Thorium Fission Power

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    I first read about Thorium as fuel for a nuclear reactor in the January/February 2012 issue of Analog Science Fiction and Fact. In the Alternate View column Jeffery D. Kooistra gave a brief description of the system and its advantages. After conducting further research I find that I believe that Thorium power can drive our civilization long enough for us to develop thermonuclear fusion as a power source. More importantly for the present Thorium power will enable us to extinguish the despicable farce of the grossly misnamed Clean Coal. I havenít put anything original into this essay: I merely include it as a vote in favor of using Thorium power to drive our civilization.

    For a heavy element Thorium has a relatively high abundance in Earthís crust; it comprises roughly 10 parts per million of common continental crust, which abundance makes it approximately three to four times more common than Uranium. Thorium itself, which consists almost entirely of the isotope Th-232, does not undergo fission, but a simple process converts Thorium into a fissile fuel; absorption of a neutron followed by a pair of beta decays does the trick. Upon absorbing a neutron, Thorium-232 transmutes into Thorium-233, which then undergoes beta-decay, with a half-life of 22 minutes, into Protactinium-233. With a half-life of roughly 27 days, Protactinium-233 undergoes beta-decay and becomes Uranium-233, which undergoes fission and has impressive properties. A thermal neutron causes a nucleus of Uranium-233 to split while producing enough neutrons to sustain the continued production of energy and the conversion of Thorium into more Uranium, provided that the reactor makes the most efficient use of its neutrons.

Thorium as a chemical:

    In the 1820's Morten Thrane Esmark discovered a strange black mineral on the island of LÝvÝy in Norway and took a sample of it to his father, the noted mineralogist Jens Esmark. Unable to identify the mineral, the elder Esmark sent a sample of it to Swedish chemist JŲns Jakob Berzelius in 1828. Berzelius analyzed the mineral and determined that it contained a new element, which he named Thorium after Thor, the hammer-wielding Norse god of thunder and lightning.

    Thorium has a primary valence of +4; that is, a Thorium atom can donate four electrons to other atoms to form a compound molecule. Compounds based on that valence are stable, especially Thorium Tetrafluoride, which we can dissolve in molten Lithium Fluoride and Beryllium Difluoride for use in molten-salt reactors.

Physical properties of Thorium

    Pure Thorium is a silvery-white metal (density = 11.7 g/cm3) which is chemically stable in air and thus retains its luster for several months. But when itís contaminated with its oxide, Thorium slowly tarnishes in air, becoming gray and finally black. The degree of contamination with the oxide influences the physical properties of Thorium. The purest specimens often contain several tenths of a percent of the oxide.

    Pure Thorium is a relatively soft metal, very ductile, and it can be cold-rolled, swaged, and drawn. In its crystalline form Thorium is dimorphic, changing at 1360 Celsius from a face-centered cubic to a body-centered cubic structure; a body-centered tetragonal lattice form exists at high pressure with impurities determining the exact transition temperatures and pressures. Thorium has one of the largest liquid ranges of any element (2946 Celsius) between the melting point (1842 Celsius) and boiling point (4788 Celsius).

    Powdered Thorium metal is often pyrophoric and will often ignite spontaneously in air; thus, it requires careful handling. When heated in air, Thorium metal turnings ignite and burn brilliantly with a white light.

    Thorium exists in nature primarily in a single isotopic form - Th-232 - which decays very slowly (its half-life is about three times the age of the Earth Ė 14.05 billion yrs.). The decay chains of natural Thorium and Uranium give rise to minute traces of Th-228, Th-230 and Th-234, but the presence of these in mass terms is negligible.

    Natural Thorium decays very slowly with the emission of an alpha particle. The emitted alpha radiation cannot penetrate human skin, which means that owning and handling small amounts of Thorium, such as a gas mantle, is considered safe. But because alpha rays can penetrate lungs and other internal organs ingestion or inhalation of Thorium dust can lead to various cancers or other diseases.

Abundance of Thorium:

    Thorium occurs in small amounts in most rocks and soils; it is about as common as lead in Earthís crust, several times more abundant in Earth's crust than all isotopes of Uranium combined, with Th-232 being several hundred times more abundant than U-235. Soil commonly contains an average of around 12 parts per million of Thorium. Thorium also occurs concentrated in several minerals including thorite (ThSiO4), thorianite (ThO2 + UO2) and monazite. Thorianite may contain up to about 12% Thorium Oxide but occurs rarely. Monazite contains 2.5% Thorium.

