Nuclear fuel cycle
The nuclear fuel cycle, also called nuclear fuel chain, is the progression of
Basic concepts
A
Some reactors do not use moderators to slow the neutrons. Like nuclear weapons, which also use unmoderated or "fast" neutrons, these fast-neutron reactors require much higher concentrations of fissile isotopes in order to sustain a chain reaction. They are also capable of breeding fissile isotopes from fertile materials; a breeder reactor is one that generates more fissile material in this way than it consumes.
During the nuclear reaction inside a reactor, the fissile isotopes in nuclear fuel are consumed, producing more and more fission products, most of which are considered radioactive waste. The buildup of fission products and consumption of fissile isotopes eventually stop the nuclear reaction, causing the fuel to become a spent nuclear fuel. When 3% enriched LEU fuel is used, the spent fuel typically consists of roughly 1% U-235, 95% U-238, 1% plutonium and 3% fission products. Spent fuel and other high-level radioactive waste is extremely hazardous, although nuclear reactors produce orders of magnitude smaller volumes of waste compared to other power plants because of the high energy density of nuclear fuel. Safe management of these byproducts of nuclear power, including their storage and disposal, is a difficult problem for any country using nuclear power[citation needed].
Front end
-
1 Uranium ore – the principal raw material of nuclear fuel
-
2 Yellowcake – the form in which uranium is transported to a conversion plant
-
3 UF6 – used in enrichment
-
4 Nuclear fuel – a compact, inert, insoluble solid
Exploration
A deposit of uranium, such as uraninite, discovered by geophysical techniques, is evaluated and sampled to determine the amounts of uranium materials that are extractable at specified costs from the deposit. Uranium reserves are the amounts of ore that are estimated to be recoverable at stated costs.
Naturally occurring uranium consists primarily of two isotopes U-238 and U-235, with 99.28% of the metal being U-238 while 0.71% is U-235, and the remaining 0.01% is mostly U-234. The number in such names refers to the isotope's atomic mass number, which is the number of protons plus the number of neutrons in the atomic nucleus.
The atomic nucleus of U-235 will nearly always fission when struck by a
Mining
Uranium ore can be extracted through conventional mining in open pit and underground methods similar to those used for mining other metals.
Milling
Mined uranium ores normally are processed by grinding the ore materials to a uniform particle size and then treating the ore to extract the uranium by chemical leaching. The milling process commonly yields dry powder-form material consisting of natural uranium, "yellowcake", which is sold on the uranium market as U3O8. Note that the material isn't always yellow.
Uranium conversion
Usually milled uranium oxide, U3O8 (triuranium octoxide) is then processed into either of two substances depending on the intended use.
For use in most reactors, U3O8 is usually converted to uranium hexafluoride (UF6), the input stock for most commercial uranium enrichment facilities. A solid at room temperature, uranium hexafluoride becomes gaseous at 57 °C (134 °F). At this stage of the cycle, the uranium hexafluoride conversion product still has the natural isotopic mix (99.28% of U-238 plus 0.71% of U-235).
For use in reactors such as
In the current nuclear industry, the volume of material converted directly to UO2 is typically quite small compared to that converted to UF6.
Enrichment
The natural concentration (0.71%) of the fissile isotope U-235 is less than that required to sustain a nuclear chain reaction in
The bulk (96%) of the byproduct from enrichment is
Fabrication
For use as nuclear fuel, enriched uranium hexafluoride is converted into uranium dioxide (UO2) powder that is then processed into pellet form. The pellets are then fired in a high temperature sintering furnace to create hard, ceramic pellets of enriched uranium. The cylindrical pellets then undergo a grinding process to achieve a uniform pellet size. The pellets are stacked, according to each nuclear reactor core's design specifications, into tubes of corrosion-resistant metal alloy. The tubes are sealed to contain the fuel pellets: these tubes are called fuel rods. The finished fuel rods are grouped in special fuel assemblies that are then used to build up the nuclear fuel core of a power reactor.
The alloy used for the tubes depends on the design of the reactor.
