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Nuclear Energy Fundamentals

Module 4: Nuclear Reactor Design

PREPARED BY

Academic Services

April 2012

© Institute of Applied Technology, 2012

ATM 1236 – Nuclear Energy Fundamentals

2 Module 4: Nuclear Reactor Design

Module 4: Nuclear Reactor Design Module Objectives After the completion of this module, the student will be able to:

Explain the fuel assembly geometry, specifications and material.

Describe the fuel assembly forms.

Explain the refueling types, frequency.

Explain the active dimensions of the nuclear reactor.

Explain the function of the moderator and the material used for

moderating nuclear reactors.

Explain the function of the reflector and the material used as

reflectors in nuclear reactors.

Describe the control system of some common types of nuclear

reactors.

Explain the principle of operation of the cooling system of some

common types of nuclear reactors.

Explain the function of the RCP.

Identify the main components of the RCP.

Identify the different types of radioactive barriers used in nuclear

reactors.

Describe the construction of the pressure vessel of BWR.

Describe the construction of the pressure vessel and the

containment building for some reactor types.

Describe the construction and the types of the steam generators.

Explain the function of the steam generator.

Describe the construction the pressurizer.

Explain the function of the pressurizer.

Explain the principle of operation of the different safety systems and

sub-systems used in NPPs including (RPS, ESWS. ECCS, Emergency

electrical systems and ventilation and radiation protection).

Explain the way to initiate the nuclear reaction in the reactor.


Module 4: Nuclear Reactor Design 3

Module 5: Nuclear Reactor Design Module Contents Topic Page No.

1. Introduction 4

2. Fuel Assembly 4

3. Moderator 8

4. Reflector 9

5. The Chain Reaction Control 9

6. The Cooling System 12

7. Protective Shield 16

8. Steam Generator 22

9. Pressurizer 24

10. Nuclear Reactor Safety Systems 25

11. Activities 31

12. References 32



1. Introduction

In this module we will focus on the primary systems and the basic

functional requirements of nuclear reactors. These include the reactor core,

reactor vessel, reactivity control, reactor coolant system, steam generators

(SG), pressurizer, and the reactor safety and protection systems.

2. Fuel Assembly

The reactor core is the main part containing the nuclear fuel. The solid fuel

material is fabricated into various small shapes called, pellets which are

usually put together and called as assemblies or bundles.

A reactor core may contain from tens to hundreds of these fuels

assemblies, held in a fixed geometrical pattern.

2.1. Fuel assembly geometry

The individual fuel rods are arranged in assemblies where there are three

basic types of fuel geometry. These are square which is used in most

reactors, hexagonal is used in VVER (a PWR developed in 1970 in the

Soviet Union) and cylindrical that is used in CANDU.

(a) (b)

(c) (d)

Fig. 5.1: Fuel assembly geometry: a) square b) cylindrical c) hexagonal d) spherical.



Some high temperature reactors such as pebble bed reactor which is one of

the promising nuclear reactor technologies known today use spherical fuel

geometry (Fig. 5.1).

In PWR for example, the fuel assembly consists of a square array of 179 to

264 fuel rods which is a long, slender tube in which nuclear fuel is

surrounded by cladding material and inserted into a reactor. There are

three types of cladding material, namely, zirconium alloy (Zircaloy),

stainless steel and magnesium alloy (Magnox). The fuel rods are assembled

into bundles called fuel elements or fuel assemblies, which are loaded

individually into the reactor core and 121 to 193 fuel assemblies are loaded

into an individual reactor.

Fig. 5.2: The PWR fuel assembly specifications.



Fig. 5.2 shows some size specifications of the PWR fuel assembly. The

cladding tube contains around 350 to 400 pellets with both ends plugged.

Those pellets are fixed with a spring.

The table below shows the basic specifications of PWR fuel assembly. As

indicated in Fig. 5.2 and table 5.1, the cross section size is about 21 cm and

the fuel assembly length is around 4 m.

Table 5.1: PWR fuel assembly specifications.

