Nuclear Energy Fundamentals
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.
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

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,
 the reactor safety
 protection systems.



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.



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 (Vodo-Vodyanoi
Energetichesky Reactor) (a PWR
developed in 1970 in the Soviet
Union) and cylindrical that is used in
CANDU.
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).
Fig. 5.1: Fuel assembly geometry: a) square b)
cylindrical c) hexagonal d) spherical.
(a)
(b)
(c)
(d)
The pebble bed reactor
(PBR) is a graphitemoderated, gas-cooled,
nuclear reactor.
It is a type of very high
temperature reactor (VHTR),
one of the six classes of
nuclear reactors in the
Generation IV initiative.

In PWR for example, the fuel assembly
consists of a square array of 179 to 264
fuel rods which is a long, slender (thin)
tube in which nuclear fuel is surrounded
by cladding material and inserted into a
reactor.

There are three types of cladding material,
namely, zirconium( Zr )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 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
Fig. 5.2: The PWR fuel assembly specifications.
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.

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
a.
b.
c.
Desired reactor power output,
Reactor core size and
Reactor power density
Fig. 5.3: The PWR fuel assembly specifications.

The fuel material is the material from which
the fuel elements (Fuel Assembly)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).
Nuclear fuel can be in many forms.
It can be in the form of
 a chemical compound,
 ceramic,
 metal alloy.
It can be also:
 pellets
 rods
 dissolved in liquid salt or liquid lead.

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),
 with significant portion while the reactor is off-load
(during refueling outages).

Examples of refueling Reactors:

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

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.






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.

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.


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
 propagate the fission process .
 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.
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.

Besides the control rods, this system includes a number of devices,:
 Sensing elements that measure the number of neutrons in the reactor.
 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).

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).

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 330
°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.

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 radioactivity, 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 shortlived, so the turbine hall can be entered soon after
the reactor is shut down.

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.

The fission reaction is accompanied by emission of radiation like α, β ,n and γ.


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.

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.

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 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
For a pressurized water reactor, the
containment also encloses
 the steam generators
 the pressurizer
 the entire reactor building

PWR containments are typically 10
times larger than a BWR

In most PWR designs the spent fuel
pool is located outside of the
containment building.
Fig. 5.14: PWR reactor 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.

In BWR's (Fig. 5.16) the containment
consists of a drywell where the
reactor and cooling equipment is
located and a wet well 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 wet well.

This condenses the steam and
reduces the pressure.
Both the drywell and the wet well are enclosed by a secondary containment
building (Fig. 5.17) which is maintained at a slight negative pressure during
normal operation and refueling operations.
•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.

Fig. 5.18: The BWR containment building design for the reactor and turbine.

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.
There are three types of steam
generators;
1.
The First Type:
Vertical u-tubes with inverted tubes
for the primary water as in PWR (Fig.
5.19).
2.
The Second Type:
The Russian VVER reactor designs use
horizontal steam generators, which have
the tubes mounted horizontally (Fig 5.20).
Fig. 5.20: Horizontal steam generator.
3.
The third type :
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).
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
 Alloy 600MA (mill annealed).

Extra Information:
Type 316 is a molybdenum-bearing grade. This addition gives the better overall corrosion resistance
properties than type 304 and higher creep strength at elevated temperatures. Type 316 gives
useful service at room temperature in sulphuric acid of concentration of lower than 15% and
higher than 85%. It also resists chloride attack and is often selected for use in marine
atmospheres.
Alloy 400 offers an exceptional blend of weldability, strength, ductility and corrosion resistance,
particularly in seawater in high-velocity applications. Alloy 400 also resists stress corrosion
cracking.
Inconel 600 is a nickel-chromium alloy used for applications that require corrosion and high
temperature resistance.
Mill annealing is a general-purpose treatment given to all mill products. It is not a full anneal and may
leave traces of cold or warm working in the microstructures of heavily worked products,
particularly sheet.

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.



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.


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
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.

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.

The ESWS is a safety-critical system because
this includes:
cooling the systems that remove heat from both
the primary system and the spent fuel rod
cooling ponds
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).
 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.
2.
3.
4.
5.
6.
High pressure coolant injection system (HPCI)
Depressurization system (ADS)
Low pressure coolant injection system (LPCI)
Core spray system
Containment spray system
Isolation cooling system
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 highpressure and low-pressure modes.
5. Containment spray system


This system consists of a series of pumps and sprayers 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).

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

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
2. Control room ventilation
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.