Thorium Based Nuclear Reactors by nooryudhi

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									         Thorium Based Nuclear Reactors
                                By

              Abdulhafed A Mohamed Elkhadrawi




A dissertation submitted to the Department of Physics, University of
Surrey, in partial fulfilment of the degree of Master of Science in
Radiation and Environmental Protection.

                   Supervisor: Dr. P.H. Regan

                        Department of Physics

               Faculty of Engineering & Physical Sciences

                         University of Surrey

                            September 2008




                         Abdulhafed Elkhadrawi 2008
In the name of Allah (God) the most gracious, the most merciful.
Abstract
 In recent years, nuclear technology has principally been based on the use of fissile 235U and
239
    Pu. Although, the naturally existing thorium isotope, 232Th, can also be transformed to
fissile 233U nucleus following thermal neutron capture reaction, inadequate attention has been
paid to the its potential use in nuclear technology. This was probably due to geological
availability of natural resources of thorium and uranium. It was established that thorium is
three to four times more abundant world-wide than uranium and can therefore be an abundant
and sustainable resource. Also it is widely distributed in nature and can easily be utilized as
an energy resource in many countries. Thorium mostly occurs in the rare earth thorium
phosphate mineral, monazite, which contains about ∼ 12% high grade thorium oxide (ThO2).
Research and development activities are conducted in a number of countries and remarkable
progress is being made. Prototype reactors such as high temperature gas cooled reactors
(HTGR) and thermal breeding in light water breeder reactor (LWBR) have been
demonstrated. The development of thorium-based fuels has been motivated by the superior
nuclear properties of the thorium fuel cycle when applied in thermal reactors. The need for
proliferation-resistant, longer fuel cycles, higher burn-up, improved waste form
characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material
has led to renewed interest in thorium-based fuels and fuel cycles in some developed
countries. The utilisation of thorium in nuclear reactor and properties of thorium and uranium
are discussed in dissertation. Also, the types of reactors for thorium fuel, proliferation and
thorium fuel cycle and reduction of radioactive waste are illustrated. It is necessary to take
the potential problems with super-criticality, production of 239Pu and long lived minor
actinides connected with this cycle into account, in the establishment of the nuclear reactors
which depend on this cycle. This work presents options for thorium based cycles, both open
and closed, and the accelerator driven reactor (ADR) with thorium fuel. There are many
disadvantages at the start and end of the thorium fuel cycle. For example, the production of
more fission gas per fission, although, experience has shown that ThO2 is superior to uranium
dioxide (UO2) in retaining these gases. Also, irradiated ThO2 and spent ThO2-based fuels are
difficult to dissolve in HNO3 owing to of the inertness of ThO2. According to the
International Atomic Energy (IAEA), certain forms of thorium, such as thotitr, may be more
resistant to weapons proliferation. Thorium-based fuels can also be utilized in all existing
reactor types exclusive of drastic modifications in the design equipment. The historical and
current status of thorium based nuclear reactor is reviewed in this work. The Russian and
Indian design for this category of reactors are also illustrated.




                                                                                              I
Acknowledgements

My deepest gratitude and appreciation goes to all those helped in completing my project.
First of all, Dr. P.H. Regan who provided guidance, support and encouragement during the
whole period of the research and the course.
Secondly, I would like to give my sincere appreciation to my parents and step mother.
I would also like to give my greatest thanks to my brothers, sisters and friends without their
endless love and trust and support throughout the whole year, this research would not have
been possible.




                                                                                            II
                                                 Table OF Contents
Abstract................................................................................................................….................I
Acknowledgements.....................................................................................................….........II
Table of contents.................................................................................................…...............III
List of Tables………………………………………………………………………………....V
List of Figures..........................................................................................................................V
Glossary of Symbols and Abbreviations…..........................................................................VI

CHAPTERS

                                                        Chapter 1
1-Introduction................................................................................................................….......1
1.1 Brief on nuclear power – Future..................................................................….......1
Nuclear applications.................................................................................................................2
* -Power generation.................................................................................................................2
* -Heat applications.................................................................................................................3
* -Hydrogen generation...........................................................................................................3
* -Waste transmutation and plutonium burning..................................................................4
* -Other applications..............................................................................................................4
1.1.2 Evolutionary developments...........................................................................................4
1.2 Thorium..............................................................................................................................5
1.2.1 Definition and discovery of the thorium................................................................…...5
1.2.2 Occurrence thorium in nature.......................................................................................6
1.2.3 The utilisation of thorium in nuclear reactor………………………………………...6
1.2.4 Brief properties of thorium and uranium.....................................................................7
1.2.5 Binding energy and fission.............................................................................................9


                                                             Chapter 2
2- Nuclear fuel cycles............................................................................................................10
2.1- Thorium-uranium nuclear fuel cycle.....................................................................…...10
2.2- The types of reactor for the thorium fuel cycle………………………………………11
2.3 -Proliferation and thorium fuel cycle………………………………………………….12
2.4- Reduction of radioactive waste………………………………………………………..15


2.5- Implementation scenarios and options………………………………………………..18


                                                                                                                                       III
2.5.1 Open and close fuel cycle……………………………………………………………..18
2.5.1.1. Open fuel cycle……………………………………………………………………...19
2.5.1.2. Close fuel cycle……………………………………………………………………...20


                                                        Chapter 3
3- Historical and current status of thorium-based nuclear reactors…………………….22
. 3.1 Historical review of thorium use in reactors.........…………………………......……22
3.2. Recent advances in thorium fuel cycles for CANDU reactors…................................23
3.2.1 Direct self-Recycle In CANDU……………………………………………………….24
3.3 The industrial chemistry of thorium and uranium……………………..........…….…25
3.4 Indian and Russian design for thorium-based nuclear reactors…..............................25
3.4.1 Indian designs……………………………………………….……….......................…25
3.4.2 Russian-U.S. Program.......................................................................................…........27


                                                        Chapter 4
4. Discussion…………………………………………………………………………………29
4.1 Positives and negatives of thorium-based nuclear reactors.........................................29
4.1.1 Thorium's advantages……….......................................................................................29
4.1.2 Technical problems with thorium-based nuclear reactors........................................31
4.2 Access to fuels…………………………………………………………………………...32
4.3 Thorium-Plutonium MOX……………………………………………………………..32
4.4 Criticality of thorium…………………………………………………………………...33
4.5 Stability of fuel supply………………………………………………………………….35


                                                         Chapter 5
5- Conclusion......................................................................................................…................36
6. References................................................................................................................….......37
7-Appendices..............................................................................................................….........40




                                                                                                                                   IV
                                         LIST OF TABLES
                                                                                                 Pages
Table-1 Reactor applications and types…………………………………………………...3
Table-3 Critical mass for different plutonium compositions......…………….…………14
Table-4 Probability of indicated yields......……………………….……………...………15
Table-5 World thorium resources (economically extractable)............…………………27
Table-6    Comparison of the effect of MOX and thorium MOX fuel......…………….…33
Table-7    Critical mass for different plutonium compositions......………………………34


Appendices
Table-2    New reactor designs……………………………………………………………..40


                                        LIST OF FIGURES


Figure1-     Regional       distribution       of     global      nuclear       capacity       from      IAEA’s
projection……………………………………………………………………………………..2
Figure-2    Nuclear reactions in the 232Th, 233U-Fuel cycle...7
Figure-4    U, S and global coal combustion (millions of tons)…………………………...16
Figure- 5 U, S and global release of Uranium & Thorium...............................................17
Figure- 6 Time evolution of the radiotoxicity of high activity radioactive wastes per unit
energy     produced      by     reactors      (Thermal         giga    Watts       x   year)      for    various
cycles................……………………………………………………………….....………..….18


Appendices
Figure-3    Seed-Blanket Fuel Assemblies..................................................................…......41
Figure-7 Time-Average Bundle Power Distributions in a High-Power Channel...........41
Figure-8     Number <n> of neutrons available for breeding in the U-Pu and the Th-U
cycles with thermal and fast neutron spectra. Breeding is impossible negative values of
<n>…………………………………………………………………………………...........…42




                                                                                                                 V
                GLOSSARY OF SYMBOLSAND ABBREVIATIONS


Glossary
η          (eta) ratio of neutron yield per fission to neutrons absorbed
ADR        accelerator driven reactor
ADS        accelerator driven system
AHTR       advance high temperature reactors
AHWR       advanced heavy water reactor
appm       atoms per million
B/A        binding energy per nuclear
BWR        boiling water reactor
CANDU      CANadian Deuterium natural Uranium reactor
CHP        Combined heat and power
CRP        coordinated research project
EC         European Commission
EPR        European Pressurised reactor
EPRI       electric power research institute
FA         fuel assembly
FBTR       fast breeder test reactor
HEU        highly enriched uranium
HF         hartee-fock
HTR        high temperature reactor
HTGR        high temperature gas cooled reactor
IAEA        international energy agency
LMFBR       liquid metal cooled fast breeder reactor
LWBR        light water breeder reactor
LWR         light water reactor
MOX          mixed uranium-plutonium oxide
NASAP       non-proliferation substitute systems



                                                                           VI
PHWR   pressurized heavy water reactor
ppm    part per million
PWR    pressurized water reactor
RMS    root mean square
RTF    Radlkowsky thorium fuel
RTR    Radkowsky thorium reactor
SBU    seed-blanket unit
SEU    slightly enriched uranium
THTR   thorium high temperature reactor
WPU    weapons-grade plutonium
WWER   Russian pressurized water reactor




                                           VII
                                         Chapter 1

1- Introduction
1.1 Brief on Nuclear Power – Future
In some countries nuclear power is accepted as a safe and economic power resource with a
proven performance record which will be a necessary contributor to meeting present energy
demand. World-wide, there are approximately 439 power plants in operation. Most of these
are water reactors and the majority of water reactors are Pressurised Water Reactors (PWRs).
On the other hand, while countries accept that the current generation of plants should
continue to operate, there are very different attitudes with respect to new build [1]. The
International Energy Agency (IAEA) is currently forecasting a global average growth rate of
3% per annum over the next 20 years; with the figures for Asia are nearer 5 % at the present
time.
In Asia, there is a continuing ‘new build' programme, both of plants of current generation
design but also some plants of revolutionary design. On the other hand within Europe, there
are extremely varied attitudes. In Finland, anew revolutionary reactor is under construction.
However, in other countries such as Italy, there is a moratorium on building new plant. In
Central and Eastern Europe there are plants at differing stages of building that will come on
line in the next couple of next years. There are no new orders in the US, yet it would appear
that the tide is turning more towards nuclear power. Pressurised Water Reactors (PWR) are
likely to be favourites for new build; however advanced high temperature reactors (AHTRs)
are also attracting attention for instance in South Africa [1, 2].
The background of new nuclear capacity suggests high-increase projections for 2030.
According to the figure (1), the greatest rise in nuclear capacity (in terms of Gigawatts of
power added) will occur in the Far East, while the strongest increase in percentage terms will
occur in the Middle East and South Asia. Net capacity will also grow albeit less dramatically,
in eastern and Western Europe, while in North America is predicted to remain essentially flat
[1, 2].