Service period
Transport of radioactive materials
Transport is an integral part of the nuclear fuel cycle. There are nuclear power reactors in operation in several countries but uranium mining is viable in only a few areas. Also, in the course of over forty years of operation by the nuclear industry, a number of specialized facilities have been developed in various locations around the world to provide fuel cycle services and there is a need to transport nuclear materials to and from these facilities.[5] Most transports of nuclear fuel material occur between different stages of the cycle, but occasionally a material may be transported between similar facilities. With some exceptions, nuclear fuel cycle materials are transported in solid form, the exception being uranium hexafluoride (UF6) which is considered a gas. Most of the material used in nuclear fuel is transported several times during the cycle. Transports are frequently international, and are often over large distances. Nuclear materials are generally transported by specialized transport companies.
Since nuclear materials are
In-core fuel management
A nuclear reactor core is composed of a few hundred "assemblies", arranged in a regular array of cells, each cell being formed by a fuel or control rod surrounded, in most designs, by a moderator and coolant, which is water in most reactors.
Because of the fission process that consumes the fuels, the old fuel rods must be replaced periodically with fresh ones (this is called a (replacement) cycle). During a given replacement cycle only some of the assemblies (typically one-third) are replaced since fuel depletion occurs at different rates at different places within the reactor core. Furthermore, for efficiency reasons, it is not a good policy to put the new assemblies exactly at the location of the removed ones. Even bundles of the same age will have different burn-up levels due to their previous positions in the core. Thus the available bundles must be arranged in such a way that the yield is maximized, while safety limitations and operational constraints are satisfied. Consequently, reactor operators are faced with the so-called optimal fuel reloading problem, which consists of optimizing the rearrangement of all the assemblies, the old and fresh ones, while still maximizing the reactivity of the reactor core so as to maximise fuel burn-up and minimise fuel-cycle costs.
This is a
The study of used fuel
Used nuclear fuel is studied in
Uranium dioxide is very insoluble in water, but after oxidation it can be converted to uranium trioxide or another uranium(VI) compound which is much more soluble. Uranium dioxide (UO2) can be oxidised to an oxygen rich hyperstoichiometric oxide (UO2+x) which can be further oxidised to U4O9, U3O7, U3O8 and UO3.2H2O.
Because used fuel contains alpha emitters (plutonium and the
The concentration of carbonate in the water which is in contact with the used fuel has a considerable effect on the rate of corrosion, because uranium(VI) forms soluble anionic carbonate complexes such as [UO2(CO3)2]2− and [UO2(CO3)3]4−. When carbonate ions are absent, and the water is not strongly acidic, the hexavalent uranium compounds which form on oxidation of uranium dioxide often form insoluble hydrated uranium trioxide phases.[8]
Thin films of uranium dioxide can be deposited upon gold surfaces by ‘sputtering’ using uranium metal and an argon/oxygen gas mixture. These gold surfaces modified with uranium dioxide have been used for both cyclic voltammetry and AC impedance experiments, and these offer an insight into the likely leaching behaviour of uranium dioxide.[9]
Fuel cladding interactions
The study of the nuclear fuel cycle includes the study of the behaviour of nuclear materials both under normal conditions and under accident conditions. For example, there has been much work on how
Normal and abnormal conditions
The nuclear chemistry associated with the nuclear fuel cycle can be divided into two main areas; one area is concerned with operation under the intended conditions while the other area is concerned with maloperation conditions where some alteration from the normal operating conditions has occurred or (more rarely) an accident is occurring.
The releases of radioactivity from normal operations are the small planned releases from uranium ore processing, enrichment, power reactors, reprocessing plants and waste stores. These can be in different chemical/physical form from releases which could occur under accident conditions. In addition the isotope signature of a hypothetical accident may be very different from that of a planned normal operational discharge of radioactivity to the environment.
Just because a radioisotope is released it does not mean it will enter a human and then cause harm. For instance, the migration of radioactivity can be altered by the binding of the radioisotope to the surfaces of soil particles. For example,
According to Jiří Hála's
In dairy farming, one of the best countermeasures against 137Cs is to mix up the soil by deeply ploughing the soil. This has the effect of putting the 137Cs out of reach of the shallow roots of the grass, hence the level of radioactivity in the grass will be lowered. Also after a nuclear war or serious accident, the removal of top few cm of soil and its burial in a shallow trench will reduce the long-term gamma dose to humans due to 137Cs, as the gamma photons will be attenuated by their passage through the soil.
