Thorium fuel cycle

Thorium fuel cycle

The thorium fuel cycle is a nuclear fuel cycle that uses thorium-232 as fertile material and uranium-233 as fissile fuel. A major advantage of the thorium fuel cycle is that production of plutonium and other long-lived actinides as radioactive waste is far less than in the uranium fuel cycle.

Thorium, as well as uranium and plutonium, can be used as fuel in a nuclear reactor. Although not fissile itself, 232Th will absorb slow neutrons to produce ( 233U), which is fissile. Hence, like 238U, 232Th is fertile.

One of the earliest efforts to use a thorium fuel cycle took place at Oak Ridge National Laboratory in the 1960s. An experimental Molten Salt Reactor technology to study the feasibility of such an approach, using thorium(IV) fluoride salt kept hot enough to be liquid, thus eliminating the need for fabricating fuel elements. This effort culminated in the Molten-Salt Reactor Experiment that used 232Th as the fertile material and 233U as the fissile fuel. Due to a lack of funding, the MSR program was discontinued in 1976.

Nuclear reactions with thorium

In the thorium fuel cycle thorium-232 absorbs a neutron (whether in a fast reactor or thermal reactor) to become 233Th. The thorium-233 normally emits an electron and an anti-neutrino (ar{ u}_e) by β decay to become protactinium-233 (233Pa):

:mathrmhbox{n}+}^2{}^{32}_{90}mathrm{Th ightarrowmathrm}^2{}^{33}_{90}mathrm{Th ightarrowmathrm}^2{}^{33}_{91}Pa}+ e^- + ar{ u}_e
Protactinium-233 then emits another electron and anti-neutrino by a second β decay to become 233U:

:mathrm}^2{}^{33}_{91}Pa} ightarrowmathrm}^2{}^{33}_{92}U}+ e^- + ar{ u}_e
Uranium-233 in turn is used as fuel. Hence, like uranium-238, thorium-232 is a fertile material.

The irradiated fuel can then be unloaded from the reactor, the 233U separated from the thorium (a relatively simple process since it involves chemical instead of isotopic separation), and fed back into another reactor as part of a closed nuclear fuel cycle.

Uranium-232 is marginally formed in this process, via ("n",2"n") reactions with 233U (yielding 232U) and with 233Pa (yielding 232Pa which decays to 232U). It is highly radioactive (the half-life of 232U is only 73.6 years), the daughter products have very short half-lives, and some emit strong gamma radiation, like 224Rn (0.54 MeV), 212Bi (0.78 MeV) and particularly 208Tl (2.6 MeV).This Uranium-232 decays to Thorium-228; and neither can be chemically separated from 233U or 232Th. As a result, both 233U and 228Th which can be recovered from reprocessing develop high-energy gamma radiation fields on standing.This by-product is a concern if reprocessing is needed. On the other hand, it has no impact on molten salt reactors (where the uranium is not separated), and may be seen as an advantage as giving an intrinsic proliferation-resistance to the cycle.

Protactinium-231 is also marginally formed in the process, via ("n",2"n") reactions with 232Th (yielding 231Th that decays to 231Pa). This unavoidable by-product (with a half life of 3.27e|4 years) is the main contributor to high-activity long-term wastes.

In a Molten Salt Reactor model, the protactinium (mainly 233Pa) is extracted from the flow (with a periodicity of a couple of days), and left for some months (decay half-life is 27 days) to decay to 233U away from the neutron flux. This is to avoid as much as possible a neutron capture that would waste two neutrons before leading to the fissile uranium-235::233Pa ("n",γ) 234Pa (β-) 234U ("n",γ) 235USuch a production of 235U has two drawbacks: it upsets the neutron balance, since two neutrons are spent before reaching a fissile nucleus; and it opens the path to minor actinides (and plutonium) through further neutron captures. [See [ Molten Salt Reactors Based on the Th-U3 Fuel Cycle] , Physique des Réacteurs du LPSC Grenoble, November 2001.]

Economical issues

In one significant respect 233U is better than the other two fissile isotopes used for nuclear fuel, 235U and plutonium-239 (239Pu): at low neutron energies it yields more neutrons per neutron absorbed (at high neutron energies, such as those found in a fast reactor the yield of neutrons from plutonium-239 increases considerably, overtaking thorium). Given a start with some other fissile material (235U or 239Pu), a breeding cycle similar to, but more efficient than that currently possible with the 238U-to-239Pu cycle (in slow-neutron reactors), can be set up.

