in these gas cooling channels. The use of graphite, which remains solid up to very high

temperatures, eliminates the need for an expensive pressure vessel around the core, the

hot exit gas then passes through steam generators. Magnox Reactor are one of the GCR

which design by United Kingdom which shown in figure 2.25. Magnox Reactor are

pressurized, carbon dioxide cooled, graphite moderated reactors using natural uranium as

fuel.

In another design known as the high-temperature gas cooled reactor (HTGR), the

fuel is packed in many fuel channels in graphite prisms. Helium coolant is pumped

through other channels through the graphite prisms. The hot exit helium gasews then goes

to a steam generator.

44

FIGURE 2.25 : Gas Cooled Reactor (GCR),Magnox Reactor Design

Liquid Metal Fast Breeder Reactors

Figure 2.26 is a fast reactor, the chain reaction is maintained by fast neutrons.

Consequently, moderator materials cannot be used in the core. To avoid materials of low

atomic mass, the core coolant is a liquid metal such as sodium or a mixture of potassium

and sodium. Liquid metals have excellent heat transfer characteristics and do not require

pressurization to avoid from boiling. However, sodium becomes radioactive when it

absorbs neutrons and also reacts chemically with water. To reduce radioactive sodium

from possibly interacting with the water or steam loop, an intermediate loop of non-

radioactive sodium is used to transfer the thermal energy from the primary sodium loop to

the water or steam loop. The great advantage of such fast liquid metal power reactors is

that it is possible to create a breeder reactor like the one in which more fissile fuel is

produced than is consumed by the chain reaction. There are two designs of the liquid

metal fast breeder reactors which are the :

Loop type, in which the primary coolant is circulated through primary heat

exchangers external to the reactor tank

Pool type, in which the primary heat exchangers and circulators are immersed in

the reactor tank.

45

FIGURE 2.26 : Liquid Metal Cooled Fast Breeder Reactor (LMFBR)

Pressure Tube Graphite Reactors

A once widely used Russian designed power reactor is the Reactory Bolshoi Moshchnosti

Kanalnye (RMBK) shown in figure 2.27 translation in English is high powered pressure

tube reactor. In this reactor, fuel is placed in the fuel channels in graphite blocks that are

stacked to form the core. Vertical pressure tubes are also placed through the graphite core

and light water coolant is pump through these tubes and into an overhead steam drum

where the two phases are separated and the steam passes directly to the turbine.

FIGURE 2.27 : Reactory Bolshoi Moshchnosti Kanalnye (RMBK)

46

Pressurized Heavy Water Reactor (PHWR)

The Canada Deuterium Uranium (CANDU) reactor shown in figure 2.28 was build by the

Canadian itself. It was the pressurized heavy water reactor invented in the late 1950s and

1960s. the CANDU reactor design are similar to most of the light water reactors. Fission

reactions in the reactor core heat pressurized water in a primary cooling loop and heat

exchanger transfer the heat to a secondary cooling loop which powers a steam turbine

with an electrical generator attached to it. The excess heat in the steam is rejected to into

the surrounding atmosphere in different was such as to the ocean, river or lake. The main

difference between CANDUs and other water moderated reactors is that CANDU uses

heavy water for neutron moderation.

47

FIGURE 2.28 : Pressurized Heavy Water Reactor (PHWR), CANDU

2.2.2 Coolant Limitations

The thermal properties of a power reactor coolant greatly affect the reactor design. By far,

the most widely used coolant is water. Water is inexpensive and engineers have

experience in using it as a working fluid in conventional fossil-fueled power plants. The

disadvantage of using water as a coolant is that it must be pressurized to prevent boiling

at high temperatures. Normal water boiling point is 100 degree Celsius. If water is below

the boiling point, it is called subcooled. Water is saturated when vapor and liquid coexist

at the boiling point, and it is superheated when the vapor temperature is above the boiling

temperature. Above the critical temperature, the liquid and gas phases are

indistinguishable, and no amount of pressure produces phase transformation.

To maintain criticality in a water moderated core, the water must remain in liquid

form. Moreover, steam is a much poorer coolant than liquid water in common sense.

Thus, for water to be used in a reactor, it must be pressurized first to prevent significant

steam formation. For water, the critical temperature is 375 degree Celsius, above which

liquid water cannot exist. Thus, in water moderated and cooled cores, temperatures must

be below this critical temperature. Typically, coolant temperatures are limited to about

340 degree Celsius. This high temperature limit for reactor produced steam together with

normal ambient environmental temperatures limit the thermal efficiency for such plants to

about 34%.

Because steam produced by Nuclear Steam Supply System (NSSS) is saturated or

very slightly superheated. Expensive moisture separators which are the devices to remove

liquid droplets and special turbines that can operate with ‘wet steam’ must be used. These

turbines are larger and the cost are more expensive than those used in power plants that

can produce superheated steam.

48

2.2.3 Evolution of Nuclear Power

Basically nuclear reactor are divided in 5 main generations which are :

a) Generation I : Early prototypes

b) Generation II : Commercial power reactors

c) Generation III : Advanced lightwater power reactors

d) Generation III+ : Evolutionary design reactors

e) Generation IV : Conceptual design reactors

FIGURE 2.29 : Evolution of Nuclear Power

49

Generation I

Generation I reactor were the early stage prototypes developed by many nations. It was

developed in the 1950s and 1960s. It was modified and enlarged from the military

reactors. Those reactors are originally usages are either for submarine propulsion or

plutonium production. Most of the Generation I reactors use natural uranium fuel with

graphite as a moderator. Generation I reactors were characterized by fundamentally

unsafe designs, and kludged layers of afterthought safety systems.Generation I reactors

are small reactor which is under 250MW.

Generation II

Generation II reactor are commercial power reactor that still under operating mode till

today worldwide. Generation II reactors were significantly improved from Generation I

reactors, but these changes were primarily evolutionary. The Three Mile Island disaster

was from Generation II design’s reactor. Most of these reactor use so-called lightwater

technology. They are moderated and cooled with ordinary water. Other Generation II

design reactor uses other coolants and moderators.

