Thorium fuel cycle

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A sample of thorium Thorium sample 0.1g.jpg
A sample of thorium

The thorium fuel cycle is a nuclear fuel cycle that uses an isotope of thorium, 232
Th
, as the fertile material. In the reactor, 232
Th
is transmuted into the fissile artificial uranium isotope 233
U
which is the nuclear fuel. Unlike natural uranium, natural thorium contains only trace amounts of fissile material (such as 231
Th
), which are insufficient to initiate a nuclear chain reaction. Additional fissile material or another neutron source is necessary to initiate the fuel cycle. In a thorium-fuelled reactor, 232
Th
absorbs neutrons to produce 233
U
. This parallels the process in uranium breeder reactors whereby fertile 238
U
absorbs neutrons to form fissile 239
Pu
. Depending on the design of the reactor and fuel cycle, the generated 233
U
either fissions in situ or is chemically separated from the used nuclear fuel and formed into new nuclear fuel.

Contents

The thorium fuel cycle has several potential advantages over a uranium fuel cycle, including thorium's greater abundance, superior physical and nuclear properties, reduced plutonium and actinide production, [1] and better resistance to nuclear weapons proliferation when used in a traditional light water reactor [1] [2] though not in a molten salt reactor. [3] [4]

History

Concerns about the limits of worldwide uranium resources motivated initial interest in the thorium fuel cycle. [5] It was envisioned that as uranium reserves were depleted, thorium would supplement uranium as a fertile material. However, for most countries uranium was relatively abundant and research in thorium fuel cycles waned. A notable exception was India's three-stage nuclear power programme. [6] In the twenty-first century thorium's potential for improving proliferation resistance and waste characteristics led to renewed interest in the thorium fuel cycle. [7] [8] [9]

At Oak Ridge National Laboratory in the 1960s, the Molten-Salt Reactor Experiment used 233
U
as the fissile fuel in an experiment to demonstrate a part of the Molten Salt Breeder Reactor that was designed to operate on the thorium fuel cycle. Molten salt reactor (MSR) experiments assessed thorium's feasibility, using thorium(IV) fluoride dissolved in a molten salt fluid that eliminated the need to fabricate fuel elements. The MSR program was defunded in 1976 after its patron Alvin Weinberg was fired. [10]

In 1993, Carlo Rubbia proposed the concept of an energy amplifier or "accelerator driven system" (ADS), which he saw as a novel and safe way to produce nuclear energy that exploited existing accelerator technologies. Rubbia's proposal offered the potential to incinerate high-activity nuclear waste and produce energy from natural thorium and depleted uranium. [11] [12]

Kirk Sorensen, former NASA scientist and Chief Technologist at Flibe Energy, has been a long-time promoter of thorium fuel cycle and particularly liquid fluoride thorium reactors (LFTRs). He first researched thorium reactors while working at NASA, while evaluating power plant designs suitable for lunar colonies. In 2006 Sorensen started "energyfromthorium.com" to promote and make information available about this technology. [13]

A 2011 MIT study concluded that although there is little in the way of barriers to a thorium fuel cycle, with current or near term light-water reactor designs there is also little incentive for any significant market penetration to occur. As such they conclude there is little chance of thorium cycles replacing conventional uranium cycles in the current nuclear power market, despite the potential benefits. [14]

Nuclear reactions with thorium

"Thorium is like wet wood […it] needs to be turned into fissile uranium just as wet wood needs to be dried in a furnace."

