Breeder reactor

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Assembly of the core of Experimental Breeder Reactor I in Idaho, United States, 1951

A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. These reactors can be fuelled with more commonly available isotopes of uranium and thorium, such as uranium-238 or thorium-232, as opposed to the rare uranium-235 which is used in conventional reactors. These materials are called fertile materials since they can be bred into fuel by these breeder reactors.

Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use. These extra neutrons are absorbed by the fertile material that is loaded into the reactor along with fissile fuel. This irradiated fertile material in turns transmutes into fissile material which can undergo fission reactions.

Breeders were at first found attractive because they made more complete use of uranium fuel than light-water reactors, but interest declined after the 1960s, as more uranium reserves were found, and new methods of uranium enrichment reduced fuel costs.

Fuel resources

Breeder reactors could, in principle, extract almost all of the energy contained in uranium or thorium, decreasing fuel requirements by a factor of 100 compared to widely used once-through light water reactors, which extract less than 1% of the energy in the uranium mined from the earth. The high fuel-efficiency of breeder reactors could greatly reduce concerns about fuel supply, energy used in mining, and storage of radioactive waste. With seawater uranium extraction (currently too expensive to be economical), there is enough fuel for breeder reactors to satisfy the world's energy needs for 5 billion years at 1983's total energy consumption rate, thus making nuclear energy effectively a renewable energy. In addition to seawater the average crustal granite rocks contain significant quantities of Uranium and Thorium that with breeder reactors can supply abundant energy for the remaining lifespan of the sun on the main sequence of stellar evolution.

Fuel efficiency and types of nuclear waste

Nuclear waste became a greater concern by the 1990s. In broad terms, spent nuclear fuel has three main components. The first consists of fission products, the leftover fragments of fuel atoms after they have been split to release energy. Fission products come in dozens of elements and hundreds of isotopes, all of them lighter than uranium. The second main component of spent fuel is transuranics (atoms heavier than uranium), which are generated from uranium or heavier atoms in the fuel when they absorb neutrons but do not undergo fission. All transuranic isotopes fall within the actinide series on the periodic table, and so they are frequently referred to as the actinides. The largest component is the remaining Uranium which is around 98.25 % Uranium-238, 1.1% Uranium-235 and 0.65% Uranium-236. The U-236 comes from the non-fission capture reaction where U-235 absorbs a neutron but releases only a high energy gamma ray instead of undergoing fission.

The physical behavior of the fission products is markedly different from that of the Actinides. In particular, fission products do not themselves undergo fission, and therefore cannot be used as nuclear fuel, either for nuclear weapons or nuclear reactors. Indeed, because fission products are often neutron poisons (absorbing neutrons that could be used to sustain a chain reaction), fission products are viewed as nuclear 'ashes' left over from consuming fissile materials. Furthermore, only seven long-lived fission product isotopes have half-lives longer than a hundred years, which makes their geological storage or disposal less problematic than for transuranic materials.

With increased concerns about nuclear waste, breeding fuel cycles came under renewed interest as they can reduce actinide wastes, particularly plutonium and minor actinides. Breeder reactors are designed to fission the actinide wastes as fuel, and thus convert them to more fission products.

After spent nuclear fuel is removed from a light water reactor, it undergoes a complex decay profile as each nuclide decays at a different rate. Due to a physical oddity referenced below, there is a large gap in the decay half-lives of fission products compared to transuranic isotopes. If the transuranics are left in the spent fuel, after 1,000 to 100,000 years, the slow decay of these transuranics would generate most of the radioactivity in that spent fuel. Thus, removing the transuranics from the waste eliminates much of the long-term radioactivity of spent nuclear fuel.

Fission probabilities of selected actinides, thermal vs. fast neutrons

Isotope	Thermal fission cross section	Thermal fission %	Fast fission cross section	Fast fission %
Th-232	nil	1 n	0.350 barn	3 n
U-232	76.66 barn	59	2.370 barn	95
U-233	531.2 barn	89	2.450 barn	93
U-235	584.4 barn	81	2.056 barn	80
U-238	11.77 microbarn	1 n	1.136 barn	11
Np-237	0.02249 barn	3 n	2.247 barn	27
Pu-238	17.89 barn	7	2.721 barn	70
Pu-239	747.4 barn	63	2.338 barn	85
Pu-240	58.77 barn	1 n	2.253 barn	55
Pu-241	1012 barn	75	2.298 barn	87
Pu-242	0.002557 barn	1 n	2.027 barn	53
Am-241	600.4 barn	1 n	0.2299 microbarn	21
Am-242m	6409 barn	75	2.550 barn	94
Am-243	0.1161 barn	1 n	2.140 barn	23
Cm-242	5.064 barn	1 n	2.907 barn	10
Cm-243	617.4 barn	78	2.500 barn	94
Cm-244	1.037 barn	4 n	0.08255 microbarn	33
n=non-fissile

Today's commercial light water reactors do breed some new fissile material, mostly in the form of plutonium. Because commercial reactors were never designed as breeders, they do not convert enough uranium-238 into plutonium to replace the uranium-235 consumed. Nonetheless, at least one-third of the power produced by commercial nuclear reactors comes from fission of plutonium generated within the fuel. Even with this level of plutonium consumption, light water reactors consume only part of the plutonium and minor actinides they produce, and nonfissile isotopes of plutonium build up, along with significant quantities of other minor actinides.

Conversion ratio, break-even, breeding ratio, doubling time, and burnup

One measure of a reactor's performance is the "conversion ratio", defined as the ratio of new fissile atoms produced to fissile atoms consumed. All proposed nuclear reactors except specially designed and operated actinide burners experience some degree of conversion. As long as there is any amount of a fertile material within the neutron flux of the reactor, some new fissile material is always created. When the conversion ratio is greater than 1, it is often called the "breeding ratio".

For example, commonly used light water reactors have a conversion ratio of approximately 0.6. Pressurized heavy-water reactors (PHWR) running on natural uranium have a conversion ratio of 0.8. In a breeder reactor, the conversion ratio is higher than 1. "Break-even" is achieved when the conversion ratio reaches 1.0 and the reactor produces as much fissile material as it uses.

The doubling time is the amount of time it would take for a breeder reactor to produce enough new fissile material to replace the original fuel and additionally produce an equivalent amount of fuel for another nuclear reactor. This was considered an important measure of breeder performance in early years, when uranium was thought to be scarce. However, since uranium is more abundant than thought in the early days of nuclear reactor development, and given the amount of plutonium available in spent reactor fuel, doubling time has become a less important metric in modern breeder-reactor design.

"Burnup" is a measure of how much energy has been extracted from a given mass of heavy metal in fuel, often expressed (for power reactors) in terms of gigawatt-days per ton of heavy metal. Burnup is an important factor in determining the types and abundances of isotopes produced by a fission reactor. Breeder reactors, by design, have extremely high burnup compared to a conventional reactor, as breeder reactors produce much more of their waste in the form of fission products, while most or all of the actinides are meant to be fissioned and destroyed.

In the past, breeder-reactor development focused on reactors with low breeding ratios, from 1.01 for the Shippingport Reactor running on thorium fuel and cooled by conventional light water to over 1.2 for the Soviet BN-350 liquid-