    At present we have only poor knowledge of the distribution of Thorium resources because insignificant demand for Thorium has led to relatively low-key exploration efforts. Two sets of estimates define world Thorium reserves, one set by the US Geological Survey (USGS) and the other supported by reports from the Organization for Economic Cooperation and Development (OECD) and the International Atomic Energy Agency (the IAEA). Under the USGS estimate, USA, Australia and India have particularly large reserves of Thorium. Both the IAEA and OECD appear to conclude that India may possess the largest share of the world's Thorium deposits. The Government of India estimates the reserve at 846,477 tonnes.

    The prevailing estimate of the economically available thorium reserves comes from the US Geological Survey, Mineral Commodity Summaries (1996Ė2010):

Country             Reserves in tonnes (2010)

United States             440,000

Australia                300,000

Brazil                   16,000

Canada                100,000

India                  290,000 to 650,000

Malaysia                 4,500

South Africa              35,000

Other Countries           90,000

World Total            1,300,000 to 1,660,000

    Another estimate of reasonably assured reserves (RAR) and estimated additional reserves (EAR) of thorium comes from the OECD report "Trends in Nuclear Fuel Cycle", published in Paris, France in 2001:

Country             RAR (tonnes)             EAR (%)

Australia             489,000               19%

USA                400,000               15%

Turkey              344,000               13%

India               319,000               12%

Venezuela           300,000               12%

Brazil              302,000                12%

Norway            132,000                 5%

Egypt              100,000                 4%

Russia              75,000                 3%

Greenland           54,000                 2%

Canada             44,000                 2%

South Africa          18,000                 1%

Other countries        33,000                 1%

World total         2,610,000

    Those estimates show that at least enough Thorium exists in the United States to provide power to the country at its current level of energy usage for over 1,000 years.

Nuclear Structure of Thorium:

    The Thorium nucleus consists of 90 protons and 133 to 144 neutrons. For the isotopes and their fundamental properties we have:

Isotope         Half-life         Decay Mode         Decay Energy (Mev)

Th-223         0.9 sec         Alpha decay             7.7

Th-224         1.05 sec        Alpha decay             7.3

Th-225         8.0 min         Alpha decay*            6.92

Th-226         30.9 min        Alpha decay            6.45

Th-227         18.5 days       Alpha decay             6.145

Th-228         1.913 yr        Alpha decay             5.521

Th-229         7340 yr        Alpha decay             5.167

Th-230         75,380 yr       Alpha decay             4.767

Th-231         25.5 hr         Beta decay              0.381

Th-232         14.05 billion yr   Alpha decay             4.08

Th-233         22.2 min        Beta decay             1.246

Th-234         24.1 days       Beta decay              0.263

* The alpha decay provides over 95 percent of this isotopeís decay. The other mode of decay, providing less than 5 percent of the isotopeís decays, consists of electron capture with an energy of 0.68 Mev.

Nuclear reactions of U-233:

    In order to obtain power from Thorium we want to use U-233 as the fissionable fuel in nuclear reactors, though it does not occur naturally. To obtain U-233 we must bombard atoms of Th-232 with neutrons. When a nucleus of Th-232 absorbs a neutron it becomes a nucleus of Th-233, which has a half-life of about 22 minutes and decays into Pa-233 through beta decay. Pa-233 has a half-life of about 27 days and decays into U-233, also through beta decay. In a normally operating reactor U-233 will generate heat at the rate of 0.28 Watt per kilogram. If made to undergo complete fission, one pound (0.45 kilogram) of U-233 will provide the same amount of energy as burning 1,500 tons (1,350,000 kilograms) of coal.

