453
T.K. Ghosh and M.A. Prelas, Energy Resources and Systems:
© Springer Science + Business Media B.V. 2009
9 NUCLEAR
Abstract
The world’s demand for energy will increase because the quality of living is tied
to its consumption and because the world’s population is increasing. It is well
known that high quality energy sources such as electricity and oil are directly
linked to economic growth and quality of life (e.g., access to food, medicine,
housing and education). As more countries participate in the benefits of globalization,
world energy use will grow. For example, in the past decade prior to the
financial collapse of world markets that occurred during October of 2008, two of
the world’s most populated countries, China and India, had unprecedented eco$150
per barrel during the summer of 2008. The year 2008 also saw an precipitous
drop in oil prices as well. It is well known that economic growth is cyclic and that
during periods of recession, demand for energy decreases. After October of 2008
the price of oil dropped under $45 dollar per barrel. As of this writing the world
was still in the recession of 2008, but history tells us that a recovery will occur and
that during the next economic growth period oil prices will once again rise. The
other factor which impacts growth in energy demand is growth in population. The
US Census Bureau has projected that the world population will increase to 9
billion people by 2042 from approximately 6.6 billion people today. The future
holds more uncertainty. The competition for scarce energy resources will only
accelerate as more people in the world participate in the economic benefits of
globalization and as world population grows.
One of the requirements for sustaining human life and progress is availability
of a clean source of energy that does not harm the environment. The release of
CO2 into the atmosphere is the main cause of global warming and various associated
climate changes. A major advantage of nuclear energy is that it doesn’t put
CO2 into the atmosphere. In addition it provides a steady source of constant electrical
power that does not suffer the whim of weather patterns. This is a critical
feature because climate change does impact weather patterns. For example, areas
Volume 1: Fundamentals and Non-Renewable Resources, 453–647.
nomic growth and this played a major role in the price of oil climbing to nearly
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454 9 Nuclear
which currently have an abundance of wind or clear skies for solar energy use
today may experience a change over the 30 or so year lifetime of a wind farm or a
solar energy plant.
Although nuclear energy generates electricity without releasing harmful gases;
SO2, NOx, CO2, etc, the issues related to spent fuel (nuclear waste) from the
nuclear reactors must be addressed. Reprocessing of spent fuels and the use of
breeder reactors can substantially minimize its impact and even can lead to zero
waste system. In this chapter, the complete nuclear fuel cycle is discussed.
9.1 Introduction
There are two fundamental ways by which energy is released from nuclear
reactions: fission and fusion of atomic nuclei. The mass of a nucleus (a nucleus is
made up of protons and neutrons) is always less than the sum of the individual
masses of the protons and neutrons which constitute it. The difference is a meautilized
for electricity generation. This binding energy can be calculated from the
Einstein relationship and is discussed below.
In 1905, a young patent clerk in Switzerland developed a theory that would
forever change the world. That young patent clerk was Albert Einstein. His theory
of relativity is one of the major achievements in modern physics (the other being
quantum mechanics).
In the 1880s, Physicists believed that waves require a medium to propagate.
Sound waves propagate in air, and waves propagate in water. It was believed that
the earth traveled through an “ether wind,” which allowed light to propagate
through space. American Physicist Albert Abraham Michelson (1852–1931) and
American Chemist Edward Williams Morley (1838–1923) designed an experiment
that attempted to examine the motion of earth relative to the ether using Michelson’s
new instrument, an interferometer. The experimental arrangement is shown in
Fig. 9.1.
A light beam is split. Part of the beam travels straight and the other part travels
90° up. Michelson and Morley assumed that the speed of light is c and the speed
of the ether wind is v. Michelson and Morley reasoned that the part of the beam
that travels anti-parallel and then parallel to the ether wind will take a longer time
to traverse the distance 2d, and the time is given by:
22
2
vc
dc
vc
d
vc
d
t (9.1)
a fission or a fusion reaction part of this binding energy is released and is
sure of the nuclear binding energy which holds the nucleus together. During either
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9.1 Introduction 455
Fig. 9.1. Michelson interferometer.
Likewise along the path perpendicular to the ether wind, the time it takes the
beam to traverse the distance 2d is given by;
22
2
vc
d
t (9.2)
The ratio of the parallel time to perpendicular time can be represented as:
2
2
1
1
c
vt
t
(9.3)
If the light source has zero velocity with respect to the ether, then the ratio is 1.
If the earth is in motion, the light source should have a velocity with respect to the
ether so the ratio should not be 1. The two beams at the detector will then be out
of phase and thus will interfere with one another. The interference fringes can then
be measured.
Michelson and Morley ran thousands of experiments over many months. In
each and every one of their experiments the ratio was 1. They had to conclude that
there was no ether wind.
From this work, other physicists tried to explain the result. George Francis
FitzGerald (1851–1901), an Irish physicist, developed the idea that objects grew
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456 9 Nuclear
shorter in the direction of their motion because the ether wind would exert a pressure
on the body. FitzGerald developed the FitzGerald length contraction formula
from this assumption and used it to explain the Michelson–Morley experiment.
The length is calculated from the following expression.
2
2
1
c
v
LL o (9.4)
In Eq. (9.4), L is the length in the moving reference frame and Lo is the length
at rest. The problem with the FitzGerald contraction is that when v > c, the L is an
imaginary number.
Hendrick Antoon Lorentz (1853–1928), a Dutch physicist, extended the work
of FitzGerald. He postulated that mass would have to increase with motion and is
given below.
2
2
1
c
v
m
m o
(9.5)
Lorentz’s work was able to predict the motion of high-speed charged particles
in electromagnetic fields. Often times these contractions are called the Lorentz
contractions.
Einstein assumed that the relative motion of bodies is impossible to sort out.
How does one determine which body is absolutely at rest? The second assumption
that he made in his new theory is that the speed of light in a vacuum is a constant
regardless of frame of reference. This was a drastic change from the universally
accepted Newtonian mechanics. In Newtonian mechanics if I am on a platform
moving at the speed of light and I throw an object off that platform with a velocity
of the speed of light, then the relative speed of that object to the rest frame is,
cVVV 221 (9.6)
where V1 = V2 = c
According to Einstein, the velocity of the object would be,
c
c
VV
VV
V
2
21
21
1
(9.7)
This forced the object to have a velocity no greater than the speed of light.
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9.1 Introduction 457
Classical physicists believed that mass could not be created or destroyed. But
the Lorentz contraction and Einstein’s special theory of relativity differed in that
mass was increased as velocity increased. The implication of this is very important
to nuclear energy. We can make the following approximation (use a Taylor series
expansion),
2
2
2
2 2
1
1
1
c
v
c
v
(9.8)
So, the mass of a body in motion, m1, becomes,
2
2
2
2
1
22
1
c
v
mm
c
v
mm ooo (9.9)
The increase in the mass as a result of motion is given by;
2
2
1
2
1
c
v
mmmm oo (9.10)
The term ½ mov2
is the kinetic energy of the rest mass. This energy is denoted
by E, therefore,
2
c
E
m (9.11)
This equation can be rearranged to give the mass energy relationship,
2
mcE (9.12)
The implication of this relationship can be seen in the calculation of binding
energy of nuclei. If we assume that the mass of an atom is the summation of the
mass of its components, then for an oxygen atom, we would assume it has eight
neutrons, eight protons and eight electrons. The sum of these components would
be,
enp
oxygen
theory mmmm 888 (9.13)
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458 9 Nuclear
where mp, mn and me are the masses of a free proton, neutron and electron. The
problem is that the actual mass of an oxygen atom is less than the sum of its com-
ponents.
oxygen
theory
oxygen
erimental mmexp (9.14)
This mass difference is,
oxygen
erimental
oxygen
theory mmm exp (9.15)
The resulting mass change must go into energy according to the mass energy
relationship,
2
mcE (9.16)
This energy is known as the binding energy. In order to break apart the atom
into its basic components, an energy equivalent to the binding energy of the atom
must be supplied.
By convention, the number of protons in an atom is represented by Z, the number
sented by Z, the number of neutrons in an atom is represented by N, and the atomic
number is the total number of protons and neutrons and is represented by A.
ZNA (9.17)
It is useful to compare the binding energy of all atoms by looking at the binding
energy per nucleon (a nucleon is either a proton or neutron). The binding energy
per nucleon is given by the following expression.
Binding Energy per Nucleon =
A
mC 2
(9.18)
We can plot a curve of the binding energy per nucleon for all known atoms as
shown in Fig. 9.2.
The binding energy per nucleon curve tells us that the most stable element in
the universe is iron. The second law of thermodynamics tells us that all systems
tend towards their minimum energy. Therefore,
All matter in the universe will tend towards iron
of electrons in an atom is equivalent to the number of protons and thus are repreLicensed
to Jiri Martinec<martinec@fme.vutbr.cz>
9.2 The Fusion Cycle 459
Fig. 9.2. Binding energy per nucleon as a function of mass number for the stable nuclides (Courtesy
[1]).
9.2 The Fusion Cycle
A point in space with near infinite energy exploded. As the energy wave propagated,
it cooled. Light particles such as baryons and leptons formed. Eventually
hydrogen began to condense from this energy wave during the cooling process.
The hydrogen began to form density perturbations in space. This spatial variation
of mass began to attract other hydrogen particles. The more particles that
were attracted to the density variations, the more gravitation attraction there was
to bring in more hydrogen. The mass began to build up. Eventually, the mass was
large enough to form a proto-star. At this point, nuclear reactions that convert
hydrogen into heavier elements produced sufficient energy that the energy produced
by the fusion reactions balanced the energy lost by radiation. The mass
grew larger and larger and eventually, the fusion energy produced was sufficient
to create a high surface temperature and the surface glowed in the visible wavelength
(e.g., stars were formed).
According to the modern cosmology, the universe started with the big bang.
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460 9 Nuclear
Some of the stars that were formed had enormous mass. The hydrogen fuel in
these massive stars was consumed at rapid rates. Other elements formed from
hydrogen fusion such as helium, and carbon began to fuse to form heavier elements.
These fusion reactions created oxygen. Oxygen began to fuse creating silicon.
Silicon began to fuse to form iron. The iron core remained inactive since it is the
element with the largest binding energy per nucleon. As the iron core of the star
grew larger and larger, the star could no longer support the pressure from the surrounding
mass. This caused the star to collapse. As the star collapsed upon itself,
the outer shells of silicon, oxygen, carbon, helium and hydrogen grew denser. This
density increase caused an upsurge of fusion activity that resulted in an explosion.
The massive star became a supernova. The massive energy release sent iron, silicon,
oxygen, carbon, and helium and hydrogen particles in all directions at very
high energy. Some of these particles collided (the supernova was like a massive
celestial high-energy accelerator) and underwent complex nuclear reactions. These
reactions produced elements heavier than iron. The remnants of the supernova
began to condense. New stars were formed along with planets from the debris.
9.3 The Fission Cycle
Among the heavy elements that condensed to form planets was uranium-235. In
the late 1930s Enrico Fermi had been bombarding uranium with thermal neutrons
(the definition of thermal neutron is discussed later). He believed that he had
created an element with an atomic mass of 93. Basically he was right, but his
explanation was not very clear.
In 1938, German Physicist Otto Hahn and his coworkers Fritz Straussman performed
an experiment planned by Hahn and Lise Meitner in which uranium was
bombarded with thermal neutrons. The chemical makeup of the uranium had
changed in that presence of barium was found. Because Meitner was a Jewish
exile from Germany, Hahn and Straussman, who worked in the Kaiser-WilhelmInstitut
in Germany, could not include Meitner as a co-author of the paper
announcing the result. Lise Meitner was widely credited as being the first to recognize
that nuclear fission had occurred. She and Otto Frisch developed a proof that
fission had occurred in these experiments and published a paper in Nature, January
16, 1939. In their paper they described how barium could be produced in a fission
reaction along with krypton and additional neutrons.
Word of the discover spread like wildfire. Niels Bohr went to the US for a conference
and discussed the possibility of fission with US physicists. Many of the
physicists went back to their laboratories and verified the fission reaction.
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9.3 The Fission Cycle 461
Hungarian-American physicist Leo Szilard began to think about the nuclear
chain reaction shortly afterwards. He pondered on the implications of the reactions
and became concerned. Given that he was a Jewish refugee from Hitler’s tyranny,
he was fully aware of the danger that the Nazis posed to the world and what it
would mean if Hitler was able to develop a weapon based on the nuclear chain
reaction. Sizlard discussed his fear with physicist Eugene Paul Wigner and
Edward Teller. They decided to bring their concerns to the attention of the US
government. In order to do so they needed the help of the world’s most prominent
scientist, Albert Einstein. They were able to persuade Einstein to write a letter
to President Franklin D. Roosevelt. In 1941, Roosevelt agreed to start a massive
research program to develop a bomb based on nuclear fission. The order was
issued on December 6, 1941, the day before the Japanese attacked Pearl Harbor.
