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Speransa 2008: Round table: Nuclear and sustainable development Aachen University of Applied Science, campus Jülich \AcUAS - Fissile resources and breeding - complete.doc 1

Erasmus Intensive Programme Project (IP)

Speransa − Stimulation of Practical Expertise in Radiological and Nuclear Safety

24/02/08 – 07/03/08

– Mol/Brussels (Belgium) –

Round table: Nuclear and Sustainable Development

Fissile Resources and Breeding

Eduardo Vera Garcia, Ohran Uluyol, Sebastian Sasonow, Roland Höll

Aachen University of Applied Science, campus Jülich, Germany

AACHEN UNIVERSITY OF APPLIED SCIENCE

CAMPUS JÜLICH

SPERANSA 2008

-FISSILE RESOURCES AND BREEDING-

Höll, Roland W.

Sasonow, Sebastian Uluyol, Orhan

Vera Garcia, Eduardo


INDEX

TABLE INDEX ................................................................................................... 2

FIGURE INDEX.................................................................................................. 2

Fissile material and breeding.......................................................................... 3

The world’s nuclear fuel market ............................................................................3 Importance of nuclear fuels .................................................................................................. 3 Uranium production .............................................................................................................. 4 Uranium demand and supply................................................................................................ 5

Abundance and distribution of uranium and thorium..........................................5 Uranium ................................................................................................................................ 5 Thorium................................................................................................................................. 6

Energetics of fission reactions ..............................................................................6 Background........................................................................................................................... 6 Conversion and breeding...................................................................................................... 7

Breeder Reactor .................................................................................................... 10 What is it? ........................................................................................................................... 10 Liquid-Metal Cooled (LMFBR) ............................................................................................ 10 Gas-Cooled Fast Breeder Reactor (GCFR) ....................................................................... 11 Molten Salt Breeder Reactor (MSBR) ................................................................................ 11 Light-Water Breeder Reactor (LWBR)................................................................................ 11

Production of fissile 239Pu and 233U ...................................................................... 11 Uranium – Plutonium cycle ................................................................................................. 12 Thorium – Uranium cycle.................................................................................................... 12

Fuels and future conclusions .............................................................................. 13 Momentary situation ........................................................................................................... 13 Outlook about the nuclear fuel build-up in the next years .................................................. 13 Conventional use of fuel ..................................................................................................... 14 Breeding option................................................................................................................... 14 Exploring possibilities ......................................................................................................... 14 Mining costs........................................................................................................................ 14 Final remarks ...................................................................................................................... 15

Bibliography.......................................................................................................... 15

TABLE INDEX Table 1 World Natural Uranium Production by Country, 2005 – 2007 ................................. 4 Table 2 Binding energy per nucleon as a function of atomic mass number......................... 6 Table 3 Critical Energies for Fission (in MeV) ...................................................................... 7

FIGURE INDEX Figure 1 Historical Uranium Price Development .................................................................. 3 Figure 2 World Uranium Production in 2006 by Country...................................................... 4 Figure 3 World Uranium Demand and Production by Country/ Region (1996 – 2007)........ 5 Figure 4 Variation of η with energy for 233U, 235U, 239Pu................................................. 9

Höll / Sasonow / Uluyol / Vera


Fissile material and breeding

The world’s nuclear fuel market

Importance of nuclear fuels

Concerns about the impact of hydrocarbon use on climate and global warming are significantly growing. Furthermore, we are all well aware that security of supply is increasingly an issue. In this context, it is now principally recognised that nuclear energy has to be back on the agenda. All in all, the prospects for nuclear power industry and thus for the nuclear fuel mining are very positive for the coming years. The economically eligible uranium reserves (defined by the maximum award per kilogram of the current state of technology) were approved by the International Atomic Energy Agency (IAEA) and the OECD Nuclear Energy Agency (NEA) in 2006 in the so-called Red Book.

Figure 1 Historical Uranium Price Development

As it can be seen in Figure 1, since 2002, the uranium prices have increased more than tenfold. The spot market price of uranium began an increase from about US$ 9/lb U3O8 in mid 2001 following a fire at the Olympic Dam mill (Australia) in October 2001 and was propelled in subsequent years by a series of interrupting events, such as the mine shaft flooding at the Mc Arthur River mine (Canada) in April 2003, the threat of the early shutdown of the Rössing mine (Namibia) and the Ranger mine (Australia) in 2003, just to mention some of themi.