    The fission of one atom of U-233 generates 197.9 MeV, which equals 3.171 ◊ 10 -11 Joule. Thus, U-233 releases 19.09 trillion Joules per mole or 81.95 trillion Joules per kilogram of heat in a reactor. The various contributions to that heat consist of:

Source Process                 Average energy released(MeV)

        Instantaneously released energy

Kinetic energy of fission fragments         168.2

Kinetic energy of prompt neutrons           4.9

Energy carried by prompt γ-rays            7.7

        Energy from decaying fission products

Energy of β -particles                   5.2

Energy of anti-neutrinos                 6.9

Energy of delayed γ-rays__________________   5.0_________________________

Sum                             197.9

    The production of U-233 begins with neutron irradiation of Th-232, which absorbs the neutrons to become Th-233. With a half-life of about 22 minutes, the Th-233 undergoes beta decay to form Pa-233, which has a half-life of 27 days. The Pa-233 undergoes beta decay to become U-233, which has a half-life of 160,000 years. In some proposed molten salt reactor designs the designers seek to isolate the Protactinium from further neutron capture before it can decay into U-233. If the Pa-233 were to absorb a neutron, it would become Pa-234, which decays to U-234. Because U-234 is not fissile, it must absorb another neutron in order to become the fissile U-235.

    U-233 usually undergoes fission when it absorbs a neutron, but sometimes it retains the neutron; that means that around 11% of the U-233 gets converted by further neutron absorption to U-235. In those cases neutron capture by U-233 yields U-234. Again the U-234 will have to capture a neutron before it can become U-235; thus, like Pa-233, U-233 must absorb two neutrons before it becomes another fissile nucleus.

Thorium as a source of nuclear fuel:

    The American public first became aware of U-233 derived from Thorium in 1946. It was described as "a third available source of nuclear energy and atom bombs" (in addition to U-235 and Pu-239). In the course of the Cold War the United States produced roughly 2 metric tons of U-233. That material was produced in reactors that were designed for the production of Plutonium-239. The production costs, estimated from the costs of Plutonium production, were $2-4 million/kg, much of the cost due to the need to separate the U-233 from other materials in the reactor. Today few reactors remain in the world with significant capabilities to produce more U-233.

    Like Uranium-238, Thorium-232 does not undergo fission, but upon absorption of a single neutron it transmutes into a properly fissile isotope. As U-238 becomes Pu-239, so Th-232 becomes U-233, which constitutes a highly fissile nuclear fuel. Any Thorium-based nuclear fuel cycle must thus involve some means to irradiate the Thorium with neutrons to convert it into the fissile material. For a breeder reactor U-233 gives us an obvious driver, though we might use U-235 or Pu-239 to get the process started.

    The basic idea behind the breeder reactor lies in the fact that Thorium can produce more U-233 than it consumes in producing it. Indeed, a thermal-neutron breeder reactor can only function with U-233 as the fissile driver and that reactor needs to make exceptionally efficient use of the neutrons (such as by preventing neutron loss through absorption or escape from the reactor). But such slow-neutron breeding stands as a unique feature of Thorium-based systems; it wonít work with Uranium.

    A reactor might start with a mix of Thorium and Plutonium. As the Plutonium gets used up the U-233 content increases and takes over the function of the driver. The reactor has then become a pure Thorium reactor, with only Thorium going in to produce power. Obtaining maximum energy output depends on the arrangement of the neutron flux, influenced by such factors as neutron absorption by the intermediate product Protactinium-233.

    Either fast reactors or thermal reactors can breed U-233 from Th-232. This differs from Uranium-based fuel cycles, which require the superior neutron economy of a fast reactor in order to breed Plutonium, producing more fissile material than the reactor consumes.

Thorium fuel cycle:

    Although it wonít undergo fission itself, Th-232 will absorb slow neutrons to produce, after two beta decays, U-233, which will undergo fission. Like U-238, then, it is fertile. But unlike Uranium, preparation of nuclear fuel from Thorium does not require isotopic separation.

    The U-233 created by the Thorium fuel cycle can be separated from the reactor's fuel and used for making nuclear weapons. But only a limited amount of U-233 ever exists in the reactor and in its heat-transfer systems, thereby preventing any ready access to weapons material. That fact gives us a compelling reason to prefer the use of a molten-salt reactor with its liquid-fuel cycle. However, the neutrons that the reactor produces can be absorbed by a blanket of Thorium or Uranium and fissile U-233 or Pu-239 produced.

    Spontaneous fission of U-233 produces a negligible neutron flux; thus, U-233 can serve as the explosive component in a simple gun-type nuclear bomb design. The first detonation of a nuclear bomb involving U-233, as part of the bomb core with Plutonium, occurred on 15 April 1955 in Operation Teapot. Although the test was successful, U-233 has been used only occasionally in nuclear weapons. The main reason behind that fact concerns how U-233 compares to Pu-239.