The start of mankind’s trek into nuclear energy began with fear while the world
was in the death grip of World War II.
Nuclei of atomic mass up to 40 remain stable as the number of protons and the
number of neutrons remain equal (Fig. 9.3). The strong force which holds the
nucleus together is stronger than the Coulombic force which causes the protons to
repel one another. The strong force has a maximum interaction length of about
1–1.5 fm, beyond which it is zero. Additionally, the strong force is repulsive as
distances become shorter than 0.1 fm.
A nucleus is on the order of 1 fm in diameter. A proton and a neutron are about
0.01 fm in diameter. The density of nucleons in a nucleus is about constant
throughout the volume. The importance of this fairly constant density is that the
distance between nucleons remains constant regardless of the number of nucleons
in the nucleus.
As the number of protons increases, the Columbic force gets larger and larger.
The distances between protons must get larger. Neutrons fill in the space as protons
move further apart. Eventually, the number of nucleons fills a volume so large
The line of stability shown in Figs. 9.4 and 9.5 is very narrow. If there are
fewer neutrons in a nucleus than the line of stability, the nucleus will undergo a
radioactive process in which a proton in the nucleus is converted to a neutron and
a positron. A positron is a positively charged electron.
11 Z
A
NZ
A
N YX (9.19)
where is a positron and is a neutrino.
that the strong forces can no longer act. At this point, the number of neutrons
becomes larger than the number of protons.
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462 9 Nuclear
Fig. 9.3. Values of protons and neutrons for 266 stable nuclides. The line Z = N represents number
of proton equals to number of neutron.
Fig. 9.4. A plot of all 250 stable elements vs. the atomic mass (Courtesy of [2]).
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9.3 The Fission Cycle 463
Fig. 9.5. Plot of neutron number versus proton number for all known nuclides (Dark squares represent
stable nuclides and light squares represent unstable nuclides (Courtesy of [2]).
In this reaction, the element X is neutron poor and lies above the line of stability.
If there are fewer neutrons than the line of stability, the decay mode is . If
there are more neutrons than the line of stability, the decay mode is .
11 Z
A
NZ
A
N YYXX (9.20)
where is an antineutrino.
The farther a nucleus is from the line of stability, the faster it decays. In nuclear
science, the rate of decay is illustrated by the half life. The half life is the time it
takes N radioactive atoms to decay to 0.5N radioactive atoms. The decay of radioactive
elements is a statistical process. This process can be modeled by a rate
equation:
kN
dt
dN
(9.21)
where k is a decay constant.
The solution to this equation is,
kt
eNtN 0 (9.22)
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464 9 Nuclear
The half life can be found by solving,
212
1
5.0
0
tk
e
N
tN
(9.23)
5.0ln2
1tk (9.24)
2
1
693.0
t
k (9.25)
Typically, in nuclear science, the decay rates of nuclei are given in half-lives.
From the half-life, the decay constant is calculated using Eq. (9.25).
There are other types of radioactive decay processes. Some heavy nuclei can
decay by giving off alpha particles (helium nucleus) and other types of ions.
One of the most interesting processes is fission. Elements like plutonium 238
can spontaneously fission. As discussed, fission is the splitting of a nucleus. On
average, each fission gives off more than two neutrons (with a high energy distribution
as shown in Fig. 9.6).
nfhflX Z
A
NZ
A
NZ
A
N 2
2
21
1
1 (9.26)
where is the statistical number of neutrons given off per fission. For U-235,
is 2.44.
Fig. 9.6. Energy distribution of neutrons given off by the fission of U-235 as calculated
(Courtesy [1]).
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9.3 The Fission Cycle 465
The release of more than one neutron per fission was key to Leo Szilard’s idea
of a chain reaction. If you consider neutron bookkeeping, you use one thermal
neutron to initiate fission. In the process the fission gives off 2.44 additional neutrons.
If each of these neutrons causes fission, then you have 2.44 × 2.44 (5.95)
neutrons in the second generation. As you can see, each additional generation
would have 2.44 more neutrons than the previous generation. This process is a
geometrical process. However, such processes can not be sustained. There are
natural processes which will cause the fission rate to plateau.
Nuclear fission occurs in several elements. The reaction begins with the capture
of a neutron in the nucleus. In general, the probability of capturing a neutron with
a low velocity is higher than capturing a neutron with a high velocity. The energy
of a neutron is given by,
2
2
1
vmE nn (9.27)
where, En is the neutron energy (Joules), mn (kilogram) is the neutron mass and
v (meters per second) is the neutron velocity.
Energy is related to temperature. For an individual neutron the temperature/energy
relationship is,
kTEn (9.28)
where k is Boltzman;s constant (1.381 × 10 J/K) and T is temperature in Kelvin.
Because the energy of a high velocity neutron can be a very small number in
Joules, it is common to express particle energy in electron Volts (eV).
1 eV = 1.6 × 10 J (9.29)
can be calculated from Eqs. (9.28) and (9.29). Room temperature is about 20o
C or
293 K. So a neutron of temperature 293 K has an energy of:
0253.0
106.1
2931038.1
106.1 19
23
19293
kT
E (9.30)
gies. The unit of a barn is used which is 1 × 10 cm2
. This is about equal to the
–19
–24
–23
eV
function of energy. The cross section is a measure of the probability of a reaction.
Fig. 9.7 shows, the cross section for fission in U-235 is very high at low enerIf
we want to know the energy of a neutron, which is at room temperature, it
Figure 9.7 shows the cross section for nuclear fission in U-235 and U-238 as a
Thus the higher the cross section, the higher is the probability of the reaction. As
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466 9 Nuclear
cross sectional area of a typical nucleus. The absorption cross sections for all of
the elements fall of as 1/v (where v is velocity) at energies below 2 eV. For light
sharp peaks at energies above 2 eV which are due to enhanced absorption in the
nucleus due to metastable levels. These sharp peaks are called resonances and they
occur roughly between 2 eV and 100 KeV. Thermal cross sections are very important
to modern light water or heavy water moderated power plants. Table 9.1
tabulates cross sections for important elements at 0.025 eV energies.
Fig. 9.7. Cross section for neutron capture (n, ) and fission for U-235 and U-238 as a function of
neutron energy (Adapted from [1]).
Table 9.1. Data for interactions with various elements at 0.025 Ev.
Element Absorption cross section (barns) Fission cross section (barns)
U-233 579 531
U-235 681 582
Natural U 7.59 4.19
Pu-239 1011 743
Pu-241 1377 1009
The concept of a cross section can be understood by looking at the illustration
in Fig. 9.8 shown below,
elements the 1/v behavior persists to high energies. For heavy elements there will be
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9.3 The Fission Cycle 467
Fig. 9.8. Illustration showing concept of a nucleus cross section. An incident beam of neutrons
intersect a target material and the non interacted beam of neutrons exits the target.
In this illustration, an incident beam of neutrons with intensity I(x) (neutrons
per square meter per second) strikes the target and the non-interacted neutrons exit
can be set up to calculate it. We use the cross section to make this calculation by
setting up the following relationship,
dxxINxdI (9.31)
3 2
The solution to this differential equation is,
xN
exI (9.32)
Terms can be grouped. The macroscopic cross section is equal to N . Using
the macroscopic cross section we can write the equation,
dx
xI
xdI
(9.33)
Thus, the term dx represents the probability that a neutron will interact in a
length of dx. The macroscopic cross section is the probability per unit path length
that a neutron will interact. Thus,
x
e
I
xI
0
(9.34)
where this represents the fraction of neutrons at the point x which have interacted
in a length of dx.
where dI(x) is the differential neutron beam intensity at the point x, N is the
the target. The neutron beam intensity is a function of x (meter), I(x) and an equation
density of target atoms (atoms per m ), is the cross section (m ), dx is a differential
length.
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468 9 Nuclear
e(- x)
is the probability that a neutron reaches the point x without interacting
with the target.
Define, p(x) dx as the probability that a neutron which reaches x has its first interaction
in a length dx. The mean free path ( ) is the average distance that a neutron
travels before interaction. The mean free path is given by,
1
0 0
dxexdxxpx x
(9.35)
When a neutron is captured by a nucleus, a compound nucleus is formed. The
additional neutron causes instability in the nucleus thus causing it to deform (Fig.
9.9). As the nucleus deforms it begins to split into two pieces. These pieces then
emit gamma rays and neutrons leaving behind two fission fragments (Fig. 9.10).
The fission fragments typically are unstable and eventually emit radiation over a
time scale of seconds to years.
Fig. 9.9. Curve showing the potential energy versus distance of the two components of the compound
nucleus as it deforms and splits into two segments. EA is the activation energy required to
split the nucleus into two fragments and the equation is the governing Coulombic repulsion equation
(Adapted from [1]).
For fission to occur, it is possible to formulate a model which shows why
(KE) and total energy (MAC2
). Total energy is equal to potential energy plus
kinetic energy,
KEVCM A
2
(9.36)
certain elements fission. Begin by defining Potential energy (V ), Kinetic Energy
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9.3 The Fission Cycle 469
Fig. 9.10. An illustration of the phases of neutron capture leading to fission. In step a the neutron
ments form and fast neutrons emitted. The fragments promptly emit gamma rays in step e as they
energetically stabilize. In step f the radioactive fission fragments decay (Adapted from [3]).
Therefore,
KECMV A
2
(9.37)
Initially, before the two fragments split, the distance between the two fragments
nucleus (Ecrit) to overcome the strong force. So,
critA ECMV 2
(9.38)
is captured by the nucleus. In steps b and c the nucleus begins to split. In step d the fission fragis
0 and the KE is 0. As the nucleus begins to deform, energy must be added to the
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470 9 Nuclear
Fig. 9.11. This figure illustrates how the nucleus fissions. E1 is the total energy of the nucleus
before splitting, E2 is the total energy plus the energy required to overcome the strong force, E3
is the total energy of the two fission fragments, Q is the change in energy from the initial to final
state and Eq is the coulomb repulsion energy.
From Fig. 9.11, the Q value is,
2
2
2
1
2
CMCMCMQ AAA (9.39)
where MA1 A2
second fission fragment.
Ecrit can be estimated as,
QEE qcrit (9.40)
and Eq by,
21
2
21
RR
eZZ
Eq (9.41)
1 2
1
2
The radius of a nucleus is estimated by,
2
3/1
A
rR e (9.42)
Mis the mass of the first fission fragment and is the mass of the
where Z is the number of protons in fission fragment 1 and Z is the number
of protons in fission fragment 2, e is the charge of a proton, R is the radius of the
nucleus before splitting and R is the radius of the nucleus when it begins to split.
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9.3 The Fission Cycle 471
where A is the atomic mass and 2
2
Cm
e
r
e
e and me is the mass of an electron.
From Equations (9.41) and (9.42) it can be seen that,
3/1
2
3/1
1
2
21
2
AA
r
eZZ
E
e
q (9.43)
and,
3/1
2
3/1
1
21
2
2
AA
ZZCm
E e
q (9.44)
The term 2meC2
is about 1 MeV. Also it can be assumed for the time being that
Z1 = Z2, therefore,
3/1
2
16.0
A
Z
Eq (9.45)
For U-238, Z = 92 and A = 238, and
Eq = 218 MeV
Q for U-238 is about 212 MeV, thus
Ecrit ~ 6 MeV
For lead, Pb-208, Z = 82 and A = 208. The value of Eq is given by,
Eq = 182 MeV
For Pb-208, Q is about 162 MeV, so,
Ecrit ~20 MeV
crit and binding energy of the last nucleon for variousIn Table 9.2 is given the E
elements.
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472 9 Nuclear
and the binding energy for the last nucleon.
Element Ecrit (MeV) Binding energy last
nucleon (MeV)
Th-232 5.9 *
Th-233 6.5 5.1
U-233 5.5 *
U-234 4.6 6.6
U-235 5.75 *
U-236 5.3 6.4
U-238 5.85 *
U-239 5.5 4.9
Pu-239 5.5 *
Pu-240 4.0 6.4
*Binding energy is not relevant for this analysis.
Fission will occur when the binding energy of the last nucleon is greater than
Ecrit. From Table 9.2, this condition applies to U-234, U-236 and Pu-240. Thus the
nuclei which capture the neutron to trigger fission are U-233, U-235 and Pu-239.
These materials are the primary practical fissile fuels.
The fission fragments do come out as a light fragment and a heavy fragment.