Uranium production

As shown in Table 1 and Figure 2, there are relatively few countries exploiting uranium, about 97.9% of the global uranium production is mined in 12 countries. The following 3 countries mine more than half of the world’s production:

• Canada with 10241 t / a (22.3%) • Australia with 8878 t / a (19.4%) • Kazakhstan with 6835 t / a (14.9%)

The largest uranium mine is the McArthur mine in Canada. This mine accounts for 18% of the world uranium production (approx. 7,200 tU/a).

2005 2006 2007

tonnes Share of tonnes Share of tonnes Share of U Total U Total U Total

Country

(%) (%) (%) Canada 11,627 27,6 9,862 24,6 10,241 22,3 Australia 9,516 22,6 7,593 19,0 8,878 19,4 Kazakhstan 4,345 10,5 5,281 13,2 6,835 14,9 Niger 3,093 7,4 3,434 8,6 3,400 7,4 Russia 3,400 8,1 3,200 8,0 3,700 8,1 Nambia 3,147 7,5 3,077 7,7 4,179 9,1 Uzbekistan 2,300 5,5 2,242 5,6 2,550 5,6 USA 916 2,2 1,699 4,2 2,030 4,4 Urkraine 900 2,1 900 2,2 900 2,0 China 750 1,8 750 1,9 750 1,6 South Africa 674 1,6 559 1,4 886 1,9 India 550 1,3 550 1,4 550 1,2 Subtotal 41,218 98,0 39,147 97,8 44,899 97,9 Others 847 2,0 892 2,2 949 2,1 Total 42,065 100,0 40,039 100,0 45,848 100,0

Table 1 World Natural Uranium Production by Country, 2005 – 2007

Figure 2 World Uranium Production in 2006 by Country

Others; 2,20%India; 1,40%

Canada; 24,60%

Australia; 19%

Kazakhstan; 13,20%Niger; 8,60%

Russia; 8%

Nambia; 7,70%

Uzbekistan; 5,60%

USA; 4,20%

Ukraine; 2,20%

South Africa; 1,40%

China; 1,90%



Uranium demand and supply

In Figure 3 is important to notice that Kazakhstan is the only country which has achieved a steady increase in annual uranium production over the most recent 10 years. As well it is possible to observe how the global demand for uranium is higher than the actual production of uraniumii.

Figure 3 World Uranium Demand and Production by Country/ Region (1996 – 2007) The gap between production and demand is covered with secondary supply. The secondary supply comprises different sources like:

• Conversion of HEU (high enriched uranium, 90% 235U enrichment) from warheads into LEU (low enriched uranium, 3 to 5% 235U enrichment).

• Plutonium and uranium regained from the reprocessing of spent nuclear fuel, which is recycled in the form of mixed oxide (MOX) fuel and reprocessed uranium (RepU).

• Re-enrichment of tails coming from enrichment facilitiesiii.

Abundance and distribution of uranium and thorium

Uranium

Natural uranium consists of 3 isotopes, 234U, 235U and 238U whose isotopic abundances are 0.0055%, 0.7200% and 99.2745% respectively. Uranium occurs naturally not as a pure metal, but in a large number of minerals (at least 60 are known). The earth’s crust contains 3-4 ppm U, which makes it about as abundant as arsenic or boron. Uranium is found at this relative concentration in the large granite rock bodies formed by slow cooling of the magma about 1.7-2.5 eons ago. It is also found in younger rocks at higher concentrations (“ore bodies”). In most minerals uranium is in the tetravalent state. The most important one is uranite (UO2+x, x= 0.01 to 0.25, also called pitchblende), in which the uranium

0

10000

20000

30000

40000

50000

60000

70000

80000

1996 1998 2000 2002 2004 2006

Demand

Others

Uzbekistan

Kazahstan

Russia

Australia

Africa

Canada

USA



concentration is 50-90%; it is found in Western Europe, Central Africa, Canada and Australia. In the USA and Russia carnotite (a K + U vanadate) is the most important mineral and contains 54% uranium. In the high grade ores the mineral is mixed with other minerals so the average uranium concentration in the crushed ore is much less: e.g. about 0.5% on the Colorado Plateu, but also can be found at much lower concentrations like 0.1-0.3%.

