Materials for nuclear technology Study Support Miroslav Kursa, Ivo Szurman Ostrava 2015 VYSOKÁ ŠKOLA BÁŇSKÁ – TECHNICKÁ UNIVERZITA OSTRAVA FAKULTA METALURGIE A MATERIÁLOVÉHO INŽENÝRSTVÍ
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Materials for nuclear technology
Study Support
Miroslav Kursa, Ivo Szurman
Ostrava 2015
VYSOKÁ ŠKOLA BÁŇSKÁ – TECHNICKÁ UNIVERZITA OSTRAVA
FAKULTA METALURGIE A MATERIÁLOVÉHO INŽENÝRSTVÍ
Title: Materials for nuclear technology
Code:
Author: Miroslav Kursa, Ivo Szlurman
Edition: first, 2015
Number of pages: 65
Academic materials for the Advanced Engineering Materials study programme at the
Faculty of Metallurgy and Materials Engineering.
Proofreading has not been performed.
Execution: VŠB - Technical University of Ostrava
STUDY INSTRUCTIONS
Students of the Materials for Nuclear Technology course in the 2nd semester of the field of
Progressive Technical Materials received a study package including lecture notes and study
instructions for the combined studies.
1. Prerequisites
The following courses are required to enroll in this course: Physics of Solids, Physical
Metallurgy, Special Alloy Technology and Basics of Degradation Processes.
2. Course aim and learning outputs
Students will become familiar with basic requirements on materials applied in nuclear
technology. These materials are usually used in the construction of nuclear reactors.
The discussed topics include the importance of nuclear purity and methods of ensuring it
for individual types of materials, namely for coating and fissile materials.
After studying this module the students should be able to:
knowledge:
Explain the basic principles of nuclear reactions leading to nuclear fission.
Assess the options of obtaining energy from individual sources, including their evaluation
and comparison.
Assess material requirements for concrete applications and modify them using the alloying
process, or alternatively using technical and mechanical processing.
Who this course is intended for
This course is offered in the Master's studies of the field of Progressive Technical Materials in
the Material Engineering study program but it can be taken by students from any other field
who meet the prerequisites.
The study materials are divided into sections, chapters, that correspond to the logical division
of the covered subject and have different lengths. The estimated time to study each chapter
can significantly differ and that is why long chapters are further divided into numbered
subchapters in accordance with the below described structure.
We recommend the following procedure for studying individual chapters:
Study each chapter thoroughly and answer the questions Any questions regarding the
covered subject can be discussed within consultations.
Communication with lecturers:
Students of the combined studies will be assigned programs and semestral projects on
lectures of the Materials for Nuclear Technology course. Communication with the lecturer
will be ensured in the form of consultations on arranged dates, alternatively via e-mail. The
requirements for passing the course will be discussed in detail during the introductory
lecture.
Contents 1. Introduction ........................................................................................................................ 9
2. Physical foundations of nuclear facilities ......................................................................... 10
2.1 Composition of the atom ................................................................................................ 10
2.1.1 Electron shell ........................................................................................................... 10
2.1.2 Atom nucleus ........................................................................................................... 11
2.2 Radioactive decay of unstable nuclei ............................................................................. 12
2.2.1 Alpha radiation ........................................................................................................ 12
2.2.2 Beta radiation .......................................................................................................... 13
2.2.2.1 Negative beta decay β- ...................................................................................... 13
2.2.2.2 Positive beta decay β+....................................................................................... 13
2.2.3 Gamma radiation ..................................................................................................... 13
2.2.3.1 Photoelectric effect ........................................................................................... 14
2.2.3.2 Compton scattering .......................................................................................... 14
2.2.4 Neutron radiation ..................................................................................................... 14
2.2.4.1 Elastic scattering .............................................................................................. 14
2.2.4.2 Inelastic scattering ............................................................................................ 14
2.2.4.3 Neutron capture ................................................................................................ 15
2.3 Nuclear reactions ............................................................................................................ 15
2.4 Atomic nuclear fission by neutrons ................................................................................ 15
2.4.1 Fission chain reaction of uranium nuclei ................................................................ 17
2.4.2 Neutron balance ....................................................................................................... 18
2.4.3 Neutron diffusion .................................................................................................... 19
2.4.4 Neutron slowing down ............................................................................................ 19
2.4.5 Neutron flow in the active zone .............................................................................. 19
2.5 Synthesis of light nuclei ................................................................................................. 20
3. Nuclear reactor ................................................................................................................. 22
3.1 Classification of reactors by neutron spectrum .............................................................. 23
3.2 Requirements on operation of nuclear reactors .............................................................. 23
3.2.1 Reactor's operating cycles ....................................................................................... 23
3.2.2 Thermal energy release ........................................................................................... 24
3.2.3 Fuel exchange .......................................................................................................... 26
3.2.4 Reactor poisoning and slagging .............................................................................. 26
3.3 Fuel elements .................................................................................................................. 27
3.3.1 Nuclear fuel ............................................................................................................. 27
3.3.2 Construction and coating materials ......................................................................... 28
3.4 Coolants .......................................................................................................................... 30
3.4.1 Gas coolants ............................................................................................................ 31
3.4.1.1 Carbon dioxide ................................................................................................. 31
3.4.1.2 Helium .............................................................................................................. 31
3.4.2 Liquid coolants ........................................................................................................ 31
3.4.2.1 Water ................................................................................................................ 32
3.4.2.2 Molten salts ...................................................................................................... 32
3.4.3 Liquid metals ........................................................................................................... 32
3.4.3.1 Sodium ............................................................................................................. 33
3.5 Moderators and reflectors ............................................................................................... 33
3.5.1 Light water .............................................................................................................. 34
3.5.2 Heavy water ............................................................................................................. 34
3.5.3 Graphite ................................................................................................................... 35
3.5.4 Beryllium ................................................................................................................. 35
3.6 Absorption materials ...................................................................................................... 36
3.6.1 Materials containing boron ...................................................................................... 36
3.6.1.1 Steels ................................................................................................................ 37
3.6.1.2 Dispersion materials ......................................................................................... 37
3.6.1.3 Powder materials .............................................................................................. 37
3.6.2 Hafnium ................................................................................................................... 37
3.6.3 Cadmium ................................................................................................................. 37
3.6.4 Lanthanides ............................................................................................................. 38
3.7 Other components .......................................................................................................... 38
3.7.1 Reactor pressure vessel ........................................................................................... 38
3.7.2 Reactor shielding ..................................................................................................... 38
4. Nuclear fuels .................................................................................................................... 40
4.1 Uranium .......................................................................................................................... 40
4.1.1 Metallic uranium ..................................................................................................... 40
4.1.1.1 Occurrence and uranium ores ........................................................................... 40
4.1.1.2 Uranium production ......................................................................................... 41
4.1.2 Physical and mechanical properties of uranium ...................................................... 46
4.1.3 Powder metallurgy of uranium ................................................................................ 47
4.1.4 Uranium alloys ........................................................................................................ 48
4.1.4.1 Uranium alpha alloys ....................................................................................... 48
4.1.4.2 Uranium gamma alloys .................................................................................... 50
4.1.5 Preparation of uranium alloys ................................................................................. 51
4.1.5.1 The preparation of uranium alloys by melting ................................................. 51
4.1.6 Uranium alloys – ceramic fuels ............................................................................... 51
4.2 Plutonium ....................................................................................................................... 54
4.2.1 Plutonium sources ................................................................................................... 55
4.2.1.1 Thermal reactors ............................................................................................... 55
4.2.1.2 Fast reactors ...................................................................................................... 55
4.2.2 Plutonium production .............................................................................................. 56
4.2.2.1 Basic methods of reprocessing irradiated fuel ................................................. 56
4.2.2.2 Metallic plutonium production ......................................................................... 57
4.2.2.3 Plutonium properties ........................................................................................ 57
4.2.2.4 Processing of plutonium and its alloys ............................................................. 58
4.2.2.5 Plutonium alloys ............................................................................................... 58
4.2.2.6 Plutonium compounds ...................................................................................... 59
4.3 Thorium .......................................................................................................................... 59
4.3.1 Occurrence, ores and their enrichment .................................................................... 60
4.3.2 Thorium production ................................................................................................. 60
4.3.2.1 Preparation of pure thorium compounds .......................................................... 61
4.3.2.2 Preparation of metallic thorium ........................................................................ 62
4.3.3 Thorium properties .................................................................................................. 63
4.3.4 Thorium alloys ........................................................................................................ 63
4.4 Dispersion nuclear fuels ................................................................................................. 64
4.4.1 Metallic dispersion fuel ........................................................................................... 65
4.4.2 Non-metallic dispersion fuels .................................................................................. 65
5. Coating and construction materials .................................................................................. 67
5.1 Aluminium and aluminium alloys .................................................................................. 67
5.1.1 Aluminium production ............................................................................................ 67
5.1.2 Aluminium processing ............................................................................................ 68
5.1.3 Aluminium properties ............................................................................................. 68
5.1.4 Aluminium alloys .................................................................................................... 68
5.1.5 Aluminium corrosion .............................................................................................. 69
5.1.5.1 Corrosion resistance in water up to 100°C ....................................................... 69
5.1.5.2 Corrosion resistance in water above 100°C ..................................................... 69
5.1.5.3 Corrosion of SAP material in water ................................................................. 70
5.1.5.4 Corrosion in water vapour ................................................................................ 70
5.1.5.5 Corrosion in gases ............................................................................................ 70
5.1.5.6 Corrosion in metal liquid melts ........................................................................ 70
5.2 Magnesium and magnesium alloys ................................................................................ 70
5.2.1 Magnesium production ............................................................................................ 70
5.2.2 Magnesium properties ............................................................................................. 70
5.2.3 Magnesium alloys ................................................................................................... 71
5.2.3.1 Mg - Be alloys .................................................................................................. 72
5.2.3.2 Mg – Zr alloys ................................................................................................. 72
5.2.4 Corrosion of magnesium and its alloys ................................................................... 73
5.3 Zirconium and zirconium alloys..................................................................................... 73
5.3.1 Zirconium production .............................................................................................. 73
5.3.1.1 Processing methods of zircon concentrates ...................................................... 74
5.3.1.2 Production of zirconium tetrachloride .............................................................. 75
5.3.1.3 Separation of hafnium from zirconium (dehafnization) ................................... 75
5.3.1.4 Production of metallic zirconium by metallothermic method .......................... 76
5.3.2 Zirconium refining .................................................................................................. 77
5.3.3 Zirconium alloys ..................................................................................................... 77
5.3.3.1 Zr – Nb alloys .................................................................................................. 78
5.3.3.2 Zr – Sn alloys ................................................................................................... 78
5.3.4 Corrosion of zirconium and its alloys ..................................................................... 79
5.3.4.1 Water ................................................................................................................ 79
5.3.4.2 Gases, liquid metals .......................................................................................... 80
5.4 Beryllium and beryllium alloys ...................................................................................... 80
5.4.1 Beryllium production .............................................................................................. 80
5.4.1.1 Compound preparation for beryl production .................................................... 81
5.4.1.2 Production of metallic beryllium ...................................................................... 81
5.4.2 Beryllium properties ................................................................................................ 81
5.4.3 Beryllium corrosion ................................................................................................. 82
5.5 Steels and nickel alloys .................................................................................................. 82
5.5.1 Corrosion resistance ................................................................................................ 83
5.6 Niobium .......................................................................................................................... 84
5.6.1 Niobium production ................................................................................................ 84
5.6.1.1 Preparation of pure niobe compounds .............................................................. 84
5.6.1.2 Separation of niobium and tantalum ................................................................ 84
5.6.1.3 Production of metallic niobium ........................................................................ 85
5.6.2 Niobium processing ................................................................................................. 85
5.6.3 Mechanical properties of niobium ........................................................................... 86
5.6.4 Niobium corrosion ................................................................................................... 86
5.7 Vanadium ....................................................................................................................... 87
5.7.1 Properties of vanadium and vanadium alloys ......................................................... 87
5.7.2 Preparation technology of vanadium ....................................................................... 87
5.7.2.1 Production of metallic vanadium ..................................................................... 88
5.7.3 Vanadium alloys and applications ........................................................................... 88
5.7.4 Vanadium corrosion ................................................................................................ 89
5.8 Yttrium ........................................................................................................................... 89
5.8.1 Yttrium production .................................................................................................. 89
5.8.2 Yttrium corrosion .................................................................................................... 89
6. Effect of radiation on material properties of nuclear reactors .......................................... 91
6.1 Precipitation processes caused by radiation ................................................................... 91
6.2 Damage zone in irradiated solid substances ................................................................... 92
6.2.1 Focusing collision mechanism ................................................................................ 94
6.3 Radiation effects on the properties of metallic uranium, its alloys and compounds ...... 94
6.3.1 Radiation growth ..................................................................................................... 94
6.3.2 Swelling ................................................................................................................... 95
1. Introduction
Time to study: 0,5 hours
In recent years the global energy concept has started to deal with the issues of
dwindling reserves of fossil fuels and increased requirements on protecting the environment.
Even though classic power plants are gradually being modernized and ecologized, their
situation in the future is uncertain due to the decrease of fossil fuel reserves.
Renewable energy sources have now come into play in the energy industry. However,
their role is more of a local importance; neither today nor in the near future will these sources
be capable of replacing classic and nuclear energy sources due to their outputs and
development and construction costs.
Utilization of energy by nuclear fission of heavy elements is still the most perspective
method of generating energy. Studies and researches carried out competent by European
organizations concluded that human society in the 21st century cannot dispense with nuclear
energy. In accordance with the World Nuclear Association, 437 reactors with a total installed
power of over 374,000 MWe were in operation at the end of 2012. Nuclear power plants today
generate approximately 16 % of the total world electricity; in the European Union nuclear
power constitutes over 30 % of electricity generation.
2. Physical foundations of nuclear facilities
Time to study: 5 hours
Aim After studying this section the student should be able to:
Describe the composition of the atom, radioactive decay of unstable nuclei,
types of radiation.
Describe the mechanism of nuclear fission by neutrons, fission chain reaction
and nuclear fusion.
Lecture
2.1 Composition of the atom
The atom is the basic building unit of elements. It consists of an atomic nucleus and an
electron shell that are mutually affected by electrostatic forces. The atomic nucleus contains
so-called nucleons (protons and neutrons), the electron shell around the nucleus and its
structure determine the properties of elements. Its structure, laws and effect on element
properties are studied by chemistry, physics, physics of metals and it forms the basis for the
periodic table of elements. However, the number of protons in the nucleus is decisive when
determining the number of electrons and their structure. The whole atom is therefore
determined by the structure and stability of the nucleus.
Even though processes in the nuclear reactor are of mostly nucleus character and are
seemingly not directly associated with the electron shell, it is necessary to know the periodic
table of elements, namely the connection between the elements in individual periods and
groups, in order to be able to study these processes.
2.1.1 Electron shell
Electrons in the electron shell are governed by quantization law and Pauli exclusion
principle that were also originally derived from properties of the electron shell. Single-wave
energy state can be occupied only by two electrons with opposite spins; other electrons need
to occupy quantum states with higher energy states. Three quantum numbers are required in
order to determine the orbital in the electron shell (the principal number n, angular
momentum number l and magnetic number m) along with the spin s. The principal number n
is connected with shells of the Bohr atomic model in the following sequence: K, L, M, N, O,
P, Q. Angular momentum quantum number determines (from zero up) the orientation of the
orbital sub-shell's shape designated as s, p, d, f, g.
Electron quantum orbits K, L, M, etc. are occupied gradually based on the nucleus
charge. The external orbit has the loosest bond of electrons to the nucleus and determines the
chemical valence of elements. In addition, radiation from the electron shell occurs most easily
in this orbital. If a particle or a quantum of electromagnetic radiation affects a neutron, the
transferred energy moves it to a more remote orbit. Such excited state of the atom is unstable
and returns to the basic state in short time (approximately 10-8
s) by radiating a quantum of
monochromatic electromagnetic radiation with frequency following from the following
relation:
E = h .
Where h is the Planck constant and E is the difference in electron energy between the two
orbits. If the energy of the incident particles is so high that the affected electron moves out of
the reach of nucleus attraction forces, a positive ion and free electron are created.
2.1.2 Atom nucleus
Atoms with the same number of protons in the nucleus and therefore the same number
of electrons and same chemical properties that differ in the neutron number are called isotopes
of a particular chemical element. Most elements in nature are composed of two or more
isotopes. For example, one can mention the isotopes of hydrogen and uranium, which are
important in nuclear technology. Table 2.1 lists isotopic composition of natural uranium and
hydrogen.
Nucleus radius rj is defined as half the width of the so-called potential well, which defines
the reach of nuclear forces. It is in the order of 10-15
m, which corresponds to the size of
elementary particles. Fig. 2.1 represents this potential well as the result of simultaneous action
of nuclear forces Ej and electromagnetic forces Ee. Stronger short-range nuclear forces prevail
significantly inside the nucleus. In the rj distance the negative potential energy of nuclear
force practically disappears and only the positive potential energy of electromagnetic forces
remains. In distance rj the potential barrier 0 therefore equals electrostatic potential Ze in
distance rj:
0 = Ze / rj
Nucleus radius with mass number A can be expressed as a linear function of the third root of
number A enumerated in 10-15
m. rj = 1,7 + 1,22 . A
1/3
A nuclear cross section is introduced instead of the radius for interaction of the nucleus with
particles. For fast neutrons it has the following size:
= . rj2
The size of this effective area can be determined from the decrease of radiation when
penetrating a substance containing the assessed nuclei. It is represented by with an index
corresponding to the concrete interaction, e.g. absorption. It highly depends namely on the
type and especially energy of incident particles and on the mutual ratio of protons and
neutrons in the nucleus. Neutron cross sections will be discussed in more details in following
chapters.
Nuclear spin was introduced in the nucleus for similar reasons as in the electron shell.
Since both nucleons have a spin of ½, the spin for nuclei with an even number of nucleons A
is a whole number. The spin of "even-even" nuclei (A even, Z even) is zero.
Tab. 2.1 Isotopic composition of natural uranium and hydrogen.
izotop obsah [%] 99,985 0,015 - 0,006 0,720 99,274
hmotnost [mu] 1,007825 2,014102 3,016030 234,041 235,044 238,051
2.2 Radioactive decay of unstable nuclei
Radioactivity is the ability of certain atomic nuclei to decay spontaneously and emit
particles or electromagnetic radiation. The speed of radioactive decay is not affected by
temperature, pressure, magnetic or electric field. The number of decays ΔN is proportional to
the number of unstable nuclei N and time interval Δt:
− ∆N = λ . N . Δt
where λ is the decay constant (s-1
), the negative sign „–“ indicates a decrease of nuclei.
Integration of this relation yields the decay law.
N = N0 . e−t
under the condition that N = N0
in time t = 0. However, this law applies only to a sufficiently
large set of particles. Decay constant λ characterizes the probability of decay of one nucleus
per second. Another decay characteristic is the half-life T1/2.
T12= ln 2
λ
It is a time interval during which an average of half of the original radioactive nuclei
undergoes the radioactive decay. When studying radioactive decay, sometimes a whole chain
of radioactive transformations occurs. In nature there are whole radioactive lines where the
preceding isotope transforms to the following isotope due to radioactive alpha or beta decay.
Ionizing radiation is characterized by its energy. This energy is expressed in electron
volts (eV). This unit is related to the basic unit of energy (Joule) by the following relation:
1 eV = 1,602 . 10−19 J
2.2.1 Alpha radiation
Alpha radiation is a direct ionizing radiation consisting of alpha particles – helium
nuclei. The particles consist of 2 protons and 2 neutrons and have therefore two positive
charge units. The source of alpha radiation are heavy radionuclides, such as isotopes Po, Ra,
Th, U or transuranium elements. These radionuclides emit (usually 1 or 2) alpha particles of
certain energy levels that are characteristic of its radioactive transformation. The starting
Fig. 2.1 Potential well as the resultant Ej nuclear and electromagnetic forces Ee.
energy of alpha particles is in ones of MeV, which corresponds to starting speeds in the order
of 107 m.s
-1.
Since alpha particles contain two positive charges, they ionize heavily when passing through
the environment and lose their energy. The reach of alpha radiation is therefore significantly
limited. In air the reach is only several centimeters, in water or tissue it is only fractures of
milimeters. Protection against this radiation therefore does not constitute a problem. Take the
transformation of protactinium to actinium as an example:
Pa → Ac + He24
89227
91231
2.2.2 Beta radiation
Beta radiation consists of high-speed electrons or positrons (particles with the same
weight and opposite charge than electron). It is created during the transformation of many
natural and artificial radionuclides. Beta radiation causes ionization or excitation of atoms
and molecules when passing through the environment. When compared to alpha radiation,
beta radiation is much lighter, it moves at much higher speed (in the order of 108 m.s
-1) with
the same energy and it is less ionizing. Beta radiation has a greater impact on the
environment. Particles are often dispersed with only small energy losses and they follow a
zigzag path. For example, beta radiation with the maximum energy Emax. = 2 MeV has a reach
of about 8 m in the air, approximately 1 cm in the water and about 4 mm in aluminium.
2.2.2.1 Negative beta decay β- During the β
- decay a negative electron is released from the nucleus. This is typical for
a nucleus with an excess of neutrons. Neutron with high energy level transforms to a free
electron and proton with lower energy level. Mass number A of the nucleus remains the same
because the nucleus mass is practically unaffected by the loss of neutron; however, proton
number A is increased by one. β- decay occurs in unstable fissile products, for example:
Te → I + e → e + Xe → e + Cs → e + Ba56135
−10
55135
−10 54
135−10
−10
53135
52135
It is clear that the mass number A is still 135; however, proton number Z changes from 52 in
unstable isotope Te up to 56 in stable Ba.
2.2.2.2 Positive beta decay β+ Positive electron called positron is released from the nucleus. This is typical for a
nucleus with an excess of protons that are converted to a neutron and a positron. The mass
number A is decreased by one. Schematic representation of the positron transformation:
X → X + e10
Z−1A
ZA
The β+
decay 𝐶.611 can be used as an example. Positron emission is always accompanied by
annihilation radiation created during the reaction of a positron with an electron that produces
2 quanta of γ radiation, both with the energy of 0.51 MeV.
2.2.3 Gamma radiation
Gamma radiation refers to electromagnetic radiation of extremely short wavelength in
orders of 10-11
to 10-13
m created in the nucleus. It is usually accompanied by alpha and beta
radiation. Certain elements emit monochromatic radiation of a single wavelength, other
elements emit a whole spectrum consisting of individual lines of certain wavelengths. This
disconnected spectrum of lines is in compliance with the quantum theory. When gamma
radiation passes through the environment the electromagnetic radiation is absorbed in
accordance with the exponential law. The photoelectric effect, or Compton scattering or
creation of electron pair occur during this process. Photodisintegration or resonance
absorption can occur during the interaction with the nucleus. Of course, gamma radiation
does not affect A or Z.
2.2.3.1 Photoelectric effect This effect is present mostly for lower energy radiation. It is a process during which a
gamma photon transfers all its energy to some of the orbital electrons, usually in internal atom
shells. As a result, a photoelectron is released and it further transfers its energy by ionization
or excitation of atoms and molecules. Afterwards, the atom is in the excited state and during
the transition to the normal state it emits a photon of characteristic radiation or an electron.
The probability of the photoelectric effect also increases with the proton number of the
material. For example, in plumbum this process is the prevailing method of interaction for
gamma radiation with energy up to 1 MeV.
2.2.3.2 Compton scattering Compton scattering occurs in free or weakly bound electrons (external atom orbits). In
this case the incident photon transfers part of its energy to the electron, sets it into motion and
then continues its path in a different direction and with lower energy. The accelerated electron
then interacts with the environment just like a photoelectron, i.e. it ionizes and excites
surrounding atoms and molecules. Compton scattering is the prevailing interaction process of
medium-energy gamma radiation, for example from 0.1 MeV in aluminium and from 1 MeV
in plumbum.
2.2.4 Neutron radiation
Neutron radiation refers to radiation of electrically neutral particles whose weight is
comparable with the weight of hydrogen nuclei – protons. Nuclear reactors are the main
sources of neutrons. Neutron radiation can be divided into several groups based on its energy.
We can distinguish for example thermal neutrons (energy lower than 0.5 eV), resonance
neutrons (0.5 – 100 eV), medium-energy neutrons (1 – 500 keV), fast neutrons (0.5 – 10
MeV) and high-energy neutrons (over 10 MeV).
The interaction of neutron radiation with mass differs significantly from the processes
described above. Since neutrons do not carry an electric charge, they do not ionize when
passing through the environment and interact almost exclusively with atom nuclei. The main
interaction methods include elastic scattering, inelastic scattering, neutron capture, charged
particle emission and nuclear fission. Probability of the concrete reaction depends on the
neutron energy and composition of the absorbing environment.
2.2.4.1 Elastic scattering Elastic scattering is one of the most common interaction methods of fast neutrons. In
this process the neutron transfers part of its energy to the atomic nucleus and sets it in motion.
The accelerated nucleus then loses its kinetic energy by ionization or excitation of atoms and
molecules in the environment. Energy transferred by elastic scattering reaches highest values
in case of collisions with light nuclei. For example fast neutrons with the starting energy of 2
MeV need approximately 18 collisions in water and up to 400 collisions in plumbum to slow
down.
2.2.4.2 Inelastic scattering During inelastic scattering the neutron also transfers part of its energy to the atomic
nucleus. The transferred energy manifests itself by changing the internal state of nucleus - its
excitation. The return to the basic state is accompanied by emission of a gamma radiation
photon.
2.2.4.3 Neutron capture In this process the neutron is absorbed by the nucleus while emitting one or more
gamma photons. It is an effective method to capture thermal neutrons for example in
cadmium nuclei. These substances are often used for shielding of neutron sources or
controlling a fission chain reaction in the reactor. Neutron capture on cadmium:
Cd + n → Cd + gama48113
01
48113
Another interaction method consists of absorption of the neutron by the nucleus while
emitting particles (proton, neutron, alpha particles). These collisions are most probable for
light nuclei and fast neutrons, for example for boron:
B + n → Li + He24
37
01
510
Thanks to this property boron (often in the form of boric acid) is used to control
fission chain reaction in various reactor constructions.
As is clear from the described interaction mechanism, the absorption of neutrons
consists basically of two steps. Fast neutrons are first slowed down by scattering on nuclei of
light elements and only then they are absorbed while emitting particles or photons. Shielding
of neutrons therefore consists of multiple components - it contains light materials (water,
paraffin) for slowing down neutrons and a substance for their effective capture (B or Cd).
Sometimes a third component is required - a heavy material to shield gamma radiation from
the capture of neutrons.
2.3 Nuclear reactions
Nuclear reactions are processes in which one nucleus transforms to another nucleus.
Nuclear reactions are symbolically represented as follows:
a + A → b + B
where a is the bombarding particle, A is the target nucleus, b is the projectile and B is the
newly formed nucleus. Full representation of a nuclear reaction contains symbols of elements
and proton and mass numbers. Notice the processes that might occur in case of collision of the
bombarding particle with the target nucleus:
a) Elastic scattering – the composition and internal energy of the nucleus remain
unchanged, only the kinetic energy is distributed between the particle and the nucleus.
b) Inelastic scattering - the nucleus composition remains the same, only part of the
kinetic energy of the bombarding particle is consumed to excite the nucleus.
c) Nuclear reaction itself - both the composition and internal energy of the nucleus
change.