However the fragments do not come out as a particular pair of nuclei. They come
Fig. 9.12. The percent yield of fission fragments manifest as two peaks, one for the heavy fragment
and one for the light fragment (Adapted from [3]).
critE
out in a distribution of paired nuclei. Figure 9.12 shows the distribution of paired
nuclei as two distinct peaks.
Table 9.2. The value of
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9.3 The Fission Cycle 473
The decay of radioactive fission products produces heat. Albeit, it is a small
amount of the total heat produced in an operating nuclear power plant (about 0.7%
of the total heat), it is enough heat to require cooling when the reactor is shut
down. The Borst-Wheeler formula calculates the decay heat of a reactor after it
has shut down,
2.02.011
1010.4, TttTtP (9.46)
where P(t,T) is the total power emitted in the form of beta and gamma rays
from the decaying fission products in a reactor which has operated T seconds at a
power of 1 W, t is the time after shutdown in seconds.
For example, the University of Missouri Research Reactor is the most powerful
University operated research reactor in the United States. Its operating power is 10
MW. During a typical week it operates continuously at full power for 150 h. The
decay heat 10 s after shutdown can be calculated accurately by the Borst-Wheeler
formula,
T = 150 h = 5.4 106
s, thus
P(10s, 5.4 106
s) = 2.29 × 1011
MeV/s/ W
Decay Heat in Reactor = 2.29 1011
MeV/s/W 10,000 W
Decay Heat in Reactor = 2.29 1017
MeV/s 1.6 10 J/MeV = 36,640 W
Once the nucleus fissions, one of the products is neutrons. Depending on which
pair of fission fragments come out, the number of neutrons given off can vary.
Over a large number of fissions, the statistical average of neutrons given off per
event for example is about 2.44 for U-235. The decay chain of U-235 is shown in
Fig. 9.13. The neutrons are born at fairly high energies (greater than 1 MeV) and
as shown in Fig. 9.12, the energies have a distribution which is dependent on the
fission fragment pair.
Nuclear reactors are designed to take advantage of the large cross section at
thermal energies. In order to slow down the fast neutrons born in fission, materials
have to be used to slow down the fast neutrons. These materials are called
moderators and are based on light elements such as graphite, water, beryllium and
heavy water.
Reactors are typically designed to use thermal neutrons (with neutron energies
below 1 eV) and fast neutrons with energies from about 100 keV up to the top of
range of the fission neutron energy spectrum. Thermal cross sections are very
important to modern light water or heavy water moderated power plants. Table 9.1
tabulates cross sections for important elements at 0.025 eV energies.
In reality, the reactor materials are not entirely made up of fissile fuels and
moderators. There are components such as structural elements, fuel cladding and
there is leakage of neutrons from the reactor.
–13
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474 9 Nuclear
Fig.9.13.Uranium-235decaychain.
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9.3 The Fission Cycle 475
In reactor physics we describe the above process by the six factor formula.
k = f p Pf Pt (9.47)
In Eq. (9.47),
thermal neutrons; f is the fraction of thermal neutrons absorbed in the fuel. Some
of the thermal neutrons are absorbed in the fuel and some in the other materials
is the fast fission factor.
Fast neutrons can produce fission, is the number of additional fast neutrons that
are created by fast fission; Pf is the probability that a fast neutron stays in the
reactor. Some fraction of fast neutrons will leak out of the reactor volume; Pt is the
probability that a thermal neutron stays in the reactor. Some fraction of thermal
neutrons will leak out of the reactor volume. This process is shown in Fig. 9.14.
The factor k is a means of neutron bookkeeping. k neutrons produced in fission
will survive to be absorbed by the fuel and will produce k additional fission reactions.
If k is greater to or equal to one, then the reaction is self sustaining. This is a
chain reaction. Reactors are designed to have k > 1. A nuclear reactor is controlled
by the use of control rods. These rods are made up of materials that absorb thermal
neutrons. When the control rods are inserted into the nuclear reactor, the k value of
the reactor is less than one. When the control rods are taken out of the reactor, k >
1 and the reactor is able to undergo a chain reaction. The control rods allow the
operator to control the reactor power level and to shut the reactor down.
The k value is a geometrical series which increases rapidly,
Neutrons in Generation n = kn
(9.48)
Figures 9.15 and 9.16 is the neutron production over multiple generations for a
k value of 2.44 and 1.1, respectively. Given that the time between generations is
on the order of microseconds, and that energy release is directly proportional to
the number of neutrons, it can be seen that energy output can multiply rather
rapidly over a short period of time.
The total number of fission reactions can be very large in a short amount of
time. Even though nuclear science had many applications which had saved millions
of lives by the time Word War II started, people associate the word nuclear
with nuclear weapons. A nuclear weapon is a formidable device which is designed
to have a relatively large value of k for the purpose of releasing as much energy as
possible in a short period of time. The weapon is designed to maintain a high k
value until the energy released blows the device apart. We can see the potential of
the weapon in the following example.
is the number of fast neutrons produced by fission with
that make up the reactor; p is the resonance escape probability. As the fast neutrons
move through materials, there are absorption resonances at high neutron energies (2 to
100,000 eV in the materials that make up the reactor. This is the probability that the
fast neutrons will slow down without being absorbed;
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4769Nuclear
of Mass Destruction. CRC Press [4]
Fig. 9.14. Neutron bookkeeping in a reactor. Printed with permission from Prelas MA, Peck MS (2005) Nonproliferation Issues for Weapons
LicensedtoJiriMartinec<martinec@fme.vutbr.cz>
9.3 The Fission Cycle 477
Fig. 9.15. Neutron production from generation to generation for a k value of 2.44 (Printed with
permission from [4]).
permission from [4]).
Fig. 9.16. Neutron production from generation to generation for a k value of 1.1 (Printed with
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478 9 Nuclear
Consider a nuclear warhead which has a k = 1.5 and holds together for 139
neutron generations, what is the total energy released? We begin by looking at the
TN = neutrons in generation 1 + neutrons in gen 2 + … + neutrons in gen i
generation 0 = n0
generation 1 = n0k
generation 2 = n0k2
generation i = n0ki
This is a power series which has an analytical solution,
1
1
0
k
k
niTN
i
Since each fission will yield neutrons in a generation the total number of
fissions will be,
TF = fissions in generation 1 + fissions in gen 2 + …+ fissions in gen i
Fissions in generation 1 =
v
kn0
Fissions in generation 2 =
v
kn 2
0
Fissions in generation i =
v
kn i
0
The formula for the total number of fissions will be,
v
k
kk
niTF
i
1
0
Thus, assuming that n0 = 1, and (for U-235)
44.2
15.1
5.15.1
139
139
0nTF ~ 2.46 x 1024
fissions
Each fission gives off about 200 MeV or 3.2 × 10 Joules, so the energy produced
after 139 generations is,
–11
total number of neutrons (TN ) that occur in 139 generations,
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9.3 The Fission Cycle 479
E = 2.46 × 1024
× 3.2 × 10-11
= 7.9 × 1013
J
This is a large amount of energy. For an explosive, we can use the relationship
that 1,000 t (or a kiloton) of TNT is 4.184 × 1012
Joules. Converting this energy to
kilotons of TNT we find,
E = 7.9 × 1013
/4.184 × 1012
~ 18.9 kt of TNT
Fortunately, it is not easy to design a nuclear weapon. It is a sophisticated
device that requires a great deal of technology. For instance, we know that Iraq
invested more than 20 billion dollars in its nuclear weapons program and were not
even close to getting any significant amounts of fissile materials.
We need to dispel a common concern about the relationship of nuclear weapons to
nuclear power. A power nuclear power plant is nothing like a nuclear weapon. A
nuclear power plant has a much different design criteria which does not allow
large power excursions, thus k is kept at values below 1.1 allowing complete and
safe control of the reaction.
It is not an easy task to design a critical assembly (k > 1) of fissile materials for
a power producing reactor. There is much that goes into the design such as choice
of materials, geometry, structural components, moderator, control systems, active
safety systems, passive safety systems etc. The US for example requires that
nuclear power plants have a negative temperature coefficient which means that the
plant will shut itself down if the fuel exceeds a safe temperature. It is beyond the
In addition to producing power, the fission reaction produces fission products.
These products are radioactive (see Fig. 9.17) and produce residual heat even
when the reactor is shut down. This residual heat is a problem which must be dealt
with in the design of the power plant. The energy that is emitted by fission reactions
is shown in Table 9.3.
Table 9.3. Emitted and Recoverable Energy for Fission of U-235
Energy form Emitted energy,
MeV
Recoverable energy, MeV
Fission fragments 168.00 168.00
Fission product decay
rays 8.00 8.00
rays 7.00 7.00
Neutrinos 12.00 0.00
Prompt rays 7.00 7.00
Fission neutrons (kinetic energy) 5.00 5.00
Capture rays 0.00 4.00
207.00 200.00Total
has a desire to learn more, there are courses on nuclear reactor engineering, nuclear
reactor physics and transport theory that you may wish to take.
scope of this course to delve into this physics and system details. If the reader
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480 9 Nuclear
Calculation of the amount of power that fission produces for a given amount of
fuel consumed is very large. A simple calculation can be made relating the reactor
Fission Rate = P (MW) x
day
x
joulex
MeV
x
MeVE
fission
x
MW
joules
R
sec86400
106.1sec
10
13
6
Fission Rate = dayfissions
E
P
x
R
23
1040.5
The burnup rate can be found by assuming that mass of a fissile nuclei is A,
Burnup Rate = 24
23
10602.0
1040.5
x
A
x
E
P
x
R
Burnup Rate = daygm
E
PA
R
/895.0
radioactive decay (Courtesy of [2]).
Fig. 9.17. Distribution of fission products. Most of the products are neutron rich and undergo
power (P), to the fuel burnup, and fuel consumption:
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9.4 Uranium Resources 481
For unranium-235, which releases 200 MeV per fission, the burnup rate is,
Burnup Rate = 1.05 P gm/day (9.49)
So, one MW-day of energy requires about 1 gm of uranium-235 per day.
The fissile material is consumed by both fission reactions and capture reactions
(in which fission does not occur). The total absorption rate factor is 1.175. The
total consumption rate is:
Consumption Rate = daygm
E
PA
R
/175.1895.0 (9.50)
Consumption Rate = 1.24 gm/day
The unit of megawatt-days per metric tonne of fuel, MWD/t, is used to describe
consumption rate. A metric tonne is 1,000 kg or 106
gm. If it were possible to fission
all uranium-235 the total energy release would be 1 MWD/gm which is equal
to 106 MWD/T. The uranium-235 is not all consumed. There are parasitic absorptions
that reduce this number slightly to 800,000 MWD/t.
9.4 Uranium Resources
with 372,000 MWe of total capacity providing 16% of the world’s electricity as
base-load power (Fig. 9.18). Also there are 284 research reactors in 56 countries,
Fig. 9.18. World electricity generation by fuel type.
About 439 commercial nuclear power reactors are operating in 30 countries,
and further 220 reactors that power ships and submarines.
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482 9 Nuclear
In Fig. 9.19 the share of electricity generated by nuclear energy by various
countries is shown. France is leading all the countries with 80%.
Fig. 9.19. Percentage of electricity generated by nuclear by various countries (Adapted from [5]).
As can be noted from Fig. 9.20, the share of nuclear towards total electricity
production is steadily increasing. Interestingly, from 1990 to 2006, world capacity
rose by 44 GWe, or 13.5%, due both to net addition of new plants and uprating of
some established ones. Total electricity production increased to 757 billion kWh
(40%). The relative contributions to this increase were: new construction 36%,
uprating 7% and availability increase 57%. Nuclear power reactors that are under
construction or planned for the near future are given in Appendix IX.
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9.4 Uranium Resources 483
Fig. 9.20. Global share of electricity production by nuclear energy (Adapted from [5]).
The total generation of electricity by nuclear energy is steadily increasing
worldwide. These data are shown in Fig. 9.21. A number of nuclear power plants
are already under construction. Also several more plants have been planned
worldwide which will further increase the share of total electricity production by
nuclear energy. The plants under construction and that are planned are given in
Appendix IX
Fig. 9.21. Yearly nuclear electricity production in the world (Adapted from [5]).
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484 9 Nuclear
9.5 Fuel Cycle
Sustainability of nuclear power production relies heavily upon the closing of the
fuel cycle, i.e., reprocessing of the spent fuel. A number of countries including,
Japan, France, and Russia have been reprocessing spent fuel from their light water
reactor fleet for many years. The United States opted for the once-through cycle in
the early 1970s, but has shifted its philosophy in recent years, as fossil fuel prices
have continued to soar. Also, India and China are planning to reprocess the spent
fuel from the nuclear reactors.