                                                                                            1
Figure (1) Regional distribution of global nuclear capacity from the IAEA's projection
(source IAEA, 2004; McDonald, 2004 [2])


Several countries have embarked on long-term national energy strategy reviews, for example
UK, Finland and France. That of the UK is reasonably typical (see DTI Energy White Paper,
2003, DTI Energy Review, HMSO, 2006). The reviews covered all forms of energy
requirement, from power generation, through to transport and industry requirements. They
took account of the key issues including increases in greenhouse gas levels, decline in the
world fossil fuel supplies and other factors. The time scales considered were out to 2050,
although there are also near term targets of 2010 and 2020. By 2020 for example, a mix of
energy supply for electricity generation is envisaged, including continued reliance of coal
and other fossil fuels, but with a significantly increased reliance on imported gas and also
some reliance on renewables. Any 'nuclear’ capacity will need to compete with all these
other options. In 2007, The UK Government cleared the way for private energy companies to
put forward proposals for building new nuclear power stations in the UK [see Energy White
Paper, May 2007; Consultation Document, The Future of Nuclear Power, DTI, 2007 [3]).


1.1.1 Nuclear Applications
* Power Generation
Electrical power generation is the most significant application of civil nuclear energy. Many
different designs have been put forward and prototypes built since the dawning of nuclear
power [4].




                                                                                           2
* Heat Applications
There is a growing interest heat application and in particular Combined Heat and Power
(CHP), to improve the utilisation of the energy resource i.e. combined heat and electricity
generation. This has already been established in Russia for example with a dedicated heating
system as district heating. There is also considerable interest in using nuclear power for high-
temperature process heat applications, for hydrogen production and for oil refinement.
Another low temperature application is desalination of the sea water [4].


          Table (1) Reactor Applications and Types [4]


Application                                      Reactor types
Power generation                                 Majority of commercial present and future
                                                 generation of reactors.
Low temperature heat applications (district CHP or dedicated small reactors
heating/desalination)
High temperature applications                    Future water reactors and high temperature
                                                 gas reactors
Hydrogen generation                              Very high gas outlet temperatures ~ 1000°C,
                                                 achievable     with    high   temperature   gas
                                                 reactors.
Waste transmutation and plutonium burning        Mainly fast reactors
Other applications                               Small dedicated reactors




* Hydrogen Generation
Hydrogen technology is attracting considerable attention for transport sector applications.
Transport is a major contributor to greenhouse gas emissions and utilising hydrogen within a
fuel cell technology represents an attractive option. There are a number of international
programmes investigating hydrogen production techniques. A key objective is to generate
hydrogen without releasing carbon. There are known technologies such as high temperature




                                                                                              3
electrolysis and thermo-chemical water splitting. The required energy could be supplied by
nuclear for example, generated from high temperature gas reactor (HTGR) [4].
* Waste Transmutation and plutonium burning
An interesting application of nuclear energy is to process nuclear waste via transmutation of
plutonium and minor actinides from the waste. There are two major incentives for doing this.
This method can be used to transmute actinides to different radionuclides with much shorter
half-lives and hence reduce the radiotoxicity burden on the environment. Another important
aspect is to reduce stocks of weapons grade plutonium while producing useful energy.
Possible reactor types for actinide management are described below; these technologies
generally require fast neutron spectra [4].
* Other applications
Nuclear energy has other applications, for instance, in experimental reactors for
investigating materials testing, isotope production and physics training. Many of these small
research reactors were built during the early days of nuclear development and are now
reaching the end of their working lives. In Europe, there are European Commission (EC)
supported research programmes to address this issue. Other more esoteric applications are for
space, both for propulsion and energy systems for other planets such as Mars [4].
1.1.2 Evolutionary Developments
Part of the IAEA’s endeavour is to foster international collaborations that strive to enhance
the economics and safety of future water-cooled nuclear power plants. A 4-year IAEA
Coordinated Research Project (CRP) started in early 2004, entitled Natural Circulation,
focuses on the use of safety programme to be of assistance to meet the safety and economic
requirements of a new generation of nuclear power plants.
New evolutionary designs have been proposed for all the main water reactor types that are in
operation today. (Evolutionary means those designs which represent relatively small
modifications compared with present generation of plants). Average development costs will
be modest for example and there will not be a need for new major experimental research
facilities to justify the new design [5].
Evolutionary approaches have been essentially, those that utilise the best of existing systems
in present day plant, (such as they continue to use active systems) but attain better
redundancy, separation and diversity with improved design. Plants of this kind include the
European Pressurised Reactor (EPR). (Some information on new reactor designs is given in
Table 2 in the Appendices [2]).



                                                                                            4
1.2 Thorium
1.2.1 Definition and discovery of the thorium.
Thorium is a metallic chemical element and radioactive with the symbol (Th) and atomic
number 90. The isotope 232Th is thought to be about three times as abundant as uranium and
                                                                                232
about as abundant as the elements lead and molybdenum. In additional,                 Th has half-life
which is three more than the earth that is 14, Giga- years and is the most abundant of the
                      232
actinide elements.          Th is radioactive and it has been considered as substitute nuclear fuel
for uranium. All the thorium in earth’s crust could have more potential energy than both
uranium and fossil fuel reserves combined [6]. Uranium and thorium being are considered to
be the primary sources of the internal heat of the earth though their radioactive decay and
also both of them are actinide elements which are found in large enough quantities to mine.
Thorium was discovered in 1828. The black mineral was found on Lovoy Island Norway and
a piece of it was sent to Professor Jens Esmark. He also sent a sample to the Swedish
Chemist Jons Jabob Berzeius who then give it the name ‘Thoria toit’ or thorium-(Thor,
Scandinavian god of war). Then between the 1860s up to 1870s D.I. Mendeleev became
interested in the study of properties of thorium. He also placed it in the group of cerite metals
in the same group such as the rare earth elements.
Thorium was originally put in a group IV of the periodic table fitted due to its uniform
tetravalency in all easily reached compounds. (Even though lower valency states have been
observed, they are irrelevant to solution chemistry). By placement in the actinide chain, it is
the analogue of cerium in the lanthanum chain and shows a close chemical resemblance to
cerium (IV). Thorium ions show a larger acid reaction than lanthanum ions, indicating a
larger degree of hydrolysis. Many double salts are formed, some of them being isomorphous
with the corresponding cerium (IV) compounds [7].
Thorium was showed to decay at a fixed rate over time in to a series of other elements by
Ernest Rtherford and Frederick Soddy between 1890 and 1903. This observation led to the
recognition of half-life in their alpha particle decay experiments that led to their discovery to
the general theory of radioactivity.
In 1925 the iodide process was discovered to produce high-purity metallic thorium, by Anton
Eduard van Arkel and Jan Hedrick. The name ionium was given in the early study of
radioactive elements to the 230Th isotope which is produced in the decay chain of 238U (which
has a half life of 7.54 ×1010 years). Refined thorium from ores will contain trace amounts of
230
      Th. Its concentration is 16.4 ppm, and is in secular equilibrium with 238U [8, 9].


                                                                                                    5
1.2.2 Occurrence thorium in nature.
Thorium is about three times more abundant than uranium, and is about as common as lead.
In additional, soil commonly contains an average of found 12% parts per million of thorium.
The most common from that thorium occurs in is the rare-earth thorium-phosphate mineral
monazite which may contain up to about 12% thorium oxide. Thorium containing monazite
can be found in Antarctica, North America, Australia, Europe, Africa, and South America
(Ce) [10].

The majority of monazite is also associated a relatively small uranium content, the thorium to
uranium ratio being of the order of 1:20. This quantity depends on the age of the mineral and
on the proportion of the thorium and uranium. Alternatively monazite belongs to the
monoclinic syngony. Its solidness varies from 5 to 5.5, its particular weight from 4.9 to 5.5.
The colour of the mineral varies from pale yellow to reddish brown; nevertheless different
forms of monazite are found which are almost colourless, or dark-colored, olive-green,
cinnamon-colored or even black [10].
                                                                 232
Natural thorium consists mostly of a single isotope,                   Th with only trace quantities of other
                                                                                                       228
more radioactive thorium isotopes. In particular, an equilibrium concentration of                            Th is
                            232
created by alpha decay of         Th with half life 1.4×1010 years followed by the beta decay of
                                  228            228
short lived daughter nuclides           Ra and         Ac. The isotope 228Th half- life is 1.9 years and is
the precursor of a series of alpha, beta, and gamma emitting nuclides. In additional there are
two other isotopes of thorium that are found in nature in trace out amounts. The isotope 228Th
                                                                232
(Radiothorium, Ra-Th) which is a decay create of                      Th and thus found in trace quantity in
                                                                                          228
all thorium ore deposits, has a half-life of 1.913 years. It is assumed that                    Th is in secular
equilibrium with 232Th, giving a concentration of 0.000137 atoms of 228Th per million atoms
of 232Th (ppm), for this isotope. The second isotope is 230Th its decay product of 238U with a
half life of 7.54×104 years [10, 11].


1.2.3 The utilisation of thorium in nuclear reactor.
With a great quantity of potential fuel resources around the globe which up to now are still
not utilised optimally, thorium-based thermal reactors have a high internal change ratio. As
such they have the potential to be designed as long-life reactors without on-site refuelling
based on thermal spectrum cores [12].




