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The thorium fuel cycle is a nuclear fuel cycle that uses the naturally abundant isotope of thorium, 232Th, as fertile material, and the artificial uranium isotope, 233U, as fissile fuel for a nuclear reactor. However, unlike natural uranium, natural thorium contains only trace amounts of fissile material (such as 231Th) that are insufficient to initiate a nuclear chain reaction. Thus, some fissile material must be mixed with natural thorium in order to initiate the fuel cycle. In a thorium-fueled reactor, 232Th will absorb slow neutrons to produce 233U, which is similar to the process in uranium-fueled reactors whereby fertile 238U absorbs neutrons to form fissile 239Pu. Depending on the design of the reactor and fuel cycle, the 233U generated is either utilized in situ or chemically separated from the used nuclear fuel and used in new nuclear fuel.

A thorium fuel cycle offers several potential advantages over a uranium fuel cycle, including greater abundance, superior physical and nuclear properties of fuel, enhanced proliferation resistance, and reduced plutonium and actinide production.

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 is the Republic of Indiamarker which is developing a three stage thorium fuel cycle. Recently there has been renewed interest in thorium-based fuels for improving proliferation resistance and waste characteristics of used nuclear fuel

Thorium fuels have been used in several power and research reactors. One of the earliest efforts to use a thorium fuel cycle took place at Oak Ridge National Laboratorymarker in the 1960s. An experimental Molten Salt Reactor technology to study the feasibility of such an approach, using thorium fluoride salt kept hot enough to be liquid, thus eliminating the need for fabricating fuel elements. This effort culminated in the Molten-Salt Reactor Experimentmarker 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 cycle, fuel is formed when 232Th captures a neutron (whether in a fast reactor or thermal reactor) to become 233Th. This normally emits an electron and an anti-neutrino (\bar{\nu}_e) by β decay to become 233Pa. This then emits another electron and anti-neutrino by a second β decay to become 233U, the fuel:

\mathrm{n}+{}_{\ 90}^{232}\mathrm{Th}\rightarrow {}_{\ 90}^{233} \mathrm{Th} \xrightarrow{\beta^-} {}_{\ 91}^{233}\mathrm{Pa} \xrightarrow{\beta^-} {}_{\ 92}^{233}\mathrm{U}


In a reactor, when a neutron hits, it either splits the fuel, or is captured and transmutes. In the case of 233U, the transmutations tend to produce nuclear fuels, not transuranic wastes.When 233U absorbs a neutron, it either fissions or becomes 234U. The chance of fissioning on absorption of a thermal neutron is about 92%, therefore the capture-to-fission ratio of 233U is about 1/10 (better than the corresponding capture vs. fission ratios for 235U (about 1/6) or for 239Pu (about 1/2) or 241Pu (about 1/4)).

Uranium-234, like most actinides with an even number of neutrons, is not fissile, but further neutron capture produces fissile 235U; if this in turn fails to fission on neutron capture, it will produce 236U, 237Np, 238Pu, and eventually fissile 239Pu.

The result is that there is very little long-lived, hazardous transuranic wastes from a Thorium reactor. The tiny amount of transuranic wastes at equilibrium are mostly 237Np. These can be economically blended back into new fuel and transmuted, even though 237Np is a neutron poison.

Radioactive fission products are also produced, of course, but they have half-lives of less than 100 years or greater than 200,000 years. So, toxicity studies indicate

that in a few hundred years, the waste from a Thorium reactor can be less toxic than the Uranium ore that would have been used in a light water reactor of the same power.

Uranium-232 is also formed in this process, via (n,2n) reactions with 233U, 233Pa, and 232Th:

\mathrm{n}+{}_{\ 90}^{232}\mathrm{Th}\rightarrow {}_{\ 90}^{233} \mathrm{Th} \xrightarrow{\beta^-} {}_{\ 91}^{233}\mathrm{Pa} \xrightarrow{\beta^-} {}_{\ 92}^{233}\mathrm{U}+\mathrm{n}\rightarrow {}_{\ 92}^{232} \mathrm{U}+2\mathrm{n}


\mathrm{n}+{}_{\ 90}^{232}\mathrm{Th}\rightarrow {}_{\ 90}^{233} \mathrm{Th} \xrightarrow{\beta^-} {}_{\ 91}^{233}\mathrm{Pa}+\mathrm{n} \rightarrow {}_{\ 91}^{232}\mathrm{Pa}+2\mathrm{n} \xrightarrow{\beta^-} {}_{\ 92}^{232}\mathrm{U}


\mathrm{n}+{}_{\ 90}^{232}\mathrm{Th}\rightarrow {}_{\ 90}^{231} \mathrm{Th} + 2\mathrm{n} \xrightarrow{\beta^-} {}_{\ 91}^{231}\mathrm{Pa}+\mathrm{n} \rightarrow {}_{\ 91}^{232}\mathrm{Pa} \xrightarrow{\beta^-}{}_{\ 92}^{232}\mathrm{U}