Even after the radioactive element arrives at the roots of the plant, the metal may be rejected by the biochemistry of the plant. The details of the uptake of 90Sr and 137Cs into
In
Release of radioactivity from fuel during normal use and accidents
The IAEA assume that under normal operation the coolant of a water-cooled reactor will contain some radioactivity[12] but during a reactor accident the coolant radioactivity level may rise. The IAEA states that under a series of different conditions different amounts of the core inventory can be released from the fuel, the four conditions the IAEA consider are normal operation, a spike in coolant activity due to a sudden shutdown/loss of pressure (core remains covered with water), a cladding failure resulting in the release of the activity in the fuel/cladding gap (this could be due to the fuel being uncovered by the loss of water for 15–30 minutes where the cladding reached a temperature of 650–1250 °C) or a melting of the core (the fuel will have to be uncovered for at least 30 minutes, and the cladding would reach a temperature in excess of 1650 °C).[13]
Based upon the assumption that a Pressurized water reactor contains 300 tons of water, and that the activity of the fuel of a 1 GWe reactor is as the IAEA predicts,[14] then the coolant activity after an accident such as the Three Mile Island accident (where a core is uncovered and then recovered with water) can be predicted.[citation needed]
Releases from reprocessing under normal conditions
It is normal to allow used fuel to stand after the irradiation to allow the short-lived and radiotoxic iodine isotopes to decay away. In one experiment in the US, fresh fuel which had not been allowed to decay was reprocessed (the Green run [2] [3]) to investigate the effects of a large iodine release from the reprocessing of short cooled fuel. It is normal in reprocessing plants to scrub the off gases from the dissolver to prevent the emission of iodine. In addition to the emission of iodine the noble gases and tritium are released from the fuel when it is dissolved. It has been proposed that by voloxidation (heating the fuel in a furnace under oxidizing conditions) the majority of the tritium can be recovered from the fuel.[4]
A paper was written on the radioactivity in oysters found in the Irish Sea.[15] These were found by gamma spectroscopy to contain 141Ce, 144Ce, 103Ru, 106Ru, 137Cs, 95Zr and 95Nb. Additionally, a zinc activation product (65Zn) was found, which is thought to be due to the corrosion of magnox fuel cladding in spent fuel pools. It is likely that the modern releases of all these isotopes from the Windscale event is smaller.
On-load reactors
Some reactor designs, such as RBMKs or CANDU reactors, can be refueled without being shut down. This is achieved through the use of many small pressure tubes to contain the fuel and coolant, as opposed to one large pressure vessel as in pressurized water reactor (PWR) or boiling water reactor (BWR) designs. Each tube can be individually isolated and refueled by an operator-controlled fueling machine, typically at a rate of up to 8 channels per day out of roughly 400 in CANDU reactors. On-load refueling allows for the optimal fuel reloading problem to be dealt with continuously, leading to more efficient use of fuel. This increase in efficiency is partially offset by the added complexity of having hundreds of pressure tubes and the fueling machines to service them.
Interim storage
After its operating cycle, the reactor is shut down for refueling. The fuel discharged at that time (spent fuel) is stored either at the reactor site (commonly in a spent fuel pool) or potentially in a common facility away from reactor sites. If on-site pool storage capacity is exceeded, it may be desirable to store the now cooled aged fuel in modular dry storage facilities known as Independent Spent Fuel Storage Installations (ISFSI) at the reactor site or at a facility away from the site. The spent fuel rods are usually stored in water or boric acid, which provides both cooling (the spent fuel continues to generate decay heat as a result of residual radioactive decay) and shielding to protect the environment from residual ionizing radiation, although after at least a year of cooling they may be moved to dry cask storage.
Transportation
Reprocessing
Spent fuel discharged from reactors contains appreciable quantities of fissile (U-235 and Pu-239), fertile (U-238), and other
Mixed oxide, or MOX fuel, is a blend of reprocessed uranium and plutonium and depleted uranium which behaves similarly, although not identically, to the enriched uranium feed for which most nuclear reactors were designed. MOX fuel is an alternative to low-enriched uranium (LEU) fuel used in the light water reactors which predominate nuclear power generation.