2.4 Atomic nuclear fission by neutrons
Fission of atomic nuclei is a type of nuclear reaction in which the nucleus is divided to
similar fragments. The nuclear fission is conditioned by the weight of the dividing nucleus,
which needs to be higher than the sum of weights of the fragments. The mechanism of
nuclear fission by neutrons can be best explained on the drop model. Neutron that enters the
nucleus causes excitation and oscillation of the nucleus, which is elongated to an ellipsoid. If
there is sufficient energy in the nucleus, the prolongation continues until the nucleus splits to
two parts that are first deformed but later obtain a round shape. The energy required for
splitting of the nucleus is referred to as activation energy Ea.
Fig. 2.2. represents the dependency of potential energy Ep of the nucleus on the
distance between fission fragments. Activation energy can be supplied to the nucleus in
different ways (bombarding nuclei with charged particle from accelerators). Nuclear fission
by neutrons is the most important method. This is related to the fact that neutron is not
repelled by positively charged nucleus, i.e. it can have any energy. In addition to the kinetic
energy of a neutron, another significant advantage of neutrons for the excitation of compound
nucleus is the bond energy that is transferred to the nucleus. Ea a Evn values of the bond
energy of neutrons for selected nuclides are listed in table 2.2.
Previously we assumed (for the sake of simplification) that the nucleus is split to two
identical fragments. In fact, nuclear fission by neutrons creates fragments of different mass
number since heavy nuclei split in more than 40 ways. Creation of 2 fragments of different
weights is the most probable. Fig. 2.3 shows experimentally determined yield of fission
fragments F1 and F2 as a function of their neutron numbers for three isotopes that are
fissionable by thermal neutrons. The yield determines the ratio of the number of fissions
during which a fragment with the relevant mass number is created to the total number of
fissions. 235
U fissions will most probably yield nuclei with mass number 95 and 139. Yield of
these nuclei is 6.4 %. The number of released neutrons during the fission depends mostly on
the fission reaction process and is usually 2 to 3.
Primary fission fragments have an excess of neutrons and are therefore radioactive.
Even if one neutron is released, the ratio of the number of neutrons to the number of protons
might be outside the stability range corresponding to the relevant mass number. The products
of fragment decay are therefore also radioactive and are transformed to a stable isotope by
gradual emission of electrons often accompanied by gamma radiation. Decay chains have
different lengths. On average, the fragment undergoes 3 decay stages before creating a stable
isotope.
Fig. 2.2 Dependence of potential energy E on distance r between fission
fragments: 1-light nucleus, 2-moderate nucleus 3 hardest nucleus.
Table. 2.2 Activation energy Ea and the neutron binding energy of EVN in heavy nuclei.
jádro Ea [MeV] Evn [MeV] 232
Th 7.5 5.4 238
U 7.0 5.5 235
U 6.6 6.8 Evn > Ea 233
U 6.0 7.0 Evn > Ea 239
Pu 5.0 6.6 Evn > Ea
The radioactivity of fission products is significant for the operation of nuclear power
plants. Heat is released from fissile material even in shutdown reactors and needs to be
dissipated. Another problem is the transport and reprocessing of heavily radioactive irradiated
fuel elements.
Thermal neutron fission is energetically plausible for 233
U, 235
U, 239
Pu and certain
transuranium with higher Z (238
U nuclei are fissionable only by neutrons with kinetic energy
exceeding 1.1 MeV). The only natural fissile material is 235
U.
2.4.1 Fission chain reaction of uranium nuclei
The fission chain reaction is very quick - fission of all nuclei in one gram of uranium
would take 10 s. In practice not all neutrons enter uranium nuclei (they are absorbed by the
construction material, they enter the moderator, coolant, etc.). In addition, not all neutrons
absorbed by uranium nuclei cause their fission. The nucleus that absorbed the neutron can
emit the excessive energy in the form of a quantum of gamma radiation. The probability of
fission and radiation capture or inelastic scattering depends on the neutron energy and energy
of the interacting nucleus. For example, the fission probability for low-energy neutrons is
much higher than the probability of radiation capture. When considering a natural uranium
compound that contains 140x more 238
U than 235
U, the probability of 235
U nuclear fission at
thermal energy of the neutron is approximately 200x higher than the probability of radiation
capture of the neutron by the 238
U nucleus (238
U is not fissionable by thermal neutrons). An
important conclusion for sustaining a fission chain reaction follows from what has been stated
above: slowing down of fast (energized) neutrons created after nuclear fission to thermal
neutrons (0.025 eV). The moderator in the nuclear reactor is used for this purpose. The
nuclear reactor is a device where the controlled fission chain reaction occurs. Nuclear fission
of uranium in the active zone of the reactor produces heat, which is then removed by a
coolant. Fig. 2.4 shows the basic processes constituting a chain reaction.
Fig. 2.3 Yield of fission fragments F1 and F2 for 3
fissionable isotopes 239
Pu, 233
U a 235
U.
2.4.2 Neutron balance
The nuclear reactor is a device where the chain reaction is sustained. It consists of a
moderator, fuel containing fissile material, coolant, absorbers and construction materials. All
these parts of the nuclear reactor affect the neutron balance. We will now focus on thermal
reactors, i.e. reactors where the fission reaction is set off by slowed down neutrons, so-called
thermal neutrons.
An important characteristic for the neutron balance is the multiplication coefficient,
which is defined as the ratio of the number of neutrons in generation n to the number of
neutron in generation n-1. Assuming an infinite system without any leakage then the
multiplication coefficient will be as follows:
𝑘 = q . p . f . r
where q - is the multiplication coefficient on fast neutrons; it represents the ratio of the
number of fast neutrons created by 238
U and 235
U fission to the number of fast neutrons
created by 235
U fission (natural uranium q = 1.03),
p - probability of resonance capture leakage; it is determined by the number of neutrons
that reached the thermal zone to the number of neutrons that started the thermalization,
f - thermal utilization coefficient of neutrons is represented by the ratio of neutrons
absorbed in 235
U nuclei to the total number of absorbed thermal neutrons,
r – regeneration coefficient r = Nf . f / c, where Nf is the average number of immediate
neutrons released during one fission, f is the macroscopic fission cross section, c is the
macroscopic absorption cross section.
However, there is both thermal neutron leakage and slowing-down neutron leakage in
a finite system. Let's introduce the probability that the neutron will not leak from the system.
In this scenario the multiplication coefficient kef for a finite system will be:
kef = k . P
Based on the value of kef we can distinguish:
1. Subcritical reactor state, kef < 1, when conditions of a fission chain reaction are not
met.
Fig. 2.4 Schematic representation of processes occurring during fission
chain reaction.
2. Critical reactor state, kef = 1, fission reaction might occur and the number of neutrons
in the volume will not change.
3. Supercritical reactor state, kef > 1, chain reaction is occurring and the number of
neutrons in the volume will increase.
All three states are used during the reactor operation. Subcritical reactor state is
adjusted to the supercritical state and after reaching the required reactor power, it is put to the
critical state.
2.4.3 Neutron diffusion
Neutrons that move in material environment (moderator, fuel, coolant, construction
elements, etc.) overcome collisions with atomic nuclei. There might be the following neutron
collisions:
a) scatter collisions, which change the energy of neutrons
b) absorption collisions, when the nucleus absorbs the neutron.
Scattering of a large number of neutrons is characterized by free medium scattering
trajectory, which represents the inverse of the macroscopic scattering cross section.
Absorption of a large number of neutrons is similarly characterized by free medium
absorption trajectory, which represents the inverse of the macroscopic absorption cross
section. These characteristics can be used to prove that neutrons scatter from a place with
higher density to a place with lower density. This process is identical to the motion of
molecules in gases.
The main task of the reactor theory is to determine the distribution of neutron density
or neutron flow in the reactor's active zone. Neutron flow equals the product of the density
of neutrons n and the absolute value of neutron velocity:
ϕ( r, vΩ , t) = v . n (r, vΩ , t)
where t is the time, 𝑟 is the position vector, v is the absolute neutron velocity, Ω is the
direction in space.
Using the diffusion theory and certain simplifications we can derive the diffusion
equation with the following symbolic representation:
∂n (r , vΩ , t)
∂= formation − leak − capture
where 𝜕n / 𝜕t is the speed of change of neutron density.
2.4.4 Neutron slowing down
Nuclear fission creates neutrons with average energy of 2 MeV. However, thermal
neutrons with energy in the order of 10-2
eV are used in reactors for 235
U nuclear fission Fast
neutrons need to be slowed down to thermal neutrons. This is achieved by a neutron
moderator in nuclear reactors.
2.4.5 Neutron flow in the active zone
Physical calculation of the nuclear reactor must be done with conditions as close to the
conditions in the nuclear reactor as possible. These are very complex calculations that must
include the most accurate calculation methods, work with accurate physical and material
constants of reactor materials and predict changes in the reactor during operation.
The main tasks include for example calculation of main dimensions of the active zone,
compiling a critical equation of the reactor, calculating material composition and distribution.
The flow of neutrons in nuclear reactor's active zone is one of the important factors for
assessing the suitability of used materials. The neutron flow intensity is decisive for material
radiation damage and determines the temperatures that the given materials must withstand.
Neutron leakage occurs in border part of the active zone and the neutron flow is
therefore not the same in the whole active zone but reduces towards the borders. Neutron
leakages from the active zone can be reduced and the neutron flow can be evened by
surrounding the active zone with material that reflects neutrons back to the active zone, a so-
called reflector. The reflector material in thermal reactors must meet the same requirements
as the moderator. The effect of the reflector on reducing the neutron leakage and thus evening
the neutron flow is obvious. Fertile materials – a so-called reproduction zone – are used as
reflectors in fast reactors.
The total unevenness of the neutron flow caused by leakage of neutrons can be
reduced by inserting fuel elements with lower enrichment to the center of the active zone and
fuel elements with higher enrichment to the peripheries at the beginning of the campaign.
Also when replacing irradiated fuel elements, new elements are placed to make the neutron
flow as even as possible. This means that fresh elements are placed at the borders of the
active zone and partially irradiated elements are placed to the center. The distribution of the
neutron flow is also significantly affected by the insertion of control rods with absorption
material. The neutron flow is uneven due to the uneven burning of fuel and uneven
distribution of temperatures in the reactor's active zone.
2.5 Synthesis of light nuclei
Nucleosynthesis (nuclear fusion, thermonuclear reaction) is the opposite reaction of
nuclear fission. This means that two lighter nuclei create one heavy nucleus. Similarly to
heavy nuclei the bond energy of light nuclei per nucleon is lower than for nuclei in the center
of the periodic system. Synthesis of two light nuclei therefore represents a reaction with
significant energy factor - the released energy relative to one nucleon participating in the
reaction might constitute several times the energy released during fission.
Even though this energy is strongly exothermic, it is extremely difficult to obtain the
conditions for its creation and sustaining. Nucleosynthesis occurs at extreme temperatures
when atoms lose all of their orbital electrons and plasma is formed from cations and electrons.
Ions can merge and release excessive bond energy only if their energy during thermal motion
is sufficient to overcome the Coulomb Barrier (repulsive forces between each other). The
height of this potential barrier determines the minimum temperature required for the reaction.
The lowest barrier is in hydrogen. The potential barrier is proportionally higher for heavier
nuclei containing more protons and the temperature requirements also increase. That is why
the nucleosynthesis is best performed in the lightest nuclei.
In this respect the following reactions can be considered as the easiest to achieve and
most efficient:
H + H → He + n + 3,26 MeV01
23
12
12
H + H → H + n11 + 4,04 MeV1
312
12
H + H → He + n + 17,6 MeV 01
24
13
12
In nature thermonuclear reaction occur on a large scale in all stars, including the Sun.
So far, people managed to successfully create uncontrolled explosive nucleosynthesis in a so-
called hydrogen bomb.
Controlled release of energy through thermonuclear reaction can be considered to be
the hope for humanity from the practical point of view. This method of energy production is
not limited by insufficient material sources and is not connected with risks of radioactive
wastes and leakages. The technical aspect of this problem proved to be more difficult than
assumed at the beginning. This method requires heating deuterium and tritium to a
temperature of approximately 200,000,000 K, at which atomic nuclei of these substance
collide with each other with energy sufficient to overcome the repulsive forces and they react
with each other, releasing nuclear energy. These substances in plasma form can be maintained
for instance by magnetic confinement (a tokamak device). However, the conditions for a
thermonuclear reaction can be created in a very short period of time by pulse concentration of
ample energy of a laser beam or by accelerating particle into a small volume, where the
reaction takes place at a so-called inertial confinement before the plasma can expand (so-
called inertial confinement methods).
The principle of tokamak devices was designed already in the 50's of the last century.
Toroidal plasma is created in a chamber filled with for instance deuterium at a low pressure
by induction, during which a massive current pulse passes through the primary transformer
winding of the equipment. Plasma can be maintained in the confined form for up to several
seconds using magnetic fields created by further winding and it can be heated in several ways
up to several hundreds of millions K. However, in order to ensure the reaction releases
significant level of energy that exceeds the energy input used to create and confine the
plasma, the reaction needs to be self-sustained for at least a certain period of time. The
conditions for sustaining the reaction with a practical energy gain are characterized by the
Lawson criterion. This criterion is determined by the product of plasma density n (number of
particles in m3), confinement time t (s) and plasma temperature T during the confinement
(expressed by kinetic energy of ions in keV). To achieve practical energy gain it is necessary
to heat the plasma to 10 up to 20 keV and reach Lawson criterion values higher than 3×1021
m-3.
s.keV. Currently the ITER tokamak is being constructed in France in international
cooperation. This tokamak should reach output values of 500 MW for a period of time of
approximately 1000 s, which would represent a tenfold of the energy required to heat the
plasma and sustain the thermonuclear fusion.
Summary of terms in this chapter (subchapter)
Atomic nucleus
Electron shell
Nuclear decay
Nuclear reaction
Questions to the covered material
Briefly describe the composition of natural hydrogen and uranium.
Name and characterize the different types of radiation.
Describe fission of nucleus by neutron, describe the basic difference between
controlled and uncontrolled reactions.
Describe the synthesis of nuclei and list the key issues in the use of nuclear energy.
3. Nuclear reactor
Time to study: 5 hours
Aim After studying this section the student should be able to:
Describe the basic types of nuclear reactors, their advantages and
disadvantages.
Describe basic parts of reactors.
Describe the nuclear fuels, coolants, moderators, etc.
Lecture
Nuclear reactor is defined as a cluster of a sufficient amount of fissile material
enabling a controlled fission chain reaction without damaging the reactor or radioactivity leak
into the environment. The part of the reactor where the fission reaction takes place is called
the active zone.
Current types of nuclear reactors are constructed with high heterogeneous structure.
Calculations of such reactors are relatively demanding.
A general scheme of a nuclear reactor is shown in Fig. 3.1. Real reactors might differ
from this scheme in several aspects based on their type of use. The main parts of a nuclear
reactor are:
Fuel element containing nuclear fuel. This is where fission and release of energy
occur. The fuel element is surrounded by a protective coating against the effects of the
coolant.
Coolant ensures safe heat dissipation from the active zone to other parts of the device,
usually to the heat exchanger. It circulates in the primary circuit.
Moderator reduces the energy of fissile neutrons in the active zone to the thermal
motion energy in order to increase the probability of a fission reaction.
Reflector surrounds the active zone of the thermal reactor, reduces the leakage of
neutrons from the system and contributes to the evenness of thermal development in
the active zone.
Reproduction zone is a layer of fertile material surrounding the active zone of a fast
reactor. New fissile material is formed in this layer and it is also used to reflect
neutrons to the active zone.
Reactor control system enables operation of the reactor at a constant output, change
of operating modes, starting and stopping of the reactor, fast shutdown of the reactor
in case of emergency.
Measurement system of operating parameters and operation safety.
Reactor vessel is a pressure vessel where the active zone and accessories are located.
Reactor shielding reduces the penetration of radioactive radiation to the active zone
and reactor's surrounding to an acceptable limit.
Fuel exchange system enables remote exchange of fuel and relocation of partially
irradiated fuel elements while the reactor is stopped or during operation.
Containment hermetically encloses the reactor and other parts of the primary circuit
and therefore significantly reduces possible leaks of fissile products and radioactive
particles into the reactor's surroundings in case of an extensive emergency.
Practically all types of nuclear reactors include these main parts. However, fast reactors are
not equipped with the moderator and the moderator's function is fulfilled by the reproduction
zone.
3.1 Classification of reactors by neutron spectrum
Neutrons can be divided based on their kinetic energy to thermal, resonance, fast, etc.
Thermal and fast neutrons are used in nuclear reactors.
Thermal reactors Thermal reactors operate with thermal neutrons (0.025 eV) that play a decisive role for
the fission. These reactors operate with moderators and the characteristic fuel concentration is
up to 100 kg.m-3
of the active zone. Currently most nuclear reactors are thermal. As already
mentioned, thermal neutrons have limited fuel utilization options.
Fast reactors The fission reaction is caused predominantly by fast neutrons that are not slowed
down in moderator nuclei. The fuel concentration is usually higher than in thermal reactors,
reaching up to 1000 kg.m-3
of the active zone. The fuel utilization coefficient is high;
however, the construction of these reactors is more demanding.
3.2 Requirements on operation of nuclear reactors
3.2.1 Reactor's operating cycles
In order for the reactor to operate, i.e. for the controlled fission chain reaction to take
place, it needs to be in the so-called critical state, i.e. it needs to have sufficient volume and
sufficient amount of suitably arranged material. The outlined geometric and material factor is
then represented in the so-called critical reactor equation. As stated above, at the beginning of
the operation the reactor needs to in the supercritical state, i.e. it needs to have an excess of
reactivity to overcome the effect of fission fragments and fuel irradiation.
When physically starting the reactor it is necessary to enter a neutron source in
addition to having a supercritical amount of fissile material. Usually a nuclear reaction is
used:
Be + He → C + n01
612
24
49
Fig. 3.1 Scheme of nuclear power reactor – thermal heterogoneous.
where polonium or radium are the source of radiation. For reactors that have been already
operated and where γ radiation is already present in the active zone from the fission fragment
decay, a photoneutron source, for instance beryllium, can be used.
When starting the reactor the multiplication coefficient k > 1 and the increase of
fission from generation to generation is proportionate to the increase of the multiplication
coefficient Δk. The speed of this change is characterized by a so-called reactor period, which
is the time in which the neutron flow increases e-times.
In addition, when starting or operating the reactor, reactivity changes based on the
active zone temperature need to be taken into consideration. Reactivity of regular reactors
decreases together with the temperature, which is caused by the decreased probability of
fission, reduced material density and by high probability of resonance absorption of 238
U. The
size of the so-called reactivity coefficient 𝜕𝜌/𝜕𝑇 is reflected as the degree of the self-
regulating effect. During emergencies and accidents it is desirable that reactivity decreases as
much as possible with the increasing temperature, i.e. to achieve as negative coefficient as
possible.
During the operation, a controlled chain reaction takes place where k = 1. However,
how does fuel irradiation manifest itself in the decrease of the number of fissile isotopes and
how are fission fragments created, causing so-called contamination of the reactor and its
poisoning, both of which needs to be compensated by the removal of control rods? Creation
of new fissile nuclei due to breeding reactions has a positive effect during this process. The
rate of this increase depends on the mutual ratio of isotopes and the reactor type.
The neutron flow is distributed in a certain way in the active zone during operation (it
is affected by the reflector and by control rods). The unevenness of the neutron flow can be
regulated by the distribution of fuel rods in the active zone. At the beginning of a campaign,
less enriched elements are placed to the center and partially irradiated elements are placed
there during operation. The placement of fuel elements of different enrichment and
irradiation is always governed by a special program.
For certain types of reactors it is possible to exchange the fuel during the reactor
operation without the need to shut it down and therefore without any time losses. The
advantage of replacing fuel while the reactor is shut down is the possibility to perform a
revision of the reactor. Delayed neutrons need to be taken into consideration in the reactor
control. There is enough delayed neutrons for restarting the operation even after
approximately 20 minutes.
The reactor is shut down by inserting control rods into the reactor, which stops the
fission chain reaction. Emergency rods are inserted to the reactor in the case of an emergency
or accident. Absorption materials for regulation, compensation and emergency systems will
be described in following sections.
3.2.2 Thermal energy release
Energy released during the fission consists mostly of energy released directly during
the fission process (180 MeV) and energy of radioactive decay of fissile products, which is
released afterwards. For 233
U fission the energy is approximately 5 MeV lower and about 5
MeV higher for Pu decay - the differences are therefore quite small.
Kinetic energy of fission fragments stops close to the place of fission, which is heated
by a so-called "fission shock" that causes an increase of temperature for hundreds or even
thousands of kelvins and relocation of a large number of atoms. The fission shock lasts for a
very short time (about 10-7
s) and affects about 107 atoms. However, it is constantly repeated
at different places in the fuel and it might therefore affect the fuel properties (see radiation
growth). Furthermore, it gradually increases the temperature of fuel used in energy reactors.
Released neutrons have the total energy of about 5 MeV and are most slowed down in
the moderator. Energy of the long-range γ radiation manifests itself as heat in the fuel, active
zone materials and reactor shielding. Short-range β radiation transfers its energy already in the
fuel. Energy of neutrinos accompanying the decay process must be considered lost, since it is
not captured by the reactor.
The division of created heat is different in FBR reactors. Approximately 75 % of
thermal energy is released in fuel elements, the rest in the reproduction zone. These values
apply if the reactor is in an equilibrium state. Thermal energy of fission fragment radiation
does not manifest itself at the beginning of the campaign but needs to be taken into
consideration after the reactor shutdown (approximately 6 % of the stabilized state).
The released thermal energy per unit of time dQ can be calculated using a simple
relation. The released thermal energy during one fission reaction equals Et. The volume unit
contains N fissile nuclei. Their fission probability is determined by the fission cross section
f and neutron flow . dQ can be calculated as follows:
dQ = Et . N . σf . ∅ . dV
This relation defines the specific heat flux per volume unit:
𝑞 = dQ
dV= Et . N . σf . ∅
which is an important variable for efficient dimensioning of nuclear reactors and their
comparison.
Integration of the first relation yields the total output of all fuel elements. Mean
neutron flow and mean fs can be introduced for an approximate calculation and the mean
specific neutron flow qs can be derived from these two values. Then the total thermal output
of the reactor is:
Q = ∫ Et
v
0
. N . σf . ∅ . dv = Et . N . σfs . ∅stř . V = qs . V
It follows from the above stated facts that the specific thermal output and therefore the
reactor's thermal output is proportionate to the neutron flow . The size of the neutron flow is
therefore unlimited from the nuclear point of view. The possibilities of heat dissipation
represent the main limitation. The heat removal is limited by the strong dependence of
strength properties and corrosion of construction of construction materials on the temperature.
This temperature dependence of reactor materials limits the maximum specific heat flux
(expressed in MWm-3
) to 10 for high-temperature reactors, 20 to 60 for boiling water reactors,
40 to 100 for pressurized water reactors and 400 to 1000 for fast breeder reactors.
Specific heat flux is also affected by fission component N. This is directly proportional
to the fuel density p and fuel enrichment r (%) and can be expressed by the following
relation:
N = LaMp . ρp . r
where La is the Avogadro constant [mol-1
] and Mp is the molecular weight of fuel. Specific
heat flux can be calculated after establishing N to the equation for calculating the specific heat
flux per volume unit. This allows us to compare used fuels and to determine the enrichment
required to achieve the same qs at constant and f. It is therefore clear than when UO2 is
used instead of uranium metal, it needs to be enriched 2.16 x and UC needs to be enriched
1.47 x. This advantage of metal fuels is cancelled out by lower radiation and thermal stability.
Specific heat flux is directly proportional to neutron flow . That is why the neutron
flow distribution in the reactor's active zone cross section can be considered as the specific
heat flux distribution. Materials in the active zone demand the heat flux to be as even as
possible. The flux can be evened by the already mentioned fuel replacement program based
on fuel enrichment and irradiation.
Distribution of temperatures along the height of the active zone, i.e. along fuel
elements, depends not only on the neutron flow but also on the position of the element relative
to the active zone axis and mainly on the coolant flow. When flowing through the active zone
the coolant is gradually heated and together with value q affects the temperature of the fuel
element. This problem is especially serious for cooling by gases where the difference of input
and output temperatures is known.
3.2.3 Fuel exchange
During operation in a nuclear reactor there occurs a change of the isotopic composition
of active zones influenced by the absorption of neutrons, fission and core breakdown. At
every moment it is necessary to know kef (the effective multiplication co-efficient) for a given
compound and the reactor state and character change kef in time. This deals with what is
called long-term kinetics. The common shape of the equation of long-term kinetics has the
form:
𝑋 = A + B + C − D − E − F
where: X – the concentration change of isotope c in time,
A – formation during fission,
B – isotope formation of neutron absorption,
C – isotope formation through breakdown,
D – isotope depletion of neutron absorption
E – depletion by fission,
F – depletion by breakdown.
It is possible to write this equation for each isotope occuring in an active zone. The
concentration dependence of heavy isotopes in time is given in Figure 3.2.
3.2.4 Reactor poisoning and slagging
The chain fission reaction creates a large number of nuclei of different elements from
the center of the periodic system. Some of these nuclei have a large microscopic cross section
of thermal neutron absorption and therefore significantly affect the balance of neutrons in the
active zone. Absorption of neutrons by short-lived isotopes is called reactor poisoning. The
most significant poisoning is caused by 135
Xe. The absorption of neutrons by stable, long-
lived isotopes is referred to as reactor slagging. Let us focus on the accumulation and effect of 135
Xe, which causes reactor poisoning. The cross section of thermal neutron absorption of this
1 – 241
Pu,
0 – 240
Pu,
5 – 235
U,
9 – 239
Pu.
Fig. 3.2 Dependence of concentration c of heavy isootopes on time.
ef. time
isotope is more than 5000x larger than the 235
U cross section. Xenon is formed in the reactor
either directly as a fission product (only about 5 %) or by decay of the unstable fission
product 135
Te in accordance with the following equation:
Te → I53135 → Xe 54
13552135
The first transformation takes 2 minutes, the second transformation takes 6.7 hours. Xenon is
created in two ways:
a) natural method (decay),
b) neutron absorption creates 136
Xe, which has a small microscopic absorption cross
section.
Reactor slagging is caused for instance by 149
Sm. Samarium is a stable isotope that is
not created directly by fission but by neodymium decay. In order to gain a full understanding
of the effect of xenon on the neutron balance a so-called iodine pit needs to be mentioned
The iodine pit can be described as accumulation of xenon after the reactor output decreases.