Thorium

Natural thorium consists of 100% of the isotope 232Th which is the parent nuclide of the thorium decay series. Thorium is somewhat more common in nature than uranium with an average content in the earth’s crust of 10 ppm (by comparison, the average abundance of lead is about 16 ppm in the earth’s crust). In minerals occurs only as oxide. The content of thorium in sea water is < 0.5 x 10-3 g/m3, which is lower than that of uraniumiv.

Energetics of fission reactions

Background

As shown in Table 2 the binding energies of nuclei per nucleon decrease with increasing atomic mass number for A greater than 50. This means that a more stable configuration of nucleons is obtained whenever a heavy nucleus splits into two parts, that is, undergoes fission. The heavier, more unstable nuclei might therefore be expected to fission spontaneously without external intervention. Although heavy nuclei do spontaneously fission, they do so only rarely.

Table 2 Binding energy per nucleon as a function of atomic mass number.

For fission to occur rapidly enough to be useful in nuclear reactors, it is necessary to supply energy to the nucleus. This, in return, is due to the fact that there are attractive forces acting between the nucleons in a nucleus; energy is required to deform the nucleus to a point where the system can begin to split in



two. This energy is called the critical energy of fission and is denoted by Ecrit. Values of Ecrit are given in Table 3. Any method by which energy Ecrit is introduced into a nucleus, thereby causing it to fission, is said to have induced the fission. The most important of these is neutron absorption. According to the data Table 3, the binding energy of the last neutron in 236U is 6.4 MeV, whereas Ecrit is only 5.3 MeV. Thus, when a neutron of zero kinetic energy is absorbed by 235U, the compound nucleus, 236U is produced with 1.1 MeV more energy than its critical energy, and fission can immediately occur.

Table 3 Critical Energies for Fission (in MeV)

Nuclei such as 235U that lead to fission following the absorption of a zero energy neutron are called fissile. Although 235U is said to be fissile, the nucleus that actually fissions in this case is the 236U. With most heavy nuclei other than 233U, 235U, 239Pu and 241Pu, the binding energy of the incident neutron is not sufficient to supply the compound nucleus with the critical energy, and the neutron must have some kinetic energy to induce fissionv. Isotopes like 232Th, which are not fissile but from which fissile isotopes can be produced by neutron absorption, are said to be fertile.

Conversion and breeding

For this discussion is important to define the parameter η, which is equal to the number of neutrons released in fission per neutron absorbed by a fissile nucleus. Since radiative capture (n, γ) competes with fission (n,f), η is always smaller than ν (total number of neutrons released per fission). In particular, η is equal to ν multiplied by the relative probability (σf/σa) that absorption leads to fission.

f

f

a

f

σσσ

νσσ

νηγ +

==

Using the “capture to fission ratio” (α)



fσσ

α γ=

ανη+

=1

For a mixture of fissile or fissile and nonfissile nuclides, η is defined as the average number of neutron absorbed in the mixture. In this case, η is given by

� ΣΣ

=i

fa

ii )()(1 νη

Where ν (i) and Σf (i) are the value of ν and the macroscopic fission cross section for the ith nuclide, respectively, and Σa is the macroscopic cross section for the mixture. All these values must be computed at the energy of the neutrons inducing the fissionvi. To become critical, a reactor must be fuelled with a nuclide having a value of η greater than 1. If η were less than 1, fission in one generation would necessarily lead to less than one fission in succeeding generations, and the reactor could never achieve a critical state. However, η can not be exactly 1 since, as already noted, some neutrons inevitably are lost either in nonfission absorption reactions (n, γ) or by escaping from the system altogether, and criticality could not be reached. With nonfissile nuclides (like 232Th and 238U), only those neutrons in a reactor with energies above the fission threshold area able to induce fission. As is the case for any natural resource, the supply of 235U is finite. Fortunately, it is possible to manufacture certain fissile isotopes from abundant nonfissile material, a process known as conversion. The two most important fissile isotopes that can be produced by conversion are 233U and 239Pu. The 233U is obtained from 232Th by the absorption of neutrons.

239Pu is obtained from reactions similar to the reaction of Th.