Ratan Kumar Sinha, former Chairman of the Atomic Energy Commission of India. [15]

In the thorium cycle, fuel is formed when 232
Th
captures a neutron (whether in a fast reactor or thermal reactor) to become 233
Th
. This normally emits an electron and an anti-neutrino (
ν
) by
β
decay
to become 233
Pa
. This then emits another electron and anti-neutrino by a second
β
decay to become 233
U
, the fuel:

Fission product wastes

Nuclear fission produces radioactive fission products which can have half-lives from days to greater than 200,000 years. According to some toxicity studies, [16] the thorium cycle can fully recycle actinide wastes and only emit fission product wastes, and after a few hundred years, the waste from a thorium reactor can be less toxic than the uranium ore that would have been used to produce low enriched uranium fuel for a light water reactor of the same power. Other studies assume some actinide losses and find that actinide wastes dominate thorium cycle waste radioactivity at some future periods. [17]

Actinide wastes

In a reactor, when a neutron hits a fissile atom (such as certain isotopes of uranium), it either splits the nucleus or is captured and transmutes the atom. In the case of 233
U
, the transmutations tend to produce useful nuclear fuels rather than transuranic wastes. When 233
U
absorbs a neutron, it either fissions or becomes 234
U
. The chance of fissioning on absorption of a thermal neutron is about 92%; the capture-to-fission ratio of 233
U
, therefore, is about 1:12 – which is better than the corresponding capture vs. fission ratios of 235
U
(about 1:6), or 239
Pu
or 241
Pu
(both about 1:3). [5] [18] The result is less transuranic waste than in a reactor using the uranium-plutonium fuel cycle.

Transmutations in the thorium fuel cycle
237Np
231U232U233U234U235U236U237U
231Pa232Pa233Pa234Pa
230Th231Th232Th233Th
(Nuclides before a yellow background in italic have half-lives under 30 days;
nuclides in bold have half-lives over 1,000,000 years;
nuclides in red frames are fissile)

234
U
, like most actinides with an even number of neutrons, is not fissile, but neutron capture produces fissile 235
U
. If the fissile isotope fails to fission on neutron capture, it produces 236
U
, 237
Np
, 238
Pu
, and eventually fissile 239
Pu
and heavier isotopes of plutonium. The 237
Np
can be removed and stored as waste or retained and transmuted to plutonium, where more of it fissions, while the remainder becomes 242
Pu
, then americium and curium, which in turn can be removed as waste or returned to reactors for further transmutation and fission.

However, the 231
Pa
(with a half-life of 3.27×104 years) formed via (n,2n) reactions with 232
Th
(yielding 231
Th
that decays to 231
Pa
), while not a transuranic waste, is a major contributor to the long-term radiotoxicity of spent nuclear fuel.

Uranium-232 contamination

232
U
is also formed in this process, via (n,2n) reactions between fast neutrons and 233
U
, 233
Pa
, and 232
Th
:

Unlike most even numbered heavy isotopes, 232
U
is also a fissile fuel fissioning just over half the time when it absorbs a thermal neutron. [19] 232
U
has a relatively short half-life (68.9 years), and some decay products emit high energy gamma radiation, such as 224
Rn
, 212
Bi
and particularly 208
Tl
. The full decay chain, along with half-lives and relevant gamma energies, is:

The 4n decay chain of Th, commonly called the "thorium series" Decay chain(4n,Thorium series).PNG
The 4n decay chain of Th, commonly called the "thorium series"

232
U
decays to 228
Th
where it joins the decay chain of 232
Th

Thorium-cycle fuels produce hard gamma emissions, which damage electronics, limiting their use in bombs. 232
U
cannot be chemically separated from 233
U
from used nuclear fuel; however, chemical separation of thorium from uranium removes the decay product 228
Th
and the radiation from the rest of the decay chain, which gradually build up as 228
Th
reaccumulates. The contamination could also be avoided by using a molten-salt breeder reactor and separating the 233
Pa
before it decays into 233
U
. [3] The hard gamma emissions also create a radiological hazard which requires remote handling during reprocessing.

Nuclear fuel

As a fertile material thorium is similar to 238
U
, the major part of natural and depleted uranium. The thermal neutron absorption cross sectiona) and resonance integral (average of neutron cross sections over intermediate neutron energies) for 232
Th
are about three and one third times those of the respective values for 238
U
.