                                                                                                                6
As mentioned previously, [IAEA, VIENNA, 1971] and shown in figure (1) the applicability
                                                                             232
of thorium as a power reactor is based on an (n, γ) reactions on                   Th. The consequent
           233                                                                                   233
nucleus,         Th, is unstable and decays by beta release with a half-life of 23,3 min into          Pa.
                                                                                                  233
This nucleus is another time beta-unstable and decays with a half-life of 27.4 d into                   U,
which is an alpha emitter with a half-life of 1.62×105 years. The         233
                                                                                U nucleus is fissile by
thermal neutrons. The average number of neutrons emitted per absorbed neutron ( ) is
smaller for 235U than for 233U at all significant neutron energies for reactors; below 40 keV
for 233U it is still as well as higher than for 239Pu [13].
                                          233
This superior nuclear properties of             U with respect to fission following neutron capture
leads to the possibility of overcoming the basic negative of thorium relative to uranium as a
reactor fuel, i.e. that natural-mined thorium has no fissile isotopes. What is more, benefits of
thorium as a reactor fuel include the fact that thorium is a metal, by virtue of its simple
crystal structure. Furthermore due to its simple stochiometry, ThO2 should resist irradiation
damage better than the corresponding uranium compounds [13].




Figure (2) Nuclear Reactions in the 232Th, 233U - Fuel cycle.[13]


1.2.4 Brief properties of thorium and uranium.
One of the densest metals is uranium (approximately 19g /cm3) being comparable to tungsten,
gold and the heavier elements of the platinum group. Uranium exists in three allotropic forms.
The alpha phase, which is stable below 668°C, is orthorhombic. Between 668° and 778°C the
beta stage, which is tetragonal, exists; at higher temperatures the gamma phase occurs. The



                                                                                                        7
beta phase, which has a compound unit cell, is fragile. The alpha phase which has certain
non-metallic features in its crystallography is highly anisotropic in most of its physical and
mechanical properties [10].
In general with other metals, many of the properties of uranium are also heavily influenced
by the structure. Because of the complications arising from this factor and from the effects of
contaminations and anisotropy, the data given below have to be treating with suitable caution.
Data to be used for calculation have to always be isotopes are found in normal uranium.
Their concentration and half-lives are as follows:
238
      U   99.28% 4.5 × 109 years
235
      U   0.71%    8   × 108 years
234
      U   0.005% 2     × 105 years
In common with uranium, thorium was not created in significant quantities in a reasonably
pure form until needed by the atomic energy programme. Additionally, the different steps of
reducing thorium and consolidating it into shapes create materials of variety properties were
not well developed. Relatively little quantities of thorium have been arranged in forms
typical of probable future uses. As a result data tend to be specific to laboratory specimens;
this is mainly relevant with regard to iodide-thorium and to a smaller extent, to cast thorium.
British expertise has been more concentrated on routes which fabricate powders and
consolidate these by controlled metallurgical techniques; these materials are less well
reported in the literature [10].
The oxidation of thorium is insignificant in still air at temperatures below approximately
200°C. The rate of oxidation rises speedily with higher temperatures. At 300°, 400°, and
500°C, respectively, weight gains of 0.03, 0.43 and 8.7 g/cm2/hr have been noted. On the
other hand, a value of 12 g/cm2 at 800°C has also been reported. While this hardly seems
consistent with the previous figures, it is possible that at moderate temperatures oxidation is
reasonably independent of temperature. This view is supported by experiments in oxygen,
[10] where it was shown that at temperatures above 450°C the furnace. Between 250° and
450°C the oxidation was found to follow a parabolic law with activation energy of 31
kcal/mol; between 350° and 450°C the oxidation was linear with time, with activation energy
of 22 kcal/mol. No information has been published on oxidation CO2 [10].
At 800°C quite high rates of oxidation in clean nitrogen and argon have been reported (4.0
and 1.5 g/cm2/hr. respectively); it is well known, nevertheless, that it is very difficult to




                                                                                             8
purify these gases to the extent required for experiments with such a reactive metal as
thorium at 800°C [10].


1.2.5 Binding Energy and fission.
One of the main application fields for fast neutrons is accelerator-driven subcrtical systems
(ADS) and fusion-fission (hybrid) reactor system for fission energy production. In these
reactor systems uranium and thorium are the nuclear fuels. The technical design of ADS and
hybrid reactor systems require much effort and input. The Hartee-fock (HF) method with an
effective interaction using Skyrme forces is widely used for studying the properties of nuclei
for instance, binding energy, root mean square (RMS) charge radii, mass radii, neutron
density, proton density, electromagnetic multiple moments and so on.
J. Belle and R. M. Berman explained that the nuclei of all heavy elements (A>90) are
energetically unstable for division into approximately equal parts. The binding energy per
nucleon (B/A) of the product nuclei is greater than of the parent nuclei. However, there is a
nuclear potential (coulomb barrier) which prevents naturally occurring heavy elements from
spontaneously decaying by fission. To induce fission of a given nucleus for a heavy element,
it must be adequately excited to overcome this nuclear potential. For several heavy nuclei
containing even numbers of neutrons and protons (234U, 236U, 240Pu, and 240Pu) the excitation
required for spontaneous fission can be provided by forming these nuclei through neutron
absorption by their odd A-Z even Z isotopic neighbours (233U,             235
                                                                                U,   239
                                                                                       Pu and   241
                                                                                                  Pu,
respectively) even when the neutron contributes zero kinetic energy to the excitation process.
This is due to the quantum mechanical pairing of nucleons in a nucleus which in an even-
even case is responsible for the formation of a large binding energy of the last neutron (Bn)
added to the nucleus as compared to that for its odd-even isotopic neighbour (note that Bn =
5.295MeV for 235U and Bn =6.5451 MeV for 236U)._ Accordingly, when an odd-even nucleus
is highly excited relative to its normal ground state.
When 232Th (Bn =6.4364MeV) absorbs a zero energy neutron, the resulting odd-even nucleus
233                                                                                             232
      Th (Bn = 4.7864MeV) will not be sufficiently excited for fission. As seen from the              Th
cross section for fission, considerable additional energy must be supplied through neutron
kinetic energy to initiate a fission reaction; The same is true for 238U [14].




                                                                                                       9
                                                       Chapter 2
2. Nuclear fuel cycles.
Nuclear fuel is created by uranium or thorium, which are naturally occurring elements.
                                                                                  232
Natural thorium contains only a single fertile isotope                                  Th; on the other hand, natural
                                                         235                                        238
uranium contains a fissile component,                          U, and a fertile element,                  U. For that reason,
                                  235
uranium, enriched in                    U, is a standard fuel for the Light Water Reactors of current
                                                                                                                          239
technology. Further fissile isotopes produced by transmutation of fertile isotopes are                                          Pu
       241                                   233
and       Pu (uranium series) and                  U (thorium series). These isotopes still play an important
role in the procedures of creation fission energy during the fuel burn-up procedures. These
artificial fissile isotopes are created and burned in situ, contributing much of the energy
generated by a nuclear fuel [17].
2.1 Thorium-uranium nuclear fuel cycle.
                                                                                          233
The chief fissile nucleus in the thorium-uranium fuel cycle is                                  U and fuel regeneration is
ensured through neutron capture on 232Th which offers lots of potential positives compared to
the better known uranium-plutonium fuel cycle. These include reduced high-activity long-
lived waste production and a reduced likelihood of nuclear proliferation. A brief description
                                                                233
of this fuel cycle is given below. The initial                        U supplies for such reactors use modern-day
pressurized water reactors and a thorium and plutonium mixed oxide fuel [15].
232
      Th in the thorium fuel cycle absorbs a neutron in either a fast or thermal reaction, The
233                             233                                        233
      Th beta decays to            Pa and then subsequently to                   U, which is used as the fission fuel.
        238        232                                         232                                               233
Like          U,         Th is a fertile material. The               Th absorbs a neutron to become                    Th which
decays to 233Pa. 233Pa in turn decays with a half-life of 27_days to 233U. The 233Pa is extracted
and protected from neutrons in order to enhance the breeding ratio [16].
Grainger explained that 233U is formed by the capture of a neutron on 232Th. 233Th decays by
beta-emission (beta-particle energy of ~ 1.24MeV, with a half life of ~ 23 minutes) to 233Pa.
         233
Next           Pa decays in additional by beta-ray and gamma-ray for total of 0.571 MeV, with a
half life of 27.4 days to 233U (half-life is 159,200 years), as shown schematically below [9].




                                                                                                                                10
                 Fertile      n + 238U → 239 U
                                   92     92                                      n + 232Th→ 233 Th
                                                                                       90     90


                                   ↓ 23.5 min                                          ↓ 22.3 min
                            239                                                  233
                             93   Np + β + ν                                      91   Pa + β + ν

                                    ↓ 2.35d                                              ↓ 27.0d
                              239                                                  233
       Fissile                 94   Pu + β + ν                                      92  U + β +ν

                            24110 years                                      ↓ 159200 years


 Other parasitic reaction:


232                   233             233                        233
      Th + Fast n →     Th * →              Pa (protactinium)→     U (fissile)

Th/U/ Pu cycle
It appears that development rate of nuclear power dose necessitates the growth of breeder
reactor concepts. After that such a growth it is possible to constitute a stock of 233U fuel that
could lead to an increase of a fleet of thorium-uranium reactors. Such a concept has many
benefits, in particular relating to the reduction of radioactive waste and the risks of nuclear
propagation [9].
2.2 The types of reactor for the thorium fuel cycle.
Fuel breeding can be achieved using both fast neutrons in and also with slow neutrons by the
thorium-uranium fuel cycle. With fast neutrons, it is more tricky than with the U-Pu fuel
cycle; the primary 233U inventory is huge, on the order of five metric tons for a 3 GW thermal
power reactor. This represents the in-core inventory, to which the uranium in the fuel
reprocessing system should be added (uranium in the processing unit and uranium on standby
pending reprocessing). As a result, fast neutron thorium-uranium reactors will also be
isotope-breeders. They possibly will use the same concept as for the fast neutron reactors
based on the uranium-plutonium fuel cycle, such as with a molten metal (sodium or lead)
coolant. [15]
Thorium-uranium breeder reactors with slow neutrons require only a small233U inventory, of
the order of one of metric ton. Their theoretical doubling time is the same as to that of
uranium-plutonium fast neutron breeder rectors. On the other hand, fission products are more
efficient at poisoning slow neutron reactors than fast neutron reactors. Therefore, to maintain
a low doubling time, neutron captures in the fission products and other elements of the
structure and coolant have to be reduced.