    U-233 has only one seventh the radioactivity of Pu-239 (159,200 years half-life versus 24,100 years), but its bare critical mass is 60% higher (16 kg versus 10 kg). The spontaneous fission rate of U-233 is twenty times higher than that of Pu-239, but because the radioactivity is lower, the neutron flux density is only three times higher. Thus, a nuclear explosive device based on U-233 is more of a technical challenge than one based on Plutonium, though the technological level involved is roughly the same. The main difference between the use of Uranium and Plutonium is the co-presence of U-232 which, because of its intense radioactivity, makes U-233 very dangerous to work on and quite easy to detect. That fact makes U-233 difficult to handle, providing a real deterrent to the proliferation of nuclear weapons.

The Thorium-based fuel cycle

    The Thorium fuel cycle, which includes the potential for breeding fuel without resorting to fast neutron reactors, holds out considerable promise in the long-term use of nuclear energy to power civilization until we discover something better.

    Because we use a liquid salt to contain the fuel, that salt also accumulates the fission products. To keep the reactor functional we must divert a certain proportion of the salt to a processing unit in order to modify its chemical composition. That modification requires a four-stage process:

    I: The first stage consists of extracting the Uranium from the diverted salt sample and reinjecting it directly in the main salt circulation. The Uranium thus does not continue on to the other levels of the processing unit. If the reactor operates as a breeder, the excess Uranium is removed in this stage.

    II: In the second stage, the Protactinium is extracted and allowed to decay into U-233 away from the neutron flux, which would transmute the Protactinium into species that do not contribute to the reactorís operation and would waste neutrons.

    III: In the third stage, as much of the fission products as feasible is extracted. Because those fission products are liable to capture neutrons, the chain reaction would not be sustainable if the fission products are not removed.

    IV: In the fourth and final stage fresh Thorium Tetrafluoride is injected into the salt to replace the Thorium that the reactor has used up.

    The rest of the salt, which remains unchanged, goes back into the reactor as is, where it will remain in the circuit indefinitely.

Nuclear Reactors using Thorium:

    Liquid-fluoride reactors use actinide fluoride salts dissolved in a carrier medium of low-absorption fluoride salt solvents. The solvents used in this application are low-melting point mixtures of Beryllium Fluoride (BeF2)(sublimes at 800 Celsius) and Lithium Fluoride (LiF)(melting point =845 Celsius) isotopically enhanced in the more-abundant isotope Lithium-7. The actinide fluorides most commonly used as the nuclear fuel are Thorium Tetrafluoride (ThF4)(melting point >900 Celsius) and Uranium Tetrafluoride (UF4)(melting point =960 Celsius). LiF-BeF2 salt mixtures have the advantages of very low neutron absorption, excellent heat capacity, stability under intense radiation, and the ability to dissolve appreciable amounts of Thorium Tetrafluoride or Uranium Tetrafluoride.

    The capture cross section of Th-232 and the fission and capture cross sections of U-233 are such that fuel conversion, and even fuel breeding can be achieved both with fast neutrons and with thermal neutrons in the Th-232/U-233 fuel cycle. With a Th-232/U-233 fuel, the cross section ratio is much better with thermal neutrons than with fast neutrons, so breeding can be achieved with less fissile matter than would be needed in a Uranium or Plutonium reactor (the fission of a U-233 nucleus yields an average of 2.5 neutrons). Alternatively the extra neutrons can be used for the transmutation of radioactive isotopes, e.g. some of the accumulating fission products.

    Neutron loss is a major factor in the design of a nuclear reactor. A too-small reactor would lose too many neutrons. A proper design requires a reactor large enough for most of the neutrons to interact with the fuel before they can leave the core. Most of the neutrons that do escape are sent back into the core by reflectors, thick graphite blocks that surround the reactor. Thanks to properly designed reflectors, neutron escapes from the reactor are effectively nonexistent.

    Sterile captures of neutrons in the graphite and the salt are small, as low as an average of 0.1 neutrons per fission. As for sterile captures in the fission products and in the Protactinium found in the Th-232 to U-233 decay tree, they have to be reduced and this can be done thanks to well designed and efficient on line fuel processing.

Liquid and Solid Fluoride Salt Mixtures: 7LiF - BeF2 - 233UF4.