Depending on various parameters (reactor power, work at this power, power after reduction,
etc.), the reactor might not be started due to the iodine pit. In this situation it is necessary to
wait until xenon undergoes a natural decay. The kinetics of the creation and existence of the
iodine pit are determined by the half-life of 135
Xe, which is larger than the half-life of 135
I; that
is why the concentration of 135
Xe will be increasing at first. At reduced reactor power new 135
I
is created only to a limited degree and the xenon concentration therefore reaches its peak and
then decreases.
3.3 Fuel elements
A fuel element is a construction unit containing fuel material, where energy is released
through fission reactions. The fuel element is the most important technical element of the
nuclear reactor since its properties determine and limit the technical and economic properties
of the nuclear production block.
Fuel element is the most exposed part of the nuclear reactor because it works under
specific hard conditions of the active zone. Physical and chemical processes and structural
changes that take place in the fuel and in construction materials of fuel elements cause
significant changes of most of their properties. Reliable and safe operation of fuel elements
under these conditions in all operation states must be ensured.
3.3.1 Nuclear fuel
Nuclear fuel is the source of thermal energy in the nuclear reactor. Heat released
through a fission reaction is dissipated through the reactor coating and coolants. Fuel consists
of fissile isotopes 235
U, 233
U and 239
Pu and breeder isotopes 238
U and 232
Th. The selection of
concrete fuel is determined by the used type of fuel cycle.
The nuclear fuel geometry is simple due to technological reasons - it consists of a
twig, rod, pipe or plate, ball (for HTGR reactor microelements) and predominantly cylindrical
tablets. Fuel materials will be discussed in more details in one of the following chapters; here
we will discuss only some of the basic data. Fuels can be divided to metal and ceramic based
on their chemical composition. Based on the isotopic composition we can distinguish natural
uranium, enriched uranium, plutonium compound with uranium or plutonium with thorium.
During operation the maximum fuel temperature needs to be taken into consideration.
This temperature must be reasonably lower than the melting or transformation temperatures of
the given fuel. These values are included in table 3.1, where TM refers to the melting
temperature and Ttr to the transformation temperature of the given fuel.
Metal fuels offer suitable physical properties for use in nuclear reactors, they have
high density of fissile and breeder isotopes and good thermal conductivity. Another advantage
of this type of fuel is its mechanical strength allowing the use of self-supporting elements and
workability with low tolerance values. It is used in the form of low-alloyed α alloys and γ
alloys, where higher alloy concentration creates an isotope grid.
However, the radiation stability of metallic uranium is limited. The volume growth
and radiation growth change during the operation and the combination of exterior load and
reactor radiation leads to the so-called radiation creep. The above listed factors limit most of
the parameters of reactors with uranium metal. That is why the utilization of uranium metal is
currently limited.
Ceramic fuels include for instance uranium oxide, plutonium oxide and thorium oxide
and their compounds, as well as carbides, nitrides, etc. Ceramic fuels have higher radiation
stability; however, their disadvantage is their lower density, thermal conductivity and strength
(the load bearing function is then transferred to the coating). These disadvantages become
most obvious in oxide fuels. Oxide fuels are used mostly in the form of sintered tables with a
diameter of 5 - 20 mm and length of 10 - 30 mm. Tablets with a small hole in the middle are
used for high load fuels. A different form is represented by microelements of high-
temperature reactors with a diameter of 300 – 800 m.
Dispersion fuel elements contain fissile isotopes dispersed in the form of a compound
or alloy from non-fuel material. The properties of these fuels in dispersion alloys depend
mostly on the matrix properties (for example, Al in research reactors) or on the properties of
both oxide fuel and ceramic matrix for non-metallic dispersion fuels.
3.3.2 Construction and coating materials
As the name indicates, the construction and coating materials are used to protect the
fuel material (coating) and for the construction of the fuel elements, as well as other
components of the active zone. These materials will be studied in more detail in one the
following chapters. This chapter will present the requirements placed on these materials, their
radiation damage and individual metals and their alloys. These include mostly zirconium,
stainless steel, magnesium, beryllium, titanium, etc., all of which need to be suitably alloyed.
Fuel element coatings are exposed to the most demanding operating conditions in the
reactor and are subject to the strictest requirements. Other construction materials must meet
the requirements derived from them. That is why we will focus mostly on coating materials,
which - after they are verified for coating - are usually also used as construction materials.
Fission reaction products remain contained in the metal fuel. In ceramic fuels these
products are gradually released during operation into the space between the fuel and the
coating. Coating materials are used based on the reactor type, especially depending on the
type of coolant or fuel. Research reactors use aluminium; energy reactors must use coatings
with a higher thermal stability. See table 3.2 for examples.
Table. 3.1 Limiting temperatures of nuclear fuels.
fuel U metallic UO2 UC UN PuO2 ThO2
temperature
[°C] Ttr = 668 TM = 2880 TM = 2350 TM = 2850 TM = 2260 TM = 3300
Thermal stability of the coating refers both to the maintenance of mechanical
properties at high temperatures, as well as the corrosion resistance at these temperatures. The
higher the temperature of the coating material, the higher the achievable efficiency of the
nuclear reactor.
In energy reactors the highest thermal output possible from the given volume is
required. With regards to the requirement of the highest possible efficiency of the heat cycle,
it is necessary to ensure the highest possible temperature of the coolant. The coolant
temperature depends on coating properties, whose limiting temperatures are summarized in
table 3.2 based on the used coolant. The requirement of the highest possible coolant
temperature and highest possible heat flux while observing the maximum fuel and coating
temperature can be met if the following is used:
a) small diameter of fuel rods,
b) the highest possible thermal conductivity of the fuel and coating (see table 3.3),
the highest possible heat transfer coefficient – this depends on the coolant properties and
pressure loss.
The requirement of observing the maximum allowed fuel and coating temperature is
closely connected with the distribution of temperatures in the active zone. If we are familiar
with the distribution of temperatures, the amount of obtainable heat is limited by the fuel
element with the highest temperature, which is usually located in the so-called hot channel in
the active zone axis. The course of temperatures of other fuel elements along their length is
similar, yet their absolute values are lower.
The distribution of temperatures along the length of the fuel element is represented in
Fig. 3.3. A large amount of heat is released in the fuel corresponding to the neutron flow,
which consists of a sinusoidal wave with the maximum value at the center of the fuel element.
The released heat gradually increases the coolant temperature Tch, while it increases the most
around the neutron flow maximum in the central part of the fuel element. The difference in
the coolant temperature at the input and at the output, which depends on the type of coolant
and the heat flux, determines the shift of maximum temperatures at the coating surface T2max,
at the fuel surface T1max and at the fuel element axis T0max from the central part towards the
coolant output. The smallest shift occurs in boiling water reactors, in pressurized water
reactors it is a little higher and the most significant shift is recorded in gas-cooled reactors,
where the difference between input and output temperatures of the coolant is quite substantial.
The above described distribution of temperatures along the length of fuel elements in the
active zone axis and along its cross section needs to be taken into consideration when
determining maximum heat fluxes in the reactor.
Table. 3.2 Limiting temperatures of the coating with regard to used coolant.
material Al Mg alloys Zr alloys stainless steel
max. temperature [°C] 150 520 350 360 600 700
coolant H2O CO2 H2O H2O water steam Na
Tab. 3.3 Thermal conductivity of materials for fuel elements construction*(
stainless).
material U UO2 UC Th Pu Al Mg Zr AC* steel graphite
λ [W.m-1
.K-1
] při 20°C 27 5.4 29.3 37.7 6.6 209 157 16.7 167 120
All of the above stated thermal dependencies are directly connected with thermal and
mechanical properties of the fuel and coating, their geometry and dimensions. The heat
transfer coefficient depends on properties of the coolant and the construction design of the
fuel element as a whole, which is to prevent excessive erosion, corrosion and vibrations.
3.4 Coolants
Coolants are used to dissipate heat from the active zone of the nuclear energy reactor
and then transfer it in the heat exchanger. The type of used coolant is one of the most
important factors when designing and operating a reactor. The main purpose of the coolant is
the dissipation of heat from the reactor. The fact that the coolant flows in the primary reactor
circuit is why its nuclear properties are so important, as well as its effect on reactor materials.
See table 3.4 for comparison of coolants.
Fig. 3.3 Temperaturt curve along the element of reactor (gas cooled), solid line –
temperature curtve T0 – element axis, Tch – coolant, T1 – surface fuel, T2 – surface
cladding. Dashed line – neutron flux distribution.
Table. 3.4 Basic thermo-physical properties of selected coolants.
coolant CO2 He H2O Na
density [kg.m-3
] 1.39 (100) 0.126 (100) 958 (100) 928 (100)
0.67 (500) 0.061 (500) 794 (250) 780 (700)
Specific heat
[J.kg-1
.K-1
]
918 (100) 5204
4230 (100) 1382 (100)
1155 (500) 4610 (250) 1277 (400)
coef. of thermal
conductivity
[W.m-1
.K-1
]
0.025 (100) 0.179 (100) 0.712 (100) 86.3 (100)
0.055 (500) 0.305 (500) 0.82 (250) 71.2 (400)
TM [°C] -56.6 -271.4 0 97.8
TV [°C] -78.5 -268.9 100 883
The value in parentheses indicates the temperature in °C.
3.4.1 Gas coolants
Nuclear properties of gas coolants are very good or even excellent. The absorption of
thermal neutrons is negligible, the induced reactivity is low, while the radiation stability is
high. Their heat transfer properties are depreciated by their low density that causes their low
specific heat capacity. However, the thermal stability is excellent. In addition, the viscosity of
gases is also satisfactory.
Reactor materials do not react with helium, whereas CO2 might react with graphite or
carbon contained in steels. The first reactors were cooled by air due to its availability.
However, the disadvantages of air can be easily deduced from its properties and the above
listed requirements.
3.4.1.1 Carbon dioxide Carbon dioxide has very good nuclear properties, low value of a, short induced
radioactivity of isotope 16
N with half-life of only 7.3 s. Heat-transfer properties of CO2 as a
gas, however, are not very good. Its thermal stability up to 1000°C is very good but at
temperatures exceeding this value CO2 is thermally disassociated to CO and O2. The study of
the impact of CO2 on reactor materials determined that it is suitable for metals – it is used in
reactors with magnox coatings, but it reacts with carbon graphite or steel:
𝐶𝑂2 + 𝐶 ↔ 2 𝐶𝑂
For graphite this reaction starts already at a temperature of 400 °C and accelerates with
increasing temperature. The thermal and pressure dependence of this reaction is known also in
classic metallurgy. CO2 is used as a coolant in GCR and AGR reactors; in the past it was used
in the KS 150 reactor in the A1 power plant in Jaslovské Bohunice. Another advantage of
this coolant is its low price.
3.4.1.2 Helium This gas is almost an ideal coolant for nuclear reactors. Its nuclear properties are
excellent, a is negligible, there is no induced radioactivity and it has complete radiation
stability. It has excelled heat-transfer properties, the specific heat capacity and thermal
conductivity are about 6x times higher than for CO2, it has an ideal thermal stability and low
viscosity. As an inert gas it does not react with any material. However, its admixtures,
oxygen and nitrogen, which represent approximately hundredths of a percent, cause corrosion
of construction materials, especially at high temperatures. At temperatures exceeding 800°C
oxygen forms a compound with carbon graphite.
Other disadvantages of helium are its high price and easy diffusion as a monatomic
gas. It is therefore more difficult to reach the required leakproofness of the cooling circuit
than when CO2 is used (on the other hand, its diffusion is used for leakproofness tests of
reactor components). Helium is still successfully used as a coolant in HTGR reactors.
3.4.2 Liquid coolants
Other coolants in nuclear reactors are used in liquid state. These include water,
alternatively heavy water, organic compounds and molten salts. Liquid metals are always
listed separately due to their different properties. Due to the fact that normal water is used to
obtain heavy water (CANDU), normal water is often referred to as light water (PWR, BWR).
Their cooling properties do not significantly differ and that is why we included both of them
together in the following section.
3.4.2.1 Water Light water is used in PWR or BWR reactors, depending on the boiling point and ratio
of created vapour. Water has good nuclear properties. Induced radioactivity of clear water is
negligible, created isotopes 19
O and 16
N have short half-life, 3H is only a β emitter. However,
induced radioactivity is caused by corrosion products and impurities. Radiation stability is
given by the radiolysis degree, during which ions H+ and OH
- are produced. Free radicals OH
react between each other and produce gases H2 and O2 and hydrogen peroxide H2O2.
Radiolysis is accelerated by certain ions and increasing temperature. That is why boiling
water reactors and heavy water reactors are equipped with a recombination device, where the
explosive mixture is recombined by combustion or catalysis.
Heat-transfer properties of water are excellent, it has the highest volumetric specific
heat from all used coolants and only liquid metals have better thermal conductivity. This
enables intensive heat dissipation from the reactor, which - together with the moderation
effects - results in smaller dimensions of the reactor. The thermal stability is limited by the
low boiling temperature of water, i.e. the change of state. That is why pressurized water
reactors (PWR) are used in addition to boiling water reactors (BWR); in PWR reactors the
pressure increases the boiling temperature of water.
The corrosive effect of water on reactor materials depends namely on their properties
and corrosion resistance - these will be discussed in corresponding chapters. Corrosion
accelerates the effect of reactor radiation. Radioactive products of corrosion, such as 59
Fe,
need to be removed continually during the reactor operation. High demands on the purity of
water used as a coolant are due to the following facts:
impurities might affect the neutron balance of the active zone,
impurities and corrosion products introduce radioactivity to the whole primary circuit,
impurities and corrosion products deposit on heat-transfer surfaces, where they reduce
the heat-transfer and compromise the operation safety.
admixtures to control corrosion or water radiolysis must not introduce any by-
products.
Materials in the primary circuit of water-cooled reactors are made from corrosion-
resistant materials - alloyed stainless steel and zirconium alloys; layers of corrosion-resistant
steel are welded to internal surfaces of carbon steel components.
3.4.2.2 Molten salts Molten salts are suitable for nuclear reactors operating at higher temperatures. LiF +
BeF2 compound with the addition of ZrF4, has a high radiation and thermal stability, low
vapour pressure and conductivity similar to water, with the melting temperature of 350°C.
3.4.3 Liquid metals
The decisive reason for the use of liquid metals as coolants are mainly their excellent
thermal properties - excellent heat dissipation and high boiling temperature. On the other
hand, their disadvantage is their very aggressive corrosion and solidification during cooling to
normal temperatures. Furthermore, radioactivity is induced in most liquid metals. Three
groups of metal coolants are used:
1) Sodium and sodium-potassium alloys – for the primary circuit of fast and thermal
reactors. Eutectic alloys consists of 22 % Na + 78 % K and melts at -11°C, sodium
melts at 98°C, potassium at 64°C.
2) Bismuth and lead-bismuth alloys – melting temperature of Bi is 271°C, for Pb it is
327°C, alloy with 44 % Pb has a melting temperature of 125°C.
3) Mercury – only for the secondary circuit of thermal reactors, alternatively for the
primary circuit of fast reactors, melting temperature of -39°C.
3.4.3.1 Sodium This metal is the most used coolant with the most favourable properties. Studies of its
nuclear, thermal and chemical properties discovered the following.
Nuclear properties
It has the required low moderation capacity for use in fast breeder reactors (FBR), low
absorption cross section for thermal neutrons in thermal reactors. However, induced reactivity
is produced, when the reactor radiation activates natural isotope 23
Na to: 22
Na – β+ E emitter = 0.54 MeV, and γ E emitter = 1.2 MeV, half-life of 2.6 years
24Na – β
+ E emitter = 1.39 MeV, and γ E emitter = 2.75 MeV, half-life of 15 hours
Isotope 24
Na is strongly prevailing. This high induced radioactivity requires the
inclusion of another, i.e. secondary sodium circuit before the tertiary water circuit in order to
prevent direct contact of active sodium with water and significant shielding of the primary
circuit. Its advantage in comparison with compounds is that it does not decay as an element.
Thermal properties
Its specific thermal capacity is sufficient and the thermal conductivity coefficient is
excellent. Its heat-transfer coefficient is high even at low speed (up to 130 kW.m-2
.K-1
) and it
has low viscosity similar to water, resulting in low pumping work. Its thermal stability is high
– a large range of the liquid phase up to the boiling point of 883°C while having an acceptable
melting temperature of 98°C, which allows lower pressure in the reactor. Upon solidification
the volume decreases by 2.7 %.
Chemical properties
Sodium is a highly reactive chemical element. Technically pure sodium contains
approximately 0.3 % of admixtures, namely K, Ca, Mg and Si. Aggressive corrosion of
sodium is however increased by even a low content of oxygen; that is why the oxygen content
must be monitored continuously and reduced if required. The reaction of sodium with water,
water vapour and air is intensive, accompanied by the creation of hydrogen and a large
amount of heat. For large contact surfaces the reaction of sodium with water has an explosive
character:
H2O + 2Na → Na2O + H2 + Q.
This reaction cannot be stopped, not even by increasing the pressure. Hydrogen
released during this reaction creates an explosive mix with air oxygen. Reaction of sodium
with oxygen creates Na2O – in solid form the reaction is slowed down by the increasing layer
of oxide until it stops. In liquid sodium the reaction occurs faster through burning.
3.5 Moderators and reflectors
Moderator is a substance that reduces the energy of fissile neutrons it the reactor's
active zone by elastic collisions up to the thermal neutron energy and thus increases the
probability of a fission reaction. Fast neutrons released by a fission reaction with a mean
energy of 2 MeV reduce their kinetic energy by so-called elastic scattering to the energy of
thermal neutrons, which is 0.025 eV - i.e. they decrease their energy by approximately eight
orders of magnitude. Moderators in thermal reactors are usually used also as reflectors.
The selection of a suitable moderator is governed by nuclear and physical properties of
the moderator and its behaviour in the reactor's active zone, which is given by its thermal,
chemical, mechanical and technological properties. Used moderators include H2O, D2O,
graphite, beryllium and polyphenylenes.
The resistance of liquid moderators to reactor radiation expressed as the radiation
stability is determined by the radiolysis degree. So-called radiation growth occurs in graphite
and transmutation needs to be taken into consideration for beryllium. The selection of
moderators plays a crucial role for the whole reactor concept and is one of the factors
determining the selection of other materials. Nuclear and technical properties of moderators
are decisive; for example, liquid moderators can also be used as a coolant.
3.5.1 Light water
The main advantage of light water is mainly the fact that it can be used as a coolant at
the same time - i.e. a substance with moderation effects is located close to the fuel. The
moderation ratio of light water is only 144 due to its high absorption cross section, which
leads to the use of enriched fuel. Additional nuclear properties of water, thermal stability and
its corrosive effect are described in next chapters.
3.5.2 Heavy water
Heavy water is the best moderator of all, having a moderation ratio of 6850, which is
almost 50x more than the ratio of light water. The main reason is the very low absorption
cross section of deuterium nucleus, which is almost three orders of magnitude lower than the
hydrogen nucleus. Unenriched natural uranium can be therefore used to operate reactors if
heavy water is used as the moderator. If heavy water is used as the coolant as the same time,
the active zone dimensions can be reduced. Normal water contains 0.015 % of heavy water.
However, obtaining heavy water from normal water is very difficult as their properties are
very similar, see table 3.5.
Thermal and physical properties of heavy water differ from those of light water in very
few aspects. Its use as a coolant and moderator at the same time is therefore governed by the
same principles as described in next text.
The similarity of properties of normal and heavy water as two different isotopes of the
same element causes a specific problem - isotope separation.
Table. 3.5 Comparison of light and heavy water.
feature H2O D2O
molecular weight mu 18.0156 20.0282
TM [°C] 0.00 3.913
TB [°C] 100.00 101.42
density [Kg.m-3
] 1000 1106
viscosity at 25°C [P.s] 891 1099
vapor pressure at 20°C [Pa] 2338.1 2031.5
Tcrit [°C] 374 371.5
pcrit [MPa] 22.12 21.46
standard electrode potential [V] 0.00 -0.00354
The electrolysis method stems from the fact that during electrolysis more hydrogen
than deuterium is released at the cathode, since deuterium remains in the electrolyte. This
method is energy-demanding and it was first used to produce heavy water. It is suitable for
enrichment of already partially enriched heavy water.
The distillation method uses the slight difference between the boiling point of heavy
and light water. Distillation must consists of several stages, whereas the residue always
contains a little more heavy water. This method places high demands on the equipment
dimension. Its efficiency can be increased by reducing the pressure. Distillation of hydrogen
containing deuterium is much more efficient, it uses the difference of boiling points of
hydrogen and deuterium ΔT = 3.1 K and the vapour pressure ratio of almost 2.5 for H2/D2.
The only obstacle is the necessity to operate at low temperatures and the explosibility of
hydrogen.
Another available method is zone freezing (inverse to the zone melting), which uses
the difference of 3.8 K between the melting point of light and heavy water. However, this
method is no longer used.
Isotope exchange reactions use the slight difference of reactivity between isotope
molecules. Even though isotopes have identical chemical properties, compounds or isotope
ions react with a slightly different speed, which is used for the separation in the following
system: hydrogen sulphite - water, water - water, water - ammonia. Isotope exchange reaction
or distillation method usually constitute the first stage followed by electrolysis.
3.5.3 Graphite
Graphite was the first used moderator (Fermi, 1942) and its use has also been recorded
for GCR, AGR, LWGR, HTGR reactors. Graphite is the only usable moderator in high-
temperature gas-cooled reactors (HTGR) due to its high thermal resistance. It takes solid form
up to a temperature of 3700 °C.
Its moderation ratio is 170, which is a little higher than the ratio of light water. The
reactor radiation causes a radiation growth for up to several percent. This growth needs to be
respected when designing the active zone. Radiation damage caused by accumulated Wigner
energy, which is harmful to plutonium-producing reactors, does not affect energy reactors due
to increased temperatures.
Thermal properties of graphite are excellent; in addition to the high thermal range of
the solid phase, it also has excelled thermal conductivity coefficient, which can reach up to
160 W.m-1
.K-1
based on the used production method. The compressive strength of graphite
ranges between 40 – 70 MPa (it increases with T). Thermal expansion and conductivity
depend on the crystallite orientation and are therefore anisotropic. Natural graphite cannot be
used for the production; reactor graphite is produced from organic substances, usually
petroleum coke, using the recrystallization method at high temperatures, which ensures the
required nuclear purity.
3.5.4 Beryllium
Nuclear properties of beryllium in terms of the requirements on moderators are
favourable. The moderation ratio of 144, i.e. the same as light water and comparable to
graphite, is caused by low a and good r. Beryllium is used as the moderator mainly in the
form of BeO in transport reactor (nuclear submarines).
Metallic beryllium is used as a construction material thanks to its high strength, low
density and favourable thermal stability. However, its wider application is not possible due to
technological reasons and its toxicity.
3.6 Absorption materials
Absorption materials consists of elements with high absorption cross sections. After
neutron absorption by the absorber nucleus, γ particle is emitted (radiative capture) or helion
(alternatively proton) is released (transmutation). Absorption materials are used to control the
reactor operation and manage reactivity reserves or to protect the reactor. Based on these
criteria we can distinguish control rods, which maintain reactor output at the required level,
compensation rods, which compensate the loss of reactivity by fuel irradiation and mainly by
creation of fission fragments, and emergency rods, which stop the chain reaction in case of an
emergency or accident. Absorber dissolved in a coolant, e.g. boric acid, is therefore
sometimes used for regulation and compensation. Requirements on absorption materials can
be listed in the usual order: nuclear, thermal, chemical, mechanical and technological.
The selection of materials for absorbers based on the high absorption requirement: Gd,
Sm, Eu, Cd, Dy, B, Ir, Hg, In, Er, Rh, Tm, Hf, Lu, Au, Re, Ag need to be connected with
other listed requirements. Some of the elements are out of the questions due to their price (Au,
Re, Rh, Ir), other due to the TM value (Hg, In). Only hafnium can be used in the elementary
metal form; binary alloy of AG with Cd or ternary Ag-Cd-In alloy can be used as metal,
which combines higher TM of silver with higher absorption cross section of alloys.
Lanthanides are used as oxides or in their compounds, usually dispersed in metal matrices;
boron is used in various forms. Selected properties of absorption elements are listed in table
3.6.
3.6.1 Materials containing boron
High absorption cross section of boron for thermal neutrons (750×10-28
m2) is caused
by isotope 10
B, which forms 19.7 % of natural boron and its a value = 3837×10-28
m2. For
isotope 11
B, which constitutes the rest of natural boron, the a value= 0.05×10-28
m2. The
disadvantage of isotope 10
B is the fact that after absorption it is transmuted to 7Li, which leads
to the loss of absorption properties (isotope 7Li has only a negligible neutron absorption), and
it also causes radiation embrittlement due to the presence of other elements in the matrix (He
and Li).
Boron is used in absorption rods in different forms - in alloys (in steels or stainless
steels), in compounds such as B4C or compounds dispersed in ceramic matrices.
Table. 3.6 Selected properties of neutron absorbers elements.
element
absorption
cross
section a
density melting
temperature
coef. of
thermal
expansion
coef. of
thermal
conductivity lattice
[10-28
m2]
[103 kg.m
-
3] [°C] [10
-6 K
-1] [W.m
-1.K
-1]
Gd 44000 7.9 1312 9.7 10 HTU
Sm 6500 7.5 1072 13.3 rhombohedral
Eu 4500 5.2 826 26 KSC
Cd 2400 8.6 321 29.8 93 HTU
B 750 2.3 2200 8 26 tetragonal
In 190 7.3 156 33 25 tetragonal
Hf 115 13.1 2222 5.9 22 HTU
Ag 60 10.5 960 19.7 418 KPC
3.6.1.1 Steels Either boron steels or stainless steels containing boron are used. Commonly used
steels contain approximately 2 % B. The admixture of boron in steel significantly worsens its
plastic properties; the elongation drops to only 2 % at 2 % content of B. Reduced plasticity of
the material causes technological problems during metal working, as well as residual tension
(especially at joints) that are associated with the risk of cracks. This embrittlement is also
increased by radiation and related transmutation neutron reactions.
3.6.1.2 Dispersion materials Corresponding borides ZrB2 or TiB2 in Ti or Zr are dispersed in a titanium or
zirconium matrix. It is necessary to ensure that the matrix has a certain relaxation capacity
due to the transmutation reaction.
3.6.1.3 Powder materials Embrittlement of boron steels and dispersion materials led to the development of
absorption rods in the form of capsules filled with powder materials. For example, even
though hot-pressed B4C disintegrates due to radiation, it still continues to fulfill its absorption
function enclosed in a stainless steel capsule.
The corrosion resistance of absorbers containing boron is low. Only boron steel
exceeds the properties of low-carbon steel. In the air boron oxidizes at only 100 °C, boron
carbide is resistant up to 1000 °C. The only boride resistant to corrosion in water is YB4.
3.6.2 Hafnium
The requirements on absorption materials are best met by hafnium, which is the only
element that can be practically used in its elementary form. Its absorption cross section is
115×10-28
m2. It would appear that the cross section is quite low but it can be utilized in full in
metal form; in addition, the absorption capacity of hafnium does not decrease during
operation since the created isotopes also work as absorbers. The advantage of hafnium is its
high transformation temperature and melting temperature. In terms of corrosion hafnium is
more resistant to corrosion in water and water vapour than zirconium alloys and can be used
up to a temperature of 400 °C.