                                                                                                      11
In the 1960s, an elegant theoretical solution to this problem was proposed, namely, a reactor
in which the fuel is a molten salt which also serves as the coolant.
Neutron capture on the fission products would be limited thanks to on-line salt reprocessing,
at the price of additional complexity. For this reason the reactor is also a chemistry factory.
Giving up the low doubling time objective opens the way to molten reactors with drastically
simplified on-line fuel procedures or to other reactor styles for example, those with in-
operation fuel loading/ unloading such as heavy water rectors (CANDU) or gas-cooled
pebble-bed reactors. A particularly interesting scheme would consist in complementing a
thorium-uranium reactor fleet with fast neutron reactors with a uranium-plutonium core
                                                            233
encircled with a thorium blanket that could fabricate the         U needed in excess to extend an
existing thorium-based rector fleet [15].
2.3 Proliferation and thorium fuel cycle.
In recent years many Th-based fuel design options have been investigated which have
demonstrated the basic feasibility of Th-based fuel cycles for LWRs of new and next
generation technology. Homogeneous and heterogeneous are the two main design variants
were considered. The heterogeneous design adopts a seed-blanket assembly where the
uranium and thorium fuel parts are spatially separated. The homogeneous design adopts a
combination of ThO2 and UO2 with sufficient enrichment and uranium volume fraction to
obtain the required cycle length and burn-up. The consequences of the detailed fuel cycle
examination confirm that the homogeneous design does not provide optional, enough
performance of the fuel cycle as considers decrease of proliferation potential, fuel application
and economics [18].
The concept of the heterogeneous design known as Radkowsky Thorium Fuel (RTF), is
based in-part on the ideas and skills of the Bettis Atomic Power Laboratory’s Light Water
Breeder Reactor (LWBR) program as implemented and successfully demonstrated at the
Shippingport reactor in the late 1970s.
The RTF is a new fuel concept, not a new reactor that builds on the successful LWBR
experience. Furthermore, to decreasing the proliferation potential of the standard nuclear fuel
cycle and reducing the needs for spent fuel storage and disposal the design is subject to the
following constraints.
   1- Retrofit table into existing PWRs/VVERs with minimum changes to existing
       systems/hardware.
   2- Operational and safety characteristics comparable to those of existing LWRs.
   3- Competitive economically.


                                                                                              12
In the seed and blanket regions the fuel-to-moderator ratios are different. These are optimized
to reduce Pu production in the seed and improve 233U production and burning in the blanket.
The uranium utilised in the seed region is enriched up to a maximum of 20%, which is
accepted as the non-proliferative limit. Two extra items enhance the non-proliferative
characteristics of the RTF fuels:
1- In addition to reducing the fabrication of Pu by a factor of approximately 5-7 relative to a
                                                                                             238
standard PWP/VVER, the resulting plutonium which is created has a high content of              Pu,
240
      Pu, and 242Pu the consequences of which are impractical for use in a nuclear weapon.
2- It employs a once-through fuel cycle with no reprocessing, with the bred 233U burnt in situ;
                    233
furthermore, the          U that is created is denatured by the initial admixed uranium isotopes in
order to force isotopic separation (which is impracticable) should extraction and use of the
bred 233U be attempted [18].
“A cross section of the currently proposed RTF seed-blanket unit SBU design is shown in
Figure (3) in Appendices. The seed-blanket unit is a one-for-one replacement for a
conventional 17×17 PWR fuel assembly; as shown in the figure(3)” [18].
Galperin [9] illustrated that for nuclear power to be the main source of energy in the future
the proliferation issue associated with spent fuels for nuclear power has to be accepted. In
additional this must be based on a fuel cycle which is highly proliferation-resistant. They
concluded that the non-proliferative nature of the nuclear power fuel cycle material flow
should be supported by a combination of administrative defence measures and by avoiding
the creation of any material of such quantity and quality in the fuel cycle as to be of potential
use as a weapon. In addition they recorded a further factor is the calculatation of the
complexity necessary to divide the fissile component from the standard material flow of the
fuel cycle.
Also, from the same reference, recalling an earlier review on proliferation and its conclusion
that “the extensive non-proliferation substitute systems assessment program (NASAP) study
concludes in 1980 that none of the existing or proposed fuel cycle schemes were immune to
the possibility of proliferation. Because the main proliferation potential is associated with
plutonium (Pu), produced by alteration of 238U, Thorium presents a natural substitute fertile
material.” Therefore, they emphasised thorium fuel can be a promising proliferation-resistant
alternative to the 238U /Pu cycle as a future energy source.
Galperin [9] also explained that the requirements for evaluating that the fissile material
weapon quality by considering these following properties:



                                                                                                   13
   1- Weapon yield degradation due to pre-initiation caused by spontaneous fission
       neutrons.
   2- Critical mass. The critical mass is different for different isotopic compositions of
       plutonium and uranium.
   3- Weapon stability degradation by heat emission.
Based on these requirements they provide an approximate comparison of the critical mass for
different materials is presented in Table 3. It is clearly demonstrated that a relatively small
critical mass is achieved with any plutonium composition, and the RTR-Pu requires 20 up to
50 percent more material compared with the weapon-grade material [9].


           Table – (3) Critical mass for different plutonium compositions



                 Pu source                                 Critical mass (Kg)
                 Weapon grade                              4.3
                 PWR grade                                 5.5
                 RTR-seed                                  5.9
                 RT-blanket                                6.5



It was reported that in the LWBR design which is currently being developed in a more
deliberately proliferation-resistant way the central seed region of each fuel assembly will
                                  235                                                           238
have uranium enriched to 20%            U [20,21]. The blanket will be thorium with some              U,
                                                                          233
which means that any uranium chemically separated from it for the               U is not useable as a
                                               232
weapon. Spent blanket fuel also contains             U which decays rapidly and has very gamma-
                                                                                               238
active daughters. Plutonium produced in the seed will have a high proportion of                   Pu
generating a lot of heat and making it even more unsuitable for weapons than normal reactor-
grade plutonium [21].
Gaperin et al [9] in describing the value of thorium fuel in this respect further asserted that
the probability that an explosive device constructed from RTR-Pu will deliver a nominal
yield is small to negligible and the probability of a fissile yield is relatively high, Thus, they
conclude, it is shown that the RTR-Pu will produce an unreliable weapon as show in table
(see Table 4).




                                                                                                     14
           Table 4: Probability of indicated Yields [9]


 Yield          Super grade     Weapon          PWR grade       RTE-seed       RTR-blanket
                (Trinity)       Grade Pu        Pu              Grade Pu       grade Pu
 Nominal        0.88            0.68            0.07            0,006          0.0002
 Fissile        0.02            0.06            0.35            0,55           0.74


An additional barrier for a possible diversion of a reactor grade material is the heat emitted
by its isotopes. The thermal power produces an increase of the temperature of a device which
makes it unusable for weapons [9].


2.4 Reduction of radioactive waste.
As the population has increased worldwide, coal combustion continues to be the dominant
fuel source of electricity. In 1970, fossil fuels′ share decreased from 76.5% to 66.3% in 1990;
however nuclear energy′s share in the global electricity distribution has increased from 1.6%
in 1970 to 17.4% in 1990. Although U.S. population growth is slower than worldwide growth,
its per capita consumption of energy is among the world’s highest.
During the first quarter of the 21st century, many plants may be retired as current nuclear
power plants age, although some may have their operation extended through license renewal.
As a consequence several nuclear plants are likely to be replaced with coal-fired plants
except where it is considered feasible to replace them with other fuel sources for example,
solar energy and natural gas. The would coal resources have been estimated to be
approximately 1500 years at the current use rate Figure (4) shows the increase in U.S. and
global coal combustion for the 50 years prior to 1988, along with projections beyond the year
2040.




                                                                                             15
` Figure (4) U, S and global coal combustion (millions of tons)(taken from ref. [22]).


Indicated above, using the concentration of thorium and uranium figure (5) demonstrates the
historical release of these elements and the releases that can be expected during the first half
of the next century, given the predicted increasing trends. Using these data, both U.S. and
worldwide fissionable 235U and fertile nuclear material releases from coal combustion can be
estimated.




                                                                                             16
  Figure (5) U.S and global release of Uranium & Thorium (from ref. [22]).
Assuming coal contains thorium and uranium concentration of 3.2ppm and 1.3ppm,
respectively each typical plant released 5.2 tons of uranium and 12.8 tons of thorium in 1982.
Sum U.S releases in 1982 from 154 typical plants amounted to 1971 tons of thorium and 801
tons of uranium. These figures account for only 74% of the release from combustion of coal
from all sources.
In the United States based on the predicted combustion of 2516 million tons of coal and
12,580 million tons worldwide in the period to 2040, cumulative releases for the 100 years of
coal combustion following 1937 are predicted o be:
   •   U.S. release from combustion of 111,716 million tons.
   •   Uranium, 145,230 tons containing 1031 tons of 235U.
   •   Thorium, 357,491 tons
   •   Worldwide release from combustion of 637,409 million tons.
   •   Uranium, 828,632 tons containing 5883 tons of 235U
   •   Thorium 2,039,709 tons [22].
The radiotoxicity of radioactive wastes after a few centuries is due essentially to that of the
heavy (Z > 92) alpha contributing radioactive nuclei that are created by consecutive neutron
capture in the heavy elements present in the nuclear fuel.