Generation III

A more advanced generation of reactors are the Generation III reactors. It is also known

as advanced lightwater technology. These reactors are design to be safety and efficiency

improved from Generation II reactors. These included improved of fuel technology,

superior thermal efficiency, passive safety systems and standardized design for reduced

maintenance and capital costs. These improvement will pro-long the operational life. The

first Generation III reactors were build in Japan.

50

Generation III+

Generation III+ reactor are the advancement of Generation III. These reactors were

designed to significantly improve on safety and economics perspective over the

Generation III advanced reactor design. Advanced CANDU Reactor (ACR), AP1000,

APR1400, European Pressurized Reactor (EPR), Economic Simplified Boiling Water

Reactor (ESBWR), mPower etc are all categories under Generation III+ reactor and

consider the latest reactor in today’s world market.

Generation IV

Generation IV are a set of theoretical nuclear reactor designs which are currently being

researched and will not be available in the market before 2030. There is an exceptional

for a version of the Very High Temperature Reactor (VHTR) called the Next Generation

Nuclear Plant (NGNP) which will be completed by year 2021. This generation of reactors

are improve base on few primary goals which are on nuclear safety, proliferation

resistance, minimize waste, natural resource utilization and to decrease the cost of

building and running a nuclear power plant.

2.2.4 World Nuclear Reactor Development

There are many different reactor designs, utilizing different fuels and coolants and

incorporating different control schemes that we have in the current world. Some of these

designs have been engineered to meet a specific need. A nuclear reactor is a device in

which nuclear chain reactions are initiated, controlled, and sustained at a steady state. The

most significant use of nuclear reactors is as an energy source for the generation of

electrical power.

51

FIGURE 2.30 :World Map of Nuclear Power Reactors 2009

There are total of 436 nuclear power reactor operating and 47 reactors under

construction worldwide. From the map above we can conclude that most of the nuclear

reactor technology is active in 3 main regions which are United State, Europe and Asia

Pacific. United State have 104 reactors, Europe have 196 reactors while Asia Pacific

owns 111 reactors.

From the world map in figure 2.30 above, we can see that most of the nuclear

reactor technology developed actively at 3 region of the world which is United States,

Europe and Asia Pacific. Below are the brief timeline for world nuclear reactor

development.

In 1951, Experimental Breeder Reactor 1 was build at Idaho National Engineering and

Environmental Laboratory (INEEL) produces the world’s first usable amount of

electricity from nuclear energy.

52

In 1953, BORAX-I, the first of a series of Boiling Reactor Experiment reactors, was built

at INEEL. The series is to designed to test the theory that the formation of steam bubbles

in the reactor core does not cause an instability problem.

In 1955, BORAX-III becomes the first nuclear power plant in the world to provide entire

town with all of its electricity.

In 1957, the first U.S. large-scale nuclear power plant begins operation in Shippingport,

Pennsylvania. The pressurized-water reactor supplies power to the city of Pittsburgh and

much of western Pennsylvania. It was then replaced by a more efficient light-water

breeder reactor in 1977.

In 1960, first Boiling Water Reactor (BWR) was built

In 1962, the first advanced gas-cooled reactor is build at Calder Hill in England. It was

intended to power up the naval vessel but is too big to install on ship and it was then

successfully used to supply electricity for British consumers.

In 1963, Canada’s CANDU reactor using natural uranium in fuel tubes surrounded by

heavy water.

In 1966, the Advanced Testing Reactor at Idaho National Engineering and Environmental

Laboratory come online for material testing and isotopes generation.

In 1969, the Zero Power Physics Reactor (ZPPR), a specially designed facility for

building and testing a variety of types of reactors. Nuclear reactor can be built and tested

in ZPPR for about 0.1% of the capital cost of construction of the whole power plant.

In 1979, Three Mile Island nuclear accident

In 1986, Chernobyl nuclear disaster

In 2000, nuclear power energy production grows, most notably in China, Korea, Japan,

and Taiwan, where more than 28 GW of nuclear power plant capacity is added since the

last decade of the century.

53

2.2.4.1 United States

FIGURE 2.31 : United States Map of Nuclear Power Reactors

Figure 2.31 above showing the location of nuclear reactor in the United States. The

United State is the world largest producer of nuclear power, generate for more than 30%

of worldwide nuclear generation of total electricity output.

There are total of 69 Pressurized Water Reactor (PWR), 35 Boiling Water Reactor

(BWR) in the United States which provide a total capacity of 100,582 MWe. Almost all

the US nuclear generating capacity comes from reactor that been build from 1967 till

1990. There is no new construction on nuclear power reactor since 1977. Construction of

new reactor stop due to the accident happen in Three Mile Island on 1979, but a further

PWR (Watt Bar 2) is expected to start up by 2013.

Westinghouse designed the first fully commercial PWR of 250 MWe capacity

which start to build on 1960 and operated to 1992. The first commercial plant, Dresden 1

which produce 250 MWe was design and started up in 1960 meanwhile a prototype BWR

ran from 1957 to 1963. Around 1960s, most of the order are being placed for PWR and

BWR reactor units of more than 1000 MWe capacity. Many order and projects of nuclear

54

power reactor was cancelled or suspended, and the nuclear construction industry went

into the doldrums for two decades ever since the Three Mile Island accident.

Since the Three Mile Island accident, in the 1970s, the US nuclear industry

dramatically improved its safety and operational performance with average net capacity

factor over 90% and all safety indicators exceeding target. US are preparing for the new

build of nuclear reactor which is ABWR, AP1000, ESBWR, APWR, and EPR for the

coming years. US federal government has significantly stepped up R&D spending for

future plants that improve or go beyond current design reactor. Next Generation Nuclear

Plant projected to develop a Generation IV High Temperature Gas Cooled Reactor, which

would be part of a system that would produce both electricity and hydrogen gas

massively.