Mechanical properties of hafnium are similar to those of zirconium; the yield strength
at 20 °C is approximately 160 MPa and decreases with temperature. The impact of hydrogen
is similar as in zirconium. However, the disadvantage of hafnium is its high density 13.1×103
kg.m-3
, which burdens the motion mechanism of control rods.
The production technology of hafnium is based on its similarity with zirconium and
will be therefore explained in the chapter dedicated to zirconium production. Pure hafnium
contains up to 3% of zirconium, which remains there after its separation from reactor-grade
zirconium.
3.6.3 Cadmium
Cadmium has a high absorption cross section – 2400×10-28
m2, neutron absorption by
radiation capture creates cadmium isotopes without the absorption capacity. A significant
disadvantage of cadmium is its low melting temperate (321 °C) and boiling temperature (765
°C) which exclude its use in elementary form. Other thermal properties are not very
favourable either.
As far as chemical properties are concerned, it is necessary to mention the toxicity of
cadmium salts and cadmium oxide vapours, as well as insufficient corrosion resistance of Ag-
Cd and Ag-Cd-In alloys.
The sheathing of absorption rods of pure cadmium is made of suitable construction
material, e.g. steel or aluminium, that protects cadmium from contact with the environment
and fulfills all functional requirements. Absorption rods Cd with Ag or Ag-In-Cd combining
the higher melting point and higher absorption of Cd are usually sheathed by stainless steel
protecting it from corrosion. These absorbers are used more often.
3.6.4 Lanthanides
These elements have the highest cross section in their natural form in the following
order Gd, Sm, Eu, Dy, Er, Tm, Lu, etc. Other nuclear properties are also excellent – they do
not undergo transmutation connected with helium production and their radiation stability is
excellent. Radiation does not reduce the absorption capacity of Eu, Dy and Er. However,
favourable thermal properties of lanthanides cannot be practically used due to their reactivity
associated with low corrosion resistance and due to unsatisfactory mechanical properties that
prevent their use in elementary form.
When producing these pseudo alloys, it is necessary to take into consideration the
compatibility of oxides with the matrix; for example, the matrix must contain the lowest
possible content of Si due to the volume growth during radiation. Then, Eu2O3, Gd2O3 and
Sm2O3 and their compounds are chemically stable in Fe, Ni, Cr matrices up to a temperature
of 1250°C.
3.7 Other components
3.7.1 Reactor pressure vessel
Reactor pressure vessels are one of the largest and heaviest devices in the whole
nuclear power plant. Pressurized water and boiling water reactors (PWR and BWR) are
typically equipped with steel pressure vessels; graphite gas-cooled reactors (GCR) typically
contain pressurized vessels made of prestressed concrete. The requirements placed on reactor
vessels – high operating pressure, temperatures, material and operational safety – are
exceptionally high and practically exceed all requirements placed on pressurized vessels used
elsewhere in industry. Furthermore, the requirements are further increased by the reactor
radiation (volume heat build-up, activation and radiation damage). Steels with better quality
and quantity characteristics than those used in conventional structures are required to produce
these vessels. Increased requirements are placed on steel purity, strength properties and
toughness. That is why extensive and costly research of the resistance of these materials
against embrittlement and fatigue failure is required. In fact, these properties are purposeful.
Nuclear requirements - the material including welded joints must be resistant to
radiation damage and the induced activity, especially from cobalt, must be reduced to
minimum. Another technological requirement is the mutual weldability even for large
thickness. That is why pressurized vessels in light water reactors can be made of low-carbon
or low-alloy steels, where the protective lining is created by welding or rolling of high-allow
austenitic stainless steels. Vessels in fast breeder reactors are made of 18/10 high-alloy steels
with low content of carbon, if possible.
3.7.2 Reactor shielding
High energy neutrons are produced during fission in the active zone. It is therefore
necessary to reduce these high levels of energy and then absorb the neutrons reduced energy
levels. Decreasing energy of fissile neutrons and energy levels by MeV units is realized by
inelastic scattering with nuclei of medium and high mass numbers, such as plumbum, iron or
barium. Secondary γ radiation is produced during this process. Neutrons are also slowed
down by elastic scattering using regular moderators and are then absorbed by elements with
high a – i.e. absorbers.
Efficient reduction of neutron energy with absorption of γ radiation can be achieved by
combining very light and very heavy nuclei - e.g. heavy metal hydrides. The most commonly
used material for shielding of nuclear power reactors is concrete. It is also used as the load-
bearing construction components thanks to its low price. Common concrete is suitable for
shielding against neutrons. The shielding properties of concrete against γ radiation can be
improved by adding admixtures, such as iron oxide, barium sulphate, etc.
Summary of terms in this chapter (subchapter)
Nuclear reactor
Reactor core
Fuel element
Coolant
Neutron moderator
Multiplication coefficient
Questions to the covered material
Briefly describe basic types of power nuclear reactors.
Describe basic difference between reactors PWR a BWR.
Compare operational safety of PWR and LWGR reactors.
Attempt to outline key gaps of development of the FBR type reactors in nuclear
power.
List the pros and cons of liquid metal used as coolants.
Justify the use of heavy water in certain types of nuclear reactors.
4. Nuclear fuels
Time to study: 5 hours
Aim After studying this section the student should be able to:
Describe the basic types of nuclear reactors, their advantages and
disadvantages.
Describe basic parts of reactors.
Describe the nuclear fuels, coolants, moderators, etc.
Lecture
Fuel materials are divided into fissile and breeder materials, depending on whether
their cross section is sufficient for fission by thermal neutrons or for neutron capture. Based
on this criterion, fissile materials include 233
U, 235
U, 239
Pu and 241
Pu, whereas 238
U, 240
Pu and 232
Th are classified as breeder materials. Important nuclear fuels that appear in the nature
include uranium and thorium. Natural uranium contains approximately 0.71 % 235
U and
0.0058 % 234
U, natural thorium consists of only a single isotope 232
Th. Fissile isotopes 233
U
and 239
Pu are created by breeding process of 232
Th or 238
U based on the following reactions:
𝑇ℎ90232 + 𝑛0
1 → 𝑇ℎ90233 + 𝛾 → 𝑃𝑎91
233 + 𝛽− → 𝑈92233
𝑈92238 + 𝑛0
1 → 𝑈92239 + 𝛽− → 𝑁𝑝93
239 + 𝛽− → 𝑃𝑢94239
4.1 Uranium
Uranium was discovered in 1789 in uraninite. Later it was discovered that it was UO2.
Metallic uranium was prepared only in 1841. Radioactive properties of U were discovered in
1896 by Becquerel. The interest in uranium has been on the rise ever since. As a result of this
effort, fission chain reaction was discovered. These discoveries lead to the creation of atomic
and later hydrogen bomb, which were first built by the US. However, it did not take long for
the Soviet Union to destroy the US monopoly. Both countries also focused on peaceful
utilization of the nuclear energy. The first nuclear power plant in the world was started in
1954 in Obninsk, 110 km south-west from Moscow. The thermal output of this power plant
was 30 MW and the electric output was approximately 6 MW. From this time on nuclear
energy has been developed for peaceful purposes, focusing mainly on the production of
electricity. At present (beginning of June 2013) there are 436 operating nuclear blocks
throughout the world with a total power output of 372,686 MW; 69 power reactors are being
constructed.
4.1.1 Metallic uranium
4.1.1.1 Occurrence and uranium ores Uranium occurs in the Earth's crust in the total amount of approximately 2×10
-4 wt. %.
This corresponds approximately to the content of Sn and Pg and it is about 100x the content
of Ag. Relatively large amount of uranium is also contained in sea water, approximately
1×10-7
wt. %. Uranium is geologically widespread. This is due to its chemical and physical
properties, namely its valence, reactivity, solubility of many of its hexavalent compounds and
relatively high frequency. These properties, however, also explain why the concentration of
uranium in ores is so low. The largest part of uranium is located in sites with uranium
concentration of approximately 0.01 wt. %.
Uranium is contained in more than 100 minerals, whereas only about 10 of them have
any economic importance. Some of the important minerals are listed in table 4.1.
4.1.1.2 Uranium production
Processing of uranium ores
Most of the locations of uranium provides only ores with low content of uranium that
need to be enriched before they can be processed to metallic uranium. Uranium ores can be
enriched by classic methods, such as separation by heavy liquids, gravitation or flotation.
Minerals such as uraninite, carnotite and torbernite are enriched by gravitation or by heavy
liquids. The flotation method is used for example for uranothorite.
Radiometric enrichment of uranium ores can be used as an example. This process can
be realized on radiometric separators. This type of separator is schematically represented in
Fig. 4.1.
Table. 4.1 Selected minerals of uranium.
mineral chemical composition U content [wt. %]
uraninite UO2 45-85
pitchblende UO2,2-UO2,67 (U3O8) proměnlivý
karnotite K2(OU2)2(UO4).nH2O 55
samarskite (U, Y, Ca, Th, Fe) (Nb, Ta)2O6 8-16
brannerite (U, Y, Ca, Th)3Ti5O6 40
coffinite U(SiO4)x.(OH)4x 60
uranothorite (Th1-x.Ux.SiO4) 10
torbernite (CuO.2UO3.Pb2O5.8H2O)
Fig. 4.1 Scheme of radiometric separator.
1 - container
2 – feeding system
3 – belt feeder
4 – system for increasing the speed and stabilizing
5 – separating mechanism
6 – display
7 – scintillation counter
8 – main transporter
9 – light source
10 – concentrate
11 – waste
12 – electrical device
Leaching of uranium ores
The most common method of uranium ore processing consists of hydrometallurgical
treatment using acid or carbonate leaching.
Acid leaching
Acid leaching of uranium ores is the most important method of converting the ores to
a solution. It is used to process ores containing the following oxides: Th, Ti, Ta, Nb and
REM. If the ore contains tetravalent uranium, the acid leaching is done using an oxidant
(MnO2, HNO3, Fe3+
, VO2+
, etc.). Before the leaching itself, the ore is annealed in order to
ensure dissolution of organic substances, oxidization of sulphides, dissolution of carbonates
and removal of arsenic and sulphur. Salts are added to ores before leaching; uranium oxides
and salts create uranates.
Such modified ore is then processed in acid, usually H2SO4. Dissolution of uranium
can be done for instance in accordance with the following reaction:
U3O8 + 4 H2SO4 + MnO2 → 3 UO2SO4 + MnSO4 + 4 H2O
Uranium is obtained from this uranium solution, which also contains a whole range of
additional metals (Fe, Mn, Ni, Al, Cu,…), by precipitation, sorption or extraction. Uranium
precipitates from acid solutions in the form of hydroxides through neutralization of the
solutions using bases, ammonia or calcium oxides or magnesium oxide. Sometimes uranium
precipitates in the form of phosphate after the reduction of uranium to quadrivalent form
using iron or aluminium.
Sorption methods consists of capturing uranium using the ionic exchange process.
This process is based on differing ability of ions to undergo ionic exchange with ionic
exchangers (artificial resin). Sorption by anion-exchange resins (anex), where complete
uranium anion /UO2(SO4)3/4-
is absorbed, is also a common method. Extraction methods are
used to obtain uranium from low-concentration solutions. Extraction agents include TBF
(tributyl phosphate), MIBK (methyl isobutyl ketone) and a whole range of other organic
dissolvents. Optimum extraction environment for H2SO4 leaching includes phosphoric acid
esters and alkylamines.
Preparation of uranium compounds
The most important starting materials for the uranium industry include UO2, UF4 and
UF6. Fluorides UF4 and UF6 are required for the production of metallic uranium and for
isotope enrichment of natural uranium with isotope 235
U.
Production of UO2
Uranium dioxide can be produced by thermal decay of UO2(NO3)2 or (NH4)2U2O7.
Thermal decay of UO2(NO3)2 to UO3 is ended by the reduction of UO3 by hydrogen.
UO2(NO3)2 . 6 H2O 300°𝐶→ UO3 + 2 NO2 + ½ O2 + 6 H2O
UO3 + H2 → (500-700°C) UO2 + H2O
Precipitation of diuranate by ammonia takes places within the pH range of 7-8 in accordance
with the following reaction: 2 UO2(NO3)2 + 6 NH3 + 6 H2O → (NH4)2U2O7 + 4 NH4NO3 + 3 H2O
Decay occurs at a temperature of 300 °C:
(NH4)2U2O7 300°𝐶→ 2 UO3 + 2 NH3 + H2O
Production of UF4
The following reactions can apply here:
UO2 (s) + 4 HF (g) 550°𝐶→ UF4 (s) + 2 H2O (g) + Q
3 UO3 (s) + 6 NH3 (g) + 12 HF (g) 500−700°𝐶→ 3 UF4 (s) + 9 H2O (g) + N2 (g) + 4 NH3 (g)
3 UO3 (s) + 6 NH4HF2 (g) 700°𝐶→ 3 UF4 (s) + 9 H2O (g) + N2 (g) + 3 NH3 (g)
Production of UF6
The main used process is fluorination of UF4 by elementary fluorine, which consumes
the least fluorine and ensures the highest product quality. Other options include:
2 UO3 + 6 F2 → 2 UF6 + 3 O2
U3O8 + 9 F2 → 3 UF6 + 4 O2
Methods of direct fluorination of metallic uranium using ClF3, BrF3, BrF5 have also been
elaborated. The reaction can be represented as follows:
U + 2 BrF3 → UF6 + Br2
Br2 + 3 F2 → 2 BrF3
The production of UF6 from UF4 without the addition of fluorine is also interesting:
2 UF4 + O2 800°𝐶→ UF6 + UO2F2
UO2F2 + H2 → UO2 + 2 HF
UO2 + 4 HF → UF4 + 2 H2O
Enrichment of uranium
A whole range of processes have been developed for uranium enrichment. Some of
these technologies are commonly used in practice (diffusion, centrifuge), others are in the
stage of pilot operation, etc.
Gas diffusion
This process is based on different diffusion speeds of the isotope compound of UF6
through porous membranes. Gaseous UF6 is compressed by compressors and the
concentration is slightly shifted towards the lighter isotope on the low-pressure side of the
membrane due to a little higher diffusion speed of lighter uranium molecules. The scope of
the separation effect can be theoretically determined using molecular ratios of isotopes 238
U
and 235
U. The maximum theoretical value of the elementary separation coefficient is 1.0043.
In practice its value is usually 1.004.
In theory the diffusion enrichment process can be described based on a different
weight and consequently also different speed of the compound passing through the
membrane. The following relations apply here.
E = M. v2
2
M1 . v12 = M2 . v2
2 v1
v2=
M2
M1= α (separační součinitel)
The following relation applies for the enrichment of uranium isotopes in form of UF6:
αU= √
M238 UF6M235 UF6
=1,0043
Since the separation coefficient is so low, large volume flows need to be processed in
order to ensure the economy of the process. Since only negligible enrichment is achieved in
one separation step, it is necessary to repeat a large number of these diffusion processes,
whereas UF6 is compressed again by compressors in each separation step and the compression
heat needs to be removed. The production of uranium enriched to 3 % 235
U, which is used in
nuclear reactors, can be used as a typical examples – about 1000 – 1500 separation steps are
needed in this process. The diffusion equipment includes a large number of separation
elements of relatively large sizes connected in a series. Repeated compression of the operating
gas and dissipation of the compression heat result in high energy demands reaching up to
2300 – 2500 kWh/kg JSP.
Gas centrifuges
Gaseous UF6 is supplied to the operating area by a high-speed rotor. Centrifuge forces
cause lighter isotopes to concentrate in the area around the spinning, while heavier isotopes
are cumulated near the rotor wall and can be separated by fraction. This separation of isotopes
is significantly supported by thermally induced counter current.
The size of the separation factor depends on the difference of molecular weight, the
squared volume speed and the rotor length. It has been observed that separation factors 1.2 –
1.5 are usually achieved. Due to this relatively high separation factor it is necessary to repeat
each separation step 10 – 15 times in order to achieve 3 % uranium enrichment by isotope 235
U. Since the mass flow volume of one centrifuge is relatively low, it is necessary to operate
a large number of parallel-connected centrifuges in order to ensure an economic and efficient
flow. Unlike diffusion cascades, centrifuge cascades are characterized by a small number of
stages but a large number of small, parallel-connected separation elements. The rotor location
practically excludes all friction and the energy consumption is therefore several orders of
magnitude lower than for diffusion equipment.
Dynamic gas processes
Separation of uranium isotopes is based on the dependence of the centrifugal force in a
fast curved current on the weight. The separation effect ranges between the values achieved in
centrifuges and in diffusions; the number of separation stages connected in a series also
decreases proportionally. Similarly to the diffusion technology, the process gas (in this case a
compound of H2 + UF6, in order to achieve a high flow speed) is compressed by compressors
and the energy consumption is therefore quite high. Separation factors achieved in practice
range between 1.015 – 1.025, energy consumption is approximately 3300 kWh/kg JSP.
Preparation of metallic uranium
Uranium in metal form is usually prepared through reduction of UF4 or UO2 using
suitable reducing agents. Used reducing agents include mainly metals such as Ca, Li, Na, Ba,
Mg. Currently the most significant metal for this purpose is magnesium. Metallic uranium is
prepared from a melting of NaCl + CaCl + KUF5 by electrolysis. Reduction of UF4 in a fine
powder form is carried out in a steel bomb (reaction vessel) using high-purity Mg or Ca.
Ongoing reactions:
UF4 + 2 Mg → U + 2 MgF2 + 351 kJ
UF4 + 2 Ca → U + 2 CaF2 + 537 kJ
When magnesium is used as the reducing agent, the reaction takes place in a reaction
vessel with lining made of dolomite or burnt lime. Finely ground reaction components, which
are ground to prevent oxidation before the reaction, are filled to the reaction vessel, sealed by
a graphite lid and fire-proof lining. After closing the reaction vessel is put to a furnace at a
temperature of approximately 600 °C. The reaction heat is not high enough to completely
melt the magnesium structure (the melting temperature is 1263 °C). In order to completely
melt the bomb contents it is necessary to supply thermal energy from the outside or through
an exothermal reaction inside the reaction vessel:
KClO3 + 3 Mg → KCl + 3 MgO + 1858 kJ
KClO3 is added in the amount of approximately 1/7 of UF4. The melting product
consists of a yellowcake containing slag, as well as hydrogen. After the end of the process the
reaction vessel is removed from the furnace and is opened only after cooling. The equipment
scheme is shown in Fig. 4.2.
Due to the large amount of heat (released during the reaction) the content of the
reaction vessel, which is lined with sintered CaF2 and is equipped with additional electric
heating, melts completely during reduction by calcium. The whole apparatus can be
evacuated. After drying it is filled with the reaction mix, closed and put into the furnace. Once
the reaction occurs, the molten uranium drains to a mould under the reaction vessel. After the
reaction is completed, the reaction vessel is closed, evacuated and filled with argon. Smelted
yellowcake contains slag residues – CaF2 and oxides.
The disadvantage of reduction by magnesium when compared to reduction by calcium
is the relatively low reaction heat; however, it also has certain advantages. This technique
spread in North America. Reduction by magnesium has the following advantages: it is
cheaper, less magnesium is required to prepare the same amount of UF4 when compared
with Ca (up to 40 % less), it has a higher purity and the melting temperature of MgF2 is lower
than the melting temperature of CaF2. Disadvantages of reduction by magnesium include the
higher reaction temperature (higher than the evaporation temperature of Mg), which requires
the reduction to be performed in closed vessels.
Fig. 4.2 Reaction vessel for reduction of UF4 by Mg.
Uranium prepared using this technique needs to be remelted and refined from the slag,
gas and other impurities. Examples of uranium purities prepared using different techniques
are listed in table 4.2.
Uranium melting and casting Due to high reactivity of this metal with the atmosphere and a whole range of other
materials (crucibles, etc.), the melting process is relatively demanding. The effect of
atmosphere can be limited or almost eliminated by melting in vacuum, in the atmosphere of a
suitable inert gas or using covering molten salts. Crucibles are selected based on the allowed
amounts admixtures in the uranium.
Uranium dioxide and fluorite spar do not react with uranium up to a temperature of
1700 °C. However, these materials cannot be used for melting crucibles. CaF2 is too soft at
high temperatures, UO2 has unsatisfactory thermal conductivity. Relatively good results have
been achieved using BeO, ZrO2, ThO2 and graphite. When graphite is used, uranium carbide
is created, the reaction is slow approximately below 1600°C and the carbon that enters
uranium in the amount of several hundreds ppm is not too harmful when the uranium is
subsequently used in reactor technology.
Graphite represents a suitable material for the mould. Copper water-cooled moulds are
also acceptable and their use enables reaching a more fine-grained structure. Moulds based
on ceramic materials need to be heated to approximately 700 °C before casting to remove all
humidity. The uranium casting temperature depends on the cast size and ranges between
1320 – 1340°C.
4.1.2 Physical and mechanical properties of uranium
Pure uranium has the form of a silver-white metal. However, it quickly oxidizes in
contact with air, creating a golden-yellow film. As the oxidation process progresses, the color
of this film is getting darker and the metal turns black after 3 – 4 days. The created oxidation
layer does not protect the metal from further oxidation
Natural uranium contains three isotopes: 238
U (99.274 %), 235
U (0.720 %), 234
U (0.006
%). All three isotopes are alpha emitters with the following half-lives: 238
U (4.23×109 a),
235U
(8.5×108 a),
234U (2.7×10
5 a). Such long half-lives result in relatively low radioactivity of
uranium However, it is still necessary to implement appropriate safety measures. Uranium
occurs in three allotropic modifications. Basic characteristics are listed in table 4.3.
Table. 4.2 Impurities in the uranium produced by reduction of UF4.
preparation
technology
impurities [ppm]
C N O Al Ca Cr Cu Fe Mg Mn Ni Si
reduction by Mg 8-13 19 5 20 5 1 20 15 1 10 15
reduction by Mg,
melting in
graphite
240-
295
10-
16 6-16 10 20 2 5 30 0,5 2 15 15
reduction by Ca 20-
50
50-
200 5-10
50-
100
reduction by Ca,
melting in Al2O3 40 7 16 20 5 3 28 5 8 5 10
4.1.3 Powder metallurgy of uranium
The most important techniques of powder metallurgy include cold pressing, hot
pressing and extrusion pressing, additionally also HIP (Hot Isostatic Pressing) and CIP (Cold
Isostatic Pressing). The following methods come into question for the production of powder
uranium: Hydrogenation of uranium at 225°C (maximum reaction speed) and following
decay of hydrides at 400°C and a pressure of 1×10-3
Pa, as well as reduction of UO2 by
calcium or magnesium, reduction of UCl4 vapours using sodium or electrolysis of molten salts
(UF4 or KUF5).
Cold pressing followed by sintering results in solidification of powder uranium and
creation of the required shape. The pressing power depends on the size and shape of particles
and on the moulding size. The moulding density ranges between 10-11.5 g.cm-3
at the
pressing power of 500 MPa. The sintering of uranium mouldings is best performed in
vacuum, which ensures low partial pressures of oxygen and nitrogen. When compared to
other materials, uranium sintering is more difficult - it starts only after approximately 85% of
the melting temperature is reached (see Fig. 4.3).
Hot pressing can yield almost theoretical density of uranium. In order to sufficiently
utilize the material plasticity, pressing is carried out in the upper section of thermal stability of
the α phase. Again, protective atmospheres or vacuum are used for these processes. The
dependence of uranium density on the pressing temperature (at a pressing power of 300 MPa)
is shown in Fig. 4.4.
Table. 4.3 The crystal structure and density of uranium.
characteristic uranium uranium uranium
stability area [°C] < 667.7 667.7 – 774.8 774.8 – 1132.3
crystal structure orthorhombic tetragonal BCC
number of atoms of
the unit cell 4 30 2
lattice parameters
[10-10
m]
25°C 720°C 805°C
a0 = 2.8541 a0 = 10.759
a0 = 3.525 b0 = 5.5692 c0 = 5.656
c0 = 4.9563
density [g.cm-3
] 25°C: 19.04
720°C: 18.11 805°C: 18.06 625°C: 18.396
Obr. 4.4 Density of different sintered metals
depending on sintering temperature. Obr. 4.4 Density of the hot pressed uranium.
4.1.4 Uranium alloys
4.1.4.1 Uranium alpha alloys This type of alloys has been developed with the aim to decrease the volume growth –
swelling. It is assumed that homogeneous dispersion solidification by fine particles can lead
to the creation of additional centres for capturing fission gas products. These dispersion
particles can be obtained by adding a small amount of elements with low solubility in alpha
uranium, such as Si, Fe, Al. These alloys, possibly also Cr, are thermally processed by
hardening from beta phase (temperatures of 720 – 730°C) into water or oil and annealing for
several hours at upper temperatures of the alpha phase (500 – 600°C). This thermal
processing leads to:
finer grains and lower risk of striation,
creation of a fine second phase precipitate,
removal of the forming texture in the alpha phase.
The increased swelling resistance is contributed to the presence of a fine precipitate.
Alloy elements dissolve completely (Fe-Si) or partially (Al) by heating up to the beta-uranium
zone. Quenching creates an over-saturated solid solution, from which fine particles (UAl2,
U6Fe etc.) of a size of approximately 100×10-10
m are released during the following tempering
(500 – 600°C). Volume density of particles is approximately 1013
– 1015
in 1 cm3, which
corresponds to the density of bubble nuclei of fissile gas in irradiated uranium. It is assumed
that this fine precipitate has a beneficial effect on fine nucleation of bubbles or prevents their
movement and accumulation by preventing the movement of dislocations and grain
boundaries.
U-Al System
This binary system (Fig. 4.5) is typical for uranium alloys with intermetallic
compounds. At a temperature of 600°C two metals are practically insoluble together.
Aluminum breakdownability in gama uranium is 0,7; 0,5; 0,35; 0,25 % of the mass at
temperatures 1100, 1000, 900, 800°C. By an additive of 0,4 % Al of the mass it is possible to
stabilize the beta phase up to normal temperatures.
Fig. 4.5 Binary phase diagram U-Al.
U-Si System
In this binary system intermetallic phases exist: U3Si, U3Si2, USi, U2Si3, USi2, USi3. Si
breakability in alpha uranium is very little, in beta and gama uranium Si breakability moves
at a range of decimals of % of the mass. At temperature 930°C and concentration 3,78 % of
the mass. Si reaches during gradual cooling towards peritectic reaction between gama
uranium and U3Si2 and forms U3Si. This reaction is carried out very slowly.
U-Zr System
Alloyed Zr also improves the extensive stability of uranium during its thermal cycle.
From the binary diagram (Fig. 4.6) it is seen that in the U-Zr system there exists complete
breakability in a solid state between gama uranium and beta zircon, as opposed to the
breakability of zircon in alfa and beta uranium which is limited – max. 2,5 % of at.. In the
diagram there exists intermedial phase ε with wide areas of homogeneity. This phase forms a
peritectic reaction between alfa uranium and mixed gama crystals in temperatures 620°C and
63 % of at. (39,5 % of the mass) Zr.