Uranium-232 has a relatively short half-life (73.6 years), and some decay products emit high energy gamma radiation, such as 224Rn, 212Bi and particularly 208Tl. The full decay chain, along with half-lives and relevant gamma energies, is:

{}_{\ 92}^{232}\mathrm{U} \xrightarrow{\ \alpha\ } {}_{\ 90}^{228}\mathrm{Th}\ \mathrm{(73.6\ a)}


{}_{\ 90}^{228}\mathrm{Th} \xrightarrow{\ \alpha\ } {}_{\ 88}^{224}\mathrm{Ra}\ \mathrm{(1.9\ a)}


{}_{\ 88}^{224}\mathrm{Ra} \xrightarrow{\ \alpha\ } {}_{\ 86}^{220}\mathrm{Rn}\ \mathrm{(3.6\ d,\ 0.24\ MeV)}


{}_{\ 86}^{220}\mathrm{Rn} \xrightarrow{\ \alpha\ } {}_{\ 84}^{216}\mathrm{Po}\ \mathrm{(55\ s,\ 0.54\ MeV)}


{}_{\ 84}^{216}\mathrm{Po} \xrightarrow{\ \alpha\ } {}_{\ 82}^{212}\mathrm{Pb}\ \mathrm{(0.15\ s)}


{}_{\ 82}^{212}\mathrm{Pb} \xrightarrow{\beta^-\ } {}_{\ 83}^{212}\mathrm{Bi}\ \mathrm{(10.64\ h)}


{}_{\ 83}^{212}\mathrm{Bi} \xrightarrow{\ \alpha\ } {}_{\ 81}^{208}\mathrm{Tl}\ \mathrm{(61\ s,\ 0.78\ MeV)}


{}_{\ 81}^{208}\mathrm{Tl} \xrightarrow{\beta^-\ } {}_{\ 82}^{208}\mathrm{Pb}\ \mathrm{(3\ m,\ 2.6\ MeV)}


The hard gamma emissions damage electronics, and make the use of Thorium-cycle fuels difficult in military bomb triggers. Because 232U cannot be easily separated from 233U in used nuclear fuel, there is no easy, chemical way to separate it. The hard gamma emissions also create a radiological hazard which requires remote handling during reprocessing. Of course, a sufficiently well-funded, determined organization could overcome these obstacles, but Plutonium production is a less-risky development path for nuclear weapons.

Further, the 231Pa (with a half life of 3.27  years) formed via (n,2n) reactions with 232Th (yielding 231Th that decays to 231Pa) is a major contributor to the long term radiotoxicity of used nuclear fuel.

Advantages of thorium as a nuclear fuel

There are several potential advantages to thorium-based fuels.

Thorium is estimated to be about three to four times more abundant than uranium in the 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. Also, unlike uranium, naturally occurring thorium consists of only a single isotope (232Th) in significant quantities. Consequently, all mined thorium is useful in thermal reactors.

Thorium-based fuels exhibit several attractive nuclear properties relative to uranium-based fuels. The thermal neutron absorption cross sectiona) and resonance integral for 232Th are about three times and one third of the respective values for 238U; consequently, fertile conversion of the former is more efficient in a thermal reactor. Also, although the thermal neutron fission cross section (σf) of the 233U is comparable to 235U and 239Pu, it has a much lower capture cross section (σγ) than the latter two fissile isotopes, resulting in fewer non-fissile neutron absorptions and improved neutron economy. Finally, the number of neutrons released per neutron absorbed (η) in 233U is greater than two over a wide range of energies, including the thermal spectrum; as a result, thorium-based fuels can be the basis for a thermal breeder reactor.

Thorium-based fuels also display favorable physical and chemical properties which improve reactor and repository performance. Compared to the predominant reactor fuel, uranium dioxide (UO2), thorium dioxide (ThO2) 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 233U produced in thorium fuels is inevitably contaminated with 232U, thorium-based used nuclear fuel possesses inherent proliferation resistance. Uranium-232 can not be chemically separated from 233U and has several decay products which 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 238U is sufficient to produce transuranic elements, whereas six captures are generally necessary to do so from 232Th. 98–99% of thorium-cycle fuel nuclei would fission at either 233U or 235U, so fewer long-lived transuranics are produced. Because of this, thorium is a potentially attractive alternative to uranium in mixed oxide fuels to minimize the generation of transuranics and maximize the destruction of plutonium.

Disadvantages of thorium as nuclear fuel

There are several challenges to the application of thorium as a nuclear fuel.