2.2.4.2 Europe

FIGURE 2.32 : Europe Map of Nuclear Power Reactors

55

As of June 2009 there is a total of 196 nuclear power reactor with an installed electric net

capacity of 169,711 MWe in operation in Europe and 16 unites with 13,625 MWe were

under construction in 6 countries. From figure 2.32, the countries that have nuclear power

plant in Europe are Belgium, Bulgaria, Czech Republic, Finland, France, Germany,

Hungary, Lithuania, Netherlands, Romania, Russian Federation, Slovakian Republic,

Slovenia, Spain, Sweden, Switzerland, Ukraine and United Kingdom. In the 1970s,

nuclear power started up in Europe, reaching a peak between 1980 and 1990, followed by

a period during which development was halted. From 1990s onwards, natural gas and

renewable became important.

TABLE 2.6 : Numbers of reactor built between 1971 and 2005

Period Commission Reactors

1971 – 1975 22

1976 – 1980 37

1981 – 1985 66

1986 – 1990 40

1991 – 1995 7

1996 – 2000 6

2001 - 2005 5

Total 183

56

TABLE 2.7 : Reactor Type on Different Generation

Today Short to Medium

Term

Long Term

Generations I and II III IV

Reactor Type PWR (92)

WWER (22)

BWR (19)

AGR (14)

GCR (8)

LWGR (1)

PHWR (1)

FBR (1)

EPR (PWR)

AP1000 (PWR)

WWER (PWR)

ABWR (BWR)

ESBWR (BWR)

HTR

GFR

LFR

MSR

SFR

SCWR

VHTR

2.2.4.3Asia

FIGURE 2.33 : Asia Map of Nuclear Power Reactors

57

From the Asia map shown in figure 2.33 above, most of the nuclear power development

happens in Japan and South Korea. There are total of 55 operating nuclear reactor in

Japan and 20 operating nuclear reactor in South Korea.

Japan stated its nuclear research in 1954. Japan imported its first commercial nuclear

power reactor from UK. It was a gas-cooled Magnox reactor. After this unit was

completed, Japan only builds LWR, BWR and PWR. Since 1970s, 28 BWRs and 23

PWRs have been brought into operation. By the end of 1970s, Japan industry had largely

established their own domestic nuclear power production and exports it to East Asian

countries. The first ABWR which started up in 1996-1997 are now in operation.

South Korea started its nuclear power program in the 1970s by licensing PWR

technology from US-based Westinghouse. Since then, as its industrial base has grown,

domestic researchers and firms have updated the System 80 PWR design originally

imported developed South Korean versions of all major components. South Korea has

imported CANDU from Canada and is developing a strategy to re-use PWR fuel in these.

Korean Hydro and Nuclear Power (KHNP) went on to develop the OPR-1000 and APR-

1400. These 2 reactor are in Generation III+ categories. The first APR-1400 units are

under construction, and operation will begun approximately in 2013 or 2014.

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CHAPTER 3

COMPARATIVE STUDY ON AP1000 AND APR1400

3.1 Introduction

3.1.1 The AP600 Reactor

AP1000 reactor are derived from AP600. AP600 are Generation III reactor designed by

Westinghouse Electric Company, US. The AP600 is power plants which produce 600

MWe are considered small power plant compared to current power demand in the world.

The main objectives in its design are the plant uses forces of nature and simplicity of

design to enhance plant safety and operations and reduce construction costs.

The AP600 obtains its emergency cooling from huge water tanks mounted above

the reactor. Electric power or operator actions are needed to start the coolant injection.

One of the large tanks above the reactor serves as a place to deposit heat. Water tanks

pressurized with nitrogen gas provide sprays to cool the atmosphere inside the

containment. There are no pumps needed in the process. AP600 design to be simple by

reducing the number of valves by 60%, large pumps by 50%, piping by 60%, heat

exchangers by 50%, ducting by 35%, and control cables by 80%. It is estimated that the

plant can be constructed in 3 to 4 years. All of these factors contribute to reducing the

cost.

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The basic design of the AP600 does not differ that much from a conventional

PWR. The main system consists of a reactor with two cooling circuits, leading hot water

under pressure to the steam generators. New in the design are the emergency core cooling

systems designs, to prevent the overheating and melting of the reactor core in case of a

Loss Of Coolant Accident (LOCA). These systems that contain thousands of cubic meters

of water should supply emergency cooling when the normal cooling fails.

Four tanks are located above the reactor with borated water. In case of a loss of

coolant accident (LOCA) this water (about 50 cubic meters) would enter the reactor. The

borated water would stop the fission reaction. Besides, a water tank with about 1,900

cubic meters is situated in the containment. This amount would be enough to flood the

whole containment building above the level of the reactor core. In that way the reactor

building would be changed into a kind of swimming pool in which the hot reactor could

cool down. Most evolutionary and also most controversial is the passive containment

cooling system shown in figure 3.1. After an accident it is important to keep the

containment intact. With too much pressure on the steam, it would burst and release

radioactivity. To keep the pressure low enough, the AP600 containment is constructed to

lead away the heat by a water-and-air-cooled system.

The AP600 is designed with a single containment. Conventional reactors are

constructed with double containments, a steal and a concrete one. The AP600 has only

one containment to provide maximum heat transport to outside air. Besides, above the

building is a water tank with 1,300 cubic meters is located to spray the iron containment

to cool it down. This water would be enough to cool the containment for three days.

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FIGURE 3.1 : AP600 Passive Containment Cooling System

61

3.1.2 The OPR1000 Reactor

APR1400 are derived from OPR1000 that are Generation III reactor. OPR1000 are design

by KHNP, Korea.OPR1000 produces 1000MWe. It has been in operation since 1998 and

has record in outstanding safety and reliability.The OPR1000 operability improved by

using two larger steam generator which can reduced plant trips due to greater capability to

accommodate the changes of steam generator level at transient conditions.