9.5.1 Once-Through Cycle
The fuel cycle begins with mining and milling of uranium, followed by conversion
and enrichment. The fuel cycle is described in details later in this chapter. The
enriched product is then sent to fabrication. The enriched product can be either in
a metal or oxide form. The fuel is encased in a protective barrier known as cladding
and arranged such that critical mass can be achieved within a core, under
proper conditions. This fuel is then cycled through a reactor core, which usually
amounts to three cycles of use, approximately 18 months each. Once the fuel has
reached the end of its useful life, it is transferred to a storage pool, where it decays
over a period of several years. Dry storage is also available for fuel that has
decayed sufficiently and is necessary when spent fuel pools reach their capacity.
This is the current stopping point of the current U.S. fuel cycle.
The next step in the once-through cycle would be the shipment of the spent fuel
to a permanent repository as high level waste (HLW). However, no repository has
been licensed for use and the only active project, Yucca Mountain, still has substantial
political and scientific hurdles to overcome. The topic of waste generation
and final storage is discussed later.
9.5.2 Closed Fuel Cycle
The front end of the closed fuel cycle is the same as that of the once-through
cycle. However, the back end of the closed fuel cycle is substantially different.
Instead of simply storing the spent fuel in a High Level Waste (HLW) disposal
facility, the fuel is reprocessed for recycle back into the fuel cycle. The reprocessing
involves separating the fuel material from the cladding material, retrieving the
fissionable and fissile materials from the mixture, mainly plutonium and uranium,
for recycle. The rest of the long lived fission products are vitrified for long term
storage. The uranium and plutonium can be recycled back into fuel for light water
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9.5 Fuel Cycle 485
and fast reactor use. Additionally, both thermal reactors and fast breeder reactors
U-238. Fast breeder reactors on the other hand produce more fissionable material
than they consume, providing a more efficient means of uranium use. The fuelresources
for nuclear reactors can be extended if reprocessing and fast breeder
reactors are used instead of once-through cycles.
Annual requirements of uranium worldwide for existing power reactors are
about 67,000 t of uranium. According to the authoritative “Red Book” [6] produced
jointly by the OECD’s (Organisation for Economic Co-operation and
Development) Nuclear Energy Agency (NEA) and the UN’s International Atomic
uranium, exploitable at below $80 per kilogram of uranium, are about 3.5 million
tonnes. This amount is therefore enough to last for 50 years at today’s rate of
usage – a figure higher than many widely used metals. Current estimates of all
expected uranium resources (including those not yet economic or properly quantified)
are four times as great, representing 200 years’ supply at today’s rate of
usage. It can be seen from Table 9.4 that Australia has a substantial amount
(about 23%) of the world’s low-cost uranium, followed by Kazakhstan 15%, and
Russia 10%.
Table 9.4. Known recoverable resources of Uranium (2007 data).
Australia 1,243,000 23% Niger 274,000 5%
Kazakhstan 817,000 15% Uzbekistan 111,000 2%
Russia 546,000 10% Ukraine 200,000 4%
Canada 423,000 8% Jordan 112,000 2%
USA 342,000 6% India 73,000 1%
South Africa 435,000 8% China 68,000 1%
Namibial 275,000 5% Mongolia 62,000 1%
Brazil 278,000 5% Other 210,000 4%
World total 5,469,000
Source: World Nuclear Association, London, UK [5]. Reasonably Assured Resources plus
Inferred Resources, to US$ 130/kg U, 1/1/07, from OECD/NEA & IAEA, Uranium 2007:
Resources, Production and Demand, “Red Book” [6].
The amount of uranium that can be extracted from the ores depends on the
uranium price. As can be seen from Fig. 9.22, in countries such as the USA and
Niger, significant amount of uranium can be recovered at a cost of US$ 130/kg of U.
Tonnes UCountry % of world Country Tonnes U % of World
Energy Agency (IAEA), the world’s present known economic resources of
reactors will convert some of the U-238 into plutonium some of which is burned
can make use of the U-238, which is present in 99% of natural uranium. Thermal
water power reactor is derived from the plutonium generated by the converted
during operation of the reactor. About 1/3 of the energy produced in a light
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486 9 Nuclear
* IAEA estimate
Fig. 9.22. Reasonably Assured Resources plus Inferred Resources, to US$ 130/kg U [6].
Note: The Australian figure had risen to 1,558,000 t U as at August 2007 and other country figures
would have risen also, but are not yet published. 1/1/05.
It is expected that the demand for uranium will increase as more and more reactors
are constructed worldwide. With this being anticipated, the investment for exploration
of uranium resources has increased significantly in recent years (Fig. 9.23).
Fig. 9.23. Investment for exploration of uranium (Adapted from [5]).
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9.6 Uranium Supply and Demand 487
The IAEA-NEA estimates that if all conventional resources are considered,
there are about 10 million tonnes of uranium beyond the 4.7 million tonnes known
economic resources, which is enough to go beyond 200 years’ at today’s rate of
consumption. However, if unconventional resources such as phosphate/phosphorite
deposits (22 Mt U recoverable as by-product) and seawater (up to 4000 Mt) are
considered, the available resources would be even greater. These unconventional
resources are currently not economical, but as the price of uranium increases these
resources may become attractive.
The use of the fast breeder reactor would increase the utilization of uranium 50fold
or more. This type of reactor can be started up on plutonium derived from
conventional reactors and operated in closed circuit with its reprocessing plant.
Such a reactor, supplied with natural or depleted uranium for its “fertile blanket”,
can be operated in such a way that each tonne of ore yields 60 times more energy
than a conventional reactor. However, reprocessing and the use of fast breeder
reactors can increase the operation of nuclear reactors significantly (Table 9.5).
Years of 2004 world
nuclear electricity
generation with
identified conventional
resources
Years of 2004 world
nuclear electricity
generation with
total conventional
resources
Years of 2004
world nuclear
electricity generation
with total
conventional and
unconventional
resources
Current once
through fuel cycle
85 270 675
Pure fast reactor 5, 000 – 6, 000 16, 000 – 19, 000 40, 000 – 47, 000
The values in the last row assume that fast reactors allow essentially all U-238 to be bred to
Pu-239 for fuel, except for minor losses of fissile materials during reprocessing and fuel fabrication.
Adapted from Rogner H-H, McDonald A (November 2007) Nuclear Energy - Status and Outlook,
20th World Energy Congress, Rome, Italy [7].
9.6 Uranium Supply and Demand
Uranium is currently extracted in 19 countries, and Canada and Australia account
for more than 50% of the worldwide production. Approximately 40,000 tons of
uranium is produced each year, supplying approximately 60% of the uranium
needed for the existing reactors. The rest of the uranium is made up by secondary
uranium sources, such as natural and enriched stockpiles, re-processing of spent
fuel and enrichment of uranium tails [7].
Cycle
Reactor/Fuel
reactors
with light water
recycling
fuel cycle with
Table 9.5. Years of uranium availability for nuclear power.
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488 9 Nuclear
Demand for uranium has gone up slowly over the past three decades. The
upward trend has been a result of expansion of nuclear power overseas and
more recently, the pending demand from mass global expansion of nuclear power.
Uranium prices have risen more dramatically in the past few years, generating
renewed interest in new exploration. Figure 9.24 shows recent trends in uranium
requirements and production.
Fig. 9.24. Global Annual Uranium Production and Reactor Requirements, 1950–2006 (Adapted
from [7]).
It is assumed that the need of uranium will be between 78,000 and 129,000 t of
uranium by 2030 [7] due to proposed and anticipated new reactors. This will certainly
force uranium prices up and encourage even more mine developments
worldwide.
9.7 Electricity Generation from Nuclear Energy
The steps or processes involved in generation of electricity from nuclear energy
may be divided into following categories. This is also called the Nuclear Fuel Cycle.
Uranium mining and extraction
Enrichment
Fuel Fabrication
Power production
Waste management
Spent fuel reprocessing (If the objective is to recycle unused/unburned
uranium)
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9.7 Electricity Generation from Nuclear Energy 489
The nuclear fuel cycle is shown graphically in Fig. 9.25.
Fig. 9.25. Nuclear fuel cycle (Adapted from [8]).
9.7.1 Uranium Mining and Extraction
Uranium can be found in nature in several ores. Uranium has been found as a significant
constituent in about 150 different minerals and in small concentration in
another 50 minerals. Primary uranium minerals include uraninite and pitchblende.
Other ores from which uranium may be extracted economically include autunite,
tobernite, coffinite, and carnotite.
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490 9 Nuclear
It is estimated that about 90% of world’s low-cost uranium reserves are in Canada,
South Africa, the United States, Australia, Niger, Namibia, Brazil, Algeria, and
France. Sandstone formations in the Colorado Plateau and Wyoming Basin of the
western United States also contain significant reserves of uranium.
9.7.1.1 Mining and Milling
Detailed exploration is necessary prior to any major investment being placed in
mining. There are several methods of exploration utilized today, including geological
mapping, airborne and surface surveys, hydro-chemical sampling, welllogging
and botanical methods [9].
Methods of Mining
Open-pit and underground mining remain the dominant extraction methods in
most of the world, accounting for 68% of production. Additionally, approximately
11% of uranium is produced from in-situ leaching technology, which is the preferred
method in several countries, such as Kazakhstan. The remainder of fresh
uranium supply is produced as a by-product material from copper and gold mining
and a small amount is produced from water treatment operations [6].
Open-pit and underground mining are essentially the same as that used for
other mineral extraction, such as coal. In-situ leaching is a more complex process.
Several holes are drilled, approximately 50 ft apart, and a leaching solution is
injected into the center hole. This solution consists of water, an oxidant and an
ionic complex agent, designed to mobilize and dissolve uranium. The solution is
pumped out from the remaining holes and uranium is removed through ion
exchange. The stripped solution is re-oxidized and treated for proper pH, and is
recycled back into the process. This method is not without problems, as it can contaminate
groundwater, if not performed properly. To prevent this, more water is
pumped out than is pumped in, creating a flow away from the water table. Once
the mine has been fully utilized, it is flushed with clean water and all holes are
plugged. If performed properly, in-situ mining has less environmental impact on
mining sites than open-pit or underground mining and is highly effective for
extracting uranium.
Following uranium mining from open-pit or underground methods, the ore is
crushed, grinded, pulverized into powder and roasted to remove organics. The
uranium is removed via ion exchange or solvent extraction. The resulting U3O8
product is then dried and packaged as yellowcake into 55 gallon drums for shipment
to an enrichment facility.
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9.7 Electricity Generation from Nuclear Energy 491
Mining Wastes
Open-pit and underground mining produce significant amounts of tailings from
the milling process. All of the material left over after the yellowcake is produced,
remains as waste. These tailings typically contain some uranium as well as radon
and other decay products. Additionally, the tailings contain acids from the milling
process, which can leach out various toxic metal from soil and cause groundwater
contamination. Strict enforcement of discharge limits are enforced by the Nuclear
Regulatory Commission (NRC) and Environmental Protection Agency (EPA) in
the United States.
Mill tailings must be sufficiently contained within tailing ponds, so that acids
and radioactive materials do not escape into the environment. Several advances
have been made in recent years, considering tailings were not regulated during the
early years of uranium production.
In-situ mining also has several disadvantages. Although this type of mining
does not produce tailings, the groundwater aquifer near these mines must be
restored to similar conditions as they were before in-situ extraction began. This
can be a rather challenging undertaking and has been the source of criticism for
this method of mining.
Another source of waste generated by all of these mining techniques revolves
around the milling treatment and the chemicals involved. Acids are employed for
separation of uranium from the ore as well as stripping of resin columns. Additionally,
solvent extraction utilizes various chemicals, which must also be disposed
of properly.
9.7.2 Conversion and Enrichment
Although the processes involved in conversion and enrichment have not changed
much over the past two decades, there are some notable exceptions and future
trends.
Prior to fuel enrichment, the U3O8 produced in the mill must be purified and
converted into uranium hexafluoride (UF6). UF6 was chosen as the enrichment
compound due to its unique physical and chemical properties. It is a solid at room
which behaves in this fashion and has made it the ideal candidate for enrichment.
Properties of various uranium compounds are given in Table 9.6. As can be seen
from the table, among all the uranium compounds, only uranium hexafluoride has
the melting point that is low enough to use in the separation process.
temperature, and its melting point is 147°F. This is the only compound of uranium
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492 9 Nuclear
Table 9.6. Melting points of uranium and uranium compounds.
Uranium compounds
Uranium metal 2071
Diboride 4289
Tetrabromide 960
Tribromide 1346
Dicarbide 4262
Tetrachloride 1094
Trichloride 1548
Hexafluoride 147
Tetrafluoride 1760
Tetraiodide 943
Mononitride 4766
Dioxide 4532
Disulfide >2012
Source: Adapted from [10].