Currently, plants in Europe are reprocessing spent fuel from utilities in Europe and Japan. Reprocessing of spent commercial-reactor nuclear fuel is currently not permitted in the
Partitioning and transmutation
As an alternative to the disposal of the
Waste disposal
Actinides[18] by decay chain | Half-life range (a) |
|||||||
---|---|---|---|---|---|---|---|---|
4n
|
4n + 1
|
4n + 2
|
4n + 3
|
4.5–7% | 0.04–1.25% | <0.001% | ||
228 Ra№
|
4–6 a
|
155 Euþ
|
||||||
244 Cmƒ
|
241Puƒ | 250 Cf
|
227 Ac№
|
10–29 a
|
90Sr | 85Kr | 113m Cdþ
| |
232Uƒ | 238Puƒ | 243 Cmƒ
|
29–97 a
|
137 Cs
|
151 Smþ
|
121m Sn
| ||
248Bk[20]
|
249 Cfƒ
|
242m Amƒ
|
141–351 a |
No fission products have a half-life | ||||
241Amƒ | 251Cfƒ[21]
|
430–900 a | ||||||
226Ra№ | 247 Bk
|
1.3–1.6 ka | ||||||
240Pu | 229 Th
|
246 Cmƒ
|
243 Amƒ
|
4.7–7.4 ka | ||||
245 Cmƒ
|
250 Cm
|
8.3–8.5 ka | ||||||
239Puƒ | 24.1 ka | |||||||
230 Th№
|
231 Pa№
|
32–76 ka | ||||||
236 Npƒ
|
233Uƒ | 234U№ | 150–250 ka | 99Tc₡ | 126 Sn
| |||
248 Cm
|
242Pu | 327–375 ka | 79Se₡ | |||||
1.53 Ma | 93 Zr
| |||||||
237 Npƒ
|
2.1–6.5 Ma | 135 Cs₡
|
107 Pd
| |||||
236U | 247 Cmƒ
|
15–24 Ma | 129I₡ | |||||
244Pu | 80 Ma |
... nor beyond 15.7 Ma[22] | ||||||
232Th№ | 238U№ | 235Uƒ№ | 0.7–14.1 Ga | |||||
|
A current concern in the nuclear power field is the safe disposal and isolation of either spent fuel from reactors or, if the reprocessing option is used, wastes from reprocessing plants. These materials must be isolated from the biosphere until the radioactivity contained in them has diminished to a safe level.[23] In the U.S., under the Nuclear Waste Policy Act of 1982 as amended, the Department of Energy has responsibility for the development of the waste disposal system for spent nuclear fuel and high-level radioactive waste. Current plans call for the ultimate disposal of the wastes in solid form in a licensed deep, stable geologic structure called a deep geological repository. The Department of Energy chose Yucca Mountain as the location for the repository. Its opening has been repeatedly delayed. Since 1999 thousands of nuclear waste shipments have been stored at the Waste Isolation Pilot Plant in New Mexico.
Horizontal drillhole disposal describes proposals to drill over one kilometer vertically, and two kilometers horizontally in the Earth's crust, for the purpose of disposing of high-level waste forms such as spent nuclear fuel, Caesium-137, or Strontium-90. After the emplacement and the retrievability period,[clarification needed] drillholes would be backfilled and sealed. A series of tests of the technology were carried out in November 2018 and then again publicly in January 2019 by a U.S. based private company.[24] The test demonstrated the emplacement of a test-canister in a horizontal drillhole and retrieval of the same canister. There was no actual high-level waste used in this test.[25][26]
Fuel cycles
Although the most common terminology is fuel cycle, some argue that the term fuel chain is more accurate, because the spent fuel is never fully recycled. Spent fuel includes
Once-through nuclear fuel cycle
Not a cycle per se, fuel is used once and then sent to storage without further processing save additional packaging to provide for better isolation from the
Plutonium cycle
Several countries, including Japan, Switzerland, and previously Spain and Germany,[.