Advantages

The primary physical advantage of thorium fuel is that it uniquely makes possible a breeder reactor that runs with slow neutrons, otherwise known as a thermal breeder reactor. [5] These reactors are often considered simpler than the more traditional fast-neutron breeders. Although the thermal neutron fission cross section (σf) of the resulting 233
U
is comparable to 235
U
and 239
Pu
, it has a much lower capture cross section (σγ) than the latter two fissile isotopes, providing fewer non-fissile neutron absorptions and improved neutron economy. The ratio of neutrons released per neutron absorbed (η) in 233
U
is greater than two over a wide range of energies, including the thermal spectrum. A breeding reactor in the uranium - plutonium cycle needs to use fast neutrons, because in the thermal spectrum one neutron absorbed by 239
Pu
on average leads to less than two neutrons.

Thorium is estimated to be about three to four times more abundant than uranium in Earth's crust, [20] although present knowledge of reserves is limited. Current demand for thorium has been satisfied as a by-product of rare-earth extraction from monazite sands. Notably, there is very little thorium dissolved in seawater, so seawater extraction is not viable, as it is with uranium. Using breeder reactors, known thorium and uranium resources can both generate world-scale energy for thousands of years.

Thorium-based fuels also display favorable physical and chemical properties that improve reactor and repository performance. Compared to the predominant reactor fuel, uranium dioxide (UO
2
), thorium dioxide (ThO
2
) has a higher melting point, higher thermal conductivity, and lower coefficient of thermal expansion. Thorium dioxide also exhibits greater chemical stability and, unlike uranium dioxide, does not further oxidize. [5]

Because the 233
U
produced in thorium fuels is significantly contaminated with 232
U
in proposed power reactor designs, thorium-based used nuclear fuel possesses inherent proliferation resistance. 232
U
cannot be chemically separated from 233
U
and has several decay products that emit high-energy gamma radiation. These high-energy photons are a radiological hazard that necessitate the use of remote handling of separated uranium and aid in the passive detection of such materials.

The long-term (on the order of roughly 103 to 106 years) radiological hazard of conventional uranium-based used nuclear fuel is dominated by plutonium and other minor actinides, after which long-lived fission products become significant contributors again. A single neutron capture in 238
U
is sufficient to produce transuranic elements, whereas five captures are generally necessary to do so from 232
Th
. 98–99% of thorium-cycle fuel nuclei would fission at either 233
U
or 235
U
, so fewer long-lived transuranics are produced. Because of this, thorium is a potentially attractive alternative to uranium in mixed oxide (MOX) fuels to minimize the generation of transuranics and maximize the destruction of plutonium. [21]

Disadvantages

There are several challenges to the application of thorium as a nuclear fuel, particularly for solid fuel reactors:

In contrast to uranium, naturally occurring thorium is effectively mononuclidic and contains no fissile isotopes; fissile material, generally 233
U
, 235
U
or plutonium, must be added to achieve criticality. This, along with the high sintering temperature necessary to make thorium-dioxide fuel, complicates fuel fabrication. Oak Ridge National Laboratory experimented with thorium tetrafluoride as fuel in a molten salt reactor from 1964–1969, which was expected to be easier to process and separate from contaminants that slow or stop the chain reaction.

In an open fuel cycle (i.e. utilizing 233
U
in situ), higher burnup is necessary to achieve a favorable neutron economy. Although thorium dioxide performed well at burnups of 170,000 MWd/t and 150,000 MWd/t at Fort St. Vrain Generating Station and AVR respectively, [5] challenges complicate achieving this in light water reactors (LWR), which compose the vast majority of existing power reactors.

In a once-through thorium fuel cycle, thorium-based fuels produce far less long-lived transuranics than uranium-based fuels, some long-lived actinide products constitute a long-term radiological impact, especially 231
Pa
and 233
U
. [16] On a closed cycle,233
U
and 231
Pa
can be reprocessed. 231
Pa
is also considered an excellent burnable poison absorber in light water reactors. [22]

Another challenge associated with the thorium fuel cycle is the comparatively long interval over which 232
Th
breeds to 233
U
. The half-life of 233
Pa
is about 27 days, which is an order of magnitude longer than the half-life of 239
Np
. As a result, substantial 233
Pa
develops in thorium-based fuels. 233
Pa
is a significant neutron absorber and, although it eventually breeds into fissile 235
U
, this requires two more neutron absorptions, which degrades neutron economy and increases the likelihood of transuranic production.