U-Cr System The additive Cr to uranium causes a small change in the thermal expandability co-
efficient, electric resistance and thermal conductivity. This effectively causes a softening of
uranium granules. Small additives also increase uranium strength at 20 and 500°C. Chromium
with uranium forms a eutectic system without intermetallic compounds.
U-Mo System
The Mo content in alpha uranium alloys does not usually exceed 3 % of the mass.
After casting the alloy is left to slowly cool or to undergo isothermic thermal processing
(cooling from the beata phase area into salt liquid alloys with the upper area temperature of
alfa uranium). The structure is composed of fine segments of alfa uranium and the delta
phase, which represents the intermedial phase of uranium and molybden. For increasing the
Fig. 4.6 Binary phase diagramU-Zr.
radiation of these alloys other additives are added into it, such as Si, Al, Sn, causing the
formation of fine precipitates.
4.1.4.2 Uranium gamma alloys Nb, Mo, Ti and Zr cause stabilization of the gamma phase uranium. Metastable
gamma phase uranium created in U-Mo and U-Nb systems is the most passive phase for the
transformation, which is why these alloys are only usable in the metastable state. They will be
discussed in more details in the chapter dedicated to Ti and Zr.
Uranium gamma alloys are typically represented by the uranium alloy with 10 – 14 wt.
% of Mo, whereas the weight content of Mo above 5 % already allows maintaining the
metastable gamma phase uranium up to normal temperatures (assuming that the cooling speed
of the alloy is not too slow). Isotropic gamma phase uranium is not subject to radiation growth
and that is why there is no risk of striation or radiation creep. The main reason of size changes
in these alloys is gas swelling. All uranium gamma alloys with solid gamma solution
containing Mo, Nb, Ti and Zr usually undergo homogenizing annealing after their production.
The annealing process lasts 24 hours and is done at a temperature of approximately 900 °C.
Another group of γ uranium alloys consists of alloys obtained by regeneration of
uranium irradiated in fast radiators. When this uranium is processed by melting under
oxidation slag, volatile elements and REM are removed almost completely, whereas elements
Mo, Ru, Rh, Pa, Nb, Te, i.e. the fission products, remain in the uranium and zirconium is
separated only partially. Elements that remain in uranium are called "fissium" and uranium
alloys with these elements are called uranium fissium alloys.
U-Mo gamma alloys
By cooling U-Mo alloys it is possible to reach a state where the gamma modification
of uranium is kept up to normal temperatures. This metastable modification brings out
favourable properties from the point of view of its use in nuclear reactors. After
corresponding annealing it transforms to a phase, which corresponds to an equilibrium
diagram. This transformation method and its products depend on the content of Mo, the
temperature and cooling speed. In gradual hardening an important role is played by the
hardening temperatures at the individual degrees and periods, after which the material is
exposed at this temperature
U-Nb alloys
Alloys from this system are similar in several directions to alloys on the basis of U-
Mo. With a large content of niobium (10, 20 % of the mass and more) the gama phase is
metastabile. According to several writers 6 % of the mass. Nb is enough to martensitically
transform alloy U-Nb, moving it below normal temperatures. In binary U-Nb stable
intermetallic compounds do not occur. Nb breakability in beta and alfa uranium is very small
(under 1,5 % of at.).
U-„fissium“ alloys
According to this concept we understand an alloy which is formed after longer
radiation of neutron fission material. Alloy compounds are independent from the level of the
burning out of fission material, while there is also the possibility of the slag purification (of
oxides) during pyrometallurgical treatment.
According to input material the formed casting contains U (or Pu) as the main
compound and and a group of elements formed by fission, whose oxides are less stable than
UO2. They are Zr, Mo, Tc, Ru, Rh, Pa. The typical chemical composition of U-„fissium“
alloys are given in table. 4.4.
4.1.5 Preparation of uranium alloys
4.1.5.1 The preparation of uranium alloys by melting
Induction melting
The basic problem of preparing alloys in induction furnaces is the selection of a
suitable melting crucible, which cannot react to liquid metal, it has to have good thermal
conductivity, and resistance against thermal impact. On a laboratory scale oxidic materials
(MgO, CaO, ZrO2 a UO2) are used very well.
Arc melting
In most cases this method of preparation is carried out using wolfram electrodes in
an atmosphere 75 % He + 25 % Ar. In comparison with melting in induction furnaces, the
lack of this method of preparation is a worsened homogeneity of liquid melt due to its bad
movement and the bad removal of impurities as a result of limited liquid melt fluidity.
4.1.6 Uranium alloys – ceramic fuels
With the development of high-power energy reactors, ever increasing requirements are
placed on fissile materials, such as high operating temperature, high irradiation, dimensional
stability and the maximum possible safety. Uranium or its alloys were not able to meet these
requirements. This is due to the fact that metallic uranium undergoes phase transformations at
temperatures exceeding 600 °C leading to significant deterioration of its mechanical
properties and volume changes caused by recrystallization. It is therefore clear that alloys can
be used for fuel elements only at temperatures not exceeding 600 °C. This shortcoming can be
removed using the following processes:
thermal treatment and alloying,
application of ceramic fuels.
Ceramic fuels have a whole range of advantages when compared to metal fuels, such
as:
significant thermal and radiation resistance,
high melting temperatures,
phase stability in a wide interval of temperatures (provided that suitable processing is
used),
weak dependence of the material strength on temperature,
relatively small expansion,
dimensional stability under the conditions of the required reactor output,
higher irradiation depth of the fissile component in comparison with metallic uranium.
Table. 4.4 Chemical composition of U-fissium alloy.
element conctent [wt. %]
Zr 0.1 – 0.2
Mo 1.6 – 3.4
Tc 0.5 – 1.0
Ru 1.2 – 2.6
Rh 0.2 – 0.5
Pa 0.1
Disadvantages of ceramic fuels include mainly:
high brittleness and low resistance to temperature shocks,
necessity to use special coating materials that are significantly more expensive than Al
or Mg alloys,
lower thermal conductivity (especially at higher temperatures) in comparison with
metallic uranium,
hard formability of ceramics.
Production of uranium dioxide UO2
In practice, uranium dioxide can be produced using three methods.
1. Thermal decay of uranyl nitrate followed by reduction by hydrogen, split ammonia or
carbon monoxide:
UO2(NO3)2 . x H2O (600°𝐶)→ UO3 + NO + NO2 + O2 + x H2O,
UO3 + H2 (600−1000°𝐶)→ UO2 + H2O.
2. Precipitation by ammonia from water solutions of uranyl nitrate while producing
ammonium diuranate, which is thermally decayed. This processes produces U3O8
which is then reduced to UO2 as in the first example.
3. Precipitation by hydrogen peroxide from water solutions of uranyl nitrate while
producing hydrated uranium peroxide, which is again thermally decayed and reduced
to UO2.
For an overview of the UO2 production see Fig. 4.7. Particles differ in shape and grain
size, their specific surface, agglomeration degree, porosity, etc. Within a certain production
technology, properties of powder UO2 can be modified by changing the decay or reduction
temperature. Properties of powder UO2 can be changed within certain limits by additional
processing, such as milling and alternating oxidation and reduction.
Fig. 4.7 Scheme of UO2 production by different methods.
Properties of uranium dioxide UO2
Selected physical properties of uranium dioxide are listed in table 4.5 and the most
important mechanical properties are included in table 4.6.
The listed values characterize uranium dioxide as hard and brittle ceramic material.
Elasticity constants decrease with increasing temperature, porosity and the O:U ratio. The
bending strength decreases with increasing grain size and porosity. Increasing temperature
causes an increase of the bending strength for fine porosity and small grains; for gross
porosity and large grain size increasing temperature causes a decrease of the bending strength.
The transition temperature of malleable – brittle fracture is 1250°C for UO2. Hardness
increases with the O:U ratio.
Mechanical properties of UO2 therefore heavily depend on structural factors and these
factors change depending on preparation methods and processing temperatures. Another
important factor influencing the properties of UO2 is the so-called oxygen index. It is the
numerical deviation from the oxygen index that should equal 2. Precise designation of oxides
is usually UO2+-x´. Changes of the oxygen index influence the thermal conductivity of fuel,
melting temperature of fuel, transition temperature of brittle - malleable fracture, swelling,
creep, etc. The effect of the oxygen index of UO2 on its behaviour under tension is depicted in
Fig. 4.8. During creep tests samples of sintered UO2,06 and UO2,16 deformed already at 800°C,
whereas the corresponding temperature for UO2 is 1600°C.
Table. 4.5 Physical properties of UO2.
attribute temperature [°C] value unit
density 25 10.968 g.cm-3
melting temperature 2880 °C
lattice parameter (BCC) 25 5.4682.10-10
m
atomic concentration of U 255 2.45.1022
atom.cm-3
thermal conductivity
100 0.07524
W.K-1
600 0.033
1000 0.025
specific heat
100 263.71
J.Kg.K-1
500 309.76
1000 326.5
Table. 4.6 Mechanical properties of UO2.
attribute hodnota jednotka
elastic modulus 233670 N.mm-2
shear modulus 85840 N.mm-2
flexural strength 84,2 N.mm-2
compressive strength 480 N.mm-2
hardness HV 580
Preparation of compact UO2
Uranium dioxide is used in two forms for fuel elements. The first form consists of
compact bodies, tablets prepared by pressing and sintering or hot-pressing. The second used
form is powder condensed by vibration or rotation forging. Pressing and sintering are the most
elaborated preparation procedures of compact UO2. Pressing is done at a pressure ranging
from 4 – 7×102 MPa with the use of plasticizers and binding materials. The sintering process
is carried out at temperatures ranging between 1600 – 1750°C for 1 – 4 hours. The process
also includes pre-annealing at 600 – 800°C if binding agents are used. Pores are reduced
during sintering through diffusion of vacancies along grain boundaries, whereas the speed of
the process is determined by the concentration gradient of these vacancies. Sintering is also
affected by UO2 stoichiometry and is accelerated by even slight increase of oxygen in sulphur
dioxide to the density of UO2,02, whereas any further increase of the oxygen content does not
significantly affect the sintering process. This effect is usually explained by the increase of the
O2-
diffusion coefficient when compared to U self-diffusion. The sintering process is also
affected by the atmosphere. High density UO2 is obtained in hydrogen; worse results are
achieved in N2 and Ar. UO2 with high density can be achieved in vacuum. The sintering
method itself has been designed so that it can be applied even without sintering additives and
yields high density values.
4.2 Plutonium
Elements of the periodic system ending with proton number 92, i.e. uranium, occur in
the nature. The only element of the artificially prepared elements, such as Np, Pu, Am, that
plays an important role in the industry is plutonium, which can be used as the energy source
in nuclear reactors or it can be used for military purposes. As has already been stated above,
plutonium practically does not occur in the nature, only traces of isotope 239
Pu were found in
uranium ores (in the amount lower than 5×10-2
% of the uranium content). Plutonium is
created by practically the same reaction during uranium irradiation in nuclear reactors:
U (n, γ) → U92239 (β, 23,5 min) →92
238 Np 94239 (β, 2,33 d) → Pu94
239
Fig. 4.8 Creep curves of sintered UO2.
Plutonium occurs namely as isotope 239
Pu, which is an emitter with half-life of
24300 years. 3 neutrons are created by the fission of plutonium atom and that is why this
element is a suitable fissile material in thermal and fast breeder nuclear reactors. Plutonium is
usually not used in its pure metallic form due to its numerous phase transformations. On the
other hand, plutonium alloys can be used both in solid and in liquid state. Another possible
utilization of plutonium as a fuel material in nuclear reactors includes ceramic materials, such
as UO2-PuO2, carbides, etc.
4.2.1 Plutonium sources
4.2.1.1 Thermal reactors Currently operated thermal reactors use the fission process of
235U, which however
forms only a small part of uranium occurring in the nature. The main component of natural
uranium is isotope 238
U, which – after being exposed to radiation in the reactor and due to
neutron capture and decay – is transformed to 239
U, which then decays to 239
Pu. Due to the
fact that thermal reactors contain both 235
U and 238
U, the production of plutonium is an
inevitable consequence of the production of electricity in thermal reactors.
Plutonium behaves similarly as 235
U and it can be used as fissile material since its
fission starts spontaneously after it is produced in the reactor. Since plutonium must be first
produced in order to be later combusted, the production of Pu from 238
U must exceed its
consumption. Typical thermal reactor balance in terms of plutonium production and
consumption is provided in table 4.7. The balance is expressed in so-called equivalent
kilograms of 239
Pu (Ekg), which represent the weight of 239
Pu, whose fissile value equals the
compound of Pu isotopes created in the reactor.
4.2.1.2 Fast reactors In regards to the nuclear properties of fuel, in this case plutonium, favorable results are
achieved, which are not possible to reach in thermal reactors. The complete balance of
plutonium in the fuel cycle of quick reactors will be the result of its combustion and formation
in the coating of active zones by capturing neutrons at 238
U. The balance of plutonium in a
fast reactor is more complicated than that for a thermal reactor, because besides active zones
there exists also a coating, which does not exist for a thermal reactor. The plutonium balance
in a fast reactor is shown in Table. 4.8. It deals with a fast reactor cooled with sodium, in
which oxides are used in the fuel.
Table. 4.7 Production of 239
Pu attributable to GW.rok of produced electricity (1 rteactor 1000 MWe, in
operation 365 days) [Ekg Pu].
amount reactor
batch 0
production +710 PWR
consumtion -440
net production +270
+617 Magnox
+493 CANDU
+173 AGR
4.2.2 Plutonium production
4.2.2.1 Basic methods of reprocessing irradiated fuel Plutonium in plutonium-producing nuclear reactors needs to be periodically separated
from uranium and fission products. The irradiation grade in energy reactors working with
natural uranium is approximately the same. Reactors working with highly enriched plutonium
allow higher irradiation grade; however, substantial part of the initial fissile material still
remains in the irradiated fuel. The fission product always needs to be completely removed
from uranium and plutonium. Due to the high reactivity of irradiated fuel elements, the
processing can start only after the fuel elements are cooled down and when the majority of
short-time fission products are dead and the activity is multiple times lower. All operations
with irradiated fuel must be still performed using efficient protection against radiation and
toxic substances.
From this point of view, the wet process appears to be most relevant for processing
irradiated fuels. However, in certain situations, the so-called dry process might be more
suitable.
Wet procedure
This technology is founded on the dissolvability of radiated fuel. The treatment of
burnt-out and cooled fuel begins with the removal of the coating material, which can cause
problems during other procedures and what´s more it leads to a greater volume of highly-
active solutions. After coat removal the fuel is dissolved in boiling nitric acid. The dissolving
of uranium in HNO3 is carried out during the formation of uranylnitrates and a mixture of
nitrate oxides. Dissolution in diluted HNO3 is carried out according to the equation:
U + 4 HNO3 → UO2(NO3)2 + 2 NO2 + 2 H2O
in concentrated acids:
U + 8 HNO3 → UO2(NO3)2 + 6 NO2 + 4 H2O
Acquiring plutonium from a solution
Plutonium is acquired and purified from fission products by bearer precipitates or
extractions using organic solvents. In the first case it is formed for distributing the products of
uranium fission and the plutonium of such conditions, in which both metals occur in various
valences. One of them can then be transferred to the precipitate, while the other stays in the
Table. 4.8 Plutonium balance of fast reactor.
burning in the core balance Production in fission breeder
batch = 1936 = 1936 0 = batch
production = 558 455 = production
total = 2494 455 = total
burning = 789 35 = burning
output = 1705 = 1705
231 421 = output
Net consumtion outside fission
breeder: 421 Maximum production in dision
breeder:
-231
231 Ekg/G.We.rok 190 Ekg/G.We.rok 421 Ekg/We.rok
Net Pt production is 190 Ekg/G.We.rok
solution. In separating uranium and plutonium it is possible to achieve a precipitation of
uranium or plutonium.
4.2.2.2 Metallic plutonium production Input materials for the production of metallic plutonium include oxides or halo alloys
(fluorides or chlorides). Due to high reactivity of plutonium to oxygen, this metal cannot be
prepared by reduction of oxides by hydrogen. Reduction by carbon at high temperatures in
vacuum does create metallic plutonium; however, it is strongly contaminated by carbon in the
form of carbides. The following preparation methods of metallic plutonium have been
elaborated:
metallothermic reduction of plutonium fluorides or chlorides,
metallothermic reduction of oxides,
electrolytic decay of plutonium salts from molten salts.
The most commonly used method from the list above is the metallothermic reduction
of fluorides, since hydroscopicity of chlorides has a negative impact and a fine plutonium
powder with increased oxygen content is created during the reduction of PuO2.
The basic advantage of plutonium fluorides is the fact that they are not hydroscopic in
contact with the air and their production is relatively simple. However, PuF5 and PuF6
fluorides are volatile to a certain degree and that is why they are reduced to PuF4. Even
though a whole range of metals can be used as a reducing agent (Ba, Mg, Al), the most
frequently used agent is calcium, which is used in many metallothermic processes.
During the reduction of plutonium compounds by calcium the compounds are mixed
with the excessive calcium and the mixture is heat up to the start of reduction. The equipment
for reduction of a plutonium soil batch ranging between 1 – 500 g differs only in size. It
consists of a steel bomb or an external container, preliminary annealed fire-proof crucible for
the charge and high-frequency source with a furnace for heating the bomb.
Plutonium produced from PuF4 by the metallothermic process using calcium usually
contains a little less or the same amount of admixtures contained in the input materials. The
content of volatile metallic admixtures usually decreases. Fire-proof crucibles produced by
caking of specially cleaned MgO or CaF2 significantly contaminate the reduced metal. The
purity of thermally prepared plutonium using calcium reaches 99.96 %; the summary content
of metallic admixtures ranges between 0.03 – 0.04 %.
4.2.2.3 Plutonium properties Pure plutonium is a metal with a density of 19.81 g.cm
-3 and a melting temperature of
640 °C. There is a whole range of known isotopes with nucleon numbers 232 – 246, all of
which are radioactive and usually alpha emitters. The only isotope with practical meaning for
nuclear technology is 239
Pu. If this isotope is exposed to strong neutron flow, heavier isotopes
and other transplutonium elements are created.
Metallic plutonium undergoes six allotropic transformations in the temperature
interval ranging from a room temperature to the melting temperature (see table 4.9). In terms
of linear thermal expansion, the α plutonium has the highest coefficient (46.85 ± 0.05).10-6
.K-
1; δ modifications have a negative expansion coefficient (-8.6 ± 0.3).10
-6.K
-1a δ´. The thermal
conductivity of plutonium is very low, even in comparison with uranium, and reaches only
about 5.44 W.m-1
K-1
. It also shows the weak temperature dependence.
4.2.2.4 Processing of plutonium and its alloys The main problems in plutonium processing are related to its toxicity, risk of oxidation
and its allotropic transformation, which are also accompanied by volume changes. This metal
is processed by melting and casting, forming and powder metallurgy.
Melting and casting
Plutonium has good casting properties – it has lower melting temperature, good
fluidity and high density. The volume changes during its solidification are relatively small.
However, significant volume changes during phase transformation complicate the production
of more complex casts due to the risk of cast failure. These negative properties can be
therefore partially eliminated by the use of suitable demountable casting moulds, which are
taken apart at the temperature of γ or β modifications. Melting and casting must therefore be
performed in high vacuum; the crucible is made of MgO, CaO, CaF2. Wolfram or tantalum
crucibles can also be used under certain conditions. Moulds for casting are usually made of
MgO; another suitable variant includes water-cooled copper moulds.
Powder metallurgy
Plutonium powder is prepared by the process of hydrogenization and Pu hydride
decay. Pieces of plutonium with a weight of approximately 0.5 g and total batch of 10 g are
exposed to the effect of hydrogen at a temperature of 150 °C. After a short period of time the
pieces of plutonium decay to powder PuH2,7-3, which is then ground to achieve even particle
size. This is followed by decay of hydride at a temperature of 420 °C in vacuum. The
obtained powder is then pressed in steel matrices covered with colloidal graphite at a pressure
of 1120 MPa. This process yields a moulding with a density of 16.8 g.cm-3
. Caking of
mouldings is performed in silica pipes at a temperature of 550 °C. Another possible variant is
hot-pressing in steel matrices at a temperature of 380 °C and a pressure of 380 MPa. Again,
the matrix needs to be protected by a film of colloidal graphite in order to prevent the creation
of low-temperature plutonium eutectic with iron and nickel.
4.2.2.5 Plutonium alloys The development of plutonium alloys suitable for use as fuels in nuclear reactors was
motivated, similarly as for uranium, by the effort to increase the corrosion resistance,
structural characteristics and unsatisfactory physical and mechanical properties of unalloyed
metal. Elements with low absorption cross section are suitable for alloying in thermal reactors
(e.g., Al, Mg, Zr and Be). When used in fast breeder reactors, higher content of Zu or U with
10 – 30 % Pu is required. Other studied systems including Pu – Fe, Pu – Cr, Pu – Co, Pu – Ni,
Pu – Mn, which form low-melting eutectics, could be used as liquid metal fuels in
homogenous reactors.
Table. 4.9 Phase transformations in metallic Pu.
phase stability [°C] density [g.cm-3
] lattice
α -186 až +122 19.81 monoclinic
β 122 - 206 17.82 body centered monoclinic
γ 206 - 319 17.14 orthorhombic
δ 319 - 451 15.92 FCC
δ´ 451 - 480 16.0 body centered tetragonal
ε 480 - 640 16.48 BCC
Pu – Al Alloys
Hardened Pu – Al alloys with a content 2 – 13% Al of at. are a mono-phase structure δ
of plutonium modification. In this conditions δ is a stable modification up to above 600°C, as
opposed to in aluminium, plutonium is almost non-disolvible at all temperature intervals.
With the structure of plutonium alloys with a higher content of Al (above 80 % of at.) there is
formed a Al matrix with a finely dispersed intermetallic phase of PuAl4.
Pu – U Alloys
As a fuel these alloys have special significance - They are also suitable for fast
reactors. With a content up to 25 at. % of U, the continuous phase of transformable compound
β´ is stabilized. With a content of uranium ranging from 3 – 17 at % in room temperature the
alloy has a mono-phase structure corresponding to the β modification of plutonium. In order
to improve the properties of these alloys, other compounds are still added, most often Mo.
4.2.2.6 Plutonium compounds
Plutonium oxides
Oxidic plutonium fuel has a number of advantages over metallic fuel. UO2 with a
content of Pu 1 – 5 % can be used in thermal reactors (we already know that 238
U changes to 239
Pu), assuming the content is 10 % PuO2 in UO2, then it is possible to use it as well in fast
reactors. Plutonium oxides with their own compounds and properties are distingusihed from
uranium oxides very much. Plutonium is made up of three oxides: PuO2, α Pu2O3, β Pu2O3.
Solid solutions of UO2 – PuO2 In this system there exists a never-ending number of solid solutions. The additive
PuO2 to UO2 stabilizes the cubic structure of UO2. With contents of PuO2 above 40 % the
cubic structure of UO2 is preserved during heating up to high air temperatures. The powder
mixtures of the UO2 – PuO2 system (with a content of 30 % PuO2) are the most suitable for
fast reactors. The preparation of the UO2 and PuO2 powder mixture is possible to carry out
using the mechanical mixing of oxides or by using common precipitation from the aqueous
solutions of the nitrates of both metals.
Plutonium carbides
In the Pu – C system, there exist PuC and Pu2C3 carbides. A characteristic feature of
PuC is its proportionally low melting temperature (1654°C), when as a result of the peritectic
reaction there is in the formation a liquid melt an enriched Pu and through carbide Pu2C3.
Another speciality of PuC is the proportionally wide area of the existence of monophase
structures at the limits of the concentration of carbons from 43.5 to 48 at. %. If the content of
carbon is above 48 at. %, it leads to the formation of Pu2C3.
4.3 Thorium
Unlike uranium, thorium does not belong to fissile materials. However, during its
irradiation by slow neutrons, nuclear reactions occur and lead to the creation of fissile isotope 233
U with long half-life. Thorium is therefore referred to as a fertile material – isotope 233
U is
created by the following reaction:
Th (n, γ) → Th (−β, 23 min) → Pa (−β, 2791233
90233
90232 min ) → U92
233
Due to the fact that thorium reserves significantly exceed uranium reserves, it is
reasonable to expect that thorium will be used much more in the future. However, a
significant disadvantage of thorium in this breeding process is the high required enrichment of
fuel by isotope 235
U or 239
Pu. Another difficulties are summarized by the following list:
low speed of fuel creation in thorium fuel cycle 𝑇ℎ → 𝑈92232
90232 ,
increased radioactivity of 233
U in comparison with 235
U and 239
Pu, which is caused by
accumulation of isotope 232
U and its daughter isotopes,
fuel poisoning by 232
Pa, which has a large cross section for neutron capture.
4.3.1 Occurrence, ores and their enrichment
In terms of occurrence in nature, thorium holds 35th place among other elements. As
far as geochemical properties are concerned, this metals has a lot in common with rare-earth
metals (REM), Zr and U, whereas their beds are usually complex. This metal occurs in rocks
in scattered state in the form of the following minerals: monazite, ortite, zircon, xenotime,
thorianite, etc. There are about 120 known minerals containing thorium.
4.3.2 Thorium production
Monazite concentrates are processed in order to obtain compounds of thorium, REM,
uranium and phosphorus. The production technology includes the following processes:
1. Decay of a concentrate to thorium and REM compounds that are soluble in inorganic
acids.
2. Conversion of thorium and REM into a solution.
3. Separation of thorium and REM from phosphorus.
4. Split of thorium and REM.
A whole range of methods have been elaborated for the first operation:
processing by concentrated sulphur acid – so-called sulphate method,
processing by a concentrated solution of sodium hydroxide – basic method,
melting with sodium hydroxide,
caking with soda ash,
melting with fluorosilicates,
chlorination,
phosphate reduction at high temperatures.
The biggest challenge in the thorium production is its separation from REM, since
their properties are very similar. These metals are separated based on small differences in
chemical properties of certain compounds. However, these methods cannot yield the purity
required for nuclear technology. That is why extraction and sorption methods are used in the
next step.
Sulphate methode
This method is founded on the breakdown of monazite concentrates by sulphuric acid
at a temperature 200 – 230°C. The following reaction is carried out:
Th3(PO4)4 + 6 H2SO4 → 3 Th(SO4)2 + 4 H3PO4
2 (REM)PO4 + 3 H2SO4 → (REM)2(SO4)3 + 2 H3PO4
ThSiO4 + 2 H2SO4 → Th(SO4)2 + SiO2 + 2 H2O
SiO2 . x H2O + H2SO4 → SiO2 + x H2O . H2SO4
For this process it is transferred into the solution together with thorium and REM as
well as uranium. The ilmenite mixture partially or completely dissolves in the formation of
dissovable titanium and ferrite sulphates. Only mixtures of zircon, rutile, silica, cassiterite and
part of non-dissolvable monazite cannot be transferred into the solution
The result of the process is a homogenous substance, which is further treated with
water. For separating thorium and REM from the solution of sulphuric and phosphoreous acid
many methods exist, the most often method used is the selective precipitation of thoric
phosphates with a certain concentration of hydrogen ions (the method of gradual
neutralization). This principle consists of a different pH solution, in which thoric phosphates
and REM are separated. For removing thorium from acidic solutions the solution has to be
neutralized to pH = 1 and heating until boiling. At the same time 99 % Th is separated in the
form of a slightly dissolvable precipitate ThP2O7 . 2 H2O (thoric pyrophosphate) according to
the reaction:
Th4+
+ 2 H2PO4- → ThP2O7 + H2O + 2 H
+
At the end of neutralization the complete volume of the solution has to be such that the
concentration of REM in the solution does not exceed 2 %. Under these conditions there are
mostly sulphates of REM in the solution and together with thorium they shrink only around
cca 5 – 7 %. Therefore the ratio of Th : REM in the thoric concentrate is approximately 1:1.