The conversion process is described quantitatively in terms of the parameter C, which is called the conversion ratio or sometimes the breeding ratio. This is defined as the average number of fissile atoms produced in a reactor per fissile fuel atom consumed. Thus, when N atoms of fuel are consumed, NC atoms of fertile material are converted into new fissile atoms. However if the newly produced fissile isotope is the same as the isotope that fuels the reactor, the new atoms may later be consumed to convert another NC x C = NC2 atoms of fertile material, and so on. So the consumption of fertile atoms results in the conversion of a total of

NC + NC2 + NC3 + … = NC/(1-C)

Fertile atoms, provided C is less than 1. If C = 1, an infinite amount of fertile material can be converted starting with a given amount of fuel. A most important



situation occurs when C is greater than 1. In this case, more than one fissile atom is produced for every fissile atom consumed, which is a process described as breeding. Reactors that are designed so that breeding will take place are called breeder reactors or simply breeders. Reactors that convert but do not breed are called converters; reactors that neither convert nor breed but simply consume fuel are called burners. Breeders actually produce more fissile fuel than they consume. While η must be greater than 1 for conversion, it must be greater than 2 for breeding. This is because, in any reactor, one fission neutron must eventually be absorbed in fuel just to keep the reactor critical and absorbed in fertile material to produce the new fissile isotope. In actual fact, η must be substantially greater than 2. In any reactor some neutrons inevitably are absorbed by no fuel atoms or lost by leakage. It is important to notice that η is not constant but it varies according to the energy of the neutron that induces the fission. This variation of η can be observed in Figure 4.

Figure 4 Variation of ηηηη with energy for 233U, 235U, 239Pu. At thermal neutron energies (E = 0.025 eV), the value of η for 233U is about 2.29, which is sufficient in excess of 2 for breeding to be possible. Thus, a properly designed reactor in which most of the fissions are induced by thermal neutrons (thermal reactor) could breed if it were fuelled with 233U. In contrast for 235U and 239Pu (2.07 and 2.14 respectively), are not sufficiently greater than 2 to permit breeding. However at higher neutron energies (100 keV), the η value rises to values substantially above 2 for all three of the fissile fuels. As far as the value of η is concerned, it should therefore be possible to breed with these fuels if the



reactor is designed in such a way that the bulk of the fissions are induced by neutrons with sufficient high energies. Reactors of this type are called fast reactors or, since they are usually designed to breed, fast breeders. The extent to which breeding occurs in a reactor is described by the breeding gain (a parameter that is described by the symbol G). Since C, the breeding ratio, is the total number of fuel atoms produced per fuel atom consumed, it follows that G and C are simple related byvii

G = C-1

Breeder Reactor

What is it?

A breeder reactor is a nuclear reactor that consumes fissile and fertile material at the same time as it creates new fissile material. These reactors were initially (1950's and 1960's) considered appealing due to their superior fuel economy; a normal reactor can consume less than 1% of the natural Uranium that begins the fuel cycle, whereas a breeder can use much more with a once-through cycle and nearly all of it with reprocessing. Also, breeders can be designed to utilize thorium, which is more abundant than uranium. Renewed interest is also due to the dramatic reduction in waste they produce and especially long-lived radioactive waste components. Production of fissile material in a reactor occurs by neutron irradiation of fertile material, particularly 238U and 232Th. In a breeder reactor, these materials are deliberately provided, either in the fuel or in a breeder blanket surrounding the core, or most commonly in both. Production of fissile material takes place to some extent in the fuel of all current commercial nuclear power reactors. Four types of breeder reactors have been developed to date:

� Liquid-Metal Cooled Fast Reactor(LMFBR) � Gas-Cooled Fast Breeder Reactor(GCFR) � Molten Salt Breeder Reactor (MSBR) � Light-Water Breeder Reactor(LWBR)

Liquid-Metal Cooled (LMFBR)

The fundamental principles underlying the fast breeder reactor concept were discovered before the end of World War II, and the potential impact of breeder reactors on future energy supplies was immediately recognized. The LMFBR operates on the uranium-plutonium fuel cycle. This means that the reactor is fuelled with bred isotopes of plutonium in the core or driver, and the blanket is natural or depleted uranium. The number of fission neutrons emitted per neutron absorbed by 239Pu, increases monotonically with increasing neutron energy for energies above 100 keV. It follows that the breeder ratio and breeding gain increase with the average energy of the neutrons inducing fission in the system.



There is no moderator, so the core and blanket contain only fuel rods and coolant. Sodium has been universally chosen as the coolant for the modern LMFBR. With an atomic weight of 23, sodium does not appreciably slow down neutrons by elastic scattering, although as noted earlier, it does moderate neutrons to some extent by inelastic scattering. Sodium is also an excellent heat transfer agent, so that an LMFBR core can be operated at high power density.