                                                                                            17
The uranium and plutonium are in general completely used up after fuel processing in
breeder reactors based on the uranium-plutonium cycle. They contribute at a very small level
to the radiotoxicity of the wastes.
The thorium-uranium fuel cycle is even better in this perspective due the fact that, since the
                 232                                     238
mass number of         Th is 6 units less than that of         U meaning that the production of minor
actinides (neptunium, americium, and curium) which are the main contributors to the
radiotoxicity of the wastes in the-uranium-plutonium cycle is drastically reduced. During the
first 10.000 years the radiotoxicity of the wastes of the uranium-plutonium cycle is much
larger than that in the thorium-uranium as shown in Figure (6). In addition, the reduced initial
radioactivity would allow significant reductions on the size and, as a result, the cost, of
geological storage [15].




     Figure (6) Time evolution of the radiotoxicity of high activity radioactive wastes per
unit energy produced by reactors (Thermal gigaWatts x year) for various fuel cycles [15].


2.6. Implementation Scenarios and Options.
2.6.1 Open and close fuel cycle
                          232
Converting the fertile      Th in to the fissile 233U is the first stage in order to use thorium for
                                                                           233
nuclear power on a large scale. The subsequent utilisation of                    U is imaginable in the
following scenarios.




                                                                                                    18
1- Open fuel cycle based on irradiation of 232Th and in situ fission of 233U, without involving
chemical separation of 233U.
2- Closed fuel cycle based on chemical reprocessing of irradiated thorium-based fuels for the
recovery of 233U and refabrication and recycling of 233U bearing fuels [23].
2.6.1.1. Open fuel cycle.
Reference [IAEA-TECDO1450, Vienna: IAEA, 2005] explain that the engineering processes
and other complication associated with reprocessing and reproduction of highly radiotoxic
233
      U-based fuels can be ignored by using the open fuel cycle. The Radkowsky proposal of the
light water reactor is an example of thorium utilization in the open fuel cycle [23]. In
addition, it is suitable to fit the Russian WWER-T (thorium) reactor concept. The essence of
the core layout of an example concept is that every fuel assembly (FA) is prepared up of a
central seed with fissile material (medium enriched uranium, plutonium) and thorium blanket.
During refuelling the seed components are regularly replaced as compared to the FAs.
Separation of seed and blanket, optimization of moderator (water) to fuel ratio and the very
long fuel campaign (900 and 2,620 effective full power days for seed and blanket
respectively ) offer the possibility of such a system, reaching up to approximately 40% of
                                             233                            233
power to be produced by the fission of             U. The in situ use of          U and the feature of
avoiding handling dirty 233U outside the core is the aim of the open fuel cycle in introducing
thorium into a nuclear power reactor [23].
The likelihood of burning of weapons-grade plutonium (WPu) grouped with thorium in light-
                                                                                     239
water reactors of WWER-1000 category to burn and not breed                             Pu is another
encouragement to use thorium in a once- through thorium fuel cycle. For this mixed thorium
plutonium oxide having nearly 5% PuO2, could be used as a driver fuel. The removal of
uranium from the fuel composition causes a substantial increase in the rate of plutonium
burning compared to the use of standard mixed-uranium plutonium-oxide (MOX) fuel. It is
not only degraded to perform the standard burn-up (nearly 40 MWdays/kg HM) of LWR fuel
from the spent mixed thorium plutonium oxide in terms of WPu content but furthermore can
                                                                     232
be converted into proliferation-resistance with the formation of           U which has very strong
gamma emitting daughter products [23].
 In addition, the stock of civil plutonium could be considerably reduced by using the same
combination with thorium in WWER-1000 type reactors. Outside of any main modification
of core and reactor operation a direct replacement of low-enriched uranium oxide fuel is
possible by mixed thorium plutonium oxide fuel. In the instance a reactor, there is no require


                                                                                                   19
ment for using burnable absorbers in the form of gadolinium, integrated into the fuel since
       240
the      Pu isotope is present in the essential quantities in civilian grade plutonium and is a
good burnable absorber. Two layouts of the WWER-1000 reactor core with full loads of
mixed thorium-plutonium fuel are taken into account. In the standard core, all fresh fuel
assemblies out of 54 are located at the periphery and in the modified core only 12 fuel
assemblies out of 54 are sited in the periphery row. Around 1694kg of civilian grade
plutonium will be consumed by typical WWER-1000 reactor core with full load of thorium
plutonium oxide. The plutonium unloaded from these reactors will have isotopic containing
28% 239Pu, which could be used of disposed only in fast reactor. Some 300kg of 233U will be
created in the spent fuel, which would contain approximately 3500ppm 232U. The decrease in
the neutron flux on the reactor vessel is one of the main positives of such thorium plutonium
mixed oxide fuel in WWER. LWRs using combination of plutonium and thorium oxides
contain enhanced safety characteristic compared to those with enriched uranium oxide [23].
Weapons-grade Pu and civilian Pu both could be efficiently incinerated in combination with
thorium as mixed-thorium-plutonium-oxide fuel containing 20-30% PuO2 in commercial
LMFBRs. 70-80% of PuO2 content in (Th, Pu) O2 fuel could be reached in small LMFBR
cores such as the demonstration category FBTR in India [23].
2.6.1.2. Close Fuel Cycle.
The important steps of the closed fuel cycle are the reprocessing of irradiated Th-based fuels
                                       233
and separation of converted                  U. In this state WWER-1000 are the same as LWRs and
                                                                 233
therefore can be considered such a converter for                       U by utilising mixed thorium plutonium
                        233                   233                                                     233
oxide fuel. The               U content in          U is a significant part for recycling the               U formed in
LWRs. For a standard burn-up of 40 MWd/kg HM for a WWER-1000 fuel, the 23U content
would be in the range of 3000ppm. The two recycling portions below are as follows:
1-The use of (232Th -233U) O2 fuel.
2-The use of (Depleted U-233U)O2 or (Reprocessed U fromWWER-233U) O2.
The second option enables a smooth change over to the thorium fuel cycle with the least
modification in reactor design and in the technology of handling the spent fuel. On the other
                                                                                                233
hand, the utilization of depleted/ reprocessed uranium in mixture with                                U is not clean
                               235                                           233
thorium cycles since             U is also being utilized along with               U and there is also build up of
239             238
      Pu from         U. Moreover, not including build-up of minor actinides and plutonium and
minimising the radiotoxicity of disposed wastes, recovery of 233U with 232U does not use the
major benefits of thorium fuel cycle. The water temperature co-efficient of reactivity will
shift to the positive region due to the substitution of 235U by 233U in WWER-1000 reactor fuel.


                                                                                                                    20
                     235
However, when              U is replaced by plutonium, the shift in the temperature co-efficient of
reactivity is in the negative region. For that reason, it is possible to sensibly join plutonium
      233
and         U in the fuel composition in such a manner that the safety requirements concerning
temperature coefficient of reactivity is met, with addition of plutonium to compensate for the
reduction of 233U [23].
                                                                   233
In contrast, with the mixed version, divided allocation of               U and Pu seems to be more
desirable regarding improved efficiency in reactor control, lower neutron flux on reactor
vessel and relatively simpler fresh fuel fabrication and reprocessing of spent fuel. Transition
                                                                          232
to a tight lattice in WWER-1000 raises the conversion ratio of                  Th -233U fuel but cannot
convert the reactor into a thermal breeder reactor like the Shipping Port Light Water Breeder
Reactor [23].




                                                                                                     21
                                       Chapter 3
3- Historical and Current Status of Thorium Based Nuclear Reactors
3.1 Historical review of thorium use in reactors
Total core demonstrations of thorium-uranium oxide fuels in LWRs, in the period between
1960s and 1970s, were examined in two sorts of preparations:
1- Mixing thorium oxide with highly enriched uranium oxide in a uniform lattice (the
BORAX_IV, Indian Point I PWR, and EIK River BWR reactors).
2- Utilising a heterogeneous preparation of blanket and seed, where the blanket and is
responsible for the in-core fissile generation (second generation core of Shippingport
Reactor). The Shippingport test was intended to confirm the feasibility of net breeding of
fissile isotopes in the core and is usually referred to as the Light Water Breeder Reactor
(LWBR) [24].
From 1977 until 1982 the Light Water Breeder Reactor (LWBR) program at the Shippingport
station confirmed the Pressurized Water Reactor concept for business power generation. The
consequences confirmed that the ratio of the fissile content of the fuel at the end of operation
to that at the beginning of operation was about 1.0139. Moreover the effort identified a few
shortcomings in the LWBR technology compared to LWR practise at the time including (i)
lower power density of the (30%); (ii) the requirement for high 235U enrichments in the early
phase of deployment; (iii) the much more complicated design of a movable seed region; (iv)
the more complicated recycling of uranium and thorium relative to recovery of uranium and
plutonium; and (V) the extra shielding needed in the fabrication process. The EPRI
commissioned a study of the possible improvements in the nuclear fuel cycle if thorium is
included but with the least modifications of recent LWRs in the mid 1970s. The balance of
opinion from the EPRI sponsored study on thorium cycle utilisation in Combustion
Engineering System 80 PWRs can be briefed as follows [24].
1-Comparison of the characteristics of uranium and thorium based cores points out that
thorium fuelling is feasible and modifications to a PWR designed to accommodate plutonium
recycle do not seem to be necessary.
2- The introduction of a totally new system of advanced converters into the US would
probably require more effort and funding than can be justified.