After the separation of thoric precipitates the solution is further neutralized by ammonia up to
pH = 2,3, while a great part of REM shrinks in the form of acidic phosphates. After filtration
they are then processed into REM. The filtrate, which contains a certain amount of REM and
all uraniums, are further neutralized by ammonia at pH = 6, while various types of uranium
and REM left in the solution are transferred into the precipitate. The content of uranium in
the precipitate is about 1 %. The precipitate is further processed by the uranium released
in HNO3 with the resultant extraction of TBF.
4.3.2.1 Preparation of pure thorium compounds Due to the fact that the prepared concentrates contain not only thorium but also a
substantial amount of REM and other admixtures, it is necessary to include a cleaning
process. Cleaning methods of thorium compounds after monazite processing can be divided
into two groups:
1. Methods of selective precipitation and solution.
2. Methods of selective extraction by organic solvents (significant application in
practice).
Methods of selective precipitation and solution
This group of methods includes the following procedures.
Fractional crystallization
This procedure uses different basicity of thorium and REM. Ammonia solution is
added in small amounts to a hot solution of these metals in order to achieve even
concentration of hydrogen ions. This process yields Th(OH)4, containing only about 1 - 2 %
of REM.
Selective precipitation of thorium compounds
Compounds with lower solubility than corresponding REM salts can be processed for
instance using the different solubility of thorium sulphates and REM. Th(SO4)2 . 9 H2O
changes its solubility from 40 g.l-1
at 45°C to 8 g.l-1
at 0°C. This value is significantly lower
than the solubility of all REM sulphates, whereas the solubility of REM sulphates increases
with decreasing temperature.
Selective dissolution:
This method is based on the creation of complex thorium compounds with oxalates
and carbonates of alkali metals and ammonia. Solubility of thorium oxalate in water and acid
solutions is lower than the solubility of REM oxalates. Precipitation of REM oxalates at
different pH can be used to separate a certain part of REM from thorium (Fig. 4.9).
4.3.2.2 Preparation of metallic thorium The reduction of compounds to metal is performed by applying metallothermic
methods or electrolysis of molten salts. In both cases the obtained thorium has a solid form,
i.e. powder or sponge. This thorium is then processed to a compact material by arc melting or
through powder metallurgy.
Metallothermic methods
ThO2 reduction by calcium
This is probably the most frequently used method. The reduction is carried out in an
inert atmosphere in the equipment depicted in Fig. 4.10. A crucible lined a molybdenum sheet
or CaO is filled, put into the furnace and closed. After air is drained from the device, it is
filled with argon and the furnace is slowly heated to 1000 - 1100 °C. After a certain period of
time at the temperature, the reaction vessel is removed from the furnace and cooled. Ground
reduction products are processed by water and diluted HC1 in order to remove slags and
calcium residues. Thorium powder is then concentrated, cleared of iron admixtures and
surface oxides using HNO3, washed by water and dried in vacuum.
80% of REM and other admixtures are transferred to the reduced metal thorium. That
is why the content of these admixtures in the input thorium oxide must not exceed the allowed
amount for use of thorium in nuclear technology.
Fig. 4.9 Solubility of thorium oxalates and KVZ at various concentrations of H2SO4.
1 – reaction vessel
2 – steel crucible
3 – feedstock
4 – cover, connection to a vacuum pump and Ar
5 – insulation
6 – furnace
Fig. 4.10 Aparatus for reduction of ThO2 by Ca.
Iodide refinement
This technology is used to completely clean thorium of non-metallic admixtures, such
as oxygen, carbon, nitrogen and hydrogen. Admixtures consisting of volatile iodides cannot
be removed. The process of iodide dissociation is analogical to the process described for
other metals (Ti, Zr). The decay is represented by the following reaction.
Th (technické) + 2 I2 (g) 455−485°𝐶→ ThI4 (g)
1200−1300°𝐶→ Th (čisté) + 2 I2 (g)
4.3.3 Thorium properties
Thorium is a soft metal with the appearance of steel. However, its hardness is similar
to silver. In term of physical properties, this metal has a whole range of advantages in
comparison with uranium – namely high thermal conductivity and low thermal expansion
coefficient. Its specific weight is significantly lower. Its main advantages include the cubic
crystallic structure, thanks to which cyclic thermal stress or radioactive radiation do not cause
any significant dimensional and shape changes that are characteristic for uranium. Its melting
temperature is approximately 1750 °C. Thorium has a strong affinity to carbon, nitrogen and
hydrogen. These elements strongly affect its properties.
4.3.4 Thorium alloys
Pure thorium does provide sufficient strength properties at increased temperatures and
that is why it is alloyed with other suitable elements. The selection of suitable alloys needs to
take into consideration its absorption cross section for thermal neutrons.
This metal has a strong tendency to create intermetallic phases and only certain metals
are dissolved in larger amounts. The most important fissile materials are Th – U alloys.
Thorium alloys are produced by melting in a crucible or an arc furnace, alternatively by
powder metallurgy. The most important alloying elements include U, Pu, Al, C, Be, V, Si,
Mo, Ni, Sn, Zr.
Thorium – uranium
The dissolvability of uranium in thorium increases with a growing temperature (Fig.
4.11). At room temperature it dissolves in thorium about 1 wt. % of of U and at 1250°C about
7,3 wt. % of U. Alloys of thorium with uranium, which are used as combined fission and
multiple materials, contain 2 – 10 % of the mass, which is a highly-enriched uranium. That is
why these alloys are bi-phase; for casting the uranium is excluded as a fine dispersion phase
in the thoric matrix.
Thorium – plutonium
The dissolvability of plutonium in thorium in a solid state is considerable (Fig. 4.12).
The formation phase of Th6Pu13 is dependent on the speed of alloy cooling. Normally cooled
alloys with a corresponding composition contain α thorium and ε plutonium. Only annealing
at a temperature around 550°C enables the formation of the intermetallic Th6Pu13 phase.
There alloys have good radiation stability up to a temperature of 450°C. From radiation at
450°C to burning out at 1,9 % of at. it causes swelling of thorium alloys and of 5 of the mass.
% Pu increases in volume by 0,8 % for each atomic percent of burn-out.
Thoric oxides
For preparing compact ThO2 it is possible to use standard techniques, such as: cold
pressing, isostatic pressing, alloying, hot pressing, etc. The most often used in practice is cold
pressing with the resultant in air alloying. In comparison with UO2 this process is not
dependent on an oven atmosphere. While alloying in the air, in acid or hydrogen, with an inert
gas or vacuum practically the same density was reached.
Thorium carbides
These carbides are prepared for the same reasons are uranium carbides. They have
better thermal conductivity than ThO2, but basically worse resistance against the effects of
water and air. From binary diagram Th – C it comes that Th is made up of two carbides, and
they are ThC and ThC2. Thorium carbide can be prepared using the same method of uranium
carbide, by using the basic preparation techniques for ThC and ThC2 but there is the reduction
of ThO2 by carbon. When heating briquettes alloyed from ThO2 and carbon at a temperature
of 2170°C in an argon or hydrogen atmosphere there is formed ThC2 according to the
reaction:
ThO2 + 4 C 2170°𝐶→ ThC2 + 2 CO (-F2170°C = 12,54 kJ)
4.4 Dispersion nuclear fuels
Favourable properties of ceramic fuels laid the foundations of the development of
these types of fuels, where the fissile components is dispersed in a suitable inactive matrix.
Each particle can be considered a micro-element, where the matrix assumes the role of a film
preventing volume changes, leaks of fissile products and corrosion of fissile materials.
Dispersion fuels therefore combine the positive properties of ceramic fissile material and high
strength characteristics and physical properties of the matrix.
Basic requirements on dispersion fissile material:
high content of uranium or plutonium,
small absorption cross section for neutrons in accompanying elements,
good compatibility with matrix material,
good technological properties.
Fig. 4.11 Part of Binary system Th-U. Fig. 4.12 Binary system Th-Pu.
Requirements on matrix material:
small absorption cross section for neutrons,
good technological properties and their low dependence on the neutron flow,
good thermal conductivity and low thermal expansion coefficient,
good compatibility with coating material,
good corrosion resistance to the effect of coolant.
Dispersion fuels can be divided based on the matrix character to metallic and non-metallic.
4.4.1 Metallic dispersion fuel
This type of fuel is applied in water-cooled fuel elements of research reactors. The
application usually consists of dispersion of highly enriched particles of UO2 or UAl4 in an
aluminium matrix. In certain special energy reactors (reactors of very small dimensions) it is
possible to use the UO2 dispersion alloys in stainless steel. The content of the fissile
component usually does not exceed 50 wt. %. The advantages of these fuels include:
the possibility to achieve high irradiation without violating the shape or dimensions
and without release of fissile gases,
high thermal conductivity,
sufficient mechanical strength even after high radiation exposure.
4.4.2 Non-metallic dispersion fuels
This type of fuels has been developed namely for high-temperature gas-cooled
reactors. They are usually based on an oxide fissile component – UO2, PuO2 or ThO2,
dispersed in ceramic matrices ZrO2, MgO and Al2O3. The main advantages of these materials
include:
high melting temperature,
low leakage of fissile gases,
low swelling.
Most common fissile material and matrix combinations used in dispersion fuels are
listed in table 4.10.
Table. 4.10 Combination of fission and matrix materials for dospersion fuels.
disperzní fáze matrice
UO2
korozivzdorná ocel, Ni-Cr slitiny, SiC-Si, MoSi2, Be, Mo, Ta, Ti, W,
Nb, Al, Zr, Zircaloy, grafit, ThO2, BeO, Al2O3, SiO2
U Mg, Th, Zr, Zircaloy
UAl4 Al
UZr2 Zr, Zircaloy
UC korozivzdorná ocel, Zr, Zircaloy
UN korozivzdorná ocel, Zr
U3Si Zr
U6Ni Zr
U2Ti Zr
Summary of terms in this chapter (subchapter)
Nuclear fuel
Uranium enrichment
Powder metallurgy of uranium
Ceramic nuclear fuel
Plutonium
Questions to the covered material
Name the nuclear fuel that you know, describe chemical form in which they are used.
Describe basic technology of metallic uranium production.
Briefly describe preparation technology of plutonium.
Briefly describe preparation technology of thorium.
Characterize basic uranium alloys.
Briefly describe basic methods used for uranium enrichment.
5. Coating and construction materials
Time to study: 5 hours
Aim After studying this section the student should be able to:
Formulate basic requirements of the cladding material in terms of physical,
mechanical, chemical and nuclear properties
Assess the suitability of the material used for the cladding material or structural
material (eg. fot the core, cooling circuit, etc.). Become acquainted with the technology of preparation and treatment of alloys
used for cladding and construction materials.
Lecture
Out of all non-fissile materials used in the construction of a nuclear reactor, fuel
element coatings are exposed to the most demanding conditions. These materials separate the
fuel from the coolant, they must prevent all leaks of fissile products and in some cases they
also play the role of the load-bearing material. These properties must remain constant despite
the fact that fuel undergoes changes and it must withstand the effect of neutron radiation.
Another important property is the low absorption cross section for thermal neutrons. Ideal
coating material should therefore offer the following properties:
1. good corrosion resistance in the used coolant,
2. impenetrability towards fissile products,
3. low absorption cross section of thermal neutrons (applicable for thermal reactors),
4. good compatibility with fissile material,
5. good thermal conductivity,
6. sufficient strength at the operating load and at temperature changes, good plasticity,
7. low sensitivity to radioactive radiation,
8. good workability and weldability,
9. acceptable price.
5.1 Aluminium and aluminium alloys
The application of aluminium is possible if its relatively low strength does not
constitute an obstacle and if its advantages, such as low weight, high electric and thermal
conductivity, good workability and significant resistance to corrosion, find a suitable
application. One of the most important application areas is also nuclear technology.
5.1.1 Aluminium production
The most widespread method of aluminium production nowadays consists of the
production of pure Al2O3 and its further processing by electrolysis of molten salt (Na3AlF6,
Al2O3, etc.).
Most important for the aluminium production are aluminium hydroxides that
constitute the main aluminium ores - bauxites. These most important raw materials for
aluminium production contain diaspore (Al2O3.H2O or AlOOH) and hydrargyrit (Al2O3.3H2O
/Al(OH)3/). In addition to aluminium hydroxide they also contain Fe, SiO2, TiO2 oxide, Cr,
V, P, CaCO3 oxides, free humidity, etc.
5.1.2 Aluminium processing
This metal can be easily processed using the usual chipless techniques, such as
forging, rolling, drawing, pressing or extrusion. These technologies can be used both for pure
aluminium and its alloys. They can also be used for special materials, such as AlNiFe, AlSiNi,
that are applicable in nuclear energetics – namely for the production of coated pipes of
complicated shapes.
Processing of SAP materials is more demanding due to their low workability by
chipless technologies. Good results are obtained by hot-extrusion. In this process cylindrical
blocks with a diameter and height of 300 mm are first cold-pressed. Then these blocks are
coated by an aluminium sheet (protection against oxidation) and heated to 550 – 600°C. The
hot extrusion process creates a perfect metal connection, in which the density after extrusion
approaches the theoretical density and the maximum strength is reached. These semi-finished
products can be then used to produce the final construction elements using all chipless
processing methods, allowing the creation of very thin walls.
5.1.3 Aluminium properties
Aluminium is a light white glossy metal, stable in dry air. It has a high thermal and
electrical conductivity. Its melting temperature is 660 °C. Both mechanical and physical
properties can be therefore controlled by the admixture content. Mechanical properties also
heavily depend on the previous mechanical and thermal processing, which significantly
influences the grain size.
5.1.4 Aluminium alloys
In comparison with aluminium, aluminium alloys have significantly better mechanical
properties. These alloys can be divided to casting and workable alloys. Casting alloys are not
applied in nuclear technology. At higher temperatures, alloys for working usually consist of
homogenous solid solutions. At lower temperatures, due to their reduced solubility, other
phases appear in the structure of these alloys. Aluminium alloys can be further divided to
hardenable alloys (AlMgSi, AlMgZn, AlCuMg) and to non-hardenable (AlMg, AlNiFe,
AlSiNi). Strength properties of non-hardenable alloys can be improved by cold working.
Pure aluminium is not applicable as a coating material for nuclear energetics due to its
low strength. Aluminium alloys or aluminium materials prepared by powder metallurgy
techniques can be used for these purposes. Improvement of mechanical properties of
aluminium by precipitation hardening is complicated by a serious obstacle. Copper
significantly reduces the absorption cross section for thermal neutrons and magnesium
negatively affects corrosive behaviour of alloys in water at increased temperatures.
Better mechanical properties, especially at higher temperatures, are provided by
aluminium pseudo-alloy Al – Al2O3 referred to as SAP (produced by powder metallurgy).
This material was developed especially for ceramic fuels (UO2) and has better creep
properties than aluminium and good strength properties up to a temperature of 480°C.
SAP materials are made of aluminium with a purity of 99.5 %, which is enriched by
aluminium oxide (up to 6 - 14 %) using controlled oxidation during the powder preparation.
During grinding the surface is ground and oxidized at the same time. The Al2O3 layer is
brittle and is therefore peeled off. The clean surface oxidizes again. Oxide-free particles are
joined due to the pressure and friction. These particles are consequently ground and oxidized
by the above described process until a balance between these two processes – grinding and
joining and simultaneous oxidation – is found.
5.1.5 Aluminium corrosion
Another significant property of aluminium is its relative stability in contact with air,
which is caused by its high affinity to oxygen. A thin protective layer of Al2O3 is created on
its surface in accordance with the following reaction:
4 Al + 3 O2 = 2 Al2O3
and protects the basic metal against further oxidation. Specific volume of produced oxides is
1.8x larger than the specific volume of metal. The oxide layer is very compact and firmly
adheres to the metal. In terms of application of aluminium in nuclear technology (e.g. the
LVR 15 reactor), the decisive property of aluminium is its corrosion stability in water.
Aluminium is usually not used in other coolants (gases, organic substances, molten metals and
salts).
5.1.5.1 Corrosion resistance in water up to 100°C Generally speaking, unalloyed aluminium and aluminium alloys without heavy metals
offer good corrosion resistance in this environment. On the other hand, alloys containing
heavy metals have low resistance in water up to 100 °C. That is why for instance hardenable
alloy AlCuMg cannot be used in this environment.
Favourable properties are provided by AlMg3 a AlMgSi. Good corrosion resistance of
these alloys is given by the creation of a protective oxide layer consisting of Al2O3.3H2O at a
temperature below 60 - 70 ° and of AlOOH at a temperature exceeding this limit.
5.1.5.2 Corrosion resistance in water above 100°C The corrosion character of conventional aluminium alloys changes between the
temperatures of 100 and 200 °C. Pure aluminium and its alloys (without heavy metals) have
good corrosion resistance up to 100 °C and it increases with the aluminium purity. At the
same time, the corrosion attack is even. As the temperature increases, the corrosion character
changes and the attack speed increases. Pure aluminium is attacked more than aluminium
containing impurities. At higher temperatures corrosion takes an intercrystallic form and even
corrosion recedes. The speed of corrosion processes also increases. Even corrosion stops
promptly, the speed of intercrystallic corrosion decreases and non-oriented pitting corrosion
starts to prevail. These relations are depicted in Fig. 5.1. The last two types of corrosion lead
to a significant material degradation. That is why the use of these materials in reactors with
high-temperature water is excluded.
Obr. 5.1 Different corrosion types for pure aluminium in water at 180°C. 1 – uniform corrosion, 2 – intergranular
corrosion, 3 – pitting.
Corrosion resistance of an alloy can be significantly improved by adding a certain
amount of heavy metals. Small additions of Ni, Fe, Cu, Co, Mo, W or Ti prevent attacks of
grain boundaries. A typical representative is the alloy containing 1 % Ni, 0.5 % Fe, 0.1 % Ti
and 0.001 % Si. Small additions of Zr can also increase the strength at increased temperatures
without reducing the corrosion resistance.
5.1.5.3 Corrosion of SAP material in water Corrosion resistance of SAP aluminium material in water above 150°C is completely
inappropriate, in a short time it leads the total breakup of material. This is the reason why
SAP type alloyed material was also tested, but it also did not find wider use.
5.1.5.4 Corrosion in water vapour While water vapour was working on aluminium material there appeared an even and
local attack, leading to heavy metal disturbance. While pure aluminium is strongly attacked
by water vapour at a temperature of 325°C, aluminium alloys resisting hot water at these
parameters are resistant even in vapour. SAP type materials are affected less by these types of
materials.
5.1.5.5 Corrosion in gases Aluminium and most of its alloys are stable in dry gases up to high temperatures. This
particularily deals with air, nitrogen, oxygen, hydrogen and carbon dioxide. Density provides
good resistance, well sealed by a protective layers of corrosive products. The exception is
hologenides, which cause heavy corrosion.
5.1.5.6 Corrosion in metal liquid melts Aluminium and its alloys are mostly strongly attacked metal liquid melts at reactor
operating temperatures. Only in sodium up to 400°C does aluminium show a surprising
consistency, but with increasing temperature corrosion grows in leap intervals.
5.2 Magnesium and magnesium alloys
The use of magnesium in nuclear technology is limited exclusively to coatings of fuel
elements of carbon dioxide-cooled reactors working with natural metallic uranium at
temperatures of 400 – 550°C. The main advantages of aluminium include its very low cross
section for thermal neutrons, low price and very good compatibility with metallic uranium.
Uranium solubility in magnesium is negligible, no metallic compounds are formed between
these two elements and therefore no alloy reaction occurs between them even at temperatures
approaching the melting temperature of magnesium. Pure magnesium is not commonly used
as construction material due to its low mechanical properties.
5.2.1 Magnesium production
In terms of production, the most important minerals include magnesite (MgCO3),
dolomite (CaCO3.MgCO3), carnallite (KCl.MgCl2.6H2O) and bischofite (MgCl2.6H2O). The
following two production processes are most common for the production of magnesium:
electrolysis of molten salt containing MgCl2,
thermal reduction of oxides.
5.2.2 Magnesium properties
Pure magnesium has a hexagonal structure, its density at 20 °C is 1.74 g.cm-3
, which
ranks it among the lightest technically used metals. Small changes during solidification (4 %)
facilitate the casting process. High vapour tension allows its cleaning by sublimation or
vacuum distillation.
Its flash temperature poses a limit to working temperatures of magnesium-coated fuel
elements. The operating temperature was therefore increased by a zirconium alloy. The flash
temperature is increased by Ca and Si, all other monitored alloys decreased this temperature.
Bi, Ni and Zn have a particularly negative effect. It is a generally known fact that magnesium
and magnesium alloys are most suitable for the CO2 environment.
Magnesium has a good thermal conductivity and is suitable for heat dissipation from
fissile materials. Thermal conductivity is affected by alloying admixtures and thermal
processing. Thermal conductivity is reduced by Sn, Al and Mn. For pure aluminium, thermal
conductivity decreases with the increasing temperature, whereas the thermal conductivity of
all alloys increases. The thermal expansion ratio of magnesium to uranium is less
satisfactory. This causes undesirable tension between the fuel and the coating.
Mechanical properties of pure magnesium are not favourable: it has low strength and
malleability. Low malleability is given by its hexagonal structure. At low temperatures (up to
100°C) magnesium also has low ductility (5 – 14 %). At higher temperatures – about 400 °C –
magnesium grains grow fast and decrease the ductility of coating materials. The plastic strain
at temperatures up to approximately 250 °C creates intercrystallic cracks and pores along
grain boundaries at a temperature of 350 – 450 °C.
5.2.3 Magnesium alloys
Mechanical properties of magnesium can be improved by alloying with other metals,
such as Al, Be, Mn, Zn, Zr, etc. The absorption cross section for thermal neutrons of the alloy
needs to be taken into consideration in the alloying process. Examples of magnesium alloys
are listed in table 5.1.
Table. 5.1 Magnesium alloys for nuclear applications.
alloy
alloying element [hm. %]
Al Zn Mn Be Zr Ca
Russian
1.0 - - 0.04-0.1 - -
- - - 0.04-0.01 0.6 -
British and American
Magnox 1 0.5 - - 0.01 - 0.2
Magnox E 1.0 - - 0.05 - 0.1
Magnox E12 0.8 - - 0.01 - -
Magnox Al80 0.8 - - 0.002-0.015 - -
Dowmetal C 9.0 2.0 0.2 - - -
French
ZA - - - - 0.5-0.7 -
ZW 1 - 0.7-1.0 - - 0.5-0.7 -
Mg pseudoalloys
PMB 2 - - - 2 - -
PMB 5 - - - 5 - -
5.2.3.1 Mg - Be alloys Addition of beryllium significantly improves the resistance to oxidation by creating an
oxide layer with suitable mechanical properties. The preparation of Mg-Be alloys is
complicated by the large difference in melting temperatures of these two metals (melting
temperature of Be is 1283 °C, which is higher than boiling temperature of magnesium which
is 1120 °C). In practice, two approaches can be taken:
1. adding Be in form of master alloys,
2. preparation of pseudo-alloys by powder metallurgy.
Ad 1. Beryllium is added to melted magnesium namely in the form of aluminium-beryllium
master alloy. This method is applied namely for Magnox alloys.
Ad 2. The powder metallurgy technology enables the introduction of practically any amount
of beryllium into magnesium (Russian PMB 2 and PMB 5). This method can be used to more
than double the low yield strength of pure magnesium at normal temperatures and even at
higher temperatures it is still significantly higher than for magnesium. This is given by the
occurrence of MgBe13 alloy in the form of needles deposited in the metal matrix. Mechanical
properties are also affected by BeO, which is formed by oxidation of Be powder. This oxide
phase causes dispersion strengthening similar as in SAP alloys.
Magnox E12
It is one of the best known alloys. It was used in the first British nuclear power plant
Calder Hall. It is used in gas-cooled GCR reactors for coating of natural uranium. The coating
temperature reaches up to 450 °C at the CO2 output temperature of up to 400°C.
Magnox E80
This alloy has proven to be the most suitable material for CO2-cooled reactors. Is
based on the Magnox E12 alloy. It contains 0.7 – 0.9 % Al and 0.002 – 0.015 % Be, which is
practically the limit content of Be achievable by melting. Grain coarsening represents a
certain disadvantage of these alloys. Certain magnesium alloys let through plutonium;
however, Magnox Al80 is suitable in this aspect as plutonium only penetrates up to 15 µm of
the coating below the contact surface.
Magnesium alloys with high content of beryllium are resistant to oxidation even at wet
CO2. The protection effect of beryllium is usually explained by the fact that the coating layer
contains multiple times higher BeO concentration than the alloy. Higher Be content enables
the creation of a metal connection of the coating and the fuel. A lawyer of UBe13 is created at
the boundary.
Pseudo-alloys of Mg-Be are intended namely for coatings of fuel elements of GCR
reactors up to surface coating temperatures of 520 – 530°C. They are affected by relatively
strong corrosion in water and water vapour.
5.2.3.2 Mg – Zr alloys This type of alloy is among the most important alloys. The ingredient Zr at the amount
0,5 – 0,6 % considerably affects mechanical properties. It leads to increased ductility in low
temperatures (grain softening). In higher temperatures it leads to a reduction of resistance
against creepu. This type of alloy is suitable for temperatues 400 – 500°C. They have a low
absorption section and a higher grain size stability for all magnesium alloys. The additive Mn
improves resistance against creep. The corrosion resistance of the Mg – Zr alloy doesn´t reach
the corrosion resistance of the Mg – Be alloy. V CO2 resists better than the Magnox type
alloy, and even at 500°C.
Carbon dioxide atmosphere
Carbon dioxide is the most important gaseous coolant for reactors of the GCR type. In
contact with magnesium, CO2 manifests a series of reactions. From a thermodynamic point of
view even at low temperatures the following reduction processes are possible:
CO2 + Mg → MgO + CO
CO2 + 2 Mg → 2 MgO + C
Besides these reactions it can lead to the formation of MgCO3:
MgO + CO2 → MgCO3
In dry CO2 they have been measured less than 0.02 mg.cm-2
. Under normal pressure at
temperatures 500°C after 1000 hours for alloys with Mg – Zr additives. Magnox type alloys
are particularly resistant. Up to 400°C the protective layer is composed of MgCO3 and at
temperatures above 500°C it consists mostly of MgO and carbon mixtures. Water vapour
works very negatively in CO2, while the effects of nitrogen and hydrogen are not shown.