Type 14×14 15×15 17×17

10ft 12ft 12ft 12ft

Cross-section size (mm2) 197 214 214

Fuel assembly length (mm) 3,473 4,057 4,057 4,058

Fig. 5.3: The PWR fuel assembly specifications.



A modern BWR fuel assembly comprises 74 to 100 fuel rods, and there are

up to approximately 800 assemblies in a reactor core (Fig. 5.3). The

number of fuel assemblies in a specific reactor is based on considerations of

desired reactor power output, reactor core size and reactor power density.

2.2. Fuel material

The fuel material is the material from which the fuel elements are made.

Typical fuel materials are uranium metal (U metal) and uranium dioxide

(UO2); however, the material could be also a mixture of uranium dioxide

and plutonium dioxide (PuO2) and thorium dioxide (ThO2).

2.3. Fuel form

Nuclear fuel can be in many forms. It can be in the form of a chemical

compound, ceramic, or metal alloy. It can be pellets, rods, or even

dissolved in liquid salt or liquid lead.

2.4. Refueling type

Refueling is the method by which the

used fuel assemblies in the core are

replaced by fresh ones. Basically, the

fuel may be replaced either individually

or in small groups during reactor

operation on-load (at rated reduced

power), or with significant portion while

the reactor is off-load (during refueling

outages). On-load refueling is typical for

the CANDU or GCR (Magnox) reactors.

AGR may be refueled either at reduced

load (still on load) or off load. Other

reactors such as PWR, BWR and FBR are

refueled off-load (Fig. 5.4).

Fig. 5.4: A nuclear fuel assembly

being loaded into the

reactor core.



2.5. Refueling frequency

Refueling frequency or sometimes called the fuel cycle is defined as the

time period in which a significant part of the core is refueled. This

characteristic applies only to reactors with off-load refueling. The fuel cycle

length is the average time period in months from the end of one refueling

to the end of the next one. Most common refueling frequencies are 12, 18

or 24 months when a quarter to a third of the fuel assemblies is replaced

with fresh ones.

2.6. Active core diameter and height/length

The active core diameter is the diameter of the circle encompassing the

active fuel assemblies in the core. Excluding the reflector or reactor vessel

shielding. Usually the reactor vessel varies from 2 to 10 meters but it can

be more than 10 meters in some reactors such as GCR (Magnox).

The active core height is the active part

of the core excluding structural

components and support. The name

height or length is based on the actual

fuel orientation (vertical or horizontal)

in the reactor core. Usually, the core

height varies from 2 to 5 meters, but it

may be also between 5 and 7 meters.

Some GCR (AGR) reactors have core up

to 8 meters height.

Fuel

Fig. 5.5: cross-sectional view of a

nuclear reactor.

3. Moderator

A moderator or neutron moderator is a material that reduces the speed of

fast neutrons, thereby turning them into thermal neutrons capable of

sustaining a nuclear chain reaction involving uranium-235. The moderator

is used in thermal reactor and the materials used as moderators include



ordinary water, heavy water, graphite, beryllium and certain organic

compounds.

The moderator should be well distributed within the fuel zone or core. In

some reactors the fuel materials and moderator materials are mixed

together.

4. Reflector

The reflector reduces the leakage of neutrons by reflecting back the

neutrons escaping from the core. The same material used for moderator

can be used for the reflectors in the case of thermal reactors. The light

water, heavy water and carbon are mostly used as reflectors.

The use of a proper reflector helps to reduce the size of the reactor core for

a given power output since the number of neutrons leaking are lesser and

help to propagate the fission process instead. It also reduces the

consumption of the fissile material.

In the fast reactors where fast neutrons are utilized for fission, nickel,

molybdenum and stainless steel reflectors are used.

5. The Chain Reaction Control

Control rods are used to control the nuclear reactor reactivity and power.

These are made from neutron-absorbing material such as silver, indium

hafnium and cadmium. Other elements that can be used include boron,

cobalt, dysprosium, gadolinium, samarium, erbium, and europium, or their

alloys and compounds. Control rods are inserted or withdrawn from the

core to control the rate of reaction, or to stop it (Fig. 5.6).