The Reactor Coolant System (RCS) has two transfer loops forming a barrier to

the release of radioactive materials from the reactor core to the secondary system and

containment atmosphere. The main components of the RCS of OPR1000 are a reactor

vessel, two steam generators, and four reactor coolant pumps. The RCS also includes the

interconnecting piping to auxiliary systems such as the chemical and volume control

system, the safety injection system, the shutdown cooling systems etc. These RCS

components are symmetrically located on opposite sides of the reactor vessel with a

pressurizer on one side, all of the RCS components are located inside the containment

building and connected by pipe assemblies.

A large pressurizer volume to enhanced capability to cope with LOCA. Adoption

of feed water storage tank. Develop Integrated Reactor Vessel Head Assembly to reduce

the refueling time and to enhance maintainability. Uses circulating water system to reduce

the numbers of pumps. Construction period has been dramatically reduced through

repeated construction of the OPR1000. Construction capital cost has been significantly

reduced through duplication and shortened construction period of the OPR1000.

OPR1000 plant arrangement consist of one Compound Building combining five buildings

which are two Secondary Auxiliary Building, two Access Control Buildings and one

Radwaste Building.

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3.2 Reactor Coolant System

AP1000

The reactor coolant system (RCS) consists of two heat transfer circuits, with each circuit

containing one steam generator, two reactor coolant pumps, and one hot leg and two cold

legs for circulating coolant between the reactor and the steam generators (SG). The

system also includes a pressurizer (PZR), interconnecting piping, and the valves and

instrumentation necessary for operational control. The RCS arrangement is shown in

Figure 3.2. The reactor containment contains all the RCS equipment. The RCS pressure

boundary provides a barrier against the release of radioactivity generated within the

reactor. The RCS pressure is controlled by the PZR, where water and steam are

maintained in equilibrium by activating the electrical heaters or a water spray, or both.

Steam is formed by the heaters or condensed by the water spray to control pressure

differences due to expansion and contraction of the reactor coolant. Spring-loaded safety

valves are installed above and connected to the PZR to provide overpressure protection

for the RCS. These valves discharge the overpressure into the containment atmosphere.

FIGURE 3.2 : AP1000 Reactor Coolant System

63

Three stages of RCS automatic depressurization valves are also connected to the PZR.

These valves release steam and water through sprinkler to the in-containment refueling

water storage tank (IRWST) of the passive core cooling system (PXS). All the steam and

water released is condensed and cooled by mixing them with the water in the tank.

APR1400

The APR1400 is a two-loop pressurized water reactor. Its NSSS is designed to operate at

a rated thermal output of 4000 MWth with an electrical output of 1455 MWe. It consist of

two primary coolant loops, each of which consists of one 42-inch hot leg, two 30-inch

cold legs, one SG, and two RCPs. One PZR with heaters is connected to a hot leg of the

RCS. The APR1400 RCS arrangement is shown in figure 3.3. The decrease in

temperature in the hot leg result in decrease in frequency of unplanned reactor trips

during normal operation so that it can enhance the operation flexibility.

FIGURE 3.3 : APR1400 Reactor Coolant System

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Additionally, the decrease in the hot leg temperature reduce the ageing of the SG tube due

to stress corrosion by using an advanced tube material, Inconel 690, which is famous to

be more resistant to stress corrosion cracking than Inconel 600, which has been used in

conventional plants in the world now days. The pilot operated safety relief valves

(POSRVs), which replace the conventional spring-loaded safety valves, are used to

perform the functions of PZR safety valves and safety depressurization valves in the same

time. These improvements result in reliable valve operation without the need to remote

manual operation the valves under post-accident conditions.

3.3 Core and Fuel

AP1000

Several important improvements are made based on existing technology. For example,

there are fuel performance improvements, such as Zircaloy grids, removable top nozzles,

and longer burnup features. This optimization of fuel is currently used in approximately

120 operating plants worldwide. AP1000 uses a standard 17 X 17 fuel assembly which

most of the current reactors are using.

FIGURE 3.4 : AP1000 Fuel Assembly

65

AP1000 has a 157 assembly high power density core and the core is 4.27 meter. The core

design is shown in figure 3.4 above. In addition, movable bottom mounted in-core

instrumentation has been replaced by fixed top mounted instrumentation. Inconel 600 is

not used in the reactor vessel welds. The refueling cycle is about 18 to 24 months.

AP1400

The core consists of 241 fuel assemblies, 93 control element assemblies (CEAs), and 61

in-core instrumentation (ICI) assemblies. The refueling cycle of the core is 18 months

with a maximum discharge rod burn-up of 60,000 MWD/MTU and the thermal margin of

the core has increased to more than 10%. The improvement of this core design has

increase the economic efficiency and safety of the APR1400. The fuel assembly is

arranged by 236 fuel rods containing UO2 pellets in a 16 X 16 array. The core design is

shown in figure 3.5. The absorber materials used for full strength control rods and part

strength control rods are Boron Carbide (B4C) pellets and Inconel 625.

FIGURE 3.5 : APR1400 Fuel Assembly

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3.4 Pressurizer

AP1000

The AP1000 PZR shown in figure 3.6 is a main component of the RCS pressure control

system. It is a vertical, cylindrical vessel with hemispherical top and bottom heads, where

liquid and vapor are maintained in equilibrium saturated conditions to control the RCS

pressure. It consist of one spray nozzle and two nozzles for connecting the safety and

depressurization valve inletheaders. Electrical heaters are installed on the bottom head.

Theheaters are removable for replacement or maintenance. The bottom head contains the

nozzle for attaching to thesurge line. This line connects the PZR to a hot leg. The main

function is to provides the flow of reactorcoolant into and out of the PZR during RCS

thermal expansions andcontractions. The PZR safety valves are spring loaded and self-

activated when the pressure in PZR exceeded the limit.

