Thorium fuel cycle
The thorium fuel cycle is a nuclear fuel cycle that uses an isotope of thorium, 232
, as the fertile material. In the reactor, 232
is transmuted into the fissile artificial uranium isotope 233
which is the nuclear fuel. Unlike natural uranium, natural thorium contains only trace amounts of fissile material (such as 231
), 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
absorbs neutrons to produce 233
. This parallels the process in uranium breeder reactors whereby fertile 238
absorbs neutrons to form fissile 239
. Depending on the design of the reactor and fuel cycle, the generated 233
either fissions in situ or is chemically separated from the used nuclear fuel and formed into new nuclear fuel.
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, and better resistance to nuclear weapons proliferation when used in a traditional light water reactor though not in a molten salt reactor.
Concerns about the limits of worldwide uranium resources motivated initial interest in the thorium fuel cycle. 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. In the twenty-first century thorium's potential for improving proliferation resistance and waste characteristics led to renewed interest in the thorium fuel cycle.
At Oak Ridge National Laboratory in the 1960s, the Molten-Salt Reactor Experiment used 233
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.
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.
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.
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.
Nuclear reactions with thorium
In the thorium cycle, fuel is formed when 232
captures a neutron (whether in a fast reactor or thermal reactor) to become 233
. This normally emits an electron and an anti-neutrino (
decay to become 233
. This then emits another electron and anti-neutrino by a second
decay to become 233
, 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, 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.
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
, the transmutations tend to produce useful nuclear fuels rather than transuranic wastes. When 233
absorbs a neutron, it either fissions or becomes 234
. The chance of fissioning on absorption of a thermal neutron is about 92%; the capture-to-fission ratio of 233
, therefore, is about 1:12 – which is better than the corresponding capture vs. fission ratios of 235
(about 1:6), or 239
(both about 1:3). The result is less transuranic waste than in a reactor using the uranium-plutonium fuel cycle.
|(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 are fissile)
, like most actinides with an even number of neutrons, is not fissile, but neutron capture produces fissile 235
. If the fissile isotope fails to fission on neutron capture, it produces 236
, and eventually fissile 239
and heavier isotopes of plutonium. The 237
can be removed and stored as waste or retained and transmuted to plutonium, where more of it fissions, while the remainder becomes 242
, then americium and curium, which in turn can be removed as waste or returned to reactors for further transmutation and fission.
However, the 231
(with a half-life of 3.27×104 years) formed via (n,2n) reactions with 232
that decays to 231
), while not a transuranic waste, is a major contributor to the long-term radiotoxicity of spent nuclear fuel.
Unlike most even numbered heavy isotopes, 232
is also a fissile fuel fissioning just over half the time when it absorbs a thermal neutron. 232
has a relatively short half-life (68.9 years), and some decay products emit high energy gamma radiation, such as 224
and particularly 208
. The full decay chain, along with half-lives and relevant gamma energies, is:
Thorium-cycle fuels produce hard gamma emissions, which damage electronics, limiting their use in bombs. 232
cannot be chemically separated from 233
from used nuclear fuel; however, chemical separation of thorium from uranium removes the decay product 228
and the radiation from the rest of the decay chain, which gradually build up as 228
reaccumulates. The contamination could also be avoided by using a molten-salt breeder reactor and separating the 233
before it decays into 233
. The hard gamma emissions also create a radiological hazard which requires remote handling during reprocessing.
As a fertile material thorium is similar to 238
, the major part of natural and depleted uranium. The thermal neutron absorption cross section (σa) and resonance integral (average of neutron cross sections over intermediate neutron energies) for 232
are about three and one third times those of the respective values for 238
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. 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
is comparable to 235
, 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
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
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, 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.
Because the 233
produced in thorium fuels is significantly contaminated with 232
in proposed power reactor designs, thorium-based used nuclear fuel possesses inherent proliferation resistance. 232
cannot be chemically separated from 233
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
is sufficient to produce transuranic elements, whereas five captures are generally necessary to do so from 232
. 98–99% of thorium-cycle fuel nuclei would fission at either 233
, 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.
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
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
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, 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 the residual 233
is a long-lived radioactive isotope in the waste.
Another challenge associated with the thorium fuel cycle is the comparatively long interval over which 232
breeds to 233
. The half-life of 233
is about 27 days, which is an order of magnitude longer than the half-life of 239
. As a result, substantial 233
develops in thorium-based fuels. 233
is a significant neutron absorber and, although it eventually breeds into fissile 235
, 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
is recycled, remote handling is necessary for fuel fabrication because of the high radiation levels resulting from the decay products of 232
. This is also true of recycled thorium because of the presence of 228
, which is part of the 232
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
complicates matters, there are public documents showing that 233
has been used once in a nuclear weapon test. The United States tested a composite 233
-plutonium bomb core in the MET (Military Effects Test) blast during Operation Teapot in 1955, though with much lower yield than expected.
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.
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.
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. Additionally, Dresden 1 in the United States used "thorium oxide corner rods".
|Name||Country||Reactor type||Power||Fuel||Operation period|
|AVR||Germany (West)||HTGR, experimental (pebble bed reactor)||15 MW(e)||Th+235
Driver fuel, coated fuel particles, oxide & dicarbides
|THTR-300||Germany (West)||HTGR, power (pebble type)||300 MW(e)||Th+235
, Driver fuel, coated fuel particles, oxide & dicarbides
|Lingen||Germany (West)||BWR irradiation-testing||60 MW(e)||Test fuel (Th,Pu)O2 pellets||1968–1973|
|Dragon (OECD-Euratom)||UK (also Sweden, Norway and Switzerland)||HTGR, Experimental (pin-in-block design)||20 MWt||Th+235
Driver fuel, coated fuel particles, oxide & dicarbides
|Peach Bottom||United States||HTGR, Experimental (prismatic block)||40 MW(e)||Th+235
Driver fuel, coated fuel particles, oxide & dicarbides
|Fort St Vrain||United States||HTGR, Power (prismatic block)||330 MW(e)||Th+235
Driver fuel, coated fuel particles, Dicarbide
|MSRE ORNL||United States||MSR||7.5 MWt||233
|BORAX-IV & Elk River Station||United States||BWR (pin assemblies)||2.4 MW(e); 24 MW(e)||Th+235
Driver fuel oxide pellets
|Shippingport||United States||LWBR, PWR, (pin assemblies)||100 MW(e)||Th+233
Driver fuel, oxide pellets
|Indian Point 1||United States||LWBR, PWR, (pin assemblies)||285 MW(e)||Th+233
Driver fuel, oxide pellets
|SUSPOP/KSTR KEMA||Netherlands||Aqueous homogenous suspension (pin assemblies)||1 MWt||Th+HEU, oxide pellets||1974–1977|
|NRX & NRU||Canada||MTR (pin assemblies)||see)20 MW; 200 MW (||Th+235
, Test Fuel
|1947 (NRX) + 1957 (NRU); Irradiation–testing of few fuel elements|
|CIRUS; DHRUVA; & KAMINI||India||MTR thermal||40 MWt; 100 MWt; 30 kWt (low power, research)||Al+233
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 MWt||ThO2 blanket||1985; in operation|
|Petten||Netherlands||High Flux Reactor thorium molten salt experiment||45 MW(e)||?||2024; planned|
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