Various physical forms of UF6 at various temperatures and pressures are
expressed by its phase diagram. The boundaries between solid, liquid, gaseous
forms detect the operating conditions. The phase diagram of UF6 is shown in
Fig. 9.26. The phase diagram of UF6 (Adapted from. [10]).
Melting point (°F)
Fig. 9.26.
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9.7 Electricity Generation from Nuclear Energy 493
A solvent extraction process, called Plutonium Uranium Extraction (PUREX),
is used for extraction and purification of uranium. The PUREX process takes
advantage of the preference of uranium to form coordinated compounds within
aqueous solutions that can be selectively extracted from the solution by an organic
solvents, such as tributyl phosphate (TBP). Uranium is next precipitated out from
the solution using ammonium hydroxide. The second method involves using
hydrogen peroxide to precipitate out uranium peroxide from weak acid solutions.
Steps involved in producing UF6 from yellowcake are shown in Fig. 9.27, and a
simplified diagram of the solvent extraction unit is shown in Figure 9.28. Once
these methods have been successfully completed, conversion to UF6 follows [9].
Fig. 9.27. Reactions pathways for UF6 production (Adapted from [10]).
Fig. 9.28. A simplified schematic diagram of the solvent extraction unit (Adapted from [10]).
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494 9 Nuclear
Uranium diuranate (ADU) that is precipitated out from the solution is converted
to UO2 by calcining it in hydrogen. UO2 is next converted into UF6 in two
steps. Both steps utilize a fluidized bed reactor, which heats the UO2 to approximately
1,000°F in the presence of HF first and then with F2. The chemical reaction
that takes place is:
UO2 + 4HF 2H2O + UF4
This UF4 is next reacted with fluorine gas to form the UF6 as follows:
UF4 + F2 UF6
The solvent extraction step removes impurities prior to the reduction step. This
results in a very pure form of UF6 at the end of the fluorination process, leaving a
product ready for enrichment [9].
9.7.2.1 Enrichment
topes. Depending on the degree of enrichment, its grade varies. Also a number of
enrichment processes have been suggested, but only few are used commercially.
Grades
Highly enriched uranium (HEU)
Low-enriched uranium (LEU)
Slightly enriched uranium (SEU)
Methods
Thermal diffusion
Gaseous diffusion
The gas centrifuge
The Zippe centrifuge
Aerodynamic processes
Electromagnetic isotope separation
Laser processes
Chemical methods
Plasma separation
Enrichment Process
Among the methods mentioned above, gaseous diffusion and gas centrifuge methods
are most widely used commercially for enrichment of uranium. Laser separation is
still in developmental stage, but is likely to compete with gaseous diffusion and
The enrichment of uranium refers to the increase in concentration of U-235 isoLicensed
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9.7 Electricity Generation from Nuclear Energy 495
gas centrifuge methods in the near future. The United States has relied primarily
on the gaseous diffusion method due to its more simplistic design and successful
operational history. However, currently, most of the countries preferring gas centrifuge
method.
Gaseous Diffusion
The first step in the diffusion process is conversion of solid UF6 to the gaseous
form by heating it above 135°F. It is then forced through a series of porous membranes,
which separates U-235 from the U-238. Since U-235 is lighter than U-238,
it diffuses at a relatively faster rate than U-238 through the membranes. Figure
9.29 illustrates the diffusion process during uranium enrichment. To achieve a U-
235 concentration from its natural 0.7% to 3% to 5%, several thousand stages are
connected in series.
Fig. 9.29. An illustration of gaseous diffusion process for uranium enrichment and a diffusion
cell (Adapted from [11]).
HIGH PRESSURE
FEED STREAM
ENRICHED
STREAM
DEPLETED
STREAMLOW PRESSURE
LOW PRESSURE
BARRIER
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496 9 Nuclear
Gas Centrifuge
In this method, solid UF6 is also heated first to form gaseous UF6, which is then
fed into the center of a rotor, spinning at very high velocity, inside a casing held at
vacuum. Due to the mass differences, the heavier U-238 tends to separate from the
lighter U-235 and is forced outward. The enriched product is then collected from
the center and fed into the next centrifuge. The depleted UF6 is fed back into the
system for further separation. In order to achieve U-235 percentages adequate for
reactor fuel, several thousand stages must be connected in series. Figure 9.30 illustrates
a typical centrifuge unit used for fuel enrichment.
The centrifuge technology has been used worldwide with great success. The
US Energy Corp (USEC) and AREVA are planning to build a new facility in New
Mexico, USA, based on the centrifuge process, and another one near Idaho Falls,
ID, USA. One of the big advantages of centrifuge technology is that it has a 5%
energy savings over gaseous diffusion. Enrichment plants are extremely energy
intensive, and savings from the centrifuge process are very significant. The US
project will be the most advanced fuel enrichment plant in the world, employing
the new AC100 centrifuge machines. Details of these units are classified, but the
USEC expects the plant to produce 3.8 million separative work units (SWU) per
Fig. 9.30. Gas centrifuge unit for uranium enrichment (Adapted from [12]).
year by 2012, using 11,500 machines, once the plant is fully on-line [13].
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9.7 Electricity Generation from Nuclear Energy 497
About 100,000–120,000 SWU is required to enrich the annual fuel loading for
a typical 1,000 MWe light water reactor. Enrichment costs are substantially related
to electrical energy used. The gaseous diffusion process consumes about
2,500kWh (9,000 MJ) per SWU, while modern gas centrifuge plants require only
about 50 kWh (180 MJ) per SWU. Also, the current trend is to use the centrifuge
technology over the diffusion technology (Table 9.7).
Table 9.7. Trends in the use of technology for uranium enrichment.
Supply source 2007 2017
Diffusion 25% 0
Centrifuge 65% 96%
HEU ex weapons 10% 4%
Source: Reference [5].
(The Separative Work Unit (SWU) is a measure of the work expended during an
enrichment process. Uranium enrichment is sold as SWU. Higher levels of U-235
The SWU may be calculated from the following equation:
SWU = P·f(Np) + W·f(Nw) – F·f(Nf) (9.51)
where P is the amount of product, Np is the product concentration, W is the
amount of waste, Nw is the waste concentration, F is the amount of feed, and Nf is
the feed concentration, and f(x) is a value function and is defined as:
x
x
nxxf
1
12 (9.52)
where x is a given concentration.
(The function f(x) is dimensionless. The unit of SWU depends on the units of P, W,
and F, which are generally expressed in units of kilograms (kg) or tonnes. The
SWU is expressed as a kg-SWU or metric tones-SWU. The performance of a centrifuge
can be expressed in terms of rate of enrichment, that is, SWUs per year or
month. Individual centrifuges might be described in terms of kg-SWUs per year.
A SWU per unit time is referred to, not as separative work, but as separative
power).
Laser Separation
This technology differs substantially from the diffusion and centrifuge methods.
Although the different isotopes of uranium have identical chemical behavior, their
electronic energies are very different, causing them to absorb energy at different
require more SWU).
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498 9 Nuclear
wavelength. This is the fundamental basis behind laser technology, enabling isotopic
separation. The laser is focused through the target, with the wavelength
tuned specifically into that for U-235. The targeted U-235 atoms will lose an elec-
235 is collected in a negatively charged surface in the liquid form, where it runs
down to a die cast. The resulting pellets can then be shipped off to fuel fabrication.
The tailings, which amount to 30% less than with other methods, are collected and
can be disposed of or used in the fast reactor. Advantages of this technology
include reduced tailings and laser separation uses just 5% of the energy required
form, eliminating the hazards associated with UF6. Figure 9.31 illustrates the laser
separation method for uranium enrichment.
Scientists at Lawrence Livermore National Laboratory had demonstrated production
scale capability of this enrichment process. However, the technology was
transferred to USEC in the early 1990s, and they discontinued further development
of the project in 1999 primarily due to declining uranium prices. Currently,
Silex System Ltd, New South Wales, Australia, is further developing a laser based
technology based on CRISLA (condensation repression by isotope selective laser
activation) for full scale commercial application. The CRISLA process was invented
by Dr. Jeff Eerkens, University of Missouri, Columbia, Missouri, USA. GE-Hitachi
recently purchased the exclusive rights to the technology and moved the effort to
their Wilmington, North Carolina location. Exelon and Entergy, which are the two
largest nuclear utility owners in the United States, have signed letters of intent
by gaseous diffusion [14]. The uranium feed for laser separation is in the metal
Fig. 9.31. Laser isotope separation method for uranium enrichment [15].
to buy laser-enriched uranium from GE [5].
AVLIS Process Laser System
Tailings
collector
Process
laser
Laser
Vapor
Vaporizer
Product
collector
Uranium
vapor flow
238
U
235
U
lonized235
U
Pump
laser
( ) Charge
collector
Laser
tron (ionize), with no affect on the rest of the feed material, mainly U-238. The ULicensed
to Jiri Martinec<martinec@fme.vutbr.cz>
9.7 Electricity Generation from Nuclear Energy 499
9.7.2.2 Waste Generated by Enrichment Processes
The tailings of uranium enrichment are a key factor that goes into the price of SWU.
Typically, tailings contain 0.2–0.3% by weight of U-235. There is approximately
700 million kilograms of depleted UF6, containing 475 million kilograms of uranium
in the United States. This is stored in 60,000 steel cylinders at the various
laboratories and enrichment facilities. Recycling of this depleted UF6 could be an
economical and environmental success, which has been looked into by scientists at
Lawrence Livermore National Laboratory [13].
9.7.3 Fuel Fabrication
Fuel fabrication varies greatly from one reactor to another, in terms of fuel and
cladding type, fuel enrichment and geometry. The types of commercial reactors
currently used for electricity generation are listed below.
Research reactors
Pressurized water reactor, (PWR)
Boiling water reactor (BWR)
Fast breeder reactor (FBR)
Pressurized Heavy Water Reactor (PHWR) or CANDU
Fuels are produced in either the metal or ceramic form. Metal fuel has the benefits
of higher thermal conductivity and ease of fabrication. Among the drawbacks
of metal fuel are lower melting temperatures and its tendency to undergo growth
upon irradiation. Ceramic fuels have the benefit of providing good retention of fission
products during irradiation; low fabrication costs, and are chemically and
structurally stable. Drawbacks include a brittle structure, susceptible to cracking
and low thermal conductivity. There has been substantial research and development
regarding advanced fuels and will be discussed further later in this chapter.
9.7.3.1 Ceramic Pellet Fuel
Commercial light water and CANDU reactors typically use UO2 pellet fuel (MOX
fuel can also be used in these reactors). There are wet and dry processes available
to convert the enriched UF6 to ceramic grade UO2. The three methods that are
most widely used are the dry conversion, ammonium diuranate and the ammonium
uranyl carbonate routes.
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500 9 Nuclear
The dry conversion route begins with UF6 decomposition using superheated
steam to form UO2F2 particles. The particles then undergo pyrohydrolysis through
interaction with steam and hydrogen. All these steps occur within a rotary kiln,
which is the most common method of conversion in the U.S. and France [16]. The
reactions that take place within the kiln are:
UF6 + 2H2O UO2F2 + 4HF
UO2F2 + H2O UO3 + 2HF
UO3 + H2 UO2 + H2O
The ammonium diuranate route involves hydrolysis of UF6, filtration of the
resulting precipitate and heating to form UO2. Ammonium hydroxide is mixed in
with the feed UF6 to form the ammonium diuranate, which can readily be filtered,
washed to remove as much entrained fluoride as possible and dried to remove
moisture and excess ammonia. The product is then passed on to a pyrohydrolysis
and reduction furnace, which removes the remaining fluorine and completes conversion
to UO2. The reactions that take place during this process are:
UF6 + 2 H2O UO2F2 + 4HF
2 UO2F2 + 8 HF + 14 NH4OH (NH4)2U2O7 + 12 NH4F + 11 H2O
The ammonium uranyl carbonate route is used by Germany and Sweden.
Advantages of this technology include production of a free-flowing, granular UO2
with uniform particle size. The afore-mentioned method requires compaction and
granulation prior to pressing into pellets. The UF6 feed is heated and directed to
a precipitation vessel, which contains a mixture of ammonium hydroxide and
ammonium carbonate. The uranium is precipitated out as ammonium uranyl carbonate.