FIGURE 3.6 : AP1000 Pressurizer

67

APR1400

The PZR is a vertically mounted, bottom supported, cylindrical pressure vessel with

replaceable electric heaters to maintain the RCS pressure as shown in figure 3.7. The PZR

is equipped with nozzles for sprays, a surge, pilot operated safety relief valves (POSRVs),

and pressure and level instrumentation. The PZR of APR1400 has increase several

improvement on operational reliability and maintenance, increase PZR capacity, improve

capability against transient. The main function of POSRV is to have both overpressure

protection and safe depressurization. By discharging the fluid in IRWST through sparger,

it can minimize contamination in containment.

FIGURE 3.7 : APR1400 Pressurizer

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3.5 Steam Generator

AP1000

The AP1000 steam generator (SG) is a vertical shell and U-tube evaporator with integral

moisture separating equipment. The AP1000 SG is shown in figure 3.8. Two model

Delta-125 steam generators are used in AP1000. There are some design enhancements on

the AP1000 SG which include nickel-chromium-iron Alloy 690 treated tubes on a

triangular pitch, improved anti-vibration bars, single-tier separators, improved

maintenance features that allows easy access by robotic tooling during maintenance. The

main function of the AP1000 SG is to transfer heat from single-phase reactor coolant

water through U-shaped heat exchanger tubes. The steam generator separates dry and

saturated steam from the boiling mixture, and delivers the steam to a nozzle that will end

up in turbine. Water from the feed water system refills the SG water inventory through

the SG ‘s feed water inlet nozzle. In addition, the secondary side of SG provides water

inventory which will continuously available to absorb heats at the primary side.

FIGURE 3.8 : AP1000 Steam Generator

69

APR1400

The SG is a vertically inverse U-tube heat exchanger with moisture separators, steam

dryers, and an integral economizer. The APR1400 SG is shown in figure 3.9. It operates

with the RCS coolant on the tube side and the secondary coolant on the shell side. The

increased in feed water inventory of the SG enhances plant safety and reduces the number

of unplanned reactor trips. In addition, the primary outlet nozzle angle of the SG is

modified so that it will improve on the stability during mid-loop operation. The SG tube

reliability is enhanced by the following design improvements:

Inconel 690, which is known to be a corrosion-resistant material, is used as the SG

tube material.

The upper tube support bar and plate are designed to prevent vibration due to the

flow of water.

Automatic control of SG water level for all operating ranges.

FIGURE 3.9 : APR1400 Steam Generator

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3.6 Safety System

AP1000

The safety systems for AP1000 include passive safety injection, passive residual heat

removal, and passive containment cooling. All these passive systems meet the NRC

regulatory and standards. The simplification of plant systems result to reduced actions

required by the operator if an accident occurs. Passive systems uses only natural forces

such as gravity, natural circulation, and compressed gas where all these are simple

physical principles we rely on every day. There are no pumps, fans, diesels, chillers, or

other rotating machinery required for the safety systems thus this eliminates the need for

safety-related AC power sources. Since there are no safety-related pumps, the increased

flow was achieved by increasing pipe size. Additional water volumes were achieved by

increasing tank sizes. These increases were made while keeping the plant footprint

unchanged.

The passive core cooling system (PXS) shown in figure 3.11 uses three sources of

water to maintain core cooling through safety injection. These injection sources include

the core makeup tanks (CMTs), the accumulators, and the in-containment refueling water

storage tank (IRWST). These injection sources are directly connected to two nozzles on

the reactor vessel. Long-term injection water is provided by gravity force from the

IRWST, which is located in the containment just above the RCS loops. Usually, the

IRWST is isolated from the RCS by squib valves and check valves. IRWST is designed

for atmospheric pressure. The RCS must be depressurized before the injection can occur.

The RCS is automatically controlled to reduce pressure to around 0.83 bars, when the

level of water in the IRWST overcomes the low RCS pressure or the pressure loss in the

injection lines.

The PXS includes one passive residual heat removal heat exchanger (PRHR HX).

The function of PRHR HX is to protects the plant against transients that will damage the

normal steam generator and feed water systems. The IRWST provides the heat sink to

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absorb heat generated for the PRHR HX. The IRWST water absorbs emitted heat for

more than one hour before the water begins to boil. Once boiling starts, steam passes

through the containment. The steam condenses on the steel containment vessel and drains

back into the IRWST by gravity force after collection. The PRHR HX and the passive

containment cooling system provide continuous heat removal capability with no operator

action required.

The passive containment cooling system (PCS) shown in figure 3.10, provides the

safety-related ultimate heat sink for the plant. The PCS cools the containment if an

accident happens so that the design pressure will not exceed the limits. Besides that the

pressure will also reduced rapidly by the PCS. Heat is removed from the containment

vessel by the continuous, natural circulation of air. During an accident, air-cooling is

supply by water evaporation. The water drains by gravity force from a tank located on top

of the containment shield building. In addition, even with failure of water drain, air-only

cooling is capable of maintaining the containment below the predicted failure pressure.

FIGURE 3.10 : AP1000 Passive Containment Cooling System

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FIGURE 3.11 : AP1000 Passive Core Cooling System

APR1400

To improve plant safety, severe accidents have been fully considered in the APR1400

73

design. The measures of the APR1400 to cope with severe accidents are divided into

prevention and mitigation. Severe accident prevention features are summarized as

follows:

Increased design margins such as a larger PZR, larger SGs, and an increased

thermal margin

Reliable engineered safety features (ESF) including the SIS, the AFWS, and the

CSS

Extended ESFs such as the SDVS with IRWST, alternate AC power, and a diverse

protection system

Containment bypass prevention

Severe accident mitigation features are summarized as follows:

Hydrogen mitigation system (HMS) such as a passive autocatalytic recombiner

and a glow plug igniter

Reactor cavity and cavity cooling system

External reactor vessel cooling system

The SDVS and the IRWST

Emergency containment spray backup system

Robust containment with a large volume

Therefore the safety systems of APR1400 consist of the safety injection system (SIS) as

shown in figure 3.12, the in-containment refueling water storage tank (IRWST), the

safety depressurization and vent system (SDVS), the containment spray system (CSS),

and the auxiliary feed water system (AFWS). The main design concept of the SIS is

simplification to achieve higher reliability and better performance. The SIS is composed

of four independent mechanical trains and two electrical divisions. Each train has one

active safety injection pump (SIP) and one passive safety injection tank (SIT) equipped

with a fluidic device (FD). Additionally, the SIS is designed for safety water to be

injected directly into the reactor vessel.