In some PWR reactors, special control rods are used to enable the core to

sustain a low level of power efficiently. Secondary shutdown systems

involve adding other neutron absorbers, usually boric acid as a fluid, to the

system.



Fig. 5.6: The mechanism of using control rods to control the reaction.

Besides the control rods, this system includes a number of devices, sensing

elements that measure the number of neutrons in the reactor, and other

devices to regulate the position of the control rods. The control rods when

lowered into the reactor absorb the neutrons to reduce the neutron

population and when raised allow the rise in number of neutrons. In some

reactors the reaction is controlled by varying the level of moderator. In the

heavy water moderated reactors like CANDU, a combination of moderator

level control and neutron absorber rods are used.

Control rods are usually combined into fuel rod assemblies (Fig. 5.7). In

commercial PWR for example, 20 control rods are used per assembly. These

are inserted into guide tubes within the fuel element. A control rod is

removed from or inserted into the central core of a nuclear reactor in order

to control the neutron flux. This in turn affects the thermal power of the

reactor, the amount of steam produced, and hence the electricity

generated. Control rods often oriented vertically within the core. In PWRs,

the control rod drive mechanisms are mounted on the reactor pressure

vessel head and the rods are inserted from above (Fig. 5.8).



Control rod guide thimble

Instrumentation guide thimble

Grid

Fuel rod

Bottom nozzle

Top nozzle

Control rods

Fig. 5.7: The control rod combined into PWR fuel rod assemblies.

While in BWRs the control rods are inserted from underneath the core due

to the necessity of having a steam dryer above the core (Fig. 5.9).



Fig. 5.8: In PWR the control rods are inserted from above.

Fig. 5.9: In BWR the control rods are inserted from the bottom.

6. The Cooling System

The rods must be surrounded by coolant (liquid or gas); otherwise

temperatures can rise to levels hot enough to melt metallic components

over a prolonged period. This opens the possibility of a serious meltdown, in

which molten, highly radioactive material from the reactor core falls

through the floor of the containment vessel and into the ground below.

This system removes the heat released from the reactor core. It consists of

pipes through which the coolant is pumped. When passing through the

reactor cores, the coolant picks up the heat, transfers the heat to another

working medium through a heat exchanger and then returns to the reactor.

Gases, heavy and light water, and liquid metals such as sodium, lithium,

potassium, can serve as coolants. In a reactor, we must be able to control

the amount of heat produced. The heat produced depends upon the

number of fissions taking place per second in the reactor, which in turn

depends upon the number of neutrons present in the reactor.

In a PWR for example (Fig. 5.10), water in the reactor core (primary loop)

reaches about 325 °C; hence it must be kept under pressure of about 16

MPa to prevent it from boiling.



The secondary loop is under less pressure (6 MPa) and the water here boils

in the steam generators. The temperature of the steam leaving the steam

generator is 280 °C. The steam drives the turbine to produce electricity,

and is then condensed and returned to steam generator in contact with the

primary loop (circuit).

The water entering the condenser from the cooling tower or sometimes

from a large body of water (tertiary loop) draws heat from the steam that

leaves the turbine and its temperature rises from 25 °C to reach 35 °C

when returning back to cooling tower or the water body.

Primary loop 330 ºC, 16 MPa

Reactor pressure vessel

Steam 280 °C, 6 MPa

Steam generator

Condenser

Reactor Coolant pump

Water pump

25 ºC

Generator35 ºC

Turbine

Secondary loop

Tertiary loop

Fig. 5.10: Cooling system in PWR.

Compared to the PWR the BWR (Fig. 5.11), has only a single circuit in

which the water is at lower pressure (about 8 MPa) so that it boils in the



core at about 285 °C. The reactor is designed to operate with 12-15% of

the water in the top part of the core as steam.

The steam passes through the steam separator to the a steam drier plates

above the core and then directly to the turbines, which are thus part of the

reactor circuit. Since the water around the core of a reactor is always

contaminated with traces of radionuclides, it means that the turbine must

be shielded and radiological protection provided during maintenance. The

cost of shielding the turbine tends to balance the savings due to the simpler

design. Most of the radioactivity in the water is very short-lived, so the

turbine hall can be entered soon after the reactor is shut down.