The solid is then filtered out of solution, washed with clean ammonium
carbonate to remove most of the fluoride, then dried and transferred to a furnace,
supplied with hydrogen and steam at 650°C. The remaining fluoride is removed
as HF and decomposition of the ammonium uranyl carbonate produces UO2. The
reactions that take place during this process are:
UF6 + 5 H2O + 10 NH3 + 3 CO2 (NH4)4UO2(CO3)3 + 6 NH4F
(NH4)4UO2(CO3)3 + H2 UO2 + 3 CO2 + 4 NH3 + 3 H2O
Most of the light water fuel used today is in the form of small cylindrical
ceramic pellets. Enriched UO2, from the previously mentioned processes, is mixed
with an organic pore-forming agent, which will later decompose during the sintering
process. The pores that are formed provide the space for collection of fission
products produced during irradiation, preventing fuel growth beyond tolerances.
A binding agent is also added to the powder prior to pressing, which results in a
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9.7 Electricity Generation from Nuclear Energy 501
“green” pellet prior to sintering. The selection of binding agent is crucial since this
determines the amount of handling damage that will occur to the pellet prior to
sintering, thus lowering rejection rates. Pressure in the range of 3 to 4 tons/cm2
is applied to the powder-mixture within a die until the resulting pellet density is
approximately 5.5–6.0 g/cm3
. The final process of pellet production is sintering,
which is performed in ovens at temperatures of approximately 1750°C [16]. The
sintering converts the green pellet into the ceramic fuel with a crystalline grain
structure while retaining the original shape. The pellets are then grounded to strict
tolerances and the removed material is recycled back into the process.
9.7.3.2 Research Reactor Fuel Assembly
The Training, Research, Isotopes, General Atomics (TRIGA) reactor is the most
widely used research reactor throughout the world. It can be built with thermal
power levels ranging from 10 kW to 10 MW. A photograph of the fuel pellet and a
fuel pin of the TRIGA reactor is shown in Fig. 9.32. Depending on the type of the
research reactor, the fuel assembly is also different and is shown in Fig. 9.33.
Fig. 9.32. Fuel pellets along with fuel pins used in TRIGA reactor (Printed with permission
from [17]).
Fig. 9.33. Typical research reactor fuel assembly (Adapted from [18]).
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502 9 Nuclear
As shown in Fig. 9.34, the assembly of the TRIGA fuel assembly contains a
number of other components.
Fig. 9.34. The details of a TRIGA fuel pin design (Adapted from [19]).
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9.7 Electricity Generation from Nuclear Energy 503
9.7.3.3 Light Water Reactor Fuel Assemblies
The fuel assembly for light water reactors starts with the loading of pellets into
fuel pins. Several materials have been utilized for these pins, including various
alloys of stainless steel and zircoloy. However, zircoloy has become the material
of choice due to its high heat transfer capabilities, low absorption cross section
and chemical stability when exposed to the coolant and fuel. One end of each pin
is welded, and fuel pellets are stacked and a spring holds them in place. This is
shown in Fig. 9.35. The pin is evacuated, pressurized with helium and the open
end is sealed and welded. PWR fuel pins are pressurized to approximately 2,000
psi and BWR fuel pins are pressurized to approximately half of that. This is due to
different operating pressures within the pressure vessels of these reactors.
The loaded fuel pins are arranged into a proper geometry for the respective
application. PWR and BWR cores utilize a variety of square fuel pin arrays. A disassemblies
have guide tubes, which allow control rods to slide in and out. BWR
control rods are a crucifix shape and are placed in the gaps between assemblies.
BWR assemblies are also surrounded by a zircoloy shroud, to prevent coolant
mixing between assemblies.
Fig. 9.35. Basic steps in fuel rod and fuel assembly construction (Printed with permission
from [20]).
tinct difference lies in the control rod accommodations of the two core designs. PWR
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504 9 Nuclear
9.7.3.4 CANDU Reactor Fuel Assemblies
CANDU reactors utilize the pressurized water design. The fuel pin design is
essentially the same, but the assembly arrangement is different from PWR. A distinct
advantage of CANDU reactor design is that it does not require enriched fuel.
This drastically reduces the front-end costs associated with fuel production. Natural
UO2 pellets are again stacked into zircoloy pins, with a spring holding them in
place. A graphite layer is added to CANDU fuel pins between the pellets and
zircoloy pin. The pins are pressurized with helium, welded and leak checked.
Fig. 9.36 shows a CANDU fuel assembly. As can be seen from the figure, it
has significantly different arrangement than PWR or BWR fuel assemblies. The
circular assemblies are 0.5 m long and are arranged horizontally within the core.
Fig. 9.36. A CANDU fuel assembly (Courtesy of Atomic Energy of Canada Limited, Mississauga,
Ontario, Canada).
9.7.3.5 Mixed Oxide (MOX) Fuel Assemblies
The objective of reprocessing is to use separated plutonium as the fuel in the PWR
or BWR reactors. Pu-239 is generated in light water reactors, which is a fissile isotope
and responsible for approximately 30% of power production in PWR and
BWR cores. Spent fuel from these reactors still contains a substantial amount of
fissile isotopes, including the Pu-239. During reprocessing of light water spent
fuel, Pu-239 is collected and mixed with natural or enriched uranium to produce
mixed oxide fuel (MOX). This fuel may contain up to 8% PuO2 and run in existing
PWR cores with up to 30–40% of the MOX fuel. No significant core modifications is
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9.7 Electricity Generation from Nuclear Energy 505
necessary [16]. The use of MOX fuel greatly enhances the utilization of reactor
fuel beyond the once-through cycle and has been utilized by Japan and France for
many years. The United States is currently in the process of licensing a facility for
MOX production as a partial solution to the vast supply of spent reactor fuel
awaiting disposal.
9.7.3.6 Metal Fuel
Metallic fuel has not been heavily utilized in the past, but may be ready to become
more prominent with the development of fast breeder reactors and new pyrometallurgical
procedures, which combine reprocessing and fuel fabrication [21].
Advantages of metal fuel include relative ease of fabrication and its extremely
high heat transfer capabilities, as compared with oxide fuels. The main disadvantage
of metallic fuels is the low melting temperature, which has been a chief concern
of regulatory agencies. However, the very high heat transfer capability of
these fuels overcomes the low melting points and ensures fuel stability.
9.7.3.7 Thorium Fuel
Thorium is far more plentiful in nature than uranium. This is especially true in
India, where the ratio of thorium to uranium reserves is approximately 6 to 1.
India’s growth in energy consumption is dramatic, and they have invested substantial
resources into finding ways to utilize their thorium reserves. Table 9.8 provides
the estimated thorium reserves worldwide at an extraction cost of $80/kg.
Table 9.8. Thorium reserves of various countries [22].
Country Tonnes % of world
Australia 452,000 18
USA 400,000 16
Turkey 344,000 14
India 319,000 13
Venezuela 300,000 12
Brazil 221,000 9
Norway 132,000 5
Egypt 100,000 4
Russia 75,000 3
Greenland 54,000 2
Canada 44,000 2
South Africa 18,000 1
Other countries 33,000 1
World total 2,492,000
RAR + Inferred to USD 80/kg Th
“Source” Data for Australia compiled by [22]
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506 9 Nuclear
Thorium (Th-232) can be fabricated as an oxide fuel for pellets, micro-particle
(for TRISO fuel) or metal fuel. As oxide fuel, thorium can be placed around
enriched uranium providing a blanket for the fissile material. The concept is
essentially the same as that of a light water breeder reactor. Th-232 is a fertile
material and absorbs neutrons to become U-233, which fissions and keeps the
chain reaction going. However, Pu-233 is also generated during this process. Also,
there are substantial concerns about reprocessing, as U-233 has some intense
gamma-emitting daughters with short half-lives.
9.7.3.8 Fuel Cladding
Fuel cladding serves as a protection device for the fuel, provides a barrier for fission
products escaping the fuel and acts as a heat transfer surface from the fuel to
the coolant. Ideal cladding material must have a small cross section for neutron
absorption and interact well with the fuel and coolant. Materials that have been
used as cladding include stainless steel and zircoloy, the latter being most heavily
utilized for most applications.
Future reactors will need advanced cladding materials, as many of these reactors
will subject the fuel to much higher operating temperatures. TRISO fuel,
which will be discussed later in this chapter, is well suited for higher operating
temperatures. The pyrolytic-carbon layers protect the fuel from the high temperatures
and the silicon carbide layer provides the extra fission product barrier.
Unlike prior fuel designs, the TRISO fuel cladding is incorporated into the fuel
particle itself. The disadvantage of this unique design is the reprocessing challenges
this represents, making once-through of TRISO a distinct possibility. Other
future reactors concepts do not have cladding at all. Molten fuel concepts have no
specific fuel structure and hence, no cladding.
9.8 Uranium Downblending
Uranium used in nuclear weapons is enriched to approximately 93% of U-235,
while uranium used in commercial nuclear power plants typically is enriched to 3–
5% U-235. Uranium enriched to more than 20% U-235 is called Highly Enriched
Uranium (HEU) and can only be used in nuclear weapons and in research reactors.
Surplus HEU can, however, be downblended with Low Enriched Uranium (LEU)
to make it suitable for use in commercial nuclear fuel.
The downblending only involves uranium. In contrast, plutonium is used for
the production of mixed oxide fuel (MOX).
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9.9 Power Production/Burn Up 507
In 1993, the U.S. and Russia signed the US-Russia HEU Agreement, under
which Russia was to supply the downblended uranium derived from 500 t of HEU
to the USA over a period of about 20 years. While the deliveries under this agreement
are still ongoing, the U.S. now have begun downblending some of their own
surplus HEU.
9.8.1 Blending Process
In a first step, the HEU and the blendstock have to be converted to the chemical
form required for the selected blending process, if not already in the appropriate
form. For the downblending process, there exist the following methods:
Mixing of liquids
Uranium in the form of uranyl nitrate hexahydrate (UNH), UO2(NO3)2·6H2O,
or molten uranium metal are mixed together to form the final mixture.
Mixing of gases
Uranium is converted to uranium hexafluoride (UF6) and mixed together in the
gaseous phase.
The existing commercial downblending facilities in the U.S. (BWXT in
Lynchburg, Virginia, and NFS in Erwin, Tennessee) are using the UNH process,
while the Russian facilities (in Novouralsk, Seversk, and Zelenogorsk) are using
the UF6 process. Historically, downblending has also been performed at the following
DOE nuclear weapons facilities in the U.S.: the Y-12 Plant in Oak Ridge,
Tennessee (UNH and molten metal processes), and the Savannah River Site (SRS)
in Aiken, South Carolina (UNH process).
After the blending, the material has to be converted to UO2, before it can be
used in the production of commercial nuclear fuel. The blending process has been
further described by Arbital and Snider [23].
9.9 Power Production/Burn Up
A nuclear reactor core is composed of a several hundred “assemblies”, arranged in
a regular array of cells, each cell being formed by a fuel or control rod surrounded,
in most designs, by a moderator and coolant, which is water in most reactors.
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508 9 Nuclear
As the fuel is consumed (under go fission), the old fuel rods must be replaced
periodically with fresh ones (this period is called a cycle). The fuel is generally
replaced every 18–24 months. Only a part of the assemblies (typically one third)
are removed since the fuel depletion is not spatially uniform, and the rest of the
fuels are rearranged in the core. The new assemblies are not placed exactly at the
same location of the removed ones. Even bundles of the same age may have different
burn-up levels, which depends on their previous positions in the core. Thus
the available bundles must be arranged in such a way that the yield is maximized,
while safety limitations and operational constraints are satisfied. Consequently
reactor operators are faced with the so-called optimal fuel reloading problem,
which consists in optimizing the rearrangement of all the assemblies, the old and
fresh ones, while still maximizing the reactivity of the reactor core so as to maximize
fuel burn-up and minimize fuel-cycle costs.
9.10 Types of Nuclear Reactors
9.10.1 Research Reactors
Most research and training reactors are pool type and are not designed for power
production. In pool type reactor, the core containing fuel rods and control rods is
immersed in an open pool of water. The water under normal pressure acts as neutron
moderator, cooling agent and radiation shield. In this type of reactors, operainstances
for process heat but not for electricity generation.
Research reactors typically range from 10 kW to 10 MW. Research reactors are
simpler than power reactors and operate at lower temperatures. These reactors
typically need far less fuel, but more highly enriched uranium, typically up to 20%
U-235; although some use 93% U-235. The core is cooled typically by natural or
forced convection with water, and a moderator is required to slow down the neutrons
and enhance fission. To reduce neutron loss from the core generally a reflecTRIGA
reactor is shown in Fig. 9.37.
tors may work above the reactor safely since water can provide adequate shielding.
Pool reactors are used as a source of neutrons and for training, and in rare
tor is used. The most common type of research reactor is called TRIGA. A typical
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9.10 Types of Nuclear Reactors 509
Fig. 9.37. The cutaway view of a 10 kW TRIGA Mark I reactor at San Diego, CA, USA (Printed
with permission from [17]).