The IRWST is located in the containment building and the arrangement is made so that

the injected emergency cooling water will returns to the IRWST. This design does not

74

require operator action to switch the SIP suction from the IRWST to the containment

recirculation reservoir. This new design lowers the susceptibility of the IRWST to

external hazards. The functions of the IRWST are as follows;

The storage of refueling water.

A water source for the SIS, the SCS, and the CSS.

A heat sink to condense steam that discharged from the PZR for rapid

depressurization if necessary in order to prevent high pressure core melting.

A coolant supply for the cavity flooding system as shown in figure 3.13 in case of

severe accidents in order to protect the core against melting

The SDVS is a dedicated safety system designed to provide safety when depressurizing

the RCS. SDVS will function if the PZR spray is unavailable during plant cool down or a

cold shutdown. The CSS is composed of two trains and takes the suction from the IRWST

by its pump to reduce the temperature and pressure of the containment during accidents

that occur in the containment. The CSS was designed to be interconnected with the SCS

and the pumps of the CSS. SCS are designed to have the same type and capacity as the

CSS.

The AFWS is designed to supply feed water to the SGs for RCS as heat removal in a case

of failure in main feed water supply. In addition, the AFWS refills the SGs following a

LOCA to minimize leakage. The AFWS consist of two motor-driven pumps, two turbine-

driven pumps and two independent safety-related emergency feed water storage tanks

located in the auxiliary building increases the performance reliability of the AFWS.

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FIGURE 3.12 : APR1400 Safety Injection System (SIS)

FIGURE 3.13 : APR1400 Cavity Flooding System

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3.7 Plant Layout

AP1000

A typical site plan for a single unit AP1000 is shown in figure 3.14 below. The power

block complex consist of five principal building structures which are the nuclear island,

the turbine building, the annex building, the diesel generator building and the radwaste

building. Each of these building structures are constructed on individual basemats. The

nuclear island consists of the containment building, the shield building, and the auxiliary

building, all of which are constructed on a common basemat.

The plant arrangement of the AP1000 consist of the containment that contains a

4.9 meter diameter main equipment hatch and a personnel airlock at the operationg deck

level and a 4.9 meter diameter maintenance hatch and a personnel airlock at grade level.

These large hatches significantly enhance accessibility to the containment during outages

and consequently reduce the potential for congestion at the containment entrances. These

containment hatches located at two different levels, allow activities occurring above the

operating deck to be unaffected by activities occurring below the operating deck. The

containment arrangement provides significantly larger laydown areas than most

conventional plants at both the operating deck level and the maintenance floor level.

Additionally, the auxiliary building and the adjacent annex building provide large staging

and laydown areas immediately outside of both large equipment hatches.

The containment building is the containment vessel and all structures contained

within the containment vessel. The containment building is an integral part of the overall

containment system with the functions of containing the release of airborne radioactivity

following postulated design basis accidents and providing shielding for the reactor core

and the reactor coolant system during normal operations. The containment vessel is an

integral part of the passive containment cooling system. The containment vessel and the

passive containment cooling system are designed to remove sufficient energy from the

containment to prevent the containment from exceeding its design pressure following

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postulated design basis accidents. The principal systems located within the containment

building are the reactor coolant system, the passive core cooling system, and the reactor

coolant purification portion of the chemical and volume control system.

The shield building is the structure and annulus area that surrounds

thecontainment vessel. During normal operations the shield building, in conjunction with

theinternal structures of the containment building, provides the required shielding for the

reactor coolant system and all the other radioactive systems and components housed in

the containment. During accident conditions, the shield building provides the required

shielding for radioactive airborne materials that may be dispersed in the containment as

well as radioactive particles in the water distributed throughout the containment. The

shield building is also an integral part of the passive containment cooling system. The

passive containment cooling system air baffle is located in the upper annulus area. The

function of the passive containment cooling system air baffle is to provide a pathway for

natural circulation of cooling air in the event that a design basis accident results in a large

release of energy into the containment. In this event the outer surface of the containment

vessel transfers heat to the air between the baffle and the containment shell. This heated

and thus, lower density air flows up through the air baffle to the air diffuser and cooler

and higher density air is drawn into the shield building through the air inlet in the upper

part of the shield building. Another function of the shield building is to protect the

containment building from external events. The shield building protects the containment

vessel and the reactor coolant system from the effects of tornadoes and tornado produced

missiles.

The primary function of the auxiliary building is to provide protection and

separation for the safety-related seismic Category I mechanical and electrical

equipmentlocated outside the containment building. The auxiliary building provides

protection for thesafety-related equipment against the consequences of either a postulated

internal or externalevent. The auxiliary building also provides shielding for the

radioactive equipment and pipingthat is housed within the building. The most significant

equipment, systems contained within theauxiliary building are the main control room,

I&C systems, electrical power systems, fuelhandling area, mechanical equipment areas,

78

containment penetration areas, and the mainsteam and feedwater valve compartments.The

primary function of the auxiliary building is to provide protection and separation for

thesafety-related seismic Category I mechanical and electrical equipment located outside

thecontainment building. The auxiliary building provides protection for the safety-

relatedequipment against the consequences of either a postulated internal or external

event. Theauxiliary building also provides shielding for the radioactive equipment and

piping that ishoused within the building.The most significant equipment, systems, and

functions contained within the auxiliary building are the following:

Main control room

Class 1E instrumentation and control systems

Class 1E electrical system

Fuel handling area

Mechanical equipment areas

Containment penetration areas

Main steam and feedwater isolation valve compartment

The main control room provides the human system interfaces required to operate

the plant safely under normal conditions and to maintain it in a safe condition

underaccident conditions. The main control room includes the main control area, the

operations staffarea, the switching and tagging room and offices for the shift supervisor

and administrativesupport personnel.