Feedwater in from the condenser

Steam out to the turbine

Reactor coolant pump

Reactor coolant pump

Reactor core

Feedwater inlet

Steam drier

Steam separator

Steam outlet



Fig. 5.11: Cooling system in BWR, 3D and 2D illustration.

6.1. Reactor coolant pump (RCP)

The nuclear reactor coolant pump is one of the main components of the

nuclear reactor cooling system. It provides the circulating force of reactor

coolant to help in transferring heat energy through the different cooling

system parts. It also provides the driving force of spray water inside of the

pressurizer. Fig. 5.12a shows the main parts of the reactor coolant pump

and Fig, 5.12b shows a real picture of the pump.

Flywheel

Reactor coolant section nozzle

CasingImpeller

Motor rotor Motor stator

Reacor coolant dischare nozzle

Motor

(a)



(b)

Fig. 5.12: a) RCP main components. b) Real picture of the RCP.

7. Protective Shield

The fission reaction is accompanied by emission of radiation like , and .

Exposure to these radiations is dangerous. In order to protect the persons

working near the reactor from these harmful radiations the reactor is

enclosed in steel and concrete which are capable of stopping these

radiations. This arrangement of protection is called Radiation shielding.

In nuclear reactors, the following components or systems are playing a role

as barriers to radioactive release:

The fuel ceramic.

The metal fuel cladding tubes.

The coolant system.

The reactor vessel.

The containment building.

Where the containment is the final barrier to the radioactive release. In the

following sections we will discuss the later two.

7.1. Reactor pressure vessel

Usually a strong steel vessel containing the reactor core, moderator and

coolant (Fig. 5.13), but it may be a series of tubes holding the fuel and

conveying the coolant through the moderator. In a typical PWR, the reactor

pressure vessel is about 13.5 m high and about 4.4 m inside diameter, and

has wall thickness exceeding 22 cm. The active length of the fuel



assemblies may be in the range of 4 m.

7.2. Containment building

The containment building is a steel or reinforced concrete structure

enclosing a nuclear reactor. It is designed, in any emergency, to contain the

escape of radiation. The containment is the final barrier to radioactive

release.

A typical containment building is a steel structure enclosing the reactor




Pressure vessel head

Fig. 5.13: BWR pressure vessel and internal components.

vessel and normally sealed off from the outside atmosphere. The

containment is designed to contain or condense steam to a maximum

pressure in the range of 410 to 1400 kPa. This done as a short term

solution but for large break accidents other systems are used for long term

heat removal.



7.2.1. Types

Since the most common nuclear power plants are the PWRs and the BWRs

we will focus on the structure of these two types containment building.

Polar crane

Steel containment liner

Reinforced concrete containment

U‐tube steam generator

Main steam line

Main coolant pump

Upper internal

Reactor core

Control rod drive mechanisms

Fuel transfercanal



Fig. 5.14: PWR reactor containment building.

7.2.1.1. Pressurized water reactors containment

For a pressurized water reactor, the containment also encloses the steam

generators and the pressurizer, and is the entire reactor building (Fig.

5.14). PWR containments are typically 10 times larger than a BWR and in

most PWR designs. The spent fuel pool is located outside of the

containment building.

Modern designs have also shifted more towards using steel containment

structures. In some cases steel is used to line the inside of the concrete,

which contributes strength from both materials. In case of accidents the

containment becomes highly pressurized, yet other newer designs call for

both a steel and concrete containment.

The containment building can be classified based on the shape, size,

generation, etc. Fig. 5.16 shows three different containments designs based

on their shapes. These are the can design, the spherical design and the

combined design.

(a) (b) (c)

Fig. 5.15: Three different shapes of the PWR containment building; a) Can

design b) Spherical design d) Combined design.