The use of a medium-scale reprocessing facility onsite, and the use of pyroprocessing rather than the present day aqueous reprocessing, is claimed to potentially be able to considerably reduce the nuclear proliferation potential or possible diversion of fissile material as the processing facility is in-situ. Similarly as plutonium is not separated on its own in the pyroprocessing cycle, rather all actinides are "electro-won" or "refined" from the spent fuel, the plutonium is never separated on its own, instead it comes over into the new fuel mixed with gamma and alpha emitting actinides, species that "self-protect" it in numerous possible thief scenarios.
Beginning in 2016 Russia has been testing and is now deploying Remix Fuel in which the spent nuclear fuel is put through a process like Pyroprocessing that separates the reactor Grade Plutonium and remaining Uranium from the fission products and fuel cladding. This mixed metal is then combined with a small quantity of medium enriched Uranium with approximately 17% U-235 concentration to make a new combined metal oxide fuel with 1% Reactor Grade plutonium and a U-235 concentration of 4%. These fuel rods are suitable for use in standard PWR reactors as the Plutonium content is no higher than that which exists at the end of cycle in the spent nuclear fuel. As of February 2020 Russia was deploying this fuel in some of their fleet of VVER reactors.[30][31]
Minor actinides recycling
It has been proposed that in addition to the use of plutonium, the minor actinides could be used in a critical power reactor. Tests are already being conducted in which americium is being used as a fuel.[32]
A number of reactor designs, like the
It so happens that the
One promising alternative from this perspective is an
Such reactors compare very well to other neutron sources in terms of neutron energy:
- Thermal 0 to 100 eV
- Epithermal 100 eV to 100 keV
- Fast (from nuclear fission) 100 keV to 3 MeV
- DD fusion 2.5 MeV
- DT fusion 14 MeV
- Accelerator driven core 200 MeV (lead driven by 1.6 GeV protons)
- Muon-catalyzed fusion 7 GeV.
As an alternative, the curium-244, with a half-life of 18 years, could be left to decay into plutonium-240 before being used in fuel in a fast reactor.
Fuel or targets for this actinide transmutation
To date the nature of the fuel (targets) for actinide transformation has not been chosen.
If actinides are transmuted in a Subcritical reactor, it is likely that the fuel will have to be able to tolerate more thermal cycles than conventional fuel. Due to current particle accelerators not being optimized for long continuous operation at least the first generation of accelerator-driven sub-critical reactor is unlikely to be able to maintain a constant operation period for equally long times as a critical reactor, and each time the accelerator stops then the fuel will cool down.
On the other hand, if actinides are destroyed using a fast reactor, such as an
Depending on the matrix the process can generate more transuranics from the matrix. This could either be viewed as good (generate more fuel) or can be viewed as bad (generation of more radiotoxic
Fissile nuclei (such as 233U, 235U, and 239Pu) respond well to
Actinides in an inert matrix
The actinides will be mixed with a metal which will not form more actinides; for instance, an
The raison d’être of the Initiative for Inert Matrix Fuel (IMF) is to contribute to Research and Development studies on inert matrix fuels that could be used to utilise, reduce and dispose both weapon- and light water reactor-grade plutonium excesses. In addition to plutonium, the amounts of minor actinides are also increasing. These actinides have to be consequently disposed in a safe, ecological and economical way. The promising strategy that consists of utilising plutonium and minor actinides using a once-through fuel approach within existing commercial nuclear power reactors e.g. US, European, Russian or Japanese Light Water Reactors (LWR), Canadian Pressured Heavy Water Reactors, or in future transmutation units, has been emphasised since the beginning of the initiative. The approach, which makes use of inert matrix fuel is now studied by several groups in the world.[37][38] This option has the advantage of reducing the plutonium amounts and potentially minor actinide contents prior to geological disposal. The second option is based on using a uranium-free fuel leachable for reprocessing and by following a multi-recycling strategy. In both cases, the advanced fuel material produces energy while consuming plutonium or the minor actinides. This material must, however, be robust. The selected material must be the result of a careful system study including inert matrix – burnable absorbent – fissile material as minimum components and with the addition of stabiliser. This yields a single-phase solid solution or more simply if this option is not selected a composite inert matrix–fissile component. In screening studies[39][40][41] pre-selected elements were identified as suitable. In the 90s an IMF once through strategy was adopted considering the following properties:
- neutron properties i.e. low absorption cross-section, optimal constant reactivity, suitable Doppler coefficient,[42]
- phase stability, chemical inertness, and compatibility,[43]
- acceptable thermo-physical properties i.e. heat capacity, thermal conductivity,[44]
- good behaviour under irradiation i.e. phase stability, minimum swelling,[45]
- retention of fission products or residual actinides,[46] and
- optimal properties after irradiation with insolubility for once through then out.[47]
This once-through then out strategy may be adapted as a last cycle after multi-recycling if the fission yield is not large enough, in which case the following property is required good leaching properties for reprocessing and multi-recycling.[48]
Actinides in a thorium matrix
Upon neutron bombardment, thorium can be converted to uranium-233. 233U is fissile, and has a larger fission cross section than both 235U and 238U, and thus it is far less likely to produce higher actinides through neutron capture.