Instrumentation and Control Systems is the protection and safety monitoring

system and theplant control system provide monitoring and control of the plant during

startup, ascent to power,powered operation, and shutdown. The instrumentation and

control systems include theprotection and safety monitoring system, the plant control

system, and the data display andprocessing system.

The Class 1E system provides 125 volts dc power for safetyrelated and vital

control instrumentation loads including monitoring and control roomemergency lighting.

79

It is required for safe shutdown of the plant during a loss of ac power andduring a design

basis accident with or without concurrent loss of offsite power.

The primary function of the fuel handling area is to provide for the handling and

storage of new and spent fuel. The fuel handling area in conjunction with the annex

building provides the means for receiving, inspecting and storing the new fuelassemblies.

It also provides for safe storage of spent fuel as described in DCD Section 9.1,Fuel

Storage and Handling.The fuel handling area provides for transferring new fuel

assemblies from the new fuel storagearea to the containment building and for transferring

spent fuel assemblies from thecontainment building to the spent fuel storage pit within the

auxiliary building.The fuel handling area provides the means for removing the spent fuel

assemblies from thespent fuel storage pit and loading the assemblies into a shipping cask

for transfer from thefacility. The fuel handling area is protected from external events such

as tornadoes and tornadoproduced missiles. Protection is provided for the spent fuel

assemblies, the new fuelassemblies and the associated radioactive systems from external

events. The fuel handlingarea is constructed so that the release of airborne radiation

following any postulated designbasis accident that could result in damage to the fuel

assemblies or associated radioactivesystems does not result in unacceptable site boundary

radiation levels.

The mechanical equipment located in radiological control areasof the auxiliary

building are the normal residual heat removal pumps and heat exchangers, thespent fuel

cooling system pumps and heat exchangers, the solid, liquid, and gaseous

radwastepumps, tanks, demineralizers and filters, the chemical and volume control

pumps, and theheating, ventilating and air conditioning exhaust fans.The mechanical

equipment located in the clean areas of the auxiliary building are the heating,ventilating

and air conditioning air handling units, associated equipment that service the maincontrol

room, instrumentation and control cabinet rooms, the battery rooms, the

passivecontainment cooling system recirculation pumps and heating unit and the

equipmentassociated with the air cooled chillers that are an integral part of the chilled

water system

80

The auxiliary building contains all of the containmentpenetration areas for

mechanical, electrical, and instrumentation and control penetrations. Theauxiliary

building provides separation of the radioactive piping penetration areas from the

nonradioactivepenetration areas and separation of the electrical and instrumentation and

control penetration areas from the mechanical penetration areas. Also provided is

separation ofredundant divisions of instrumentation and control and electrical

equipment.The main steam and feedwaterisolation valve compartment is contained within

the auxiliary building. The auxiliary buildingprovides an adequate venting area for the

main steam and feedwater isolation valvecompartment in the event of a postulated leak in

either a main steam line or feedwater line.The annex building provides the main

personnel entrance to the powergeneration complex. It includes accessways for personnel

and equipment to the clean areas ofthe nuclear island in the auxiliary building and to the

radiological control area. The buildingincludes the health physics facilities for the control

of entry to and exit from the radiologicalcontrol area as well as personnel support

facilities such as locker rooms. The building alsocontains the non-1E ac and dc electric

power systems, the ancillary diesel generators and theirfuel supply, other electrical

equipment, the technical support center, and various heating,ventilating and air

conditioning systems. No safety-related equipment is located in the annexbuilding.The

annex building includes the health physics facilities and provides personnel and

equipmentaccessways to and from the containment building and the rest of the

radiological control areavia the auxiliary building. Provided are large, direct accessways

to the upper and lowerequipment hatches of the containment building for personnel

access during outages and forlarge equipment entry and exit. The building includes a hot

machine shop for servicingradiological control area equipment. The hot machine shop

includes decontamination facilitiesincluding a portable decontamination system that may

be used for decontamination operationsthroughout the nuclear island.The diesel generator

building houses two identical slide along dieselgenerators separated by a three-hour fire

wall. These generators provide backup power forplant operation in the event of disruption

81

of normal power sources. No safety-related equipmentis located in the diesel generator

building.

The radwaste building includes facilities for segregated storage of

variouscategories of waste prior to processing, for processing by mobile systems, and for

storingprocessed waste in shipping and disposal containers. No safety-related equipment

is located inthe radwaste building. Dedicated floor areas and trailer parking space for

mobile processingsystems is provided for the following:

− Contaminated laundry shipping for offsite processing

− Dry waste processing and packaging

− Hazardous/mixed waste shipping for offsite processing

− Chemical waste treatment

− Empty waste container receiving and storage

− Storage and loading packaged wastes for shipment

1. containment2. Turbine3. Annex4. Auxiliary5. Cooling tower7. Radwaste10. Diesel generator13. Fire water storage tank16. Transformer17. Condensate storage tank18. Diesel generator oil storage

tank19. Dematerialized water storage

tank20. Boric acid storage tank22. Turbine laydown area25. Passive containment cooling26. Diesel driven fire pump

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FIGURE 3.14 : AP1000 Plant Layout

APR1400

The general arrangement of the APR1400 was developedbased on the twin-unit concept

using a slide-alongarrangement with common facilities. The layout of the APR1400 can

be divided into a nuclearisland (NI) and a turbine island (TI). The NI consists of the

reactor containment building (RCB), the auxiliary building (AB), and the compound

building (CB). The TIconsists of the turbine building (TB) and the switchgearbuilding

(SB).The RCB is wrapped around by the AB and is founded on a common basemat with

the AB. The AB accommodatesemergency diesel generators (EDGs) and the fuel

handlingarea (FHA). The layout of the AB, particularly the physicalseparation of the

safety equipment, is designed to improveplant safety. As examples, four-train of safety

injection system (SIS) and two sets ofEDGs are arranged so that each one is placed in a

physicallyseparated division of the AB. This configuration designprevents the

propagation of system damage by internaland external events such as fire, flooding,

security incidents,and sabotage. Other internal structures are also arrangedto improve

maintainability, accessibility, and convenienceof equipment replacement.