9.10.2 Commercial Reactors for Electricity Generation
continuously undergoing design changes. Several types of the nuclear reactors are
in use either commercially or for research purpose. The power reactor deployment
timeline is shown in this figure, from past reactors to anticipated future reactors,
although it has shifted to the right multiple times.
As can be seen from Fig. 9.38, nuclear reactors for electricity generation are
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5109Nuclear
Fig. 9.38. Timeline of deployment of various generations of nuclear power reactor. Adapted from U.S. Department of Energy (2002) A technology
roadmap for Generation IV nuclear energy systems. U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International
Forum, Decemfber 2002, GIF-002-00. http://www.ne.doe.gov/GenIV/documents/gen_iv_roadmap.pdf [24].
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9.11 Generation II Light Water Reactors 511
9.11 Generation II Light Water Reactors
439 reactors are operating worldwide at the present time, with 104 in the U.S.
been made over the years in order to increase the efficiency and improve safety.
However, the basic design of the reactors has not changed substantially. Although
both of these reactor designs utilize water as a moderator and coolant, there are
distinct differences between these two reactors.
The light water design necessitates the use of enriched uranium fuel, typically
3–5% U-235. After a typical three cycles of use, the fuel reaches its maximum
burn-up and is no longer suitable as fuel for these reactors. The fuel is allowed to
decay in spent fuel pools, which provide radiation shielding and decay heat
removal capabilities. Once sufficient decay time has been reached, the fuel can be
placed in dry storage, sent to a final repository site, or re-processed to retrieve useful
fuel isotopes and minimize the high level waste requiring ultimate disposal.
Descriptions of Generation II reactors can be found in a number of nuclear engineering
text books [25–31] and NRC reactor concept manuals [32–34].
9.11.1 Pressurized Water Reactor (PWR)
Pressurized water reactors (PWR) account for approximately 60% of U.S. operating
reactors. The primary coolant system consists of large pumps, which circulate
the water through the reactor, where the energy is transferred from the fuel elements
to the coolant, then on to a secondary coolant circuit to produce steam. The cooler
water is then circulated back to the reactor and the process repeats. Another essential
component within the primary coolant system is a high pressure pump, which
maintains primary coolant pressure at approximately 2,000 psi. This prevents the
coolant from flashing to steam at its normal operating temperature of 600°F .
The steam is used to run both high and low pressure turbines to generate the
electricity. The exhausted steam is directed to a condenser, where it is cooled
through interactions with a third coolant (usually river or sea water). The condensed
secondary water is then pumped back to the steam generator and the cycle
repeats.
A schematic flow diagram of a PWR nuclear power plant is shown in Fig. 9.39.
The PWR utilizes UO2 pellets within zircoloy pins. These pins are arranged into a
square assembly, typically 14 × 14 to 18 × 18 arrays as shown in Fig. 9.40. The
assembled core is shown in Fig. 9.41. There are approximately 120–190 of these
assemblies in each reactor core. A PWR vessel with the fuel assembly and other
accessories is shown in Fig. 9.42.
Current nuclear reactors used for power production are Generation II reactors, and
PWRs and BWRs belong to Generation II reactor designs. Several advances have
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512 9 Nuclear
Fig. 9.39. Schematic of a PWR system (Printed with permissionfrom [35]).
Fig. 9.40. Initial fuel loading of PWR and their arrangements in the fuel assembly (Courtesy of
Westinghouse Electric Corp. Printed with permission from [25]).
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9.11 Generation II Light Water Reactors 513
. .
Fig. 9.41. A PWR fuel assembly for 17 × 17 array of rods (Courtesy of [33]).
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514 9 Nuclear
Fig. 9.42. A PWR pressure vessel with fuel assembly and various reactor components (Courtesy
of [34]).
9.11.2 Boiling Water Reactor (BWR)
Boiling water reactors (BWR) encompass approximately 40% of operating commercial
reactors in the U.S. A schematic diagram of a BWR nuclear power plant is
shown in Fig. 9.43. Although water is still utilized as both coolant and moderator,
primary coolant is circulated into the pressure vessel and directed to the bottom of
the reactor core, similar to PWR designs. However, BWR coolant is at a much
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9.11 Generation II Light Water Reactors 515
lower pressure than their PWR counterparts and mass boiling is allowed in the
fuel region. The steam is collected at the top of the pressure vessel, where the
quality is increased prior to direct feed into the main turbine for electricity generation.
Similarly to a PWR, the exhausted steam is fed through a condenser where
secondary water (lake, river, sea water, etc.) cools the primary coolant and condenses
it back to a liquid. This liquid is then returned to the pressure vessel and
the process repeats.
Fig. 9.43. Flow diagram of a BWR direct cycle system (Courtesy of [36]).
BWR fuel consists of oxide pellets in zircoloy pins, arranged in typically 9 × 9
or 10 × 10 arrays, as shown in Fig. 9.44. These assemblies are much smaller and
lighter than PWR assemblies, but a far greater number (up to 800) is required to
make up the larger BWR core.
Fig. 9.44. Fuel rod arrangement in a BWR core (Printed with permission from [25]).
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516 9 Nuclear
The BWR fuel assembly and the cutaway view of the pressure vessel are shown
in Figs. 9.45 and 9.46, respectively.
Fig. 9.45. A BWR fuel assembly (Adapted from [37]).
A set of pumps re-circulates water within the pressure vessel to help pre-heat
the incoming coolant water. Start-ups and shut-downs are primarily accomplished
by manipulating the control rods into and out of the reactor core, which are
inserted from the top of the core.
The transfer of heat from one coolant medium to another is a source of efficiency
loss in any reactor. Since the BWR design incorporates core cooling and
steam generation into one loop, this provides an advantage over the PWR design.
BWR pressure vessels are not subjected to higher pressures and do not require the
thicker vessel walls, drastically reducing the cost and difficulty of forging. However,
these economical advantages are countered by the radiological issues associated
with BWR designs. Since the steam is in direct contact with the turbine, it
becomes contaminated and requires massive and costly shielding during operation.
Accident analysis becomes more challenging as well, since a portion of the
primary loop must exit containment to drive the turbine.
inserted through the bottom of the core. In contrast, in a PWR control rods are
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9.11 Generation II Light Water Reactors 517
Fig. 9.46. Cutaway view of the model BWR pressure vessel (Courtesy of GE Nuclear Energy).
A major difference between the PWR and the BWR design is that in the PWR
the primary and secondary coolant systems have boundaries between one another,
which are not crossed under normal conditions. This allows containment of the
PWR primary system within one building and prevents radioactive contamination
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518 9 Nuclear
of the steam turbine. Isolation of the primary coolant within a single containment
structure offers an additional benefit with regards to accident analysis, such as a
loss of coolant accident.
To reiterate, the PWR design is not without its disadvantages. The pressure
vessel of the PWR is thicker and costlier than that of the BWR due to the high
pressure of the primary coolant. Additionally, the use of three heat transfer mediums
in the PWR lowers the overall plant efficiency.
9.11.3 CANDU Reactors
Heavy water (D2O) is an extremely good moderator; it slows down neutrons without
absorbing them. This efficient use of neutron inventory allows the use of natural
uranium to continue the chain reaction, where it otherwise would not be possible
with the use of other moderating materials.
As mentioned previously, Canada has a vast supply of uranium deposits and is
one of the world’s leading suppliers. Canada has never pursued a nuclear weapons
program and thus did not feel the need to build a fuel enrichment facility. However,
during WWII, Canada assisted the U.S. by developing the first heavy-water
moderated reactor in Montreal, which could be used to produce plutonium. After
the war ended, Canada continued to develop the heavy water moderated reactor,
and it has now evolved into a pressurized heavy water moderated reactor, called
the CANada Deuterium Uranium (CANDU) reactor. The reactor system is shown
Fig. 9.47. Layout of a CANDU reactor system (Printed with permissionfrom [38]).
The CANDU design is similar to the PWR design in many respects. It is also
pressurized to prevent boiling of the coolant and steam generators convert secondary
coolant into steam for the turbine. In a CANDU reactor the fuel assembly is
in Fig. 9.47.
Steam
generator
Steam
line
Concrete
shield
Heavy
water
Fuel
elements
Pressure tubes
Calandria
Control
rods
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9.11 Generation II Light Water Reactors 519
horizontally mounted, allowing for fuel bundles to be replaced during operation
without any need to shut down the reactor completely. These bundles consist of 37
zircoloy fuel rods, containing uranium oxide pellets. Twelve bundles lie end to
end in each channel, and are simply forced out of one end by new bundles being
inserted into the opposite end. The inside view of a CANDU is shown in Fig. 9.48.
Fig. 9.48. The cutaway view of a CANDU reactor assembly (Adapted from [39]).
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520 9 Nuclear
CANDU technology remains popular within Canada and has been partially
adopted in India due to its ability to utilize thorium fuels. Flexible fuel capabilities
are one of the chief advantages of this design. Additionally, the ability to re-fuel
during operation pushes capacity factors much higher than is usually seen in PWR
and BWR designs. Utilizing natural uranium eliminates enrichment costs, drastically
reducing the price of fuel. The biggest drawback of the CANDU design is
the high cost of heavy water. The CANDU system is shown in Fig. 9.49.
Fig. 9.49. A CANDU reactor (Printed with permission from [35]).
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9.11 Generation II Light Water Reactors 521
9.11.4 RBMK Reactors
The Russian water-cooled, graphite moderated reactors were developed in the
former Soviet Union and 27 have been built and operated (Fig. 9.50). RBMK
cores have some similarities to Advanced Gas Reactor (AGR) and BWR designs.
The overall layout is similar to an AGR, but each fuel assembly is housed in a
pressurized tube, containing light water as a coolant. Despite the pressure, the
coolant is allowed to boil and directly feeds the turbine. The fuel is made up of 18
zircoloy rods, containing enriched uranium oxide, to make up a 3.5 m long bundle
[16]. The fuel bundles are stacked in the core.
Negative void temperature coefficients are inherent in the design of most
nuclear reactors and add a large degree of safety. A detailed analysis of the reactor
safety system has been given by IAEA [40]. As temperature rises within the core,
the reactor essentially shuts itself down as the void co-efficient takes on a greater
influence on core reactivity. This is not the case with the RBMK design, as graphite
has a positive void co-efficient and any increase in temperature adds positive
reactivity to the core. This design instability became infamous in 1986, when
Chernobyl reactor number 4 catastrophically failed, during planned testing, in which
operators had disregarded critical operational procedures and safety systems were
over-ridden. Since then, the RBMK design has been modified to prevent a similar
occurrence. The most significant change has been utilization of higher enriched
fuels to help absorb thermal neutrons and create a negative void feedback.
Fig. 9.50. A RBMK reactor system (Printed with permission from [38]).
Concrete shield
Graphite
Fuel elements
tubes
moderator
Pressure
Steam
drums
Control rods
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522 9 Nuclear
9.12 Generation III & III+ Reactors
Generation III and III+ reactors are the next generation of reactors, some of which
have already been built and are operating worldwide. Industry experience has
prompted drastic alterations in the design, construction, licensing and operation of
nuclear power plants. In the past, reactors were built to suit individual companies
and even individual sites. This will no longer be the case, as each of the next
generation designs will be built more modularly, requiring the customers to prepare
their sites to suit the reactor design. The shift in design philosophy is intended to
reduce construction time and costs, stream-line licensing procedures for regulators
and standardize operational procedures.
Many of these new designs have incorporated passive or inherent safety features,
some requiring no active controls or operator action to mitigate design accidents.
This important characteristic will allow the new reactors to operate more efficiently
and produce more power than their predecessors. Additionally, these reactors have
been designed for longer operational lifetimes (60 years) and higher fuel burnup.
The Generation III and III+ reactors are listed in Table 9.9.
Table 9.9. Generation III and III+ reactors.
Reactor Type Abbreviation
Used
Design life Vendor
Advanced Boiling Water
Reactor
ABWR General Electric, USA
Advanced Passive-1000 AP1000 60 Westinghouse Electric
Co., USA
US Evolutionary Power
Reactor
USEPR 60 Areva, France
US Advanced Pressurized
Water Reactor
USAPWR 60
Economic Simplified
Boiling Water Reactor
ESBWR General Electric, USA
IRIS International Team
System 80+ System 80+ Westinghouse Electric
Co., USA
9.12.1 Advanced Light Water Reactors
Several different designs of advanced light water reactor are in the process of
being deployed worldwide. A focus will be made on the most likely designs to be
deployed and the regions of the world in which they will be located.