7.2.1.2. Boiling water reactors

In BWR's (Fig. 5.16) the containment consists of a drywell where the



reactor and cooling equipment is located and a wetwell that is also known

as a torus or suppression pool. During accidents, the reactor coolant

converts to steam in the drywell and the pressure builds up quickly. Vent

pipes or tubes from the drywell direct the steam below the water level in

the wetwell. This condenses the steam and reduces the pressure. Both the

drywell and the wetwell are enclosed by a secondary containment building

(Fig. 5.17) which is maintained at a slight negative pressure during normal

operation and refueling operations.

Wetwell

Drywell


Spent fuel pool

Secondary concrete shield wall

Fig. 5.16: Cross-section sketch of a typical BWR containment.

In some BWR designs the spent fuel pool is inside the containment but in

most of them it is located outside of the containment building.



The BWR containment shape looks like a cuboid which is very different from

PWR shape (Figs. 5.17 & 5.18). Since the steam going through the turbines

is coming directly from the reactor it is still slightly radioactive the turbine

building has to be considerably shielded as well. This leads to two buildings

of similar construction with the taller one housing the reactor and the short

long one housing the turbine hall and supporting structures.

Secondary containment

Primary containment

Fig. 5.17: Primary and secondary containment buildings in BWRs.

Fig. 5.18: The BWR containment building design for the reactor and turbine.



8. Steam Generator (SG)

Steam generators are heat exchangers used to convert water into steam

from heat produced in a nuclear reactor core. They are used in pressurized

water reactors between the primary and secondary coolant loops and they

are considered as a part of the cooling system.

In commercial power plants steam generators height can be up to 21 m

and weigh as much as 800 tons. Each steam generator can contain

between 3,000 to 16,000 tubes, each about 2 cm in diameter (Fig. 5.19).

The coolant, which is maintained at high pressure to prevent boiling, is

pumped through the nuclear reactor core. Heat transfer takes place

between the reactor core and the circulating water and the coolant is then

pumped through the primary tube side of the steam generator by coolant

pumps before returning to the reactor core (primary loop). That water

flowing through the steam generator boils water on the shell side to

produce steam in the secondary loop that is delivered to the turbines to

make electricity. These loops also have an important safety role because

they are primary barriers between the radioactive and non-radioactive

sides of the plant as the primary coolant becomes radioactive from its

exposure to the core. For this reason, the integrity of the tubing is

essential in minimizing the leakage of water between the two sides of the

plant. There is the potential that, if a tube bursts while a plant is operating,

contaminated steam could escape directly to the secondary cooling loop.

Thus during scheduled maintenance outages or shutdowns, some or all of

the steam generator tubes are inspected.

In other types of reactors, such as the pressurized heavy water reactors of

the CANDU design and liquid metal cooled reactors, heat exchangers are

used between primary coolant (metal or heavy water) and the secondary

water coolant.



Tube bundle

Tube support plate (7 total)

Secondary moisture separator

Primary moisture separator

Fig. 5.19: Vertical u-tube steam generator.

8.1. Steam generator types

There are three types of steam generators; vertical u-tubes with inverted

tubes for the primary water as in PWR (Fig. 5.19). The Russian VVER



reactor designs use horizontal steam generators, which have the tubes

mounted horizontally (Fig 5.20). The third type is the Once-through steam

generators that convert water to steam in a single pass (where water

enters at one end and steam exits at the other end (Fig. 5.21).

Fig. 5.20: Horizontal steam generator.

8.2. Steam generator tube material

Many types of materials are used to

manufacture the SG tubes. These can

be high-performance alloys and super

alloys such as type 316 stainless steel,

Alloy 400 and Alloy 600MA (mill

annealed).

SG

Fig. 5.21: Once-through steam

generator.

9. Pressurizer

The pressurizer (Fig. 5.22) is used to control the pressure in the reactor

cooling system so that boiling does not occur in the reactor. The

pressurizer also is used to act as a surge tank for the system taking up the

level variations in the system. Heaters are installed at the bottom of the

pressurizer for heating the water inside the pressurizer to about 345 ºC to



produce a bubble of steam. The steam bubble is used to maintain the

pressure in the sealed primary system at around 16 MPa.

Manhole

Spray hole

Nozzle for safety valves

Nozzle for control valves

Spray protection

Nozzle for volume compensation pipe valves

Heating rod Drain nozzle

Fig. 5.22: The pressurizer.