The layout of the NI improvesthe structural safety margin against external events

suchas a seismic event.The RCB of the APR1400 is a pre-stressed concretestructure in the

shape of a cylinder with a hemispherical dome specified as seismic category I. It is placed

on acommon basemat with the AB. Theinterior surface of the RCB is steel-lined for leak-

tightness.A protective layer of concrete covers the portion of theliner over the foundation

slab. The IRWST is situated inthe RCB in an annular-shape configuration between

thesecondary shield wall and the containment wall. The safety injection pump

(SIP)always take water from the IRWST without switching itssuction from the IRWST to

the containment sump forlong-term cooling following a leak of coolant accident (LOCA).

As measures to mitigate severe accidents, the reactorvessel cavity is designed in a

manner that allows the moltencore materials to spread out so that the heat transfer area

isnot less than 0.02 m2/MW and so that these materials arecooled and solidified on the

cavity floor. In addition, theconvoluted vent path of the reactor vessel cavity

83

preventsmolten core debris from being released into the containmentatmosphere. In order

to improve the convenience ofmaintenance, an equipment hatch, the structural

arrangement,and a polar bridge crane are designed so that an steam generator (SG) canbe

replaced in one piece. Work platforms are installed toenhance the convenience of in-

service inspections of theSGs and maintenance of the reactor coolant pump (RCP).

The AB is a reinforced concrete structure specified asseismic category I. It wraps

around the RCB in a quadrantarrangement. The AB houses the main control room

(MCR),EDGs room, FHA, and the various components related tosafety, such as the

SIS.The systems and internal structures in the AB arearranged to provide physical

separation so as to minimizethe danger from internal and external events such as fireand

flooding without adversely affecting accessibility. Toimprove the actuation reliability, the

safety equipment isspatially separated. Each train of the SIS which consistsof four trains

is located in a separate division. The EDGsare also separated on opposite sides. The

internal layoutof the AB is designed to provide sufficient space and alifting rig to replace

heat exchangers and to replace agenerator of the EDG without removing the outer

wall.This design improves the convenience of operation andmaintenance. The internal

arrangement of components isdivided into a radiation area and a clean area to reducethe

occupational exposure dose.

The TI consists of the TB and the SB arranged in adirection radial to the RCB.

Both buildings are situated ona common basemat and are designed with a steel

structureand a reinforced concrete turbine pedestal specified asseismic category II. The

TB encloses the components thatconstitute the heat cycle and produce the electricity.

TheSB houses the electrical distribution equipment. To reducethe construction schedule,

an underground common tunnelis designed to accommodate underground facilities in

thebase floor of the TB. In addition, demineralizers arearranged at the same level for

effective maintenance.As a common facility for both units, the CB is designedwith a

reinforced concrete structure specified as seismiccategory II. It accommodates an access

control area, aradwaste treatment area, primary and secondary samplinglaboratories, and

a hot machine shop. This arrangementmakes access from each unit more convenient and

contributesto reducing the size of the power block due to its compactdesign.

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FIGURE 3.15 : APR1400 Common Basemat

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CHAPTER 4

CONCLUSION

4.1 Conclusion

There are several similarities and differences between AP1000 reactor and APR1400

reactor. AP1000 are from the Westinghouse U.S. while APR1400 are from the Korea

Hydro and Nuclear Power.

Their similarities are the plant design life are 60 years, the design of the

pressurizer and steam generator. The material used to manufacture the steam generator

are Inconel 690. Both designs come from the basic Pressurized Water Reactor (PWR)

design.

AP1000 reactor is more considerable if we are comparing on construction cost and

safety wise due to its plants simplification which reduces overall construction time and

cost and its safety system which uses passive safety system. Passive safety system means

no pumps, diesel or AC related source for the safety system, it only uses force of nature

such as gravity and circulation of air. Hence this reduces the cost of maintenance and

reduces the risk of safety equipment failure. There are spring loaded valves are used in

the AP1000 and this reduces the dependability on operator.

APR1400 reactor is considerable if we are comparing in the aspect of generating

capacity and safety valves that can greatly reduce the risk of accident happens in the

reactor. The generating capacity of APR1400 are 1400 MWe which are 400 MWe

86

comparing to AP1000. The safety valve which are the Pilot Operated Safety Relief

Valves which replace the spring loaded vales that are used by AP1000 can greatly reduce

the risk of malfunction when accident happens. The control room will detect a signal

from the Pilot Operated Safety Relief Valves and hence notify the operator to manual

override the Pilot Operated Safety Relief Valves if automatic system failure.

Generally AP1000 have more advantages compare to APR1400 in the aspect of

cost of construction, cost of production, time of construction and safety wise. AP1000

will have more market value compare to APR1400 because in the future, safety will be

the main concern in the world.

4.2 Future Work

This project is about the comparative study on nuclear reactor technology for nuclear

power. The main challenges in doing this project is the lack of information on reactor

technology due to it is a national security issue in the world. For future work, the study on

advanced nuclear reactor in Generation IV can be done and more information should be

gathered to make the project more informative for the references of student in Universiti

Tenaga Nasional or for the public.

87

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[13] Jacques Bouchard, Ralph Bennett. September-October 2008. Nuclear

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APPENDICES

90

APPENDIX

Model of Reactor Coolant System

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