Problems include the high cost of fuel fabrication due partly to the high radioactivity of the traces of the short-lived 232U that contaminates the 233U fuel; the similar problems in recycling thorium due to highly radioactive 228Th; some weapons proliferation risk of 233U; and the technical problems (not yet satisfactorily solved) in reprocessing. Much development work is still required before the thorium fuel cycle can be commercialised, and the effort required seems unlikely while (or where) abundant uranium is available.

Nevertheless, the thorium fuel cycle, with its potential for breeding fuel without fast neutron reactors, holds considerable potential long-term benefits. Thorium is significantly more abundant than uranium, and is a key factor in sustainable nuclear energy.

Nuclear waste issues

When 233U absorbs a neutron, it either fissions or becomes the next heavier isotope, 234U. The chance of not fissioning on absorption of a thermal neutron is about 1/7 (or even less than 10% according to another source), which is less than the corresponding capture/fission ratios for 235U (about 1/6) or for 239Pu or 241Pu (about 1/4). Uranium-234, like most actinide nuclides with an even number of neutrons, is not easily fissionable with slow neutrons, but further neutron capture produces fissile 235U; if this in turn fails to fission on neutron capture, it will produce uranium-236, neptunium-237, Pu-238, and eventually fissile Plutonium-239.

Thus production of heavy transuranic nuclides (the minor actinides other than neptunium) is far less than in the 238U/239Pu cycle, because 98–99% of thorium-cycle fuel nuclei would fission before reaching even 236U. On the other hand, the thorium cycle produces some protactinium-231 (half-life 33,000 years) via the ("n",2"n") reaction on 232Th. [ cite web| url=| title= Status of Nuclear Data for the Thorium Fuel Cycle |page= 4 |quote= Relative number of nuclei n = N(231Pa)/N(233U): Fast reactor 0.8e|-2 Thermal reactor 1.9e|-3... The problem is exacerbated by the fact that the secondary heavy nuclei produced in this cycle possess, as a rule, extremely unpleasant nuclear physics characteristics from the experimentalist’s point of view |year= 1997 |coauthors= Kuz’minov B.D.; Manokhin V.N. |accessdate= 2008-01-15 ] [ cite web |url= |title= Method of increasing the deterrent to proliferation of nuclear fuels - Patent 4344912 |quote= protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor |accessdate= 2008-01-15 ] [ cite web |url= |title= AN OVERVIEW OF R&D IN FUEL CYCLE ACTIVITIES OF AHWR |quote= the higher energy (n, 2n) reactions encountered by Th-232 during the irradiation in Th-U233 fuel also lead to the formation of long lived Pa-231... Pa-231 is of special concern in HLLW of AHWR because the pentavalent Pa-231 is capable to migrate much more in water/soil compared to other ions... the amount of Pa-231 produced in (Th-U233)O2 MOX fuel is ~ 3 gms/te at 20,000 MWd/t of burn-up... removal of protactinium has to be established using suitable solvents that is yet to be tested |author= Bhattacharjee B. |publisher= Bhabha Atomic Research Centre |accessdate= 2008-01-15 ] [ cite web |url= |title= Nuclear Energy With (Almost) No Radioactive Waste? |quote=according to computer simulations done at ISN, this Protactinium dominates the residual toxicity of losses at 10 000 years |month= July |year= 2001 |coauthors= Brissot R.; Heuer D.; Huffer E.; Brun, C. Le; Loiseaux, J-M; Nifenecker H.; Nuttin A.] Because the thorium/uranium-233 cycle produces a smaller amount of long-lived actinide isotopes, the long-term radioactivity of the spent nuclear fuel is less. Common fission products have half-lives up to 30 years (strontium-90, caesium-137) or more than 200,000 years (technetium-99), and radioactivity in the period intermediate between these two scales is chiefly from actinide wastes. Another positive, if a solid-fuel reactor is used, is that thorium dioxide melts around 3,300 °C compared to 2,800 °C for uranium dioxide cycle. [ cite web| url= |title= Perspectives of the Thorium Fuel Cycle |work= NUCLEAR TECHNOLOGIES AND NON-PROLIFERATION POLICIES |publisher= Analytical Center for Non-Proliferation |year= 2004 |accessdate= 2008-01-15]

With fuel reprocessing, the Thorium fuel cycle, so impractical in other types of reactors, produces 0.1% Fact|date=July 2008 of the long-term high-level radioactive waste of a light-water reactor without reprocessing (all modern reactors in the U.S.).