Actinides in a uranium matrix
If the actinides are incorporated into a uranium-metal or uranium-oxide matrix, then the neutron capture of 238U is likely to generate new plutonium-239. An advantage of mixing the actinides with uranium and plutonium is that the large fission cross sections of 235U and 239Pu for the less energetic delayed neutrons could make the reaction stable enough to be carried out in a critical
Mixed matrix
It is also possible to create a matrix made from a mix of the above-mentioned materials. This is most commonly done in fast reactors where one may wish to keep the breeding ratio of new fuel high enough to keep powering the reactor, but still low enough that the generated actinides can be safely destroyed without transporting them to another site. One way to do this is to use fuel where actinides and uranium is mixed with inert zirconium, producing fuel elements with the desired properties.
Uranium cycle in renewable mode
To fulfill the conditions required for a nuclear renewable energy concept, one has to explore a combination of processes going from the front end of the nuclear fuel cycle to the fuel production and the energy conversion using specific fluid fuels and reactors, as reported by Degueldre et al. (2019[49]). Extraction of uranium from a diluted fluid ore such as seawater has been studied in various countries worldwide. This extraction should be carried out parsimoniously, as suggested by Degueldre (2017).[50] An extraction rate of kilotons of U per year over centuries would not modify significantly the equilibrium concentration of uranium in the oceans (3.3 ppb). This equilibrium results from the input of 10 kilotons of U per year by river waters and its scavenging on the sea floor from the 1.37 exatons of water in the oceans.[citation needed] For a renewable uranium extraction, the use of a specific biomass material is suggested to adsorb uranium and subsequently other transition metals. The uranium loading on the biomass would be around 100 mg per kg. After contact time, the loaded material would be dried and burned (CO2 neutral) with heat conversion into electricity.[citation needed] The uranium ‘burning’ in a molten salt fast reactor helps to optimize the energy conversion by burning all actinide isotopes with an excellent yield for producing a maximum amount of thermal energy from fission and converting it into electricity. This optimisation can be reached by reducing the moderation and the fission product concentration in the liquid fuel/coolant. These effects can be achieved by using a maximum amount of actinides and a minimum amount of alkaline/earth alkaline elements yielding a harder neutron spectrum.[citation needed] Under these optimal conditions the consumption of natural uranium would be 7 tons per year and per gigawatt (GW) of produced electricity. The coupling of uranium extraction from the sea and its optimal utilisation in a molten salt fast reactor should allow nuclear energy to gain the label renewable. In addition, the amount of seawater used by a nuclear power plant to cool the last coolant fluid and the turbine would be ~2.1 giga tons per year for a fast molten salt reactor, corresponding to 7 tons of natural uranium extractable per year. This practice justifies the label renewable.[citation needed]
Thorium cycle
In the thorium fuel cycle thorium-232 absorbs a neutron in either a fast or thermal reactor. The thorium-233 beta decays to protactinium-233 and then to uranium-233, which in turn is used as fuel. Hence, like uranium-238, thorium-232 is a fertile material.