Gas-Cooled Fast Breeder Reactor (GCFR)

This reactor concept is a logical extrapolation from HTGR technology. It is helium-cooled reactor fuelled with a mixture of plutonium and uranium. The core of the GCFR is similar to that of an LMFBR, with mixed PuO2 and UO2 pellets in stainless steel pins in the GCFR have a roughened outer surface to enhance heat transfer to the passing coolant.

Molten Salt Breeder Reactor (MSBR)

This is thermal breeder that operates on the 233U-232Th cycle. It is recalled that 233U is the only fissile isotope capable of breeding in a thermal reactor. The MSBR concept is a unique design among reactors in that the fuel, fertile material, and coolant are mixed together in one homogeneous fluid. This is composed of various fluoride salts that, at an elevated temperature, melt to become a clear, no viscous fluid.

Light-Water Breeder Reactor (LWBR)

For many years, it was the accepted view that is not possible to build an ordinary light-water reactor that will breed even if it is fuelled with 233U. For one thing, it was thought that too many neutrons would be lost at thermal energies owing to the large absorption cross-section of water. It was recognized that this problem could be avoided by reducing the amount of water relative to fuel in the core. In so doing, the energy spectrum of the neutrons would be shifted to higher energies due to the decreased moderation from the water. This shift would, in turn, mean that a larger fraction of the neutrons would be absorbed in the 233U at intermediate energiesviii.

Production of fissile 239Pu and 233U

As previously stated it is possible to convert the nonfissile isotopes 232Th and 238U into fissile 233U and 239Pu respectively. It is a relatively simple matter to achieve such reactions. Naturally occurring thorium is entirely 232Th. Therefore, it is merely necessary to introduce thorium, in one form or another, into a critical reactor where it is exposed to neutrons. After a suitable irradiation time, when the 233U has build up to a desired level, the thorium is withdrawn from the reactor and the 233U is



extracted from the thorium. This can be done by chemical means since thorium and uranium are two entirely different chemical elements. In order to obtain 239Pu, the fertile isotope 238U is irradiated in a reactor. This however, occurs automatically in most power reactors of current design, given that the nuclear fuel in common reactors is made of a mixture of 95-97% 238U and 3-5% 235U. Then the conversion from 238U to 239Pu takes place during the normal operation of the reactor. The plutonium is later extracted chemically from the fuelix.

Uranium – Plutonium cycle

During the working life of traditional reactors like PWRs (pressurized water reactors) or BWRs (boiling water reactors), approximately once every 2 years, the reactors are shut down and a portion of the fuel (between one third and one quarter) is removed and placed in a spent fuel pool adjacent to the reactor containment building. Most of the spent fuel in these reactors is fertile 238U, non-negligible amounts of the fissile plutonium isotopes 239Pu and 241Pu. Thus a normal 1,000 MWe nuclear plant discharges about 180 kg of fissile plutonium plus 220 kg of 235U at each refuelling. This fissile material, if consumed in a reactor, would release the energy equivalent to that of about 1 million tons of coal. The plutonium and uranium from spent fuel can be utilized if it is recycled. The spent fuel is reprocessed; this means that the plutonium and uranium are chemically extracted from the fuel. The plutonium, in the form of PuO2, is then mixed with UO2

*, fabricated into what is called mixed oxide fuel (MOX), and returned to the reactor. It is estimated that the adoption of a uranium-plutonium recycle would reduce the cumulative U3O8 requirements of LWRs (light water reactors) by about 40% over the 30 year lifetimes of these reactors. The fuel cycle for a breeder reactor is similar to the one for traditional reactors. Either natural or depleted uranium (depleted means uranium from which 235U has been extracted) can be used in the blanket. This uranium, together with the uranium and plutonium from the reprocessing plant, is fabricated into core and blanket assemblies and introduced into the reactor. The spent fuel, after a cooling period in the spent fuel pool, is sent to be reprocessedx, closing this way the cycle.