Industries, Japan
Mitsubishi Heavy
Canada Ltd., Canada
Atomic Energy
Reactor
Advanced CANDU CANDUACRS
Innovative and Secure
International Reactor
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9.12 Generation III & III+ Reactors 523
Utilities within the U.S. have several designs under consideration, although
none have yet been built. NRG Energy and South Texas Project are set to begin
construction on the first two new generation reactors in the U.S, and already
submitted the combined license application to the NRC. These will be of the
that have been operating in Japan since 1996, with four more under construction.
This is a proven design, which will dramatically reduce the construction time and
enhance operational confidence. The ABWR (Figs. 9.51 and 9.52) features internal
circulating pumps for improved safety, fully digital control systems and demonstrated
reduction in construction and operational costs [41].
Fig. 9.51. The layout of an ABWR power plant (Courtesy of GE Hitachi Nuclear Energy.
www.ge-energy.com/nuclear).
design was selected mainly due to operational experience of the four similar units
advanced boiling water reactor (ABWR) designed by General Electric. This
ABWR Plant Layout
1. Reactor Pressure Vessel
2. Reactor Internal pumps
3. Fine Motion Control Rod Drives
4. Main Steam Isolation Valves
5. Safety/Relief Valves (SRV)
6. SRV Quenchers
7. Lower Drywell Equipment Platform
8. Horizontal Vents
9. Suppression pool
10. Lower Drywell Flooder
11. Reinforced Containment
Concrete Vessel
12. Hydraulic Controls Units
13. Control Rod Drive Hydraulic
System Pumps
14. RHR Heat Exchanger
15. RHR Pump
16. HPCF Pump
17. RCIC Steam Turbine and pump
18. Diesel Generator
19. Standby Gas Treatment Filter
and Fans
20. Spent Fuel Storage Pool
21. Refueling Platform
22. Shield Blocks
23. Steam Dryer and Separator
Storage pool
24. Bridge Crane
25. Main Steam Lines
26. Feedwater Lines
27. Main Control Room
28. Turbine-Generator
29. Moisture Separator Reheater
30. Combustion Turbine Generator
31. Air Compressor and Dryers
32. Switchyard
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524 9 Nuclear
Fig. 9.52. The cutaway view of the ABWR reactor vessel (Courtesy of GE Hitachi Nuclear
Energy. www.gepower.com/prod_serv/products/nuclear_energy/en/downloads/abwr_callouts.pdf).
9.12.2 Advanced Passive-1000 Reactor (AP-1000)
Westinghouse has received final design certification from the NRC for its
Advanced Passive-1000 (AP-1000) reactor, the first Generation III+ design to do
so. The AP-1000 design has improved upon proven PWR technology to include
Advanced Boiling Water Reactor Assembly
1. Vessel Flange and Closure Head
2. Vent and Head Spray
3. Steam Outlet Flow Restrictor
4. RPV Stabilizer
5. Feedwater Nozzle
6. Forged Shell Rings
7. Vessel Support Skirt
8. Vessel Bottom Head
9. RIP Penetrations
10. Thermal Insulation
19. Feedwater Sparger
20. High Pressure Core Flooder
(HPCF) Sparger
21. (HPCF) Coupling
22. Low Pressure Flooder (LPFL)
23. Shutdown Cooling Oulet
24. Steam Separators
25. Steam Dryer
26. Reactor Internal Pumps (RIP)
27. RIP Motor Casing
28. Core and RIP Differential
Pressure Line
29. Fine Motion Control Rod Drives
30. Fuel Assemblies
31. Control Rods
32. Local Power Range Monitor
11. Core Shroud
12. Core Plate
13. Top Guide
14. Fuel Supports
15. Control Rod Drive Housings
16. Control Rod Guide Tubes
17. In Core Housing
18. In-Core Insturment Guide Tubes
and Stabilizers
2
1
25
24
4
22
3
5
19
6
23 20
13
30
31
11
14
7
9
18
12
28
26
16
8
17
15
10
27
29
32
21
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9.12 Generation III & III+ Reactors 525
enhanced safety features, as the most serious accidents require no operator action
or AC power to shut down the plant or remove decay heat. This AP-1000 (Fig. 9.53)
is an 1,100 MWe, scaled-up version of the AP-600 originally designed [42–47].
The steam generators (Fig. 9.54) are located within the containment building. The
additional safety features shown in Fig. 9.55 makes it very attractive among other
reactor designs.
Fig. 9.53. AP 1000 reactor pressure vessel assembly (Printed with permission from [47]).
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526 9 Nuclear
Fig. 9.54. Steam generator assembly of AP1000 (Printed with permission from [47]).
Fig. 9.55. Safety features of AP1000 (Printed with permission from [35]).
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9.12 Generation III & III+ Reactors 527
9.12.3 Advanced Pressurized Water Reactor (APWR)
Mitsubishi has designed a large 1,700 MWe Advanced PWR (APWR), which
boasts one of the largest outputs of any commercial reactor designs to date and is
shown in Fig. 9.56. This plant is a larger and more efficient version of the PWR
designs currently in operation throughout the U.S. An application for standardized
design certification was submitted to the NRC in December, 2007 and is awaiting
approval. The fuel assembly of APWR is shown in Fig. 9.57, and the pressure
vessel is shown in Fig. 9.58.
Fig. 9.56. Layout of a APWR systems (Printed with permission from [48]).
9.12.4 Evolutionary Power Reactor (EPR)
France has become the nuclear industry leader over the past two decades. The
European Pressurized Reactor (EPR), shown in Fig. 9.59, was designed by Areva,
which is a 1,700 MWe unit and uses four redundant safety loops and passive systems.
Two of these plants are currently under construction in Europe and more are
scheduled for construction worldwide. AREVA has submitted for regulator
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528 9 Nuclear
Fig. 9.57. Fuel assembly of USAPWR (Printed with permission from [48]).
Fig. 9.58. The cutout view of US APWR designed by Mitsubishi Heavy Industries for US market
(Printed with permission from [48]).
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9.12 Generation III & III+ Reactors 529
Fig. 9.59. Cut out of the EPR pressure vessel. US EPR (Adapted from [49]).
1
2
3
4
5
6
7
8
9
11
10
Section through reactor pressure vessel of EPR showing RPV internals
1 Control rod drive mechanism
2 Liquid level probe
3 RPV closure head
4 Control rod guide assembly
5 Coolant inlet nozzle
6 Core barrel
7 Fuel assembly
8 Lower core support grid
9 Flow distribution plate
10 Coolant outlet nozzle
11 Fuel assembly with inserted control rod
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530 9 Nuclear
approval in several countries including the USA. This modified design for the
USA has been renamed as the Evolutionary Power Reactor (EPR). The layout for
an USEPR is shown in Fig. 9.60 and the cutaway of the reactor to be used in the
USEPR is shown in Fig. 9.61.
Fig. 9.60. The layout of the US-EPR power plant (Adapted from [49]).
Licensed to Jiri Martinec<martinec@fme.vutbr.cz>
9.12 Generation III & III+ Reactors 531
Fig. 9.61. The reactor pressure vessel for US-EPR (Adapted from [49]).
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532 9 Nuclear
9.12.5 Economic & Simplified Boiling Water Reactor (ESBWR)
General Electric has designed a 1,390 MWe reactor, called the Economic & Simplified
Boiling Water Reactor (ESBWR) to compete with other Generation III+ reactors
[50–53]. The technical data of an ESBWR system is given in Table 9.10. The
ESBWR is designed with natural circulation cooling and will have better safety
features then the current BWR (Figs. 9.62–9.64). A comparison of various safety
features between BWR and ESBWR is given in Table 9.11. This reactor is behind
the ABWR and AP-1000 in terms of projected deployment, but orders have been
already placed for the ESBWR.
Fig. 9.62. A schematic of the reactor pressure vessel of Economic Simplified Boiling Water
Reactor (ESBWR) (Printed with permission from [51]).
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9.12 Generation III & III+ Reactors 533
Fig. 9.63. Cutaway of ESBWR reactor (Courtesy of GE Hitachi Nuclear Energy).
Table 9.10. ESBWR, Technical fact sheet
Parameters Values
Plant life (years) 60
Thermal power 4,500 MW
Electrical power 1,560 MW
Plant efficiency 34.7 %
Reactor type Boiling water reactor
Core
Fuel type Enriched UO2
Fuel enrichment 4.2% c
No. of fuel bundles 1,132
Coolant Light water
Moderator Light water
Operating cycle lengtha
12–24 months
Outage durationb
~14 days
Percent fuel replaced at refueling See footnote d
Average fuel burnup at discharge ~50,000 MWd/MT
Number of Steam Lines 4
Number of feedwater trains 2
Containment Parameters
Design temperature 340°F
Design pressure 45 psig
(Continued)
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534 9 Nuclear
Table 9.10. (Continued)
Parameters Values
Reactor Parameters
Design temperature 575°F
Operating temperature 550°F
Design pressure 1,250 psig
Nominal operating pressure 1,040 psia
Feedwater & Turbine Parameters
Turbine Inlet/Outlet temperature 543/93°F
Turbine Inlet/Outlet pressure 985/0.8 psia
Feedwater temperature 420°F
Feedwater pressure 1,050 psia
Feedwater flow 4.55 x 104 gpm
Steam mass flow rate 19.31 x 106
lbs/h
Yearly waste generated
High level (spent fuel) 50 t
Intermediate level (spent resins, filters, etc.) and
Low Level (compactables/non-compactables)
Waste
1,765 ft3
a
Days of operation between outages; b
For refueling only
c
For a 24 month cycle; d
20% for a 12 month cycle, 42% for a 24 month cycle
Source: Reference [53].
Table 9.11. A comparison of safety features between BWR and ESBWR.
ESBWRFunction Current BWR reactors
safety systems Safety systems Nonsafety
High-pressure
inventory
control
Motor and/or steam driven
pumps with some vessel
inventory loss and
containment heat up
Isolation condensers
conserve coolant inventory
and avoid containment
heat up
Multiple motordriven
pumps
Depressurization
and low
pressure
inventory
control
Automatic depressurization
system with complex
cooling water systems
Diverse/redundant
automatic depressurization
system using pool
with gravity flow for
inventory control
Diesel generator–
driven pumps
Containment
decay Heat
removal
Diesel generator–driven
pumped systems with
complex cooling water
systems and ultimate heat
sink
Completely passive
condensers with simple
transfer of heat to pools
that can boil off to the
atmosphere
DG-driven pumps
and cooling water
Fission product
control and offsite
doses
Double containment barriers
and motor-driven
filter and purge systems
Numerous incontainment
natural
removal mechanisms
HVAC systems
Severe
accident
features
Inserting or igniters for
hydrogen control and other
features to limit corium
impact. Containment vent
added as backup in ABWR.
Lower drywell flooder.
External reactor building
connection to RPV.
Inert containment Core catcher and
passive lower
drywell flooder to
systems
Source: Reference [52].
connect portable
limit corium
impact and the
ability to easily
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9.12 Generation III & III+ Reactors 535
9.12.6 International Reactor Innovative & Secure (IRIS) Reactor
International Reactor Innovative & Secure (IRIS) reactor is designed by Westinghouse
as a modular 335 MWe PWR. The IRIS pressure vessel contains the steam
generators and coolant circulating pumps, adding significant accident mitigation
features to the design (Fig. 9.64). The design certification for the IRIS is expected
in 2010, with a first of a kind plant built by 2015 [54–57].
Fig. 9.64. The design of the steel containment building for IRIS system (Printed with permission
from [56]).
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536 9 Nuclear
The layout of the pressure vessel and the power plant are shown in Fig. 9.65
and Fig. 9.66, respectively.
Fig. 9.65. Layout of reactor pressure vessel of IRIS reactor (Adapted from [57]).
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9.12 Generation III & III+ Reactors 537
Fig. 9.66. Lyaout of a IRIS plant (Printed with permission from [56]).
9.12.7 APR-1400
South Korea’s regulatory agency awarded design certification to the APR-1400 in
9.12.8 VVER-1200
Russia signed an agreement with Natsionalna Elektricheska Kompania and Atomstroyexport
in January 2008 to build two new AES-92 plants in Bulgaria, utilizing
VVER-1000 reactors, with a combined output of 2,000 MWe. These reactors are
also being built in China and India. Russia has improved upon this design to create
the VVER-1200, which has an output of 1150–1200 MWe with higher efficiency
and longer anticipated lifetime. Russia is projecting to begin operating these reactors
in 2012 and is hoping to connect an additional 20–25 of these reactors to
the grid by 2020 [58].
2003. This 1,450 MWe design evolved from a U.S. design developed in the 1900’s,
known as the US System 80+. This design incorporates the common attributes of
generation III and III+ reactors in that it utilizes burnable poisons within the fuel to
increase burn-up to 60 GWD/t, longer plant lifetime and increased safety features.
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