As thorium-232 captures neutrons, it first becomes 233Th, which quickly decays to 233Pa. Protactinium-233 in turn decays to 233U with a half-life of 27 days. Uranium-233 is an excellent reactor fuel.

As 233U is bombarded by neutrons with a thermal spectrum of speeds, each absorbed neutron either splits the uranium or (with a probability of about 1/7 or less) produces uranium-234, which will absorb another neutron to become fissile uranium-235, which will fission to elements similar to those from 233U or (with a probability of about 1/6) become uranium-236. These fission products almost all have half-lives less than 30 years.

The only source of high-radioactivity, long-lived, transuranic elements (TRUs) is neptunium-237 produced from the tiny fraction of 236U (about 2–3% of the original 233U) produced at the tail-end of this process. The neptunium can be separated by the fuel-salt reprocessing and disposed as waste.

The neptunium can also be left in the salt where it continues to absorb neutrons becoming successive isotopes of plutonium with mass 238 to 242 and a majority chance of fissioning at each isotope with an odd number of neutrons, or eventually with even-lower probability, isotopes of the next heavier minor actinides, americium and curium. If all actinides (including protactinium-231) are left in the reactor and eventually fissioned, the remaining wastes are only fission products and activation products.

Fission products have half-lives less than about 30 years (137Cs, 90Sr), 90 years (151Sm) or over 200,000 years (Long-lived fission products). Reprocessed waste from thorium is therefore less radioactive than the original thorium ores after about 300 years when 137Cs and 90Sr have declined to about 1/1000 of original levels, though mobility of some long-lived fission products such as technetium-99 in the environment may be greater.

The use of thorium with liquid fluoride salts is now known as the Liquid Fluoride Thorium Reactor or LFTR. This term is becoming more commonFact|date=July 2008 in describing the MSR run with thorium suspended in fluoride salts.

With continuous reprocessing, a molten-salt-fueled reactor has more than 97% burn-up of fuel. This is very efficient, compared to any system, anywhere. Light water reactors burn up about 2% of fuel on a once-through fuel cycle (current practice, 2007).

Nuclear reactors

After starting the reactor with existing 233U or some other fissile material such as 235U or 239Pu, a breeding cycle similar to but more efficient than that with 238U and plutonium can be created. The 232Th absorbs a neutron to become 233Th which quickly decays to 233Pa. Protactinium-233 in turn decays with a half-life of 27 days to 233U. In some molten salt reactor and Liquid fluoride reactor designs, the 233Pa is extracted and protected from neutrons (which could transform it to 234Pa and then to 234U), until it has decayed to 233U. This is done in order to improve the breeding ratio.

Uranium-233 is an excellent reactor fuel for reactors that utilise thermal neutrons. At low neutron energies 233U is superior to 235U and 239Pu, because it produces more neutrons per neutron absorbed (it has a high "beta" coefficient). Its absorption of neutrons (cross-section) also varies less with temperature and neutron energy than 239Pu or 235U. This stability suggests potential for high burnup, higher operating temperatures, and therefore more efficient conversion of heat to electricity. [ cite web |url= |format=PDF |title= Thorium – Fuelling a Sustainable Future for Nuclear Power] This also enables the possibility of constructing a thermal breeder reactor.

List of thorium-cycle reactors

References and links


See also

* Thorium
* Nuclear fuel cycle
* Radioactive waste

External Links

* [ FactSheet on Thorium] , World Nuclear Association.
* [ Thorium fuel cycle — Potential benefits and challenges] , IAEA, may 2005.
* [ Thorium Fuel Links]
* [ Perspectives of the Thorium Fuel Cycle]
* [ The Use of Thorium as Nuclear Fuel] American Nuclear Society, Position Statement, November 2006
* [ Revisiting the thorium-uranium nuclear fuel cycle] , © European Physical Society, EDP Sciences 2007.

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