    Despite providing some degree of neutron moderation, 4LiF-BeF2 mixtures are not terribly good neutron moderators, thus liquid-fluoride reactors generally employ solid moderating materials in order to moderate neutrons to thermal energies. Graphite is most commonly employed, being abundant, relatively inexpensive, and chemically compatible with the salt. Graphite is not "wetted" by the fluoride salt and can be sealed in ways that limit the intrusion of fission product gases (especially xenon) into the structure of the graphite. Further, the salt does not corrode the graphite, which fact makes graphite ideal to form the core of the reactor.

    The core of the reactor consists of a cylindrical block of graphite. Holes bored in the graphite serve as channels that allow the molten salt to circulate. As the solid structure of the reactor the graphite also serves as a neutron moderator. The molten salt that circulates in the channels, containing both the Thorium and Uranium needed to sustain the chain reaction, works as both the fuel and the coolant.

    The volume of Carbon, in the form of graphite, makes up 70 % of the total core volume. The volume of salt is distributed with 44.57% in the core, 22.17% in reservoirs, and 33.26% in the heat exchanger. The composition of the salt consists of: Lithium Fluoride (7LiF) with 70% of the atoms, Beryllium Fluoride (BeF2) with 17.5% of the atoms, and Th-232 and U-233 Fluorides accounting for 12,5% of the atoms, plus a small quantity of fission products. The nuclear fuel consists of 98.57% Thorium and 1.43% Uranium. For every Gigawatt of electricity generated the reactor consumes 2.6 kg of Thorium per day or about one metric ton per year. In 2010 the United States generated electricity at a rate of 1,137.3 Gigawatts, so the United States has enough known reserves of Thorium to last about four centuries at that rate of energy production.

    In the basic design, a molten-salt reactor, compared to other kinds of reactors, generates heat at higher temperatures, continuously, and without refueling shutdowns, so it can provide hot air to a more efficient Brayton-Cycle turbine. A molten-salt reactor run this way is about 30% better in thermal efficiency than common thermal plants, whether of the combustive or traditional solid-fueled nuclear type.

    The amount of power that we want to produce sets the reactor dimensions. If the salt temperature in the heat exchanger reaches 700 Celsius and declines to 600C, then the reactor has a conversion efficiency of 40% (if the exhaust temperature from the Brayton turbine is 300C); which means that for a yield of 10,000 Megawatts of thermal power coming out of the reactor we get 4,000 Megawatts of electric power.

Starting the Reactor

    U-233 does not occur in nature and, yet, a Thorium based molten salt reactor cannot be started without fissile matter. However, the Plutonium taken from a pressurized-water reactor's spent Uranium Oxide fuel can be used to kick-start a Thorium-based molten-salt reactor. The spent fuel is allowed to cool for five years after removal from the reactor, the Plutonium is extracted, and a Fluoride is made (PuF3) which is mixed with the Thorium Fluoride in the molten salt. Making the ratio of heavy nuclei in the salt equal to that of a Th-232/U-233 salt ensures that the reactor will reach the Th-232/U-233 equilibrium situation as quickly as possible. As the reactor incinerates the Plutonium, the fuel is replaced with Thorium. The U-233 produced from the Thorium thus progressively replaces the Plutonium as the fissile isotope in the reactor. After 15 years of operation, the fuel composition in the reactor is practically the same as that of a genuine Th-232/U-233 reactor. There is a difference, however, in that the irradiation of Plutonium generates more minor actinides than does irradiation of U-233. Those actinides are progressively incinerated in the reactor, certainly, but it takes more than a hundred years to reduce the minor actinides to the amount in a Th-232/U-233 reactor that was started directly with U-233. This difference also affects, obviously, the radiotoxicity of losses.

    After 200 years of production and 1000 years of cooling, the radiotoxicity of the losses from a Th-232/U-233 reactor started with Plutonium is about 4 times higher than in a reactor started directly with U-233. After the elapse of 1200 years the radiotoxicity of the fission products becomes constant because the medium lived products (those whose half-life is on the order of 30 years) have decayed to stable nuclei. We can compare that to the radiotoxicity induced by a pressurized-water reactor that generates the same power over the same time period. The radiotoxicity induced by a molten salt reactor is more than 10,000 times less than that due to a pressurized-water reactor similar to those presently operated in France, if only the heavy nuclei losses are considered in the case of a molten-salt reactor. If the molten-salt reactorís residual inventory is added to the heavy nuclei leaked during processing, the radiotoxicity is 100 times less than that of the pressurized-water reactor. The residual inventory would have to be included in the event that the reactor were stopped after 200 years of operation, presumably to let alternative energy sources take over.