                                                                                             22
3- Use of thorium with recycling can eventually raise the energy output per mined ton of
uranium by about 85% beyond the once-though uranium cycle and by 22% beyond
plutonium cycle.
4- The thorium cycle may not be economically attractive to cycles with poor fuel
conservation features even with the above aspects taken into account, owing to the early
years’ fuel demand being high and any savings occurring in later years [24].
Some reactor types other than the LWRs have experimented with the use of thorium. The
mainly notable of these is the first gas cooled, graphite moderated reactor in the US (Peach
Bottom, 40MWe, 1967-1969) and the first pebble bed reactor in Germany (AVR, 15MWe,
1966-72). The AVR reactor confirmed the capability of the thorium fuel encapsulated by
pyrolytic graphite to operate up to burn-ups about 100MWD/kg. The industrial follow-up
reactors (the 300MWE Fort St. Vrain in the US and THTR in Germany) did not prove to be
successful enough to generate future orders.
By contrast today, gas cooled reactor experiments are being constructed in China and in
Japan, but they are not stressing the thorium cycle as greatly as the plant technology. In
South Africa the gas –cooled electricity generating plant under study does not use thorium in
the cycle. Only in India are thorium-fuelled cores being stressed (to utilize the great reserves
of thorium in India), but not in graphite moderated reactors, [24].
3.2. Recent Advances in Thorium Fuel Cycles for CANDU Reactors.
An evolutionary approach to exploiting the energy potential of thorium can be provided by
the ‘once through’ thorium fuel cycle in CANDU reactors. The central 8 elements in a
CANFLEX1 fuel bundle in the mixed bundle contain thoria; however the outer most 35
element contain slightly enriched uranium (SEU). In existing CANDU reactors, detailed fuel-
core fuel-management simulations have shown that this approach can be successfully
implemented. The natural uranium fuel cycle is higher than uranium requirements. In
additional, by recycling the irradiated thorium fuel containing 233U, the energy can be derived
from the thorium as is, without any processing into the centre of a new mixed bundle.
Recycling of the central 8 thoria elements results in an additional burn-up of 20MW. d/kgHE
for each cycle from the thoria elements. These thoria elements are remains remarkably
reactivity constant over irradiation for each recycle [25].
The uranium requirements are 35% after the first recycle, lower than for the natural uranium
cycle and remain fairly constant with further recycling (the total uranium requirement
averaged over a number of cycles is 30% lower than natural uranium fuelled CANDU
                                                              233
reactor). This thorium cycle stage creates a stockpile of           U, which is safeguarded in the


                                                                                               23
spent fuel and which could be recovered in the future in depending on or predicated by
economic or resource considerations.
High-neutron efficiency enables the maximum energy to be derived from the thorium fueld,
thus reducing the uranium requirements. In addition, high neutron efficiency also opens the
door to a variety of fuel cycle strategies that would not otherwise be possible.
On the other hand, by minimising the amount of structural material associated with the fuel,
the simple fuel bundle design contributes to the high neutron efficiency of the reactor. Also
the simplicity of the fuel design increases the fuel cycle flexibility. Since thorium has no
fissile isotope, neutrons must be initially provided by adding a fissile component, either
                                                                                        232
directly to the ThO2 itself, or outside as separate driver fuel, to transmute the             Th to
valuable fissile 233U. The manner in which this is done defines a variety of thorium fuel-cycle
option in CANDU reactors. [25].
There is the opportunity to build into the design of these cycles a very high degree of
proliferation resistance since the thorium fuel cycle is not commercially employed right from
the start. This would apply to all parts of the cycle, from the supply of the fissile material
(used to initiate the cycle), to design of the recycle technology, to the supply of any fissile
components required as additional for the recycled material. What is true for the thorium fuel
cycle could also be employed to effectively burn up surplus weapons-material (Plutonium or
HEU), while at the same time creating a valuable source of fissile material for future
generations which is safeguarded in the spent fuel [25].
3.2.1 Direct self-Recycle In CANDU.
As stated by Kalakrishnan et al., [25], “This paper extends the previous work by examining
the effect of reusing the central 8 thoria elements after irradiation, into the center of new
mixed bundle containing fresh SEU in the outer 2 ring. Therefore, it is an addition of the
once-through cycle, to a recycle option that does not involve reprocessing. It is called direct
self-recycle, owing to the fact that irradiated fuel elements would be straight transferred into
a new fuel bundle with out any modification to the elements”.
This is analogous to DUPIC cycle involving as an alternative to reprocessing from a PWR
into a CANDU, direct self-recycle in the CANDU. This recycle selection would have the
highest degree of proliferation resistance exclude any chemical separation and does not
require access to the fuel pellets or substitution of the fuel elements. In addition, it would be
cheaper than reprocessing technology. Figure (7) in appendices gives information of the
time-average bundle power distributions in a high-power channel [25].



                                                                                                24
3.3 The Industrial Chemistry of Thorium and Uranium.
The motivation for a renewed study of the elements in fission residues from the discovery of
atomic fission revived the importance of inorganic chemistry, which had been overshadowed
for many years by improvements in physical and organic chemistry [7].
Apart from these practical connections between the elements, there is some theoretical
justification for considering thorium and uranium at the same time as the lanthanons. The
latter are sometimes calledƒ–type transition elements because the increase of atomic number
from one element to the next is accompanied by the addition of electrons to states inside the
4-ƒ, rather than to the 5-d atomic shell despite the fact that the 5-s and 5-p levels have
already been filled. The complete recognition of every member of the lanthanon family has
taken nearly 160 years [7].
Mendeleev's periodic classifications of the elements have done much to explain the situation.
It does not account for all the trivalent elements into which crude yttria and crude ceria had
been resolved and puts them together in one gap in the table, giving no guide to their total
number. However, it was obvious that scandium and yttrium occupied two well-defined
places in group III and was followed by lanthanum with an uncertain number of other
elements, while zirconium, titanium and thorium occupied positions in group IV and niobium
and tantalum belonged to group V.
With regards to uranium and the thorium, their chemistry is sufficiently distinct as to require
separate treatment [7].


3.4 Indian and Russian Design for Thorium-based Nuclear Reactors.
3.4.1 Indian designs.
Indian has relatively modest uranium resource (nearly 50,000tons) however, is gifted with
one of the largest deposits of thorium in the earth (approximately 360.000tons) in the beach
sands of Southern India. One of the vital elements in improving the fuel utilization in the
Indian strategy is closed fuel. Pressurized Heavy Water Reactors (PHWRs) based on pressure
tube technology; natural uranium and heavy water were what the indigenous nuclear power
program that India was using. In the second stage of the Indian nuclear power program the
Plutonium from the natural uranium-based PHWRs will be used in fast Breeder Reactors for
multiplying the fissile base. In the third stage of the Indian nuclear power program,
considering the large thorium reserves in India the future systems, which will be based on the
thorium-233U fuel cycle. In the third stage of the Indian nuclear power program, there is a



                                                                                            25
requirement for the timely advance of thorium-based technologies for the entire thorium fuel
cycle [25].
The AHWR is a 300 MWe, water cooled natural circulation reactor by boiling light pressure
tube type, vertical, heavy water moderated (Th-233U) O2 and (Th-Pu) O2 pins are the two
main fuels. The fuel cluster is manufactured to generate energy out of 233U, which is bred in
situ from thorium and has a somewhat negative void coefficient of reactivity. For the AHWR,
the well-proven pressure tube technology has been conducted many passive safety features
have been included. In addition, several pioneering passive safety systems have been
included in the plan for decay heat removal under shut down situations and alleviation of
postulated accident conditions. The manufacture of these reactors has increasingly undergone
modifications and developments based on the feedback from the experimental research and
development (R&D) [24].
R&D programmes devoted to thorium fuel cycle development were a natural requirement
due to the fact that thorium plays such an important role in Indian’s nuclear power
programme. In the Indian power reactors, an actively pursuing research and development
programme in fabrication, characterization and irradiation testing of ThO2-PuO2 and ThO2-
UO2 fuels is taking place. In the new PHWRS, for flux flattening in the initial Core Fuel
bundles containing high density ThO2 fuel pellets are being used.ThO2 pins and sub-
assemblies are also to be utilised as axial and radial blankets in the Fast Breeder Test Reactor
(FBTR) functioning at Kalpakkam. KAMINI, a neutron source reactor, is operating with
233
      U– AI alloy fuel. In the centre, ThO2-PuO2 and ThO2-233UO2 are proposed as fuel for the
Advanced Heavy Water Reactor (AHWR)._The under study fields are Development of novel
fuel fabrication processes and techniques related to automation and remotization needed for
233
      U based fuel fabrication [25].
India has made six times more the thorium than uranium in the utilization of thorium for
large-scale energy production a major goal in its nuclear power program, utilizing a three-
stage concept [26].
1- Pressurized Heavy Water Reactors (PHWRs), sometimes known as CANDUs fuelled by
natural uranium, plus light water reactors, manufacture plutonium.
                                                                                      233
2- Fast Breeder Reactors (FBRs) employ this plutonium-based fuel to strain                  U from
thorium. The blanket in the region of the core will have uranium as well as thorium, so that
further plutonium (ideally high-fissile Pu) is produced as well as the 233U.
                                                                233
3- Advanced Heavy Water Reactors (AHWRs) burn the                     U_and this plutonium with
thorium, obtaining about 75% of their power from the thorium.


                                                                                                26
The exhausted fuel will then be reprocessed to recuperate fissile substances for recycling.
In order to give greater efficiency, the Indian program has moved from aiming to be
sustained simply with thorium to one driven with the addition of further fissile uranium and
plutonium. While continuing with the PHWR and FBR programs, the subcritical Accelerator-
Driven System (ADS) is another option for the third stage [26].
As states in reference [26], “Although Indian’s embrace of thorium as its future nuclear fuel
is based mostly on necessity; the thorium fuel cycle itself has many attractive features. To
begin with, thorium is much more abundant in nature than uranium. Soil commonly contains
an average of around six parts per million (ppm) of thorium three times as much as uranium”.
Table 5 below gives figures of thorium resources in some of countries of the world [26].