Alternatively, if solid thorium is used in a closed fuel cycle in which 233
U
is recycled, remote handling is necessary for fuel fabrication because of the high radiation levels resulting from the decay products of 232
U
. This is also true of recycled thorium because of the presence of 228
Th
, which is part of the 232
U
decay sequence. Further, unlike proven uranium fuel recycling technology (e.g. PUREX), recycling technology for thorium (e.g. THOREX) is only under development.

Although the presence of 232
U
complicates matters, there are public documents showing that 233
U
has been used once in a nuclear weapon test. The United States tested a composite 233
U
-plutonium bomb core in the MET (Military Effects Test) blast during Operation Teapot in 1955, though with much lower yield than expected. [23]

Advocates for liquid core and molten salt reactors such as LFTRs claim that these technologies negate thorium's disadvantages present in solid fuelled reactors. As only two liquid-core fluoride salt reactors have been built (the ORNL ARE and MSRE) and neither have used thorium, it is hard to validate the exact benefits. [5]

Reactors

Thorium fuels have fueled several different reactor types, including light water reactors, heavy water reactors, high temperature gas reactors, sodium-cooled fast reactors, and molten salt reactors. [24]

List of thorium-fueled reactors

From IAEA TECDOC-1450 "Thorium Fuel Cycle – Potential Benefits and Challenges", Table 1: Thorium utilization in different experimental and power reactors. [5] Additionally, Dresden 1 in the United States used "thorium oxide corner rods". [25]

NameCountry Reactor type PowerFuelOperation period
AVR Germany (West) HTGR, experimental (pebble bed reactor) 15 MW(e)Th+235
U
Driver fuel, coated fuel particles, oxide & dicarbides
1967–1988
THTR-300 Germany (West) HTGR, power (pebble type) 300 MW(e)Th+235
U
, Driver fuel, coated fuel particles, oxide & dicarbides
1985–1989
Lingen Germany (West) BWR irradiation-testing 60 MW(e)Test fuel (Th,Pu)O2 pellets1968–1973
Dragon (OECD-Euratom)UK (also Sweden, Norway and Switzerland) HTGR, Experimental (pin-in-block design) 20 MWtTh+235
U
Driver fuel, coated fuel particles, oxide & dicarbides
1966–1973
Peach Bottom United States HTGR, Experimental (prismatic block) 40 MW(e)Th+235
U
Driver fuel, coated fuel particles, oxide & dicarbides
1966–1972
Fort St Vrain United States HTGR, Power (prismatic block) 330 MW(e)Th+235
U
Driver fuel, coated fuel particles, Dicarbide
1976–1989
MSRE ORNL United States MSR 7.5 MWt233
U
molten fluorides
1964–1969
BORAX-IV & Elk River Station United StatesBWR (pin assemblies) 2.4 MW(e); 24 MW(e)Th+235
U
Driver fuel oxide pellets
1963–1968
Shippingport United States LWBR, PWR, (pin assemblies) 100 MW(e)Th+233
U
Driver fuel, oxide pellets
1977–1982
Indian Point 1 United States LWBR, PWR, (pin assemblies) 285 MW(e)Th+233
U
Driver fuel, oxide pellets
1962–1980
SUSPOP/KSTR KEMA NetherlandsAqueous homogenous suspension (pin assemblies) 1 MWtTh+HEU, oxide pellets1974–1977
NRX & NRU CanadaMTR (pin assemblies) 20 MW; 200 MW (see)Th+235
U
, Test Fuel
1947 (NRX) + 1957 (NRU); Irradiation–testing of few fuel elements
CIRUS; DHRUVA; & KAMINI IndiaMTR thermal 40 MWt; 100 MWt; 30 kWt (low power, research)Al+233
U
Driver fuel, ‘J’ rod of Th & ThO2, ‘J’ rod of ThO2
1960–2010 (CIRUS); others in operation
KAPS 1 &2; KGS 1 & 2; RAPS 2, 3 & 4 India PHWR, (pin assemblies) 220 MW(e)ThO2 pellets (for neutron flux flattening of initial core after start-up)1980 (RAPS 2) +; continuing in all new PHWRs
FBTR India LMFBR, (pin assemblies) 40 MWtThO2 blanket1985; in operation
Petten NetherlandsHigh Flux Reactor thorium molten salt experiment 45 MW(e)?2024; planned