Automated pressure control valves (called power operated relief valves) and

safety valves, connected to the top of the pressurizer, can open to control

and maintain pressure. As explained in module 4, the pressurizer is part of

the PWRs. The pressurizer is about 13 meters tall and weighs 80 tones.

10. Nuclear Reactor Safety Systems

Safety is one of the most important design aspects of nuclear reactors.



Various systems are included in the different designs of NPPs to ensure

maximum safety measures.

The following safety systems are used to ensure maximum safety:

Reactor protection system (RPS) Essential service water system (ESWS) Emergency core cooling system (ECCS) Emergency electrical systems Ventilation and radiation protection

10.1. Reactor protection system (RPS)

A reactor protection system is composed of systems that are designed to

immediately terminate the nuclear reaction. All plants have some form of

the following reactor protection systems:

1. Control rods

Control rods can be quickly inserted into the core to absorb neutrons and

rapidly terminate the nuclear reaction.

2. Safety injection

In this system the nuclear reaction can be stopped by injecting a liquid that

absorbs neutrons directly into the core. In BWRs for example, the liquid

usually consists of a solution containing boron which can be injected to

displace the water in the core.

10.2. Essential service water system (ESWS)

The essential service water system (ESWS) circulates the water that cools

the plant’s heat exchangers and other components before dissipating the

heat into the environment. Because this includes cooling the systems that

remove heat from both the primary system and the spent fuel rod cooling

ponds, the ESWS is a safety-critical system.

Since the water is frequently drawn from an adjacent river, the sea, or

other large body of water (Fig. 5.23), the system can be endangered by

large volumes of seaweed, marine organisms, oil pollution, ice and debris.



In locations without a large body of water in which to dissipate the heat,

water is recirculated via a cooling tower (Fig. 5.24).

Fig. 5.23: The cooling water is drawn from large body of water.



Fig. 5.24: A cooling tower is used to reduce the temperature of the cooling water.

10.3. Emergency core cooling system (ECCS)

An emergency core cooling system comprises many systems that are

included to safely shut down a nuclear reactor during accident conditions.

The condenser is not used during an accident, so other methods of cooling

are required to prevent fuel rods melt down.

In most plants, ECCS is composed of the following systems:

1. High pressure coolant injection system (HPCI)

This system consists of a pump or pumps that have sufficient pressure to

inject coolant into the reactor vessel while it is pressurized. This system is

normally the first line of defense for a reactor since it can be used while the

reactor vessel is still highly pressurized.

2. Depressurization system (ADS)

The function of this system is to depressurize the reactor vessel and allows

lower pressure coolant injection systems to function, which have very large

capacities in comparison to high pressure systems. It consists of a series of

valves which open to vent steam under the surface of water in the wetwell

incase of BWRs or directly into the primary containment structure, in other

types of containments. Some depressurization systems are automatic in

function but can be inhibited, some are manual.

3. Low pressure coolant injection system (LPCI)

This system consists of a pump or pumps which inject additional coolant

into the reactor vessel once it has been depressurized.

4. Core spray system

This system reduces the generation of steam by using special spray nozzles

called “spargers” to spray water directly onto the fuel rods. Reactor designs



can include core spray in high-pressure and low-pressure modes.

5. Containment spray system

This system consists of a series of pumps and spargers which spray coolant

into the primary containment. It is designed to condense the steam into

liquid water within the primary containment structure to prevent

overpressure.

6. Isolation cooling system

This system is often driven by a steam turbine, and is used to provide

enough water to safely cool the reactor if the reactor building is isolated

from the control and turbine buildings. As it does not require large amounts

of electricity to run, and runs off the plant batteries, rather than the diesel

generators, it is a defensive system against the total loss of ac power (i.e.

failure of both offsite and onsite ac power sources). This condition is known

as station blackout (SBO).

10.4. Emergency electrical systems

These electrical systems usually consist of diesel generators and batteries.

These are not needed during normal operating conditions because nuclear

power plants receive electrical power to power their systems from off-site.