Thorium and accelerator driven systems

    An accelerator driven system produces high-energy neutrons through the spallation reaction of high-energy protons from an accelerator striking heavy target nuclei (such as Lead, Lead-Bismuth or other material). Spallation denotes the process whereby high-energy particles strike a heavy nucleus and cause the ejection of nucleons. In this case, a high-energy proton beam directed at a heavy target causes the expulsion of a number of spallation particles, including neutrons.

    The system can then direct those neutrons to a subcritical reactor containing Thorium, where the neutrons breed U-233 and promote the fission of it. That system therefore provides the possibility of sustaining a fission reaction which can readily be turned off, and used either for power generation or destruction of actinides resulting from the Uranium/Plutonium fuel cycle. The use of Thorium instead of Uranium reduces the quantity of actinides that the reaction produces.

Advantages of Thorium reactors:

    Unlike Uranium-based or Plutonium-based reactors, a Thorium-based reactor, with its on line fuel processing, uses up 100% of its fuel. Further, because fuel processing occurs continuously on site, the reactor needs only one fuel inventory (whereas, a reactor with solid fuel needs two inventories so that the reactor can continue to operate while its spent fuel is being processed). Moreover, the inventory is small: 70 metric tons of Th-232 with one metric ton of U-233 for the production of 1 Gigawatt of electric power. Finally, on site processing reduces to a minimum the transport of radioactive materials.

    If fuel processing includes Protactinium extraction, as in the liquid-fluoride reactors, excess neutrons will be available. Those neutrons can be used for one of two things:

    a) for breeding: the reactor can produce 5% excess U-233, producing a doubling time of 25 years; thus, after 25 years' operation the reactor has produced enough excess Uranium to start a new, similar, reactor.

    b) For the transmutation of radioactive fission products, such as Tc-99 and I-129, into non-radioactive isotopes.

    Thorium also makes a very effective radiation shield, although it has not been used for this purpose as much as Lead or depleted Uranium. Thus Thorium can play roles other than that of fuel in a Thorium-based reactor.

    We can summarize some of the benefits of Thorium as a nuclear fuel when compared with Uranium as follows:

    1. Weapons-grade fissionable material (U-233) is harder to retrieve safely and clandestinely from a Thorium reactor due to the intense radioactivity in the reactor. Separated U-233 is always contaminated with traces of highly radioactive U-232 (68.9 year half-life but whose daughter products such as Tl-208 are strong gamma emitters with very short half-lives). This fact confers proliferation resistance on the fuel cycle by making U-233 hard to handle and easy to detect, though it results in increased costs;

    2. A Thorium reactor produces 10 to 10,000 times less long-lived radioactive waste than do Uranium or Plutonium reactors;

    3. Thorium comes out of the ground as nearly 100% pure Th-232; thus, it does not require enrichment, whereas natural uranium contains only 0.7% fissionable U-235.

    4.Thorium cannot sustain a nuclear chain reaction by itself, so fission stops by default if Protactinium is not removed from the reactor and the resulting U-233 put back.

    5.Unlike the Uranium in breeder reactors, Thorium requires irradiation by external neutron sources and reprocessing to produce fissionable fuel. That fact makes Thorium-based fuels initially more expensive than Uranium fuels. But the first Thorium reactor may activate a second Thorium reactor, then a third, and so on.

    6. The known abundance of Thorium guarantees that, at current levels of demand for electricity, Thorium power can satisfy that demand for about 400 years (and the estimated abundance extends that time to 1000 years).

Recycling the Wastes of Fission

    Nuclear fission inevitably generates a variety of fission products, various elements in a variety of their isotopes, most of which are neutron-rich and, thus, radioactive. As each of these fission products tends to have many more neutrons than it needs for nuclear stability, rapid beta decay generally follows fission and most fission products assume a stable form quite quickly. When all of the isotopes of an element reach stability we can logically ask whether or not they are worth chemical extraction and recycling to other, non-nuclear uses.

