User:Ionium Dope/Thorium fuel cycle

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 resource 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 India 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 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 ²³²Th captures a neutron (whether in a fast reactor or thermal reactor) to become ²³³Th. It normally emits an electron and an anti-neutrino (${\bar {\nu }}_{e}$) by β⁻ decay to become ²³³Pa. It then emits another electron and anti-neutrino by a second ß⁻ decay to become ²³³U:

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

When ²³³U absorbs a neutron, it either fissions or becomes ²³⁴U. The chance of fissioning on absorption of a thermal neutron is about 92%, and the capture-to-fission ratio of ²³³U is about 1/10, which is greater than the corresponding capture/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.

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

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

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

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

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:

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

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

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

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

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

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

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

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

Because ²³²U cannot be easily separated from ²³³U in used nuclear fuel, these hard gamma emitters create a radiological hazard which necessitates remote handling during reprocessing.

Further, the ²³¹Pa (with a half life of 3.27×10⁴ 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 section (σ_a) 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 (σ_f) of the ²³³U is comparable to ²³⁵U and ²³⁹Pu, it has a much lower capture cross section (σ_γ) 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 radiological hazard of conventional uranium-based used nuclear fuel is dominated by plutonium and other minor actinides generated through neutron capture. A single neutron capture in ²³⁸U is sufficient to produce transuranic elements, whereas six captures are generally necessary to do so from ²³²Th. In fact, 98–99% of thorium-cycle fuel nuclei would fission before reaching even ²³⁶U. As a result, fewer long-lived transuranics are produced in thorium fuel. 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.

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.

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 Station and AVR respectively, there are challenges associated with achieving this burnup in light water reactors (LWR), 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 substantial 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, although 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 reactors, sodium-cooled fast reactors, and molten salt reactors.

List of thorium-fueled reactors

Name and Country	Type	Power	Fuel	Operation period
AVR, Germany	HTGR, Experimental (Pebble bed reactor)	15 MW(e)	Th+235U Driver Fuel, Coated fuel particles, Oxide & dicarbides	1967 – 1988
THTR-300, 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
Dragon, 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 Bottom, USA	HTGR, Experimental (Prismatic Block)	40 MW(e)	Th+235U Driver Fuel, Coated fuel particles, Oxide & dicarbides	1966 – 1972
Fort St Vrain, USA	HTGR, Power (Prismatic Block)	330 MW(e)	Th+235U Driver Fuel, Coated fuel particles, Dicarbide	1976 - 1989
MSRE ORNL, USA	MSBR	7.5 MWt	233U Molten Fluorides	1964 - 1969
Shippingport & 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 & NRX, Canada	MTR (Pin Assemblies)		Th+235U, Test Fuel	Irradiation–testing of few fuel elements
KAMINI; CIRUS; & 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 &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)	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

See also

• Thorium
• Nuclear fuel cycle
• Nuclear power
• Nuclear fission
• Radioactive waste
• World energy resources and consumption
• Uranium depletion
• Peak uranium
• Fuji MSR
• Energy amplifier
• Alvin Radkowsky

External links

• FactSheet on Thorium, World Nuclear Association.
• Thorium fuel cycle — Potential benefits and challenges, International Atomic Energy Agency, May 2005.
• Thorium Fuel Links
• 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.

Category:Nuclear chemistry Category:Nuclear power Category:Nuclear reprocessing Category:Nuclear technology Category:Actinides Category:Environmental issues with energy