Table (5) World Thorium Resources (economically extractable) (from ref. [26]).
        Country                                   Reserves (tons)
        Australia                                 300,000
        India                                     290,000
        Norway                                    170,000
        USA                                       160,000
        Canada                                    100,000
        South Africa                              35,000
        Brazil                                    16,000
        Other countries                           95,000
        World total                               1,200,000


3.4.2 Russian-U.S. Program.
Moscow’s Kurchatov Institute is the place where Russia has developed thorium-uranium fuel
since 1990s. The Russian and US governments both provide funding to design fuel for the
conventional Russian VVER_1000 reactors that involve thorium power. By contrast to the
typical nuclear fuel, which utilises in enriched uranium oxide, the new fuel congregation
design also has the plutonium in the canter as the seed, in a demountable arrangement, with
the thorium and uranium around it as a blanket.
A standard VVER-1000 fuel congregation has 331 fuel rods, every one of 9-millimeter
diameters, structuring a hexagonal congregation 235-mm wide. The center portion of each
congregation is 155-mm across and holds the seed substance, made of metallic plutonium-



                                                                                              27
zirconium alloy in the shape of 108 twisted three-section rods with a cladding of zirconium
alloy and is 12.75-mm wide.
The blanket is made of uranium-thorium oxide fuel pellets in 228 cladding tubes of
zirconium alloy, each of 8.4-mm diameters. These pellets are in four layers around the canter
portion. The blanket substance accomplishes 100gigawatt-days burn-up. Jointly together fuel
congregation, the seed and blanket have the same geometry as a standard VVER-100 fuel
assembly. Thorium fuel burns three quarters of the initially loaded weapons-grade plutonium,
compared with a 31%burn for mixed oxide (MOX) fuel which is made of a mixture of
uranium and plutonium. However thorium fuel does not create more plutonium and has cost
advantages over MOX. Grae et. al concludes: Thorium fuel shows a potential means for the
disposal of overload weapons-grade plutonium in Russian VVER-reactors.
Utilising the thorium fuel technology, plutonium can be disposed of up 75% as fast as MOX
at a considerably lesser cost. Spent thorium fuel would be more proliferation- resistant than
spent MOX fuel and will not require major and expensive reactor adjustments. In addition,
Thorium fuel offers additional benefits in conditions of minimised weight and volume of sent
fuel and consequently lower disposal costs [24].
WER-1000 reactors provide an alternative to the use of the enriched uranium oxide, and this
system creates fuel for Russia. A demountable centre mixture has the plutonium and the
blanket arrangement with uranium/thorium around it. The seed and blanket together are the
same size as a standard WER fuel assembly. The central seed fuel rods use extensive
experience of Russian navy reactor design and are burned for three years (as standard for
WERs). The blanket material stays in the reactor for close to ten years.            The process
manufactures about half the used fuel of MOX (mixed oxide) fuel plants and has less fissile
plutonium [25].
As explained in reference [13] [IAEA-TECDO1450, Vienna: IAEA, 2005] the computation
carried out by Russian specialists suggest a possibility to obtain self-sufficiency in the 232Th -
233
      U fuel cycle by means of a breeding ratio ≥1.0 in a BN-800 type sodium cooled LMFBR.
The same outcomes have been also reported from France. The calculations demonstrate the
likelihood of breeding ratio to approaching 1.0 but not exceeding it in other kinds of reactors
also, namely HTGRs or Heavy Water Reactors.




                                                                                               28
                                         Chapter 4
4. Discussion
4.1 Positives and Negatives of Thorium-based Nuclear Reactors.
4.1.1 Thorium's advantages.
One of the benefits of the Thorium-based fuel has over MOX say US and Russian nuclear
authorities, is that spent thorium fuel doe not contain out weapons- or energy-usable
plutonium components that can be separated out in reprocessing. In thorium-based fuel the
plutonium cannot be reprocessed for any energy or weapons purpose. The fuel could then be
stored and buried as standard spent nuclear fuel.      Thorium fuel creates over 80% less
plutonium than MOX and the small quantity of plutonium that is created is denatured and
diluted with other isotopes making it unsuitable for use in weapons or for energy utilisation
[23].
It is also reported that replacing standard U fuel by RTR fuel may have positives of
decreasing, or even elimination from the fuel cycle, the potential for lessening the spent fuel
storage/disposal needs. In referring to two international projects, the Inventive Nuclear
Reactors and Fuel Cycles programme (INPRO) initiated by the IAEA and the US-led
Generation IV international from (GIF) aiming for future nuclear systems, the IAEA
document has also identified and summarised the benefits and challenges of thorium fuel
cycles which are listed below[10, 13].
*-The mining and extraction of thorium monazite is relatively easy and significantly different
from that of uranium and its ores. Most of the commercially exploited sources of monazite
are from that beach or river sands along with heavy minerals. The overburden during mining
is much smaller than in the case of uranium and the total radioactive waste production on
mining operation is about 2 orders of magnitude lower than the of uranium. The so-called
Radon impact is also much smaller than in the uranium case due to the short lifetime of
thoron as compared to that of radon and needs therefore, much simpler tailings management
than in the case or uranium, to prevent long term public doses. As far as occupational doses
are concerned, there is no need to control ventilation with respect to (Rn-220) inhalation
because monazite extraction is done in open pit. However, the inhalation and ingestion dose
factors are high for thorium and thoron [23].
*-Between three and four times more and more thorium is abundant than uranium, for this
reason thorium has had a renewed of interest in recent years.



                                                                                            29
*-Thorium dioxide is chemically more stable and has higher radiation resistance than
uranium dioxide. Fission creation release is rare for ThO2-based fuels and one order of
magnitude lower than that of UO2. ThO2has favourable thermo-physical properties because
of the higher thermal conductivity and lower co-efficient of thermal expansion compared to
UO2. Therefore ThO2-based fuels are expected to have a better in – pile performance than of
UO2 and UO2-based mixed oxide.
*- Thorium fuel cycle is a good-choice way to produce long term nuclear energy with low
radiotoxicity waste. What is more, the transition to thorium could be done though the
incineration of weapons grade plutonium (WPu) or civilian plutonium.
                                                                    232
*- The absorption cross-section for thermal neutrons of                   Th of 7.4 barns is nearly three
times that of 238U to 239Pu. As a consequence, thorium is a better fertile material than 238Th in
thermal reactors [23].
*- ThO2 is comparatively inert and dose not oxidize, unlike U which oxidizes to U3O8 and
UO3. For this reason, long term interim storage and permanent disposal in repository of spent
ThO2-based fuel are simpler without the problem of oxidation.
*- For burning of WPu or civilian Pu once- through cycle, (Th, Pu)O2 fuel is more attractive,
as compared to (U,Pu)O2 since plutonium is not bred in the former and 233U formed after the
‘once-thorium’ cycle in the spent fuel ensures proliferation resistance [23].
*- There are positive features of thorium are related to accelerate driven system (ADS) and
energy amplifier (AE).
*- There is a lower proportion of delayed neutrons during the thorium cycle (due to the 233U,
the effective beta factor of the 232Th only affects the fast field where there are a small amount
of reaction in the PWR spectrum) which accelerates the reactor kinetics.
In 232Th -233U fuel cycle, a lesser amount of plutonium and long-lived Minor Actinides (MA:
                                                       238
Np, AM and Cm) are formed as compared to the                 U -239Pu fuel cycle, thereby minimizing
the radio toxicity associated in spent fuel [23].
The advantages of the thorium fuel cycle, breeder nuclear reactor for example Super Phenix
are based on the 238U-239Pu fuel cycle [15]. In this cycle, the plutonium, whose fission is the
source of energy released in the reactor is replaced by new plutonium, obtained though the
capture of a neutron by a 238U nucleus:
               238            239         239                 239
                92   U + n→    92   U →    93   Np + e − →     94   Pu + 2 e −

In the 232Th-233U cycle, 232Th plays the role of 238U and 233U that of 239Pu:



                                                                                                      30
                       232
                         Th + n→233U →233Pa + e − →233Pu + e − + 233U + 2e −
                        90       90    90           91            92

As Figure 8 in appendices illustrates, while the 238U-239Pu fuel cycle needs fast neutrons to be
                         232
sustainable, the               Th-233U fuel cycle is sustainable with either fast neutrons or thermal
neutrons.
4.1.2 Technical Problems with Thorium Based Nuclear Reactors.
There are several technical challenges in the front and back end of the thorium fuel cycles:
*-ThO2 and ThO2-based mixed oxide fuels do not dissolve easily in concentrated nitric acid.
Additions of small quantities of HF in concentrated HNO3 are necessary which can lead to
corrosion. This is the reason for the use of stainless steel equipment and pipes in reprocessing
plants. The problem of corrosion is mitigated with the addition of aluminium nitrate. Boiling
THOREX solution [13M HNO3+0.05M HF + 0.1 M Al (NO3)3] at ~ 393K and long
dissolution period are required for ThO2-based fuels [25].
*- The melting point of ThO2 is (3350°C) which is the much higher compared to that of UO2
(2800°C). Consequently, a much higher sintering temperature >2000°C is essential to
produced high density in ThO2 and ThO2-based mixed oxide fuels. Admixing of ‘sintering
aids’ such as CaO, MgO, Nb2O5, etc is requisite for achieving the desired pellet density at
lesser temperature [25].
*- An important feature of the thorium cycle is that a small but significant proportion of the
233
      U undergoes an (n, 2n) reaction to form 232U.
                                         232           233        233
*- In the conversion chain of                  Th to         U,         Pa is formed as an intermediate, which has a
relatively longer half-life(~ 27 days) compared to 239Np (2.35 days) in the uranium fuel cycle
thereby requiring longer cooling time of at least one year to complete the decay of all the
233           233
      Pa to         U. Normally, Pa (protactinium) is passed into the fission product waste in the
THOREX process, which could have a long term radiological impact. It is essential to
separate Pa from the spent fuel solution prior to solvent extraction process for separation of
233
      U and thorium [25].
*- The database and experience of thorium fuels and thorium fuel cycles are very limited,
compared to UO2 and (U,Pu)O2 fuels, and need to be augmented before large investments are
made for commercial utilisation of thorium fuels and cycles.
*- The three stream method of separation of uranium, plutonium and thorium from spent (Th,
Pu)O2 fuel, though viable, is yet to be developed.