Thorium – Uranium cycle

In general thorium is 3 to 4 times more abundant than uranium, widely distributed in nature as an easily exploitable resource in many countries and has not been exploited commercially so far. Thorium fuel cycle is an attractive way to produce long term nuclear energy with low radiotoxicity waste. This is because in 232Th-233U fuel cycle, much lesser quantity of plutonium and long lived Minor Actinides (MA: Np, Am, and Cm) are formed as compared to the 238U-239Pu fuel cycle, thereby minimizing the radiotoxicity associated in spent fuel. * Natural uranium, depleted uranium, or the slightly enriched uranium from the reprocessing plant can be used in mixed-oxide fuel. The proportions of fissile plutonium, 235U, and 238U in the fuel would amount to between 3 to 3.5%.



The absorption cross section (cross section= probability of having a nuclear interaction) for thermal neutrons of 232Th (7.4 barns) is nearly three times that of 238U (2.7 barns). Hence, a higher conversion (to 233U) is possible with 232Th than with 238U (to 239Pu). Thus, thorium is a better fertile material than 238U in thermal reactors but thorium is inferior to depleted uranium as a fertile material in fast reactor. For the fissile 233U, the number of neutrons liberated per neutron absorbed (η) is greater than 2.0 over a wide range of thermal neutron spectrum, unlike 235U and 239Pu. Thus, contrary to 238U-239Pu cycle in which breeding can be obtained only with fast neutrons spectra, the 232Th-233U fuel cycle can breed with fast, epithermal or thermal neutron spectra. Although the 232Th-233U has many advantages it has not been commercially developed due to the discovery of new uranium deposits and their improved availability. However, in recent times, the need for proliferation resistance, 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 fuelsxi.

Fuels and future conclusions

Momentary situation

In 2006, 437 nuclear power plants (npps) produced 300 GWea, this means 16%xii of the worldwide used nuclear power for watching TV, using air conditioners, freezers etc. In order to supply these 300 GWea, the npps needed 66.000 tons of Unat. Natural uranium (Unat) composed by 99.2739% 238U, 0.7205% 235U and 0.0056% 234U. The total amount of mined uranium during this year was 40,000. The rest of the needed fuels came from secondary supplies likes military warheads. Weapons-grade uranium is highly enriched, to over 90% 235U (the fissile isotope)xiii which then is processed and sold as low enriched uranium.

Outlook about the nuclear fuel build-up in the next years

In the World Nuclear Association (WNA) Reference Scenario the growth of nuclear energy production is expected to be 377 GWea in 2010, 454 GWea in 2020 and 529 GWea in 2030. This represents a supply of fuel of 64,700 t U in 2010, 81,000 t U in 2020 and 109,100 t U in 2030xiv. If we consider that the average abundance of natural uranium in the earth’s crust is about 2.7 ppm, then we can calculate that the total uranium in the earth’s crust is 73E+12 t. On the other hand oceans as well contain dissolved uranium in a concentration of 0.0033 ppm. On the earth are about 1.34E+18 t of water. This means there are about 4.4E+09 t Unat in the oceansxv. Presently the costs of uranium filtration from the oceans are too high, however with the raise in uranium demand and thereof uranium’s prices, this filtration process will become economically attractive. Right now the uranium reserves are about 4,743,000 t Unat considering a mining costs of 130 US$/ kgxvi.



Conventional use of fuel

If we consider that the uranium reserves are approx. 4,743,000 t and that they are used for energy production, we can clearly observe that there is a limited life time range for conventional npps (burners). In years this amount of uranium would represent 45 to 50 years of fuel for the npps. However, this is a calculation with many unknown facts and influences, which might prove to be wrong. One important factor affecting these calculations is the expansion or contraction of the nuclear energy market.

Breeding option

One possibility of increasing the supply of fuels for npps is breeding. By use of breeding the new fuels for the npps will become 239Pu and 233U. Resources for the breeding process are 232Th and 238U. This makes a new type of fuel reserve available and by producing more fissionable material than consumed (through the breeding process), the amount of fuel for npps rises about 50-60 timesxvii than by the conventional application in burner reactors. This represents around 2500 years of fuel for npps.

Exploring possibilities

The countries with the highest uranium mining rates are Canada, Australia and Kazakhstan. Together they produced 56.6 %xviii of the Unat in 2007. In Russia, Ukraine, Uzbekistan and Kazakhstan, the uranium resources are about 25% of the global quantity of the proven reserves. This gives a total of 800,000 t U. There are also 20% of the estimated and speculative resources, in total 2.4 million t Unat

xix. In Australia is possible to find exploitable uranium resources. There are 747,000 t Unat of recoverable uranium at a cost of less than US$130/kgxx. The speculation about uranium resources includes the United States, Canada, Niger, South Africa and a few others. Presently for 232Th, there are approximately 1 million tons for mining under economical circumstancesxxi.