However, during an accident a plant may lose access to this power supply

and thus may be required to generate its own power to supply its

emergency systems.

1. Diesel generators

Diesel generators are used to power the NPP during emergency situations.

They are usually designed so that one can provide all the required power

for a facility to shutdown during an emergency situation. A facility can have

multiple generators as spares. Additionally, systems which are not required

to shutdown the reactor have separate electrical sources (often their own

generators) so that they do not affect shutdown capability.



2. Motor generator flywheels

Loss of electrical power can occur suddenly, and it can damage or

undermine equipment. To prevent damage, motor-generators can be tied to

flywheels which can provide uninterrupted electrical power to equipment for

a brief period of time. Often they are used to provide electrical power until

the plant electrical supply can be switched to the batteries and/or diesel

generators. The flywheel used in the RCP in Fig. 5.12a is an example.

3. Batteries

Batteries are often used as the final backup electrical system and are also

capable of providing sufficient electrical power to shutdown a plant.

Electrical inverter is needed to convert the DC power generated by batteries

to AC power to run AC devices such as motors.

10.5. Ventilation and radiation protection

In case of a radioactive release, most plants have a system designed to

remove radiation from the air to reduce the effects of the radiation release

on the employees and public. This system usually consists of the following:

1. Containment ventilation

This system is designed to remove radiation and steam from primary

containment in the event that the depressurization system was used to vent

steam into primary containment.

2. Control room ventilation

This system is designed to ensure that the operators who are required to

operate the plant are protected in the event of a radioactive release. This

system often consists of activated charcoal filters which remove radioactive

isotopes from the air.



4. Activities

The class will be divided into four groups and each group will do one of the

following activities and then groups will discuss and share findings.

Activity 1

Conduct a research on the recent accident in Fukushima nuclear power

reactors in Japan due to tsunami. Present your work in form of a power

point presentation and share your findings with your classmates.

Activity 2

Conduct a research on the accident in Three Mile Island nuclear power

reactors in the USA. Present your work in form of a power point

presentation and share your findings with your classmates.

Activity 3

Conduct a research on the accident in Chernobyl nuclear power reactors in

the former Soviet Union. Present your work in form of a power point

presentation and share your findings with your classmates.

Activity 4

Conduct a research to answer the following question: What initiates the

nuclear reaction when starting the nuclear reactor? Discuss your findings

with your classmates.



6. References

1. Why Science Matters, Using Nuclear Energy, John Townsend,

Heinemann Library.

2. Nuclear Energy (Fuelling the Future), Chris Oxlade and Elizabeth

Raum (Heinmann Library, 2008).

3. Future Energy, Improved, Sustainable and Clean Option For Our

Planet, Edited by Trevor M. Letcher, Elsevier.

4. Nuclear Safety, Gianni Petrangeli, Elsevier.

5. http://www.bwr-pr.com/

6. http://en.wikipedia.org/wiki/Boiling_water_reactor

7. http://www.nuclearfaq.ca/cnf_sectionA.htm

8. http://commons.wikimedia.org/wiki/Category:Schemata_of_pressuriz

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9. http://www.nuc.umr.edu/~ans/poor.html

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12. http://www.solcomhouse.com/nuclearpowerplants.htm

13. http://www.ecology.at/nni/index.php?p=type&t=gcr

14. http://www.mnf.co.jp/pages2/pwr2.htm

15. http://www.ask.com/wiki/Nuclear_fuel

16. http://wiki.answers.com

17. http://nuke1010.blogspot.com

18. http://www.brighthub.com/engineering/mechanical

19. http://www.animatedsoftware.com/hotwords/nuclear_reactor/nuclear



_reactor.htm

20. http://accessscience.com/content/Nuclear-Fuel-Cycle/459600

21. http://news.isc.vn/en/politics/nuclear-illinois-helped-shape-obama-

view-on-energy-in-dealings-with-exelon.html

22. http://maecourses.ucsd.edu/mae198/content/turbine.shtml

23. http://www.eskom.co.za/nuclear_energy/

24. http://cr4.globalspec.com/

25. http://wn.com/BWR

26. http://www.ask.com/wiki/Containment_building

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