                                                                                                                 31
*- One of the reasons which disqualified thorium-uranium breeders as compared to uranium-
                                                   233
plutonium breeders was, the non-existence of a           U stockpile to start with. By contrast the
PWR, BWR heavy water, and graphite moderated reactors were producing large amounts of
                                                                                                233
plutonium. The advent of thorium-plutonium MOX fuels would change the picture;                        U
could be burned in standard reactors without entailing the production of large amounts of
plutonium or minor actinides [25].
4.2 Access to Fuels
The knowledge of future thorium resources on earth is poor owing to inadequate exploration
efforts arising out of insignificant interest. However the rise in interest expressed by some
countries in advance of Fast Breeder Reactors utilising thorium, is expected to increase this
demand greatly. For the same reason, the likelihood of discovering new finds to add to the
known resources contained in the deposits should increase. The known world reserves of
thorium in RAR type are estimated at ~1.16 million tones. Approximately 31% of this (0.3Mt)
is known to be available on the beaches and inland places of India. The likelihood of finding
primary occurrences in the alkaline and acidic rocks in India is high. Other countries having
large Th reserves include Canada, China, Brazil, Norway, U.S.S.R., U.S.A., Burma,
Indonesia, Malaysia, Thailand, Turkey and Sri Lanka [27].
4.3 Thorium- Plutonium MOX
Typically, MOX fuels are comprised of ~ 5% plutonium for 95% natural or depleted uranium.
Although the irradiation tends to reduce the primary quantity of plutonium, the presence of
238
      U, gives a partial reconstitution of the plutonium stockpile. The damage of plutonium can
be rapidly up by replacing the uranium in the MOX with thorium, the result being thorium-
plutonium MOX.
Nowadays a PWR burning uranium-plutonium MOX incinerates ~ 544kg of plutonium per
year for the simplicity of the argument, we assume here a “fully MOGED PWR” however,
only third of fuel a PWR is “MOXed”. A “fully MOGED PWR” PWR loaded with thorium-
                                                                233
uranium MOX should produce approximately 280kg of                     U per year while it incinerates
approximately 800kg of plutonium, producing about 20% less minor actinides. However, the
quality of the plutonium, isotopes, would be somewhat degraded. These results are
summarized in Table (6) [15].




                                                                                                  32
Table (6) Comparison of the effect of MOX and thorium MOX fuels. The plutonium loaded
                                                                         235
comes from reprocessed UOX fuel and the small proportion of                    U remaining in the
uranium-thorium MOX is neglected. (Results from on order of magnitude simulation of a
1GWe reactor operating 91% of the time) [15].


Nuclear       At loading (kg/8TWhe)        MOX              Production Th MOX Production
                                           (kg/8TWhe)
233
      U       0                            0                               278.72
233
      Pu      1008                         -500                            -650
Total Pu      1900                         -544                            -800
MA.           0                            139                             199


4.4. Criticality of Thorium
In the High Temperature Gas Cooled Reactors studies, it was found that the plutonium was
more effective in maintaining a low critical mass than had been primarily anticipated.
Specifically, the sum of burnable poisons added to the reactor cores for the two cycles were
estimated on the basis of attaining an average fuel burn-up of approximately100 MWd/kg. As
it turned out, the fuel exposure with use of the thorium cycle was about 85 MWd/kg,
although it was ~ 115 MWd/kg with the uranium cycle; this implies that the bred 233U in the
thorium cycle was less effective than expected in maintaining criticality, whereas the bred
plutonium in the uranium cycle was more effective than originally expected. The same may
apply to the Radkowsky Thorium Reactor (RTR) [28].
A comparison of the critical mass for different materials is presented in Table (7).The values,
obtained by SCALE calculation of metal spheres with water reflectors, have nothing to do
with the actual weapon design and are used here only for the comparison of RTR and PWR
grade plutonium mixtures with weapon grade plutonium. It is demonstrated, that a relatively
little critical mass, is performed with any plutonium composition, and that RTR-Pu requires
20 to 50 percent more material compared with the weapon-grade material [29].
However, accelerator-driven systems offer interesting further parameters of freedom by
removing the criticality constraint and increasing the safety margin to prompt criticality. The
latter feature is particularly important for MA burners, which are difficult o control as critical




                                                                                               33
systems because the effectual delayed-neutron fraction is only about half of that of a normal
fast rector [30].


          Table (7) Critical mass for different plutonium compositions.
Pu source                                        Critical mass (kg)
Weapon grade                                     4.3
PWR grade                                        5.5
RTR-seed                                         5.9
RTR-blanket                                      6.5


4.5. Stability of Fuel Supply
The thorium terrestrial reserves are estimated to be about 4 times those of uranium.
Particularly, India, Madagascar and Brazil boast huge thorium beds. Also, thorium is not
very soluble in water so that its extraction from sea water is not being considered, contrary to
that of uranium. It should be stressed, nevertheless, that breeder reactor technology is
economical in its use of the fuel, even small content of ore could be worked profitably,
ensuring that fuel would remain available over some thousand years for both uranium and
thorium based breeder reactors. Since thorium reserves are larger, this is not an important
factor [15].
Thorium fuel can also remain in the reactor for longer before refuelling was essential. On the
seed-blanket fuel technology, the blanket, which is ~ 60% of the assembly, will last nine
years in the reactor. The seed is approximately 40% of the fuel cycle, and will go on the
exact schedule that the reactor currently uses. Uranium is still in the fuel assembly, but less
would be utilized. The decrease of uranium could address potential future shortages and the
rising costs of uranium. Relying on thorium, which is abundant in more countries than
uranium, would create a more reliable supply of the fuel [31].




                                                                                             34
5- Conclusion
For the next two decades electricity generation will continue in its reliance on nuclear power.
Asia has a number of evolutionary new build water reactors as well as Finland. Novel high-
temperature reactors were also considered by other countries such as US and South Africa.
Up to now, reactor development has been based mainly on the U/Pu cycle; the Th/U cycle is
less developed even though a number of options for energy production with thorium were
studied in the period 1950-1970. In addition, advanced nuclear systems like HTGRs and
MSBRs can be considered in the coming generation of high temperature heat applications, if
sufficient attention and investments is provided.


                                           235
Commonly, reactors fundamentally burn            U which uses only ∼ 1% of the natural uranium.
For this reason uranium servers are approximated to provide about a century of reactor
operation. The actual time span depends on the number of reactors in operation around the
world and on the cost of natural uranium. However, thorium is an alternative which makes it
possible to protract the life of source of nuclear energy and to diversify and steady supply. It
would be beneficial to take a new look at this sort of fuel, in the light of several studies which
have already been carried out and taking into account recent technological evolution, in
particular in the MOX industry with regards to remote controlling and shielding.
Nuclear energy is the only technological option now that has been fully developed and which
is able to supply significant amount of energy not including giving increase to significant
greenhouse gases but the associated troubles, mainly proliferation and radiotoxic waste,
could deter its expansion. Thorium fuel cycles have demonstrated many benefits to offer in
solving these worries. Owing to its intrinsic properties, thorium fuel is not only limited in the
generation of electricity but as well as realises high temperature heat application, which may
support in saving the clime via hydrogen production.


There are a range of future designs that collectively offer great flexibility in energy
applications: traditional electrical power generation, low and high temperature process heat
applications, hydrogen generation, waste toxicity decrease through actinide management and
burning of weapons grade plutonium.
The interest of nuclear community in thorium based nuclear fuel cycles have been varied in
the last 50 years. The main reason for this is the slow technical development due to the
geographical distribution of thorium and uranium reserves. The technological progresses



                                                                                               35
would certainly be in favour of thorium rather than uranium. Today, the nuclear technology
has been interrogated since of proliferation of nuclear weapons, long-lived wastes, and
technical inconveniences of excess Pu-239 either in dismantled nuclear weapons or in stocks.


Thorium will possibly be a nuclear material more valuable than uranium in the future. For
this reason, all developing countries having thorium reserves should focus their technological
attentions to the assessment of their national thorium resources (like in the case of India) and
cooperate with each other in this field to combine their efforts.




                                                                                             36
BIBLIOGRAPHY
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(2001).
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4- DTI Energy White paper, 2003, Our Energy Future-Creating a Low Carbon Economy,
TSO (The Staionery Office), Crown Copyright
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Copyright c 1967. Oxford. London Edinbrough. New York.
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proliferation   potential   of   nuclear    power    fuel   cycle,    Science    &    Global
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(Library online, Science Direct) Accessed 13-07-2008
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TECDOC-1450(2005)
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Nuclear Fuel Cycle. Europhysics News-Features, Volume 38, Number 2, 24-27, 2007
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17- http://www.google.co.uk/search?hl=en&q=Proliferation+and+thorium+fuel+cycle&meta
Nuclear power fuel cycle. Accessed 22-07-2008




                                                                                          37
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Design, 203 (2001) 65-68
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nuclear.org/info/inf26.htm [Accessed 02-08-2008]
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1319, Vienna: IAEA, 2002, 257-265
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OECD/NEA, Paris (France), 2001, Pp.31-49



                                                                                         38
31- Jenny Weil, Progress Being Made on Development of Thorium-Based Fuel, Company
Says. Nuclear Fuel, p.5,2007




                                                                              39
7-Appendices

Table (2) New Reactor Designs [2]

Reactor       Vendor            Approximate      Reactor     Certification        Target
Design                          Capacity         Type        Status in US         Certificati
                                (MWe)                                             on
EPR*          AREVA NP          1600             PWR         Pre- certification   2009
AP600         Westinghouse      650              PWR         Certified            Certified
AP1000*       Westinghouse      1117             PWR         Certified            Certified
System80+     Westinghouse      1300             PWR         Certified            Certified
US APWR       Mitsubishi        1600             PWR         Undergoing           2011
                                                             certification
IRIS          Westinghouse      360              PWR         Pre-certification    2010
              Et al
ESBWR         GE                1550             BWR         Undergoing           2007
                                                             certification
ABWR*         GE et al          1371             BWR                              Certified
ACR Series AECL                 700-1200         Modified    Pre- certification   Not
                                                 PHWR                             available

PBMR          Westinghouse,     180              HTGR        Pre-certification    Not
                                                                                  available
GT-MHR        General           325              HTGR        Research             Not
              Atomics                                        prototype planned    available
4S*           Toshiba           10-50            Sodium-     Potential            Not
                                                 cooled      construction         available



Also supported by electricity generating firms or organisation publicly investigating possible
construction in the U.S.




                                                                                              40
   Figure 3 Seed-Blanket Fuel Assemblies [18].




Figure 7 Time-Average Bundle Power Distributions in a High-Power Channel [25].




                                                                                 41
Figure. 8, Number <n> of neutrons available for breeding in the uranium-plutonium and the
thorium-uranium cycles with thermal and fast neutron spectra. Breeding is impossible for
negative values of <n> [15].




                                                                                      42

								
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