    Xenon provides a good example. Xenon accounts for a fair fraction of the mass of fission products from the fission of Uranium. Xenon has a variety of isotopes but the longest lived one (Xe-133) has a half-life of merely 5.2 days. Therefore, if we proceed on the rule-of-thumb of "ten half-lives and itís gone", after a little less than two months of storage the Xenon remaining from fission would have become essentially non-radioactive. In a molten fluoride reactor we can easily extract the Xenon: because itís a noble gas (which doesnít combine chemically with anything) the Xenon simply bubbles out of the fluoride liquid like some strange nuclear fizz. Rather than discard the Xenon, we can capture it and sell it. We can see its value in the fact that NASA and commercial satellite operators, for example, use Xenon as the propellant in the ion engines of spacecraft.

    Neodymium gives us another valuable fission product, the third-most-common element generated from fission by mass. It decays into a stable form relatively quickly; its longest-lived isotope (Nd-147) has a half-life of 10.9 days. By aging the radioactive waste from the distillation process in fluoride reactors for several months, we can subsequently extract the Neodymium Trifluoride from the other fluorides and convert it to a metallic form through electrolysis or metallic reduction. The Neodymium would then be available to sell to the burgeoning market. For example, a Neodymium-Iron-Boron alloy can be used to make super-strong, super-light magnets that can be used in the large electrical generators used in wind turbines.

    Xenon and Neodymium represent two opportunities in which a period of aging must elapse before the isotopes stabilize and marketing becomes possible. But other isotopes occur in the waste stream of a molten-fluoride reactor, isotopes whose radioactive form is the desirable and economic product. For example, consider the life-saving medical isotope Molybdenum-99. Currently, specially-designed medical isotope production reactors in Canada generate Mo-99, which is then rushed to medical facilities across North America. Mo-99, with a half-life of 66.69 hours, decays to Technetium-99m, which is then extracted and introduced into human patients in order to facilitate diagnostic procedures. The market for Mo-99 is quite large, but in solid-fueled reactors, the Mo-99 produced by fission is not accessible until the fuel is reprocessed, an infrequent event that allows most of the Mo-99 to decay into Tc-99m, which then decays, with a half-life of 6 hours, to Tc-99 by isomeric transition, rendering the Technetium useless. In a molten-fluoride reactor, on the other hand, the fluid nature of the reactor makes it possible to extract Mo-99 continuously along with the other isotopes of Molybdenum. Molybdenum forms a volatile hexafluoride just as Uranium does, so when the fuel salt is fluorinated, Uranium, Molybdenum, and several other elements come out of solution as gaseous hexafluorides. These can then be separated one from another by distillation in much the same way as crude oil is refined. The Mo-99 can then be shipped to medical facilities.

    Xenon, Molybdenum, and Neodymium are three of the most common fission products, but many other fission products have value too. The fluid nature of the fluoride reactor makes recycling of the waste quite likely to be economically attractive in many circumstances.

    On the other hand, the production of transuranic nuclear waste gives us one of the biggest concerns about todayís approach to nuclear power generation. Reducing the amount of transuranic waste sent to any repository provides an important challenge to the future development of civilian nuclear power generation. The use of Thorium in a liquid-fluoride reactor makes meeting that challenge a straightforward proposition. A suitable combination of the inherent properties of the Thorium fuel approach and of the flexibility of the liquid-fluoride fuel form can drastically reduce the production of transuranic waste.

    Thorium-232 begins the process of generating nuclear energy at least five neutron absorptions removed from the first transuranic isotope that it can generate. In that process Th-232 absorbs a neutron, transmuting into Pa-233 and then decaying into U-233, which is fissile. Bombarded by thermal neutrons, U-233 tends to fission 90% of the time it absorbs one of those neutrons. The other 10% of the time a nucleus of U-233 becomes one of U-234. A further neutron absorption by the U-234 produces a nucleus of U-235. U-235 fissions in a thermal neutron spectrum approximately 85% of the time; the other 15% of the time it becomes a nucleus of U-236. Upon absorbing another neutron U-236 becomes the first transuranic isotope of this process, Np-237. The Neptunium can be removed from the fluoride salt mixture readily by fluorination, which changes it from NpF4 to gaseous NpF6. Thus, in the Thorium-based approach the fuel lies five neutron absorptions away from the production of a transuranic isotope and in the course of those absorptions roughly 98.5% of the original fuel is removed by fission. Contrast that fact with the Uranium-Oxide approach to nuclear power, in which the majority of the fuel (97% U-238) is a single neutron absorption away from the production of the first transuranic isotope (Pu-239). Thus the use of Thorium in the liquid-fluoride reactor rather than the use of Uranium in the solid-oxide reactor reduces the amount of transuranic material generated by a very large factor.