5.2.4 Corrosion of magnesium and its alloys
Due to its non-noble character, magnesium is located left of aluminium in the
electrochemical line with a negative potential of 1.7 V. It also has a high affinity to oxygen. A
layer of MgO on the surface is incoherent and does not provide nearly any protection. In a wet
atmosphere magnesium quickly loses its metallic gloss, becomes matt and is gradually
covered by a grey layer of Mg(OH)2. With the exception of Ce, La, Ca and Be, alloying
elements reduce the oxidation stability of magnesium, e.g. Ni, Al, Zn and Mn have a
particularly negative effect. A comparison of basic magnesium alloys that can be used in
nuclear energetics clearly shows that corrosion resistance of Mg – Zr alloys corresponds to
the resistance of pure magnesium, whereas Magnox alloys containing 0.01 % Be behave
much better even in a wet atmosphere.
Water effects
This metal is not possible to use as coating material in water cooled reactors, because
it gradually passes in pure water into the solution, or it flakes.
Effects of metal liquid melt (Na, Na – K, Hg)
These metals are their alloys are heavily attacked in the liquid state of magnesium.
Because of this it is not possible to use magnesium in such types of reactors.
5.3 Zirconium and zirconium alloys
Previously zirconium did not have any significant application; today this metal plays a
crucial role for nuclear technology. Non-alloyed zirconium is used as gas absorber,
deoxidizing agent and alloy for many other metals and alloys. The main application area of
zirconium alloys is represented by coating materials.
5.3.1 Zirconium production
There are two main approaches to the preparation of zirconium:
Reduction of zirconium tetrachloride (Kroll process) producing a metallic sponge,
electrolysis of K2(ZrF6) in an environment of halogen compounds of alkali metals,
which create powder metal.
High-purity zirconium is required for its application in nuclear technology. This
concerns namely very low contents of hafnium, which has a very negative impact on nuclear
properties of zirconium due to its high absorption cross section for thermal neutrons. Since the
atomic structure of Zr and Hf is very similar, resulting in very similar properties of these
elements, their perfect separation is technologically challenging.
The Earth's crust contains about 0.028% of zirconium, which is more than Ni, Cu, Pb
or Zn. There exists a whole range of minerals, the most important of which are listed below.
Zircon ZrSiO4 – most widespread mineral of yellowish color. It does not disintegrates
in acids but it does during melting in alkaline compounds.
Baddeleyite – monoclinic form of ZrO2 of yellowish-brown color, it disintegrates in
compounds of sulphuric and hydrofluoric acids.
Most of zircon is mined from sea sands. It is accompanied by other minerals, such as rutile,
ilmenite, monasite. The composition of zircon and baddeleyite concentrates is similar to
monomineral materials. Zircon concentrate contains 65 % and baddeleyite concentrate over
90 % of ZrO2 + HfO2. The hafnium content reaches up to 0.8 – 2.5 % from the sum of Zr +
Hf. Sea sands are usually processed using the gravitation method. The final cleaning is done
by magnetic or electrostatic separation.
5.3.1.1 Processing methods of zircon concentrates Zircon is one of the minerals that are difficult to disintegrate and that is why various
pyro-methods are used for processing:
melting with sodium hydroxide creating water-soluble sodium metazirconate.
caking with oxides and carbonates of alkali metals.
Chlorination of zircon and carbon compounds creating zirconium tetrachloride,
melting with potassium fluoride or potassium fluorosilicate creating potassium
fluorozirconate,
thermal disintegration of zircon.
Melting of zircon concentrate with NaOh
This universal process can be used to obtain not only metallic zirconium (hafnium) but
also their compounds, including ZrO2. Leaching of sinter that consists of silicate and sodium
zirconate yields zirconium (hafnium) solutions and these are separated by extraction. The
following reaction occurs during caking:
ZrSiO4 + 4 NaOH = Na2SiO3 + Na2ZrO3 + 2 H2O
The process begins at a temperature of 250 – 300°C and usually occurs at temperatures
below 700°C. At least 130 % of the excessive agent is used. After the melt solidifies, it is
leached in water in an iron vessel or it is leached simultaneously with grinding. During the
solution process, sodium silicate enters the solution and zirconate is hydrolyzed:
Na2ZrO3 + 2 H2O = ZrO(OH)2 + 2 NaOH
As the solution alkalinity decreases, it is necessary to keep the content of NaOH at 3 – 5 % in
order to prevent hydrolysis of Na2SiO3 connected with its transformation to a precipitate. The
precipitate, which consists of 80 – 84 % ZrO2 (other contained impurities include SiO2,
Na2O), is leached in HCl or H2SO4 and the following reactions occur:
Na2ZrO3 + 4 HCl = ZrOCl2 + 2 NaCl + 2 H2O
Na2ZrO3 + 2 H2SO4 = ZrOSO4 + 2 NaCl + 2 H2O
ZrOSO4 + H2SO4 = H2/ZrO(SO4)2/
HCI is used for high-purity Zr compounds. Part of silicic acid is released during the solution,
whereas the precipitation of voluminous silicic acid is sped up by a coagulant. After filtering,
pure Zr compounds can be obtained from the filtrate. However, these compounds still contain
Hf. Hf is separated from Zr by a whole range of methods – most importantly extraction in
organic liquids.
5.3.1.2 Production of zirconium tetrachloride These processes are commonly used in the preparation of zirconium to prepare the
input material, i.e. zirconium tetrachloride, for production of zirconium using the Kroll
process. These methods are used to process the zircon concentrate, zirconium oxide;
alternatively zircon might be first transformed to carbon nitride and then chlorinated.
It is possible to carry out chorination in various equipment, high furnaces are suitable,
and chlorination in whirling layers. After chorination follows the fractional condensation of
chlorides, their purification, the separation of ZrCl4 a HfCl4, the transfer to aqueous solutions
for further hydometallurical treatment, the preparation of ZrO2 and HfO2.
Carburizing is carried out at temperatures 2000 – 2500°C During the joint reaction of
zircon and carbon depending on the amount of silicon reagent we acquire it in the form of
SiO mono-oxide, SiC carbide or in elementary forms and zircon in the form of ZrO2 or ZrC.
In regards to following chorination the process is led so that SiO and ZrC are formed:
ZrSiO4 + 4 C = ZrC + SiO + 3 CO
SiO mono-oxide is slightly volatile and in carburizing temperatures it easily flows out of a
furnace. The carburizing product is the carbonitride Zr and this deals with the solid solution
ZrC and ZrN with a content of 15 – 20 % ZrN. The joint reaction of ZrN and ZrC with
chlorine is carried out from 450°C to develop greater amounts of heat:
ZrC (s) + 2 Cl2 (g) = ZrCl4 (g) + C (s) + 846 kJ
ZrN (s) + 2 Cl2 (g) = ZrCl4 (g) + 0,5 N2 (g) + 670 kJ
The released heat is sufficient enough for maintaining the necessary temperature process. An
important operation is the purification of ZrCl4 distillation, by which absorbent gases are
removed (Cl2, HCl), as well as lightly volatile Ti, Si chlorides and moisture. This can also
lead to secondary effects through the reaction, for example the transfer of several non-volatile
oxides, and other compounds into volatile chlorides as the consequence of exchanged
reactions.
5.3.1.3 Separation of hafnium from zirconium (dehafnization) Hafnium as an admixture in zirconium is very undesirable in nuclear technology due
to its high absorption cross section for thermal neutrons. That is why separation of hafnium
from zirconium needs to be included in the preparation process of zirconium compounds for
production of metallic zirconium. The separation is usually performed by a
hydrometallurgical process. Despite the similarity of chemical behaviour of both metals,
certain properties differ significantly and can be therefore used to separate these metals. Used
methods include:
fractional precipitation – this process is based on solubility of the same salts that are
produced by adding a suitable agent to a solution with precious metal ions,
fractional crystallization,
ion exchange,
extraction in organic solutions – this method has the widest application.
5.3.1.4 Production of metallic zirconium by metallothermic method Metallic zirconium is produced predominantly by metallothermic reduction of Zr
compounds – the most important is reduction of ZrCl4 using Mg. Another method is
electrolytic production from molten alkali halides and K2ZrF6. Fig. 5.2 shows dependence of
the Gibbs free enthalpy of the temperature for the formation of various oxides, chlorides and
fluorides of Zr, Hf and certain other metals. From Figure 5.2 it is seen that for the reduction of
oxides, calcium may be used, for the reduction of chlorides and fluorides calcium, sodium and
magnesium should be used.
Reduction of ZrCl4 by magnesium
This reaction constitutes the basis of the Kroll process. The reduction occurs in steps
in accordance with the following reactions:
ZrCl4 + Mg → ZrCl2 + MgCl2
ZrCl2 + Mg → Zr + MgCl2
These reactions apply only if the input substances are in gaseous state. Mutual reaction of Mg
with ZrCl4 starts at 410 – 470°C; ZrCl4 is reduced only to lower chlorides at temperatures
below 650°C.
Reduction of ZrCl4 by magnesium is a complex heterogeneous process consisting of
several liquid and several solid phases, where the most important processes include wetting,
evaporation, heat and matter exchange.
After the process is completed, the basic mass of the obtained metal consists of a
porous block located at the bottom of the reactor and may have the following composition: Zr
58 wt. %, Mg 32 %, MgCl2 10 %. The next step is vacuum distillation of Mg and MgCl2 at a
temperature of 1000°C. The last step consists of arc melting of the produced material.
The production technology of Zr using reduction of ZrCl4 by magnesium consists of
the following operations:
1. preparation of input substances,
2. reduction,
3. vacuum separation,
Fig. 5.2 Dependence of Gibbs free enthalpy on temperature for selected compounds
Temperature [K]
4. modification of the sponge block (grinding)
5.3.2 Zirconium refining
This metal (similar to Hf) of high purity is possible to prepare using the iodide
method. This is a typical example of a transport reaction. Zircon has just the optimal
properties for this type of reaction.
Me (s) + 2 I2 (g) 200−300°𝐶→ MeI4 (g)
1300−1500°𝐶→ 4 I2 (g) + Me (s)
The principle of the refining process is that raw metal is heated up in vacuum equipment at
the formation temperature of ZrI4. This iodide is transferred into the gaseous phase and when
in contact with an incandescent fibre (a wire from refined Zr or other heavy meltable metal) it
disassociates from the iodine, which returns back to the reaction with raw Zr into ZrI4, and
into metal, eliminating the incandescent fibre. A diagram of the apparatus is given in Figure
5.3.
5.3.3 Zirconium alloys
Zr alloying requires a complex approach, just like for other metals. For instance, if the
goal of the alloying process is to increase the corrosion resistance, it is necessary to consider
the mechanical properties of the concrete alloy, as well as its absorption cross section for
thermal neutrons. The most important requirements placed on Zr alloys for nuclear reactors
are:
1. The alloying elements must have a low absorption cross section for thermal neutrons.
2. The alloying element must ensure sufficient corrosion resistance of products intended
for the reactor's active zone for the whole duration of their function.
3. The alloying element must ensure mechanical reliability of fuel elements in all
possible operating modes of the reactor, including emergency situations.
4. The alloying element must not produce long-time radionuclides with strong α
radiation, as this would extend the shutdown period during repairs and would increase
the costs for processing irradiated fuel elements.
The main alloying elements used in Zircaloys is tin and even though it compensates
the negative effect of nitrogen, it increases corrosion. No other group IV element besides tin
can be used. Titanium has a highly negative effect on the corrosion behaviour, Hf has a high
cross section for thermal neutrons, Pg is volatile and negatively affects the corrosion
1 – vakuum system
2 – electric contacts
3 – peephole
4 – rubber gaskets
5 – raw zirconium
6 – grid
7 – heated wire for thermal decomposition
8 – thermocouple
Fig. 5.3 Schéme of device for iodide refining of zirkonium and hafnium.
properties, Si and Ge have significantly different diameters and are practically insoluble in α
Zr and in β Zr.
The only group V.A element that comes into question is Nb. Vanadium increases the
corrosion of pure Zr even in small amounts and can only be used in polycomponent alloys for
superheated vapour. Tantalum has a 100x higher absorption cross section for thermal
neutrons.
Potential elements in group VI.A are Cr and Mo. These elements are suitable for
alloys operating in superheated vapour at temperatures ranging from 400 – 500 °C.
The only interesting element in group VIII is Fe, since Ni significantly increases the
hydrogenization of Zr during corrosion and Co has high absorption cross section for thermal
neutrons and a strong γ emitter 60
Co with long half-life is created in the reactor. The main
alloys for iron are alloys for superheated vapour with a temperature of 400 – 500 °C.
It is clear from what has been stated above that the only element from groups V, VI A and
VIII suitable for alloying zirconium used in water and water vapour at a temperature range of
300 – 350 °C is Nb.
5.3.3.1 Zr – Nb alloys In the former Soviet Union there were developed two alloys: one with a mass content
of 1 % Nb (the alloy N-1) for covering fuel cells, and an alloy with a content of Nb 2,5 % (N-
2,5) – for sewerage pipes, the holder plates of PWR type reactors and other details of reactor
active zones of the type PWR and LWGR. Alloys were developed for work in water and
vapour-aqueous mixtures at temperatures 300 – 350°C. The alloy N-1 shows very little
oxidized growth (less than the alloy N-2,5 and an alloy with 3 and 5 % Nb).
5.3.3.2 Zr – Sn alloys In addition to niobium, Sn can also improve mechanical properties and corrosion
resistance. Studies of Zr – Sn alloys arrived at the conclusion that Sn is highly soluble in α
zirconium (max. 9 % at 980°C; at 300 – 350°C the solubility is negligible). Independently
from these studies, corrosion tests proved that tin mitigates the effect of many harmful
admixtures. Fig. 5.4 shows the effect of Sn content on the corrosion resistance of spongy Zr
alloys. All alloys were prepared from the same input material. It shows that up to a content of
0.5 %, tin slows down the corrosion process by eliminating harmful admixtures. The
corrosion speed increases if the tin content exceeds 0.5 %. Reduction of the corrosion
resistance of unalloyed Zr is caused by the presence of a very small amount of nitrogen (0.006
% N2). Acceptable nitrogen contents in relation to the tin content in the alloy is shown in table
5.2.
1 – 360°C, 86 days
2 – 315°C, 162 days
3 – 315°C, 44 days
Fig. 5.4 Influence of Sn content on corrosion behaviour of Zr alloys in water.
5.3.4 Corrosion of zirconium and its alloys
5.3.4.1 Water Thanks to their properties, zirconium materials are suitable especially for water-cooled
reactors. The coolant – water, vapour or their mixture – transfers water molecules to the
oxide layer. These molecules are absorbed by the layer, whereas part of electrons is also
absorbed while producing oxygen and hydrogen ions. Oxygen ions penetrate through the
oxide layer and create ZrO2 molecules in contact with metal. This increases the coating
thickness. The created oxide layer consists of monoclinic ZrO2, it has sufficient density and
adheres well to the metal surface.
The kinetics of zirconium and zirconium alloys oxidation is not governed by one
specific law. The initial oxidation period is usually characterized by a parabolic dependence
that expresses the inverse proportionality of the growth speed to its thickness. The parabolic
dependence changes to cubic at the layer thickness of approximately 1 µm. If the layer is 2 –
3 µm thick, the growth kinetics of the oxide layer changes to linear dependence. Micropores
of various diameters and longitudinal or transverse cracks are uncovered in the layer.
Until the break of growth speed the layer has good adhesion properties, black color
and smooth, glossy surface. Its corrosion stability is also high. Oxide contained in the layer is
substoichiometric, its formula is ZrO2-x, where x≤ 0.05. At the break, when the layer thickness
reaches 2 – 3 µm, the layer color turns grey and if the thickness increases to 50 – 60 µm, the
color changes to white.. This layer is stechiometric and is a sign of corrosion failure. This type
is referred to as "breakaway". Failure occurs even at lower thickness of the layer if tension or
impurities are present. This phenomenon is represented in Fig. 5.5.
The oxidation zirconium and its alloys is an exceptionally complex process due to the
dependence of the kinetics and oxidation character on various factors:
Table. 5.2 Allowable content of nitrogen in alloys based on Zr – Sn.
Sn content [wt. %] max. N content [wt. %]
0.5 0.02
1.0 0.03
2.0 0.06
2.5 0.07
3.0 0.08
Fig. 5.5 Corrossion of pure and impure Zr, breakaway effect.
1. chemical composition of zirconium alloy in terms of admixtures and alloying
elements,
2. structural composition of the alloy, which is given by a whole range of operations
during the melting processing into a finished product (casting, forging, pressing –
including thermal processing),
3. quality of the product surface,
4. coolant composition (water purity in terms of admixtures, oxygen and hydrogen
content); coolant characteristics – input and output temperature of the coolant in the
active zone, real temperature of the fuel element surface, boiling character, speed,
flow,
5. integral batch of fast neutrons at the given moment of corrosion, duration of the
product's stay in the coolant at presence and absence of a neutron field,
mechanical composition of the zirconium product – tension, cyclic stress, friction, shocks,
contact with other materials (stainless steel).
5.3.4.2 Gases, liquid metals Zr corrosion in gases such as nitrogen, air or carbon dioxide, which are lost in
consideration of coolants or their impurities, in the course of time resemble zircon corrosion
in water or aqueous vapour. Zircon and especially alloys of the type zircaloy-2 are unsuitable
for use in reactors cooled by CO2 in the case the operational temperature exceeds 400°C (this
then leads to great growth in oxidation speed). In moist CO2 the situation is still worse. At
temperatures above 500°C the following reaction can have significance:
2 Zr + CO2 → ZrC + ZrO2
The formed ZrO2 has an unfavorable effect on corrosion resistance. Nitrogen contained in
metal at an amount up to 250 ppm does not worsen resistance against CO2, while Al and Ti
have proven to be harmful.
5.4 Beryllium and beryllium alloys
Beryllium has been industrially utilized in nuclear technology since the 50's. Be was
used mostly as the alloying element in copper alloys. It was developed during World War II –
namely for military application of nuclear energy. This metal can be used as coating material,
moderator and reflector of fast neutrons.
5.4.1 Beryllium production
Beryllium occurs in nature in the form of minerals. The best known minerals include:
Beryl – 3BeO.Al2O3.6SiO2, contains about 5 % Be and 14 % BeO – most important for the
industry.
Phenakite – 2BeO.SiO2.
Chrysoberyl – BeO.Al2O3.
Metallic beryllium is processed from rich concentrates with a beryllium content
exceeding 94 %, which are prepared by various enrichment methods, including floatation.
Beryllium is prepared by electrolysis of molten beryllium chloride or by metallothermic
reduction of beryllium fluoride. The input ore is first used to prepare oxide, which is then
used to produce chloride or beryllium fluoride.
5.4.1.1 Compound preparation for beryl production
Sulphate method of producing BeO
The beryl concentrate is melted together with sodium or calcium in an electric arc
furnace at a temperature of 1700°C. Through this method beryl oxide is transferred into
moulds, which is leached in H2SO4. The liquid melt is granulated by casting into water and
the formed granulate is dried at a temperature of 900°C. 90 % of BeO passes into the leached
mould. During sintering this reaction is carried out:
3 BeO.Al2O3.6SiO2 + 6 MeCO3 → 3 BeO.Al2O3.6MeCO3 + 6 SiO2
The leaching of H2SO4 leads to beryl sulphurization according to the equation:
3 BeO.Al2O3.6MeSiO3 + 12 H2SO4 → BeSO4 + Al2(SO4)3 + 6 MeSO4 + 6 SiO2 + 12 H2O
After leaching H2SO4 follows leaching using the water, which has separated a large
part of SiO2. Aluminium oxide is removed by adding ammonia in forming a sulphate solution
of ammonia aluminium. After solution cooling at 20°C the leached crystals from the parent
solution centrifuge (using this method removes 75 % Al from the input material). For the
further removal of aluminium to the parent solution acid is added, which creates dissolvable
complex compounds with several mixtures. At the same time NaOH is also added to the
solution, through which BeO is transferrd as well as the remaining Al2O3 into dissolvable
berylnatan and sodium aluminate. In boiling the sodium berylnatan solution it hydrolyzes into
forms of non-solvable berylnate hydroxides. This hydroxide is consequentally thermally
processed into BeO.
Berylnate fluoride production
This compound is used for producing metals using metallothermic methods. The first
stage is preparing a sodium fluoroberylnatane reaction of BeO or Be(OH)2 with NH4HF2 in
forming ammonia fluoroberylnatane (NH4)2BeF4. Berylnate fluorides are then acquired using
the thermal breakdown of anmmonia fluoroberlynatane at 900 – 950°C.
5.4.1.2 Production of metallic beryllium Beryllium in metallic form is prepared mainly by electrolysis of molten salts BeCl2
together with NaCl and by metallothermic reduction of beryllium salts using Mg, Ca or Na.
Beryl metallothermic reduction
Metal beryllium is possible to prepare for the reduction of fluorides by using beryl
magnesium, calcium or sodium. The processis carried out according to the equation:
BeF2 + Mg → Be + MgF2
The reduction is carried out at temperatures 900 – 1000°C in a high-frequency furnace in
a graphite crucible. After reduction has finished the temperature increases in the crucible to
1350°C and the liquid melt is casted out into another crucible, where it hardens. A purity of
97 % Be is reached. Further refining is possible by electrolytic methods.
5.4.2 Beryllium properties
Beryllium is a grey coloured metal. Its melting temperature is 1283 °C, it has an HTU
structure and density of 1.85 g.cm-3
. The structure of a technical metal is heavily dependent
on its chemical composition and the method of thermal and mechanical processing.
This metal has the smallest absorption cross section for thermal neutrons, 0.0092×10-28
m2
and high slow-down ability. Mechanical properties of beryllium show signs of many
irregularities, which are associated with difficulties during processing. Its low malleability at
normal and increased temperatures is particularly disadvantageous. Melting and casting
therefore yields almost exclusively coarse-grain material with unsuitable properties for further
processing.
Better properties are manifested by materials prepared by powder metallurgy, further
processed by pressing or rolling. The best isotropic properties are provided by materials
produced by powder metallurgy and formed by hot-pressing.
5.4.3 Beryllium corrosion
Corrosion properties of this metal are similar to aluminium. Whereas at normal
temperature beryllium is resistant to chemical impacts, it is more reactive at higher
temperatures. In contact with air, polished beryllium retains its gloss for a long period of time.
The dependence of beryllium oxidation speed on the temperature is depicted in Fig. 5.6.
Be is stable in carbon dioxide up to 500°C – under the condition that CO2 does not
contain humidity. At temperature exceeding 650 °C Be is attacked by both dry and wet carbon
dioxide.
5.5 Steels and nickel alloys
Stainless noble steels and nickel alloys are used in nuclear energetics as construction
materials of the primary circuit. Their good corrosion resistance in gaseous environment,
high-temperature water and liquid metals makes these materials suitable for coating of fuel
elements.
Another important property is their strength at increased temperatures, which is
significantly higher than in other materials, such as Al, Mg, Zr, Be and their alloys. Working
temperatures of austenitic steels and special nickel alloys ranges between 700 – 750°C (which
for approximately 200 – 300°C more than for classic coating materials). Another
unquestionable advantage of this group of materials is the well-developed processing and
production technologies. Their relatively low price (in comparison with Zr) also plays an
important role.
The main disadvantage of noble austenitic steels is their high absorption cross section
for thermal neutrons, which is 2.88×10-28 m2. Their use in the reactor's active zone is
therefore conditioned by fuel enrichment. However, their high absorption cross section for
thermal neutrons is significant only for thermal reactors (PWR, etc.) and they are therefore
more suitable as a coating material for FBR reactors.
When noble steels are used in thermal reactors, it is necessary to lower the contents of
elements with significantly higher absorption cross sections for thermal neutrons, namely Mn
Fig. 5.6 Beryllium oxidation rate in air depending on the temperature.
and Ta. Steels must also be characterized by low content of elements creating isotopes with
long half-lives and hard γ radiation. These include namely Co, the content of which should
not exceed 0.03 %.
An overview of highly alloyed steels and alloys suitable for nuclear energetics is
included in table 5.3.
5.5.1 Corrosion resistance
The corrosion resistance of stainless steels of the type 18/8 (18Cr, 8Ni, Cmax 0.08) in
pressurized water or vapour (a vapour-aqueous mixture) is very good up to temperatures
around 600°C. For example pitting is very dangerous, which can lead to a higher
concentration of Cl ions in water. Other types of corrosive attack occuring on stainless steels
is inter-cystallic corrosion, which is possible to effectively prevent by stabilizing steel with
titanium or niobe. The corrosion resistance of stainless steels is higher than the resistance of
zircon and its alloys. These steels are most suitable as reactor covering material working with
vapour at temperatures 430 – 600°C and a pressure up to 35 MPa.
Table. 5.3 Selected pecial steels and nickel alloys.
Material
alloying element [hm. %]
Cmax Mnmax Simax Cr Ni Al, other
Feritic steels and alloys with high Cr content
A/S/ 502 SS 0.10 1.00 1.00 4-6
A/S/ 405 SS 0.08 2.00 1.00 11.5-13.5
0.3
A/S/ 406 0.07 0.40 0.48 13.5 0.12 3.90 (Mo)
Fe-Cr-Al 0.03 0.10 0.10 24 0.1 5.60 (Ti)
Austenitic steels 18/8 type
18 Cr 13 Ni (A/S/ 304) 0.08 2.00 1.00 18-20 8-12
18 Cr 12 Ni Mo (A/S/
316) 0.08 2.00 1.00 16-18 10-14 Mo (2-3)
Austenitic steels 16/13 type
16 Cr 13 Ni Nb 0.10 1.50 0.50 17.5 13 0.015 (Nb+Ta)
17 Cr 13 Ni Mo Nb 0.13 2.00 1.00 17-19 13-15 Mo 1.75-2.75 Nb
A/S/ 318 4988 0.10 1.50 0.50 15.5-17.5 12.5-14.5 Mo 1.10-1.50 Nb, V, N2
12 R 72 HV 0.10 1.80 0.50 15 15 Mo 1.20, Ti, B
Incoloy alloys
Incoloy 800 0.10 1.00 0.60 21 32
Inconel alloys
Inconel 600 0.08 1.00 0.50 16 74 Ti, Al
Inconel 702 0.08 0.50 0.50 15.5 79 Ti, Al
Nimonic alloys
Nimonic 75 0.15 1.00 1.00 18-21 76 Ti
Nimonic 80 0.10 1.00 1.00 18-21 70 Ti, Al
Hasteloy alloys
Hasteloy A 0.15 2.00 1.00 - 57 18-22 Mo, W
Hasteloy N 0.03 1.00 1.00 7.5 71 16 Mo, W, V, Ti, Al, B
5.6 Niobium
This metal has a whole range of interesting properties for use in high-temperature
reactors. Its mechanical properties at higher temperatures are significantly better than for
classic coating materials. It has a relatively good malleability. The melting temperature of
niobium is high (2415°C) at relatively low density (8.57 gxcm-3
). The cross section for
thermal neutrons is also satisfactory (1.15×10-28
m2). It has good corrosion resistance in
molten metals (Na, Li, Hg, Sn, Zn, Bi) and it practically does not react with U and Pu.