After starting the reactor with existing U-233 or some other
One of the earliest efforts to use a thorium fuel cycle took place at
Thorium was first used commercially in the Indian Point Unit 1 reactor which began operation in 1962. The cost of recovering U-233 from the spent fuel was deemed uneconomical, since less than 1% of the thorium was converted to U-233. The plant's owner switched to uranium fuel, which was used until the reactor was permanently shut down in 1974.[53]
Current industrial activity
Currently the only isotopes used as nuclear fuel are
Virtually all ever deployed
The term nuclear fuel is not normally used in respect to fusion power, which fuses isotopes of hydrogen into helium to release energy.
See also
References
- ^ "Why Nuclear – Generation Atomic". January 26, 2021. Retrieved June 27, 2021.
- ^ "Nuclear Waste May Get A Second Life". NPR.org. Retrieved June 27, 2021.
- ^ "How much depleted uranium hexafluoride is stored in the United States?". Depleted UF6 Management Information Network. Archived from the original on December 23, 2007. Retrieved January 15, 2008.
- ^ "Susquehanna Nuclear Energy Guide" (PDF). PPL Corporation. Archived from the original (PDF) on November 29, 2007. Retrieved January 15, 2008.
- ^ "Nuclear Fuel Cycle | World Nuclear Transport Institute". Wnti.co.uk. Retrieved April 20, 2013.
- ^ A good report on the microstructure of used fuel is Lucuta PG et al. (1991) J Nuclear Materials 178:48-60
- ^ V.V. Rondinella VV et al. (2000) Radiochimica Acta 88:527–531
- ^ For a review of the corrosion of uranium dioxide in a waste store which explains much of the chemistry, see Shoesmith DW (2000) J Nuclear Materials 282:1–31
- ^ Miserque F et al. (2001) J Nuclear Materials 298:280–290
- ^ Further reading on fuel cladding interactions: Tanaka K et al. (2006) J Nuclear Materials 357:58–68
- ^ P. Soudek, Š. Valenová, Z. Vavříková and T. Vaněk, Journal of Environmental Radioactivity, 2006, 88, 236–250
- ^ Generic Assessment Procedures for Determining Protective Actions During a Reactor Accident, IAEA-TECDOC-955, 1997, p. 169
- ^ Generic Assessment Procedures for Determining Protective Actions During a Reactor Accident, IAEA-TECDOC-955, 1997, p. 173
- ^ Generic Assessment Procedures for Determining Protective Actions During a Reactor Accident, IAEA-TECDOC-955, 1997, p. 171
- ^ A. Preston, J.W.R. Dutton and B.R. Harvey, Nature, 1968, 218, 689–690.
- ^ "Russia's Nuclear Fuel Cycle | Russian Nuclear Fuel Cycle - World Nuclear Association".
- ISSN 0029-5493.
- ^ Plus radium (element 88). While actually a sub-actinide, it immediately precedes actinium (89) and follows a three-element gap of instability after polonium (84) where no nuclides have half-lives of at least four years (the longest-lived nuclide in the gap is radon-222 with a half life of less than four days). Radium's longest lived isotope, at 1,600 years, thus merits the element's inclusion here.
- thermal neutron fission of uranium-235, e.g. in a typical nuclear reactor.
- .
"The isotopic analyses disclosed a species of mass 248 in constant abundance in three samples analysed over a period of about 10 months. This was ascribed to an isomer of Bk248 with a half-life greater than 9 [years]. No growth of Cf248 was detected, and a lower limit for the β− half-life can be set at about 104 [years]. No alpha activity attributable to the new isomer has been detected; the alpha half-life is probably greater than 300 [years]." - sea of instability".
- ^ Excluding those "classically stable" nuclides with half-lives significantly in excess of 232Th; e.g., while 113mCd has a half-life of only fourteen years, that of 113Cd is eight quadrillion years.
- ISBN 0-08-044462-8, Amsterdam, 315 pp. (2005).
- ^ Conca, James (January 31, 2019). "Can We Drill a Hole Deep Enough for Our Nuclear Waste?". Forbes.
- .
- .
- ISBN 978-1849710732.
- ^ Dyck, Peter; Crijns, Martin J. "Management of Spent Fuel at Nuclear Power Plants". IAEA Bulletin. Archived from the original on December 10, 2007. Retrieved January 15, 2008.