Mining costs

With the rising price of the uranium and the price for electricity, the mining of ores with lower uranium content is becoming more attractive. By rising the mining expenditure from 40 US$/kg to 130 US$/kg the amount of Unat that can be extracted has been doubled. With this in mind it is speculated about the time when filtration of Unat from the oceans will become economically attractive. Presently uranium ocean filtration costs are around 300 US$/kg, but more important for this is the energy balance of the process. The world’s supply of electricity and generation of health and wealth will make necessary the expansion of Unat mining, development of the breeding technology and thorium mining. Also the geophysical prospecting of uranium and thorium deposits must continue. The costs of this expansion will be influenced by the world’s need of fuel and energy.



Final remarks

The world's reserves of 235U are not adequate to indefinitely support the needs of a nuclear power industry based only on burner or converter reactors. With the introduction of breeder reactors the fuel base switches from 235U to 238U and 232Th, both of them are considerably more plentiful than 235U. Furthermore, all of the depleted uranium that is, the residual uranium, mostly 238U, remaining after the isotope enrichment process can be utilized as breeder fuel. Breeders are capable of satisfying the electrical energy needs of the world for thousands of years. There are considerable differences of opinions in the world regarding the desirability or urgency to develop breeders, owing to a divergence of views on the adequacy of uranium resources and the potential for nuclear proliferation. Some authorities feel that sufficient amounts of uranium will be discovered and produced, albeit at increasing cost, to satisfy the requirements of the domestic and export markets well into the first half of the next century. Other experts are less optimistic about uranium supplies and see the development of the breeder as insurance against future shortages of uranium. Yet still others express deep concern over the possible diversion of plutonium to weapons use resulting in the proliferation of nuclear weapons. The major industrial nations Great-Britain, France, Germany, and Japan have essentially no indigenous uranium resources, and these countries must utilize to the fullest extent possible all of the uranium they import. In addition, these nations possess or can lay claim to large quantities of depleted uranium. France, for instance, has enough depleted uranium on hand to satisfy its electrical needs for the next 100 years if the uranium is used for fuel in breeder reactor power plants. France and Japan have active breeder programsviii. In the future thorium and thorium based fuel as metal, oxide or carbide, can be utilized with fissile 235U or 239Pu in nuclear research and power reactors for conversion to fissile 233U, thereby enlarging the fissile material resources. As well thorium fuel cycle is an attractive way to produce long term energy with low radiotoxicity waste.

Bibliography

i Recent Activities and Trends in the Uranium Market, R. Kwasny, F. Aul and K. Lohrey, Alzenau. ATW 52. Jg. (2007) Heft 11 – November, page 696. ii Ibidem, pages 697, 698. iii Ibidem, page 700. iv Radiochemistry and Nuclear Chemistry, Third Edition, Gregory Choppin, Jan-Olov Liljenzin, Jan Rydberg. Butterworth-Heinemann 2002, pages 99, 103-105. v Introduction to Nuclear Engineering, Third Edition, John R. Lamarsh, Anthony J. Baratta. Prentice Hall 2001, pages 74 – 77. vi Ibidem page 85. vii Ibidem pages 119 -124. viii Ibidem pages 168-180. ix Ibidem page 120.



x Ibidem pages 187-190. xi Thorium Fuel Cycle- Potential Benefits and Challenges, International Atomic Energy Agency (IAEA), IAEA-TECDOC-1450, May 2005. xii ATW 52. Jg. (2007) Heft 11 – November, page 718. xiii World Nuclear Association. xiv ATW 52. Jg. (2007) Heft 11 – November, page 710. xv Ibidem, page 718. xvi OECD: Uranium 2005 – Recourses, Production and Demand xvii http://www.schneller-brueter.de/sites/was_ist_snr.htm xviii Estimate of NUKEM xix Atw 52. Jg. (2007) Heft 11 – November Page 703 xx Research Note no. 17 2006–07, Parliamentary Library, Australia xxi Merkel B., Dudel G. et al.: Untersuchungen zur radiologischen Emission des Uran-Tailings Schneckenstein, 1988

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