Reusing Nuclear Fuel

    In a Uranium reactor the solid-oxide fuel rods must be replaced periodically by new fuel rods. The spent fuel rods are sent to a cooling pond where decay heat can be removed. The spent fuel still contains large amounts of unused fuel in the form of both Uranium and other actinides, but that fuel cannot be used until itís reprocessed. The reprocessing involves changing the solid Uranium Oxide into liquid Uranium Nitrate by dissolving it in strong Nitric Acid. Then a combination of chemical processes in a variety of solvents takes place to separate fission products, transuranics, and Uranium from one another. Many recycles of the fuel would be needed to incinerate most of the U-238 present in the original spent fuel, but the costs involved in reprocessing necessitate that spent nuclear fuel is rarely subjected to more than one or two recycles before it is sent to a nuclear-waste repository.

    The liquid-fluoride Thorium reactor presents an entirely different approach to fuel management that makes repeated recycling easy and economical. That is because the nuclear fuel is already in a chemically stable form as a fluoride. Thus the fuel-laden salt is protected from chemical attack, combustion, and corrosion. As a fluid the salt is in a form in which chemical processes can be used to remove fission products or to add new fuel to replace that lost to fission. In addition, the ionic nature of liquid-fluoride salt renders the fuel essentially impervious to radiation damage: the fuel remains chemically unaltered and completely retains its physical properties, in spite of the passage of intense alpha, beta, gamma, and neutron radiation.

    In order to purify the fuel salt, the first step is to remove the uranium fuel by fluorination. Then the carrier salt (LiF-BeF2) can be distilled from fission product fluorides in a high-temperature still at over 1400C. The remaining fission product fluorides constitute the equivalent of "high-level waste" from fluoride reactor reprocessing. The extracted LiF-BeF2 is recombined with the uranium and reinserted into the reactor core for another cycle of power generation.

    In the salt, made molten by the heat of the core, newly-formed U-233 forms soluble Uranium Tetrafluoride (UF4). Bubbling Fluorine through the blanket converts that tetrafluoride into gaseous Uranium Hexafluoride (UF6). That process does not chemically affect the less-reactive Thorium Tetrafluoride. The Uranium Hexafluoride fizzes out of the salt solution to be captured and then reduced back to soluble UF4 by hydrogen gas in a reduction column. Finally the UF4 is put back into the core to serve as fissile fuel.

    An interesting feature is that all heavy nuclei remain in the reactor, except for those that leak out during processing: a few heavy nuclei (Uranium and transuranics) are taken along with the fission products when they are extracted; chemical reactions are never perfect. These losses are estimated to be about 10-5 (one heavy nucleus taken along for each 100,000 fission product nuclei removed). Because they remain in the salt indefinitely, all heavy nuclei (other than those lost) will eventually fission. The actinides produced in the reactor are thus incinerated in the reactor. Excluding losses, the only radioactive nuclei that escape the reactor are the fission products.

    Radiotoxic losses can be further reduced with off line processing of the fission products: the transuranics and the Uranium that are taken with the fission products could be retrieved thanks to more thorough processing in a specialized unit, and put back into the reactor to be incinerated.


    Once we have Thorium reactors up and running, we can use them to generate electricity and/or to provide process heat for chemical syntheses.

    In one such application we can use the reactorís waste heat to drive the Sabatier reaction. That reaction produces methane (natural gas) and water from hydrogen and carbon dioxide, ideally at temperatures between 300 and 400 Celsius and at elevated pressure. The only special requirement is that of a suitable catalyst, such as nickel or, more effectively, ruthenium on alumina. If we were to use the exhaust heat from a Thorium reactor to manufacture methane and do it with enough reactors, then we might be able to put an end to the vile practice of fracking, which many people claim has polluted their drinking water.

    Taking the heat at the outlet temperature of 700 Celsius, we can use it as process heat for the production of ammonia, which we can then use as fuel in cars, trucks, boats, and trains (see Ammonia-Based Fuel Cells).

    If we take the heat at an average of 650 Celsius and reduce it to 300 Celsius in a turbine, we will get electricity at a theoretical efficiency of 37.92%. We can then use the waste heat to drive the Sabatier reaction, hopefully producing methane at a price that will make the vile practice of fracking economically unviable.


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