See also

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Fissile material

In nuclear engineering, fissile material is material capable of sustaining a nuclear fission chain reaction. By definition, fissile material can sustain a chain reaction with neutrons of thermal energy. The predominant neutron energy may be typified by either slow neutrons or fast neutrons. Fissile material can be used to fuel thermal-neutron reactors, fast-neutron reactors and nuclear explosives.

Nuclear fuel cycle Process of manufacturing and consuming nuclear fuel

The nuclear fuel cycle, also called nuclear fuel chain, is the progression of nuclear fuel through a series of differing stages. It consists of steps in the front end, which are the preparation of the fuel, steps in the service period in which the fuel is used during reactor operation, and steps in the back end, which are necessary to safely manage, contain, and either reprocess or dispose of spent nuclear fuel. If spent fuel is not reprocessed, the fuel cycle is referred to as an open fuel cycle ; if the spent fuel is reprocessed, it is referred to as a closed fuel cycle.

Mixed oxide fuel, commonly referred to as MOX fuel, is nuclear fuel that contains more than one oxide of fissile material, usually consisting of plutonium blended with natural uranium, reprocessed uranium, or depleted uranium. MOX fuel is an alternative to the low-enriched uranium (LEU) fuel used in the light water reactors that predominate nuclear power generation.

Breeder reactor type of fast neutron reactor that produces more fissile material than it consumes

A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use, by irradiation of a fertile material, such as uranium-238 or thorium-232 that is loaded into the reactor along with fissile fuel. Breeders were at first found attractive because they made more complete use of uranium fuel than light water reactors, but interest declined after the 1960s as more uranium reserves were found, and new methods of uranium enrichment reduced fuel costs.

Uranium-238 Isotope of uranium

Uranium-238 is the most common isotope of uranium found in nature, with a relative abundance of 99%. Unlike uranium-235, it is non-fissile, which means it cannot sustain a chain reaction in a thermal-neutron reactor. However, it is fissionable by fast neutrons, and is fertile, meaning it can be transmuted to fissile plutonium-239. 238U cannot support a chain reaction because inelastic scattering reduces neutron energy below the range where fast fission of one or more next-generation nuclei is probable. Doppler broadening of 238U's neutron absorption resonances, increasing absorption as fuel temperature increases, is also an essential negative feedback mechanism for reactor control.

Fast-neutron reactor Type of nuclear reactor

A fast-neutron reactor (FNR) or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons, as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.

Integral fast reactor

The integral fast reactor is a design for a nuclear reactor using fast neutrons and no neutron moderator. IFR would breed more fuel and is distinguished by a nuclear fuel cycle that uses reprocessing via electrorefining at the reactor site.

Molten salt reactor Type of nuclear reactor cooled by molten material

A molten salt reactor (MSR) is a class of nuclear fission reactor in which the primary nuclear reactor coolant and/or the fuel is a molten salt mixture. Key characteristics are operation at or close to atmospheric pressure, rather than the 75-150 times atmospheric pressure of typical light-water reactors (LWR), hence reducing the large, expensive containment structures used for LWRs and eliminating a source of explosion risk; and higher operating temperatures than in a traditional LWR, hence higher electricity-generation efficiency and in some cases process-heat opportunities. Design challenges include the corrosivity of hot salts and the changing chemical composition of the salt as it is transmuted by reactor radiation.