- ^ Archived at Ghostarchive and the Wayback Machine: "Historical video about the Integral Fast Reactor (IFR) concept". Nuclear Engineering at Argonne.
- ^ "Nuclear Fuel Fabrication - World Nuclear Association".
- ^ "REMIX fuel pilot testing starts at Balakovo reactor - World Nuclear News".
- ^ Warin D.; Konings R.J.M; Haas D.; Maritin P.; Bonnerot J-M.; Vambenepe G.; Schram R.P.C.; Kuijper J.C.; Bakker K.; Conrad R. (October 2002). "The Preparation of the EFTTRA-T5 Americium Transmutation Experiment" (PDF). Seventh Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation. Retrieved January 15, 2008.
- ^ Gudowski, W. (August 2000). "Why Accelerator-Driven Transmutation of Wastes Enables Future Nuclear Power?" (PDF). XX International Linac Conference. Archived from the original (PDF) on November 29, 2007. Retrieved January 15, 2008.
- ^ Heighway, E. A. (July 1, 1994). An overview of accelerator-driven transmutation technology (PDF). LAMPF user`s group meeting. Washington, DC. Retrieved January 15, 2008.
- ^ "Accelerator-driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles" (PDF). Nuclear Energy Agency. Retrieved January 15, 2008.
- ^ Brolly Á.; Vértes P. (March 2005). "Concept of a Small-scale Electron Accelerator Driven System for Nuclear Waste Transmutation Part 2. Investigation of burnup" (PDF). Retrieved January 15, 2008.
- ^ C. Degueldre, J.-M. Paratte (Eds.), J. Nucl. Mater. 274 (1999) 1.
- ^ C. Degueldre, J. Porta (Eds.), Prog. Nucl. Energy 38 (2001) 221.
- ^ Hj. Matzke, V. Rondinella, Th. Wiss, J. Nucl. Mater. 274 (1999) 47
- ^ C. Degueldre, U. Kasemeyer, F. Botta, G. Ledergerber, Proc. Mater. Res. Soc. 412 (1996) 15.
- ^ H. Kleykamps, J. Nucl. Mater. 275 (1999) 1
- ^ J.L. Kloosterman, P.M.G. Damen, J. Nucl. Mater. 274 (1999) 112.
- ^ N. Nitani, T. Yamashita, T. Matsuda, S.-I. Kobayashi, T. Ohmichi, J. Nucl. Mater. 274 (1999) 15
- ^ R.A. Verall, M.D. Vlajic, V.D. Krstic, J. Nucl. Mater. 274 (1999) 54.
- ^ C. Degueldre, M. Pouchon, M. Dobeli, K. Sickafus, K. € Hojou, G. Ledergerber, S. Abolhassani-Dadras, J. Nucl. Mater. 289 (2001) 115
- ^ L.M. Wang, S. Zhu, S.X. Wang, R.C. Ewing, N. Boucharat, A. Fernandez, Hj. Matzke, Prog. Nucl. Energy 38 (2001) 295
- ^ M.A. Pouchon, E. Curtis, C. Degueldre, L. Tobler, Prog. Nucl. Energy 38 (2001) 443
- ^ J.P. Coulon, R. Allonce, A. Filly, F. Chartier, M. Salmon, M. Trabant, Prog. Nucl. Energy 38 (2001) 431
- ^ Claude Degueldre, Richard James Dawson, Vesna Najdanovic-Visak Nuclear fuel cycle, with a liquid ore and fuel: toward renewable energy, Sustainable Energy and Fuels 3 (2019) 1693-1700. https://doi.org/10.1039/C8SE00610E
- ^ Claude Degueldre, Uranium as a renewable for nuclear energy, Progress in Nuclear Energy, 94 (2017) 174-186. https://doi.org/10.1016/j.pnucene.2016.03.031
- ^ a b See thorium fuel cycle
- ^ See Thorium occurrence for discussion of abundance.
- ^ "Thorium Reactors: Their Backers Overstate the Benefits" (PDF). Retrieved March 8, 2021.,
- ^ Chidambaram R. (1997). "Towards an Energy Independent India". Nu-Power. Nuclear Power Corporation of India Limited. Archived from the original on December 17, 2007. Retrieved January 15, 2008.