Its disadvantages include poor resistance to oxidation at temperatures exceeding
200°C and heavy dependence of its malleability on the content of gaseous and non-metallic
impurities.
5.6.1 Niobium production
This metal occurs in nature together with tantalum. Tantalum and niobium also share
similar properties. The production technology of both these metals is identical, whereas the
chemical compounds of both elements are separated during the last phase before production
of metallic Nb. Perfect separation of these elements is necessary for use in nuclear
technology due to the negative nuclear properties of Ta (σa = 21.4×10-28
m2).
Niobium is not a very common metal, its average content in the Earth's crust is
approximately Niob 1×10-3
%. There are more than 130 minerals in total; however, only
several of them are used in the industry. The most important minerals include: Columbite
and tantalite – (Fe, Mn) (Nb, Ta)2O6, that represent an isomorphic compound of tantalate and
iron niobate; Loparite – a compound of titanate and sodium niobate, calcium and REM (Na,
Ca, Ce)2 (Ti, Nb)2O6. Ta and Nb ores are usually poor and contain 0.03 – 0.2 % of Me2O5
oxides; that is why they need to be concentrated. The basic used methods are gravitation
methods.
5.6.1.1 Preparation of pure niobe compounds The concentrate is treated for pure compounds – complex fluorides, chlorides or
oxides. These are then the basic imputs for the production of metallic niobium. Columbite
concentrates are treated most often by melting with alkaline compounds, such as NaOH,
Na2CO3, or by breakdown using HF. During melting with NaOH the following reactions are
carried out:
Fe (NbO3)2 + 10 NaOH = 2 Na5NbO5 + FeO + 5 H2O
Mn (NbO3)2 + 10 NaOH = 2 Na5NbO5 + MnO + 5 H2O
During leaching in water there is carried out a reaction for the formation of niobium
precipitates, through which mixtures of SiO2, SnO2, Al2O3, FeWO4 are transferred into the
solution:
12 Na5NbO5 + 55 H2O = 7 Na2O . 6 Nb2O5 . 32 H2O + 48 NaOH
After filtration this precipitate dries at a temperature of 100°C. It is necessary to be aware that
in the given equations niobium can be substituted by tantalum.
5.6.1.2 Separation of niobium and tantalum The similarity of these metals poses an obstacle for their separation. The original
separation method used different solubility and crystal structure of K2NbOF5.H2O and
K2TaF7. Potassium fluorotantale has lower solubility than the similar niobium compound and
it can therefore be filtered from the solution after cooling. Solution containing K2NbOF5
evaporates and K2NbOF5.H2O is released from the solution. These crystals are cleaned by
further crystallization. Other applicable methods include:
use of ion exchangers,
fractional distillation of Nb and Ta chlorides,
extraction by liquids (cyclohexanone, MIBK, etc.).
The process of liquid extraction found wide application in practice. This process is
characterized by three stages:
1. extraction of niobium and tantalum in organic solvents and their separation from
admixtures contained in the solution,
2. re-extraction of niobium from the organic solvent using water,
3. re-extraction of tantalum from the organic solvent by water solution of ammonium
fluoride.
5.6.1.3 Production of metallic niobium Metallic niobium is produced either by reduction of Nb2O5 by sodium or by niobium
carbide. Due to high thermodynamic stability of Nb2O5 it can be reduced only by Na, other
metals (Mg, Ca, Al) cannot be used. The disadvantage of this method consists in the difficulty
of separating oxides and the reduced metal, which results in insufficient purity of the
produced metal.
The carbidothermal process of niobium preparation is based on reduction of Nb2O5
using NbC at a temperature of 1600 – 1700°C in accordance with the following equation:
Nb2O5 + 5 NbC = 7 Nb + 5 CO
The process is performed in a graphite tube furnace in hydrogen or argon atmosphere. The
advantage of this method is the use of cheap reduction material (soot) and the high achieved
reduction degree.
Reduction of NbCl5 chlorides is also possible. It is performed in a steel bomb using Na
with an addition of CaCl2, which reduces the reaction speed and the amount of released heat.
The electrolytic production method is applied particularly for tantalum but it can also be used
for niobium. Electrolysis is performed from molten alkali fluorides or chlorides with
admixture of K2NbF7 or K2NbF5. These methods are used to prepare powder niobium and the
following processes are used to achieve compact metal:
1. powder metallurgy,
2. arc melting in vacuum or inert atmosphere,,
3. electron melting.
5.6.2 Niobium processing
Powder niobium is processed similarly as Mo and W. The process is divided into the
following two operations: powder pressing and caking. The pressing depends on the grain size
and pressure ranges between 250 and 750 MPa. Caking is realized in vacuum (simultaneously
with cleaning due to volatilization of certain admixtures). The caking temperature is 1400 –
1500°C and lasts for 2 hours. The final caking is done at 2300°C.
Niobium melting is performed in an electric arc furnace with a consumable electrode
in a copper water-cooled crystallizer. The electrode is produced by caking powder niobium.
Electron melting has a wide range of advantages: the metal can be overheated and maintained
in melted state in high vacuum, the metal can be used in any form (powder, sponge, etc.), it is
possible to prepare super alloys. The use of high vacuum also leads to evaporation of certain
admixtures, which results in refining of the metal.
Niobium produced by powder metallurgy or electron melting in particular has good
malleability. It can be easily processed with all chipless processing methods, such as pressing,
drawing, rolling, etc. Cast niobium might contain inclusions. Common impurities, such as
carbon or nitrogen, reduce the niobium workability and increase its hardness.
5.6.3 Mechanical properties of niobium
Mechanical properties of niobium are highly influenced by the content of impurities. It
is strongly affected by elements that create interstitial solid solutions with niobium. These
include namely carbon, oxygen and nitrogen that might cause brittleness even at normal
temperature. The effect of oxygen is represented in Fig. 5.7. The deformation ability of
niobium is very negatively affected already at 0.03 % of carbon, 0.3 % of oxygen and 0.1 %
of nitrogen. Suitable alloying elements can be used to improve the mechanical properties of
niobium. Systems relevant for nuclear technology are binary systems Nb – V, Nb – Zr, Nb –
Mo, Nb – Ti and ternary systems Nb – Ti – Cr, Nb – Ti – Mo.
5.6.4 Niobium corrosion
In terms of corrosion, niobium is not very resistant in contact with air at higher
temperatures. Starting at 200 °C, a thin layer of oxides is created on the surface and it is stable
up to 400 °C. At a temperature exceeding 400 °C, the created Nb2O5 is porous and does not
protect the metal against oxidation. In addition to oxidation, oxygen is also diffused in the
metal, causing embrittlement of niobium. Niobium forms NbN and Nb2N nitrides together
with nitrogen. Absorption of nitrogen causes significant embrittlement of niobium already at
400 °C. Created nitrides are very stable and can be removed only by annealing in vacuum at
temperatures of 2000 °C. Hydrogen is also absorbed by niobium, which results in
embrittlement.
However, it is possible to significantly improve the resistance to oxidation by adding
V, Ti, Cr, Mo or W. Further improvement is also achieved by ternary systems. These alloys
still need to be protected by silicide coatings for long-term use in oxidation environments at
increased temperatures. Niobium alloys with zirconium or vanadium are suitable to water
vapour environment, where mostly zirconium alloys with relatively low strength values are
used. Ternary alloys Nb – Ti – Cr can also be used.
The behaviour of niobium in molten metals is very interesting – see table 5.4. Especially
important is the good corrosion resistance of niobium in liquid sodium or in the sodium –
potassium system. However, the corrosion resistance is conditioned by low oxygen content
(up to 5 ppm).
Fig. 5.7 Influence of oxygen content on hardness of unalloyed Nb
5.7 Vanadium
5.7.1 Properties of vanadium and vanadium alloys
Vanadium is a steel-grey, hard metal that can be ground and polished. Physical and
namely mechanical properties are greatly affected by impurities constituting interstitial solid
solutions. These include namely oxygen, nitrogen, hydrogen and carbon. These elements
cause brittleness of vanadium. In order to ensure vanadium is malleable, the maximum
content of oxygen and nitrogen combined must not exceed 0.2 wt. %. Carbon does not
significantly affect the hardness up to a content of 0.25. Mechanical properties of vanadium
may be modified by alloying. The disadvantage of most of the alloying elements is the fact
that the resulting alloys are usually brittle. Malleable alloys are V – Ti and V – Zr.
Vanadium has a relatively low density of 6.1 g.cm-3
and a high melting temperature
1910°C. It has a KPC lattice. In compact state, vanadium does not react with air, water and
alkaline hydroxides at a normal temperature. It is resistant to acids with oxidizing effects –
with the exception of HF. It dissolves in aqua regia and HNO3. It easily creates basic and
acidic radicals that can form the central atom in polyacid together with elements of groups IV
V. VI. A VIII. of the period table. Its most important compounds include oxides, chloride,
sulfates and sulfides.
5.7.2 Preparation technology of vanadium
The content of vanadium in the Earth's crust (0.2 %) is higher than the content of
copper, zinc and lead. However, the disadvantage is that it rarely occurs in rich finding sites.
Usually it is accompanied by other minerals and it often creates complex minerals, most
importantly:
Roscoelite KV2/AlSi3O10/(OH)2 – a mica containing approximately 32.4 % of V2O3. It
occurs in certain poor dikes, namely in the US.
Patronite V2S5 – contains 19 – 25 % of V2O5, can be found namely in Peru.
Vanadite – Pb5(VO4)3.Cl – contains 19.4 % of V2O5, located in the oxidizing zone of
lead-zinc ores.
Carnotite K2O.2UO3.V2O5.3H2O – contains 19.8 % of V2O5. Largest sites of these
minerals are located in the US.
Due to the fact that it accompanies iron ores in hundredths, it can also be produced
from slag, to which it passes during the iron industry and metallurgic processes. Various
production processes are used due to the different character of vanadium ores and materials.
Table. 5.4 Corrosion resistence of Nb in liquid metals.
liquid metal
melting
temperature
[°C]
immunity at temperature [°C]
300 600 800
Na, K, Na-K -12.3 – 98.3 good good good
Li 186 good good good
Mg 651 - good unknown
Hg -38.8 good good unknown
Ga 29.8 good bad unknown
Pb 327 good good good
Bi-Pb 125 good good good
Bi-Pb-Sn 97 good good unknown
Bi 271.3 good good good
5.7.2.1 Production of metallic vanadium One of the most used methods is the reduction of V2O5 using calcium. It is performed
in accordance with the following reaction:
5 Ca + V2O5 = 5 CaO + 2 V + 1463 kJ
The reaction is strongly exothermic and is moderated by CaCl2, which at the same time
transforms CaO to liquid slag. The process itself is realized in a steel bomb in a magnetized
crucible. The prepared vanadium is melted by the released heat and creates semi-melted
grains. The purity of the prepared metal is 99.9 %.
The purest vanadium may be prepared by thermal dissociation of VI2. The equipment
and principles are similar as for Ti and Zr. The reaction temperature for creation of VI2 is
900°C and the dissociation temperature is 1400°C.
Other preparation methods include reduction of VCl3 by magnesium. The reduction is
performed in a steel vertical retort (see fig 5.8). The equipment casing is cooled by water in
the upper part. Input VCl3 is added to the reactor from a reservoir located in the upper part.
Magnesium is put to the retort and it is heated to a temperature of 750 – 800°C in an argon
atmosphere. Then, VCl3 is supplied in such an amount so as to maintain constant reaction
temperature. After the process is finished, the crucible is quickly put to the furnace, where
vanadium is separated from other reaction products and is cooled at the same time. The
obtained metal in the form of a sponge is malleable.
Vanadium obtained by one of the above described methods may be further processed
by melting or by powder metallurgy. The selection of crucible material for processing of
vanadium by melting poses a certain difficulty. Common materials cause contamination of
vanadium and thus increase its hardness and reduce its malleability. It seems that the most
suitable method is melting in vacuum or arc melting under the protective atmosphere in a
copper, water-cooled mould. Powder vanadium is usually pressed at a pressure of 250 – 300
MPa and sintered at temperatures of 1400 – 1510°C. This method yields malleable metal.
5.7.3 Vanadium alloys and applications
Vanadium is applied namely in the production of alloys and compounds. It plays an
important role in steel production, where it is added in the form of ferrovanadium to
construction, tool and fire-resistant steels. In construction steels it creates finer grains (0.15 –
0.25 %), in tool and high-speed steels it creates very hard carbidic phases (1 – 2.5 %). Other
Fig 5.8 Scheme of equipment for the production of vanadium by reduction of
VCl3 by magnesium.
important application areas include namely powder metallurgy and chemistry, where it is used
as catalyzer in the form of V2O5.
This metal can also be used in nuclear technology due to its favourable properties.
Important properties of this metal include high melting temperature and good resistance in the
environment of various molten metals. Other important properties include namely low
sensitivity of mechanical and physical properties to radiation, good mechanical properties at
increased temperature and good thermal conductivity. The combination of these vanadium
properties creates good conditions for its use as a coating material.
5.7.4 Vanadium corrosion
From the resistance point of view of non-alloyed vanadium against oxidation absorbed
gases have a significant effect, mainly oxygen, nitrogen and hydrogen, which is manifested in
a temperature above 300°C. There is special significance in the joint reaction of vanadium
and oxygen. There exist a number of oxides, from which the greatest V2O5 oxide is melted at
a temperature of 675°C. Above this temperature there then does not exist any protective
effects of the surface layers, preventing oxidation. In a CO2 environment this metal is only
slowly attacked.
5.8 Yttrium
5.8.1 Yttrium production
This metal accompanies a series of minerals, mostly rare soils. In the form of
phosphates it occurs in xenotine and monazite, in the form of silicates in gadolinite, among
other important materials are samarskite and euxeniete.
After the electro-magnetic enrichment of oxides the separation of other KVZ is carried
out. The result of separation is Y2O3 with a purity of 99,9 %.
The production of metallic yttrium comes from pure halogen compounds of yttrium.
The following reduction reaction has the greatest significance:
2 YF3 + 3 Ca → 2 Y + 3 CaF2
This reduction is carried out after preliminary degasification or by the melting treatment of
fluorides in a vacuum. The reduction itself is carried out at a temperature of 1000°C and a
pressure of 1.33.10-2
Pa in a crucible made from tantalum.
The thermal breakdown of the YI3 Za formation of elementary yttrium is not possible
from a thermal dynamic point of view. For yttrium production fusible electrolysis is not also
suitable. The high content of oxygen in the metal is possible to reduce its distillation in a
vacuum at a high temperature, by purifying imput material before reduction, or through metal
treatment in YCl3 or YF3 – CaCl2 liquid melts.
5.8.2 Yttrium corrosion
Under normal conditions this metal is very stable in the air. This is because of the
presence of thin oxidized layers, which create very good prevension against oxidation. In
increased temperatures this leads a heavy metal attack, which further grows with the
increasing content of impurities in the metal. One oxidation mechanism does not occur for
higher temperatures. Yttrium is resistant against melted uranium and a whole series of its
alloys.
Summary of terms in this chapter (subchapter)
Cladding material
Compatibility
SAP
Dehafnization
Corrosion
Zircalloy
Questions to the covered material
Explain the main function of the cladding material and characterize its basic
properties.
Name the materials that are used as cladding material.
Briefly describe preparation technology of zirconium.
Why must be zirconium for cladding materials cleaned from hafnium.
6. Effect of radiation on material properties of nuclear reactors
Time to study: 5 hours
Aim After studying this section the student should be able to:
Formulate types of damage that may occur in the material due to irradiation.
Assess the impact of irradiation on fuel materials. Evaluate the impact of irradiation on cladding and construction materials.
Lecture
6.1 Precipitation processes caused by radiation
An electrically charged particle that enters the material loses its energy in two ways:
through excitation and ionization,
by transferring kinetic energy to individual building particles.
Quantity assessment of this phenomenon has shown that there is a certain critical energy for
interaction of heavy charged particles with the electron shell. If the "projectile" energy
exceeds this limit, then energy is transferred mainly by excitation and ionization. If it is below
this limit, then energy is transferred exclusively by elastic collisions. The ratio of energy loss
through excitation and ionization to energy loss through elastic collisions for charged particles
with energy in orders of MeV is approximately 103.
The amount of energy transferred at an elastic collision of the flying particle with the
nucleus is determined in the rest state (in resting, centre of mass system) based on the
collision law in accordance with the following expression:
∆E = E1 .4 M1 .M2
(M1+ M2) . sin2
ϑ
2
where: E1, M1 – energy and weight of the moving particle,
ϑ – angle at which the particle departs from its direction,
M2 – weight of stationary (rest) particles
The maximum energy that can be transferred during a head-on collision (ϑ ´180°) is
determined by the following relation:
∆𝐸𝑚𝑎𝑥 = 4E1 . M1 . M2
(M1 + M2)2
and the average energy transferred to grid particles during isotropic scattering is represented
by the following equation:
∆𝐸 = 2E1 . M1 . M2
(M1 + M2)2
This expression can be further modified for the most common situation, i.e. collision of a
neutron with the atomic nucleus. If M2 >>1, M1 = 1 (applies for neutrons), the average change
of neutron energy during the collision with atomic nucleus can be determined by the
following relation:
∆E = 2E1 .1 + M2(1 + M2)2
= 2E1M2
The above listed equations clearly illustrate the importance of the weight of particles entering
the mutual interaction. At identical energy, electrons can therefore transmit only a small part
of their energy due to their low weight.
The dependence of the transfer of average relative energy in an elastic collision of
neutrons, protons, alpha particles and aluminium nuclei with atomic nuclei up to a weight of
250 is illustrated in Fig. 6.1a. Similar dependences for electrons are represented in Fig. 6.1b.
Curves for protons and neutrons are identical due to their same weight. For heavy elements,
represented by aluminium in the figure, the transferred energy steeply increases to the
maximum value as the target nucleus weight increases. With further increase of the target
nucleus weight the energy transferred during the collision decreases.
Energy transferred during mutual collisions between incident particles and target
nuclei can cause displacement of atoms from grid positions, alternatively it may cause grid
vibration causing increase of thermal energy. Displacement of atoms from grid positions can
occur only if the energy transferred during the mutual collision is higher than the bond energy
that keeps the atom in equilibrium position in the grid. The average value of this bond energy
for metal, ion and covalent grids has been determined to approximately 25 eV. Based on this
value it can be claimed that heavy particles with energy for example of 1 MeV, can displace
atoms from their equilibrium positions and that these atoms can then cause displacement of
secondary, tertiary and possibly other atoms due to their relatively high energy. Electrons with
input energy of 1 MeV are capable of displacing atoms only in light nuclei (up to atomic
weight of 90). In general, these displaced atoms are not capable of creating secondary, not to
mention tertiary grid defects.
6.2 Damage zone in irradiated solid substances
If the energy transferred during collisions of a displaced atom is larger than the bond
energy of the grid atom, the displaced atoms might then release other atoms from their
positions by mutual collisions.
The trajectory of such displaced grid atom is marked not only by Frenkel defects. This
also causes local heating due to mutual collisions, during which energy lower than EV (bond
energy) is transferred. The estimated value of this local heating reaches several thousands K,
Fig. 6.1 a) Mean relative energy transmitted during mutual collisions of different particles with the atomic
nuclei of different masses, b) similar dependence for incident electrons.
whereas this temperature is produced for only a short period of time (10-10
– 10-11
s). Local
heating creates thermal tension, whereas Frenkel defects cause mechanical stress in the grid.
Since free paths are shortened between two consecutive mutual collisions, the
displaced atom creates a large amount of vacancies at the end of its path, which leads to the
creation of a dilution zone. The original theory assumed that this dilution zone is surrounded
by a zone that is saturated with atoms in interstitial positions (see Fig. 6.2). The created
configuration is transformed (bonding) due to the pressure of this shell and a grid similar to
the original one is created, whereas some of the atoms are relocated from their original grid
positions.
A comparison of the actual size of this affected zone with the calculated value led to
the conclusion that the actual dimensions of the affected zone are larger than the theoretical
assumptions. That is why the theory has been completed by so-called extended interstitials
and focusing collision mechanisms.
The original concepts assumed that mutual collisions were independent. The newer
theory stemmed from the possibility of existence of focused, as well as scattered collisions.
Both cases for a KPC grid are schematically represented in Fig. 6.3. The space between
vacancies and interstitials increases by focusing collisions, which leads to less frequent
mutual reactions. The defect penetration depth is affected by discontinuities of the grid,
mainly by grain boundaries.
interstitial atom
substitutional atom
Fig. 6.2 Crystal lattice defects at the end of track of stamped particles.
Fig. 6.3 Schematic representation of the collision mechanism in FCC lattice; a) focusing collisions, b)
scatering collisions.
a) b)
6.2.1 Focusing collision mechanism
In a simpler case the focusing collisions have the direction of close-packed
arrangement of atoms, for instance the <110> direction in KPC grid. In the last stages of the
cascade of collisions the particle diameter "grows" as the energy reduces. The diameter
gradually becomes comparable to the interatomic distance, which prevents "channel"
movement of atoms between grid planes and so-called focused grids start to appear. Mutual
collisions in close-packed atomic chains in crystals depend on the direction.
The theoretical model studies the effect of primarily displaced atoms during the
collision with fast neutrons, as is schematically represented in Fig. 6.4. The displaced atom
first uses part of its energy to excite electrons. Then it displaces other atoms from grid
positions on its path; it loses its energy by these collisions and towards the end of its path it
collides with almost every atom it meets. Around point P, where the displaced atoms finally
stops, an area with high density of vacancies is created. This vacancy slide is referred to as
the dilution zone.
6.3 Radiation effects on the properties of metallic uranium, its alloys and
compounds
6.3.1 Radiation growth
For polycrystalline material with a texture in the direction /010/, parallel with the
direction of deformation, it leads to a growth in uranium patterns in the same direction. The
dimensional change is possible to characterize by a co-efficient of radiation growth G:
𝐺𝑖 = ln(1/𝑙0)
𝑏
where: l, l0 – final and original pattern length,
b = number of fission atoms / total number of atoms = proportion of all fission atoms.
If it is possible to extend Δl a little, Gi is approximately given in the formula:
𝐺𝑖 = ∆𝑙/𝑙
𝑏
Fig. 6.4 Schematic representation of radiation damage of copper by fast neutrons, Seeger´s model.
Besides its given dependence of crystall direction, the co-efficient of radiation growth is
dependent on a series of factors, for example, burnt-out, the degree of material deformation,
temperature deformation, grain size, thermal treatment, etc. The dependence of the co-
efficient of uranium radiation growth in the direction /010/ at a temperature is captured in Fig.
6.5. There areas are evident here:
a) the area of great radiation growth in temperatures lower than room temperature,
b) the area of the negligible effect of temperature on G/010/ in a temperature range of 50 –
250°C,
c) the area of sharp G/010/ drop to zero at temperatures 400 – 500°C. This area is
considerably dependent on the fuel´s thermal load and at higher temperatures the
upper temperature limit of radiation growth increases.
6.3.2 Swelling
As already mentioned, swelling is an isotropic increase in volume, caused by the
gathering of fission products, mostly gaseous ones. It is considerable in high temperatures,
approximately above 450 °C.
It is necessary to be aware that uranium swelling prevents the deep burn-out of nuclear
fuels. Swelling is low at low radiating temperatures and even during a higher burn-out of 0.5
- 1 % it usually does not exceed 3 % of the original volume. In areas 400 – 500 °C, which is a
common operational temperature, most power reactors especially show considerable swelling
in higher values of thermal load (above 15 kW.kg-1
), which is already connected with crack
formation along grain edges. This dependence is documented in Fig. 6.6. In the given thermal
range areas of radiation growth and swelling overlap and it is supposed that great voluminous
growth is caused by the joint effects of both of these phenomena. Internal tension, brought out
by radiation growth in the individial grains, is released by a shift along the grain edges,
connected with the formation of intercrystalline cracks. From the surroundings fission, gases
gather into bubbles and expand in extreme cases up to a pressure equilibrium with the external
environment.
Fig. 6.5 Dependence of radiation-induced growth factor on temperature of irradiation.
Summary of terms in this chapter (subchapter)
Zone of damage
Focusing collision
Radiation growth
Swelling
Questions to the covered material
Characterize the basic forms of dimensional unstability of uranium. Causes,
consequences for the fuel material.
Briefly describe the nature of the damage zone in solids.
Briefly describe swelling in uranium.
Used literature that may be used for further study
KURSA, M., KUCHAŘ, L.: Kovy jaderných reaktorů. VŠB Ostrava, 1984, 399 s.
KUCHAŘ, L., DUŽÍ, P.: Základy jaderné metalurgie. VŠB Ostrava, 1985, 186 s.
Construction technologies for nuclear power plants. International atomic energy
agency, Vienna, 2011, ISBN: 978-92-0-119510-4.
MAJER, V.: Základy jaderné chemie. Praha, 1981, 612 s.
BEČVÁŘ, J.: Jaderné elektrárny. Praha, 1981, 634 s.
DUŽÍ, P.: Základy jaderné metalurgie – návody do cvičení. VŠB Ostrava, 1987, 122 s.
HABRMAN, P., KUCHAŘ, L.: Základy jaderné energetiky a bezpečnosti. VŠB
Ostrava, 1988, 72 s.
BEŇADIK, A., a kol.: Technologie jaderných paliv – kapitoly z keramických paliv.
VŠB Ostrava, 1989, 163 s.
KUSALA, J.: Miniencyklopedie jaderné energetiky. Energetická společnost ČEZ,
2004.
Journal of Nuclear Materials Murray R.L.: Nuclear Energy (Fifth Edition). Elsevier.
2001. ISBN: 978-0-7506-7136-1.
Nuclear Power Reactors in the World, 2012 Edition, International atomic energy
agency, Vienna, 2012, ISBN 978–92–0–119510–4.
Fig. 6.6 Influence of temperature od irradiation on swelling of uranium.
Fast reactor database, International atomic energy agency, Vienna, 2006, ISBN 92–0–
114206–4.
Zirconium in the Nuclear Industry, 15th International Symposium, eds. Bruce
Kammenzind and Magnus Limbäck, ASTM International, ISBN: 978-0-8031-4514-6.
Jaderná energie a energetika, Simopt, 2013, ISBN 978-80-87851-01-2.
Pressurized water reactor [online]. [25.8.2013]. Dostupné z <
http://www.if.pw.edu.pl/~pluta/pl/dyd/mtj/zal00/Zberecki/reaktory.htm>
Reaktor PHWR [online]. [25.8.2013]. Dostupné z
<http://proatom.luksoft.cz/jaderneelektrarny/index.php?akce=reaktor&idtypbloku=20
>
GCR, Magnox [online]. [26.8.2013]. Dostupné z
<http://jaderneinfo.webnode.cz/news/gcr-magnox/>
Gas Cooled & Advanced Gas Cooled Reactors [online]. [27.8.2013]. Dostupné z
<http://www.nucleartourist.com/type/gcr.htm>
Reaktor FBR [online]. [27.8.2013]. Dostupné z
<http://proatom.luksoft.cz/jaderneelektrarny/index.php?akce=reaktor&idtypbloku=10
>
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