Uranium-233 (233U) is a fissile isotope of uranium that is bred from thorium-232 as part of the thorium fuel cycle. Uranium-233 was investigated for use in nuclear weapons and as a reactor fuel. It has been used successfully in experimental nuclear reactors and has been proposed for much wider use as a nuclear fuel. It has a half-life of 160,000 years.

Uranium (92U) is a naturally occurring radioactive element that has no stable isotope. It has two primordial isotopes, uranium-238 and uranium-235, that have long half-lives and are found in appreciable quantity in the Earth's crust. The decay product uranium-234 is also found. Other isotopes such as uranium-233 have been produced in breeder reactors. In addition to isotopes found in nature or nuclear reactors, many isotopes with far shorter half-lives have been produced, ranging from 215U to 242U. The standard atomic weight of natural uranium is 238.02891(3).

Plutonium (94Pu) is an artificial element, except for trace quantities resulting from neutron capture by uranium, and thus a standard atomic weight cannot be given. Like all artificial elements, it has no stable isotopes. It was synthesized long before being found in nature, the first isotope synthesized being 238Pu in 1940. Twenty plutonium radioisotopes have been characterized. The most stable are plutonium-244 with a half-life of 80.8 million years, plutonium-242 with a half-life of 373,300 years, and plutonium-239 with a half-life of 24,110 years. All of the remaining radioactive isotopes have half-lives that are less than 7,000 years. This element also has eight meta states; all have half-lives of less than one second.

Spent nuclear fuel nuclear fuel that has been irradiated in a nuclear reactor (usually at a nuclear power plant)

Spent nuclear fuel, occasionally called used nuclear fuel, is nuclear fuel that has been irradiated in a nuclear reactor. It is no longer useful in sustaining a nuclear reaction in an ordinary thermal reactor and depending on its point along the nuclear fuel cycle, it may have considerably different isotopic constituents.

Weapons-grade nuclear material substance that is pure enough to be used to make a weapon

Weapons-grade nuclear material is any fissionable nuclear material that is pure enough to make a nuclear weapon or has properties that make it particularly suitable for nuclear weapons use. Plutonium and uranium in grades normally used in nuclear weapons are the most common examples.

Uranium-232 is an isotope of uranium. It has a half-life of around 68.9 years and is a side product in the thorium cycle. It has been cited as an obstacle to nuclear proliferation using 233U as the fissile material, because the intense gamma radiation emitted by 208Tl makes the 233U contaminated with it more difficult to handle.

Uranium-236 (236U) is an isotope of uranium that is neither fissile with thermal neutrons, nor very good fertile material, but is generally considered a nuisance and long-lived radioactive waste. It is found in spent nuclear fuel and in the reprocessed uranium made from spent nuclear fuel.

Long-lived fission products (LLFPs) are radioactive materials with a long half-life produced by nuclear fission of uranium and plutonium.

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The liquid fluoride thorium reactor is a type of molten salt reactor. LFTRs use the thorium fuel cycle with a fluoride-based, molten, liquid salt for fuel. In a typical design, the liquid is pumped between a critical core and an external heat exchanger where the heat is transferred to a nonradioactive secondary salt. The secondary salt then transfers its heat to a steam turbine or closed-cycle gas turbine.

Traveling wave reactor type of nuclear fission reactor

A traveling-wave reactor (TWR) is a proposed type of nuclear fission reactor that can convert fertile material into usable fuel through nuclear transmutation, in tandem with the burnup of fissile material. TWRs differ from other kinds of fast-neutron and breeder reactors in their ability to use fuel efficiently without uranium enrichment or reprocessing, instead directly using depleted uranium, natural uranium, thorium, spent fuel removed from light water reactors, or some combination of these materials. The concept is still in the development stage and no TWRs have ever been built.

Nuclear transmutation conversion of an atom from one element to another

Nuclear transmutation is the conversion of one chemical element or an isotope into another chemical element. Because any element is defined by its number of protons in its atoms, i.e. in the atomic nucleus, nuclear transmutation occurs in any process where the number of protons or neutrons in the nucleus is changed.

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Further reading