Unlike uranium, natural thorium contains no fissile isotopes; fissile material, generally 233U, 235U, or plutonium, must be supplemented to achieve criticality. This, along with the high sintering temperature necessary to make thorium-dioxide fuel, complicates the fuel fabrication process. Oak Ridge National Laboratorymarker experimented with thorium-tetrafluoride as fuel in a molten salt reactor from 1964 - 1969, which was far easier to both process and separate from fuel poisons (contaminants that slow or stop the chain reaction.)

If thorium is used in an open fuel cycle (i.e. utilizing 233U in-situ), higher burnup is necessary to achieve a favorable neutron economy. Although thorium dioxide has performed well at burnups of 170,000 MWd/t and 150,000 MWd/t at Fort St. Vrain Generating Stationmarker and AVRmarker respectively, there are challenges associated with achieving this burnup in light water reactors , which compose the vast majority of existing power reactors.

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

Alternately, if thorium is used in a closed fuel cycle in which 233U is recycled, remote handling is necessary for fuel fabrication because of the high radiation dose resulting from the decay products of 232U. This is also true of recycled thorium because of the presence of 228Th, which is part of the 232U decay sequence. Further, although there is substantial worldwide experience recycling uranium fuels (e.g. PUREX), similar technology for thorium (e.g. THOREX) is still under development.

Although the presence of 232U makes it a challenge, 233U can be used in fission weapons, but this has been done only occasionally. The United States first tested 233U as part of a bomb core in Operation Teapot in 1955. However, unlike plutonium, 233U can be easily denatured by mixing it with natural or depleted uranium. Another option is to judiciously mix thorium fuels with small amounts of natural or depleted uranium during fabrication to ensure that 233U concentrations at the end of cycle are acceptably low.

Despite the fact that thorium-based fuels produce far less long-lived transuranics than uranium-based fuels,

there are some long-lived actinides produced that constitute a long term radiological impact, especially 231Pa.

Reactors

Thorium fuels have been demonstrated in several different reactor types, including light water reactors, heavy water reactors, high temperature gas reactor, sodium-cooled fast reactors, and molten salt reactors.

List of thorium-fueled reactors

Name and Country Type Power Fuel Operation period
AVRmarker, Germany HTGR, Experimental (Pebble bed reactor) 15 MW(e) Th+235U Driver Fuel, Coated fuel particles, Oxide & dicarbides 1967 – 1988
THTR-300marker, Germany HTGR, Power (Pebble Type) 300 MW(e) Th+235U, Driver Fuel, Coated fuel particles, Oxide & dicarbides 1985 – 1989
Lingen, Germany BWR Irradiation-testing 60 MW(e) Test Fuel (Th,Pu)O2 pellets Terminated in 1973
Dragonmarker, UK OECD-Euratom also Sweden, Norway & Switzerland HTGR, Experimental (Pin-in-Block Design) 20 MWt Th+235U Driver Fuel, Coated fuel particles, Oxide & Dicarbides 1966 - 1973
Peach Bottommarker, USA HTGR, Experimental (Prismatic Block) 40 MW(e) Th+235U Driver Fuel, Coated fuel particles, Oxide & dicarbides 1966 – 1972
Fort St Vrainmarker, USA HTGR, Power (Prismatic Block) 330 MW(e) Th+235U Driver Fuel, Coated fuel particles, Dicarbide 1976 - 1989
MSREmarker ORNLmarker, USA MSBR 7.5 MWt 233U Molten Fluorides 1964 - 1969
Shippingportmarker & Indian Point 1, USA LWBR PWR, (Pin Assemblies) 100 MW(e), 285 MW(e) Th+233U Driver Fuel, Oxide Pellets 1977 – 1982, 1962 – 1980
SUSPOP/KSTR KEMA, Netherlands Aqueous Homogenous Suspension (Pin Assemblies) 1 MWt Th+HEU, Oxide Pellets 1974 - 1977
NRU & NRXmarker, Canada MTR (Pin Assemblies) Th+235U, Test Fuel Irradiation–testing of few fuel elements
KAMINI; CIRUSmarker; & DHRUVA, India MTR Thermal 30 kWt; 40 MWt; 100 MWt Al+233U Driver Fuel, ‘J’ rod of Th & ThO2, ‘J’ rod of ThO2 All three research reactors in operation
KAPS 1 &2marker; KGS 1 & 2; RAPS 2, 3 & 4marker, India PHWR, (Pin Assemblies) 220 MW(e) ThO2 Pellets (For neutron flux flattening of initial core after start-up) Continuing in all new PHWRs
FBTR, India LMFBR, (Pin Assemblies) 40 MWt ThO2 blanket In operation
(IAEA TECDOC-1450 "Thorium Fuel Cycle - Potential Benefits and Challenges", Table 1. Thorium utilization in different experimental and power reactors.)

References and links

References

  1. [1] Le Brun, C., “Impact of the MSBR concept technology on long-lived radio-toxicity and proliferation resistance”, Technical Meeting on Fissile Material Management Strategies for Sustainable Nuclear Energy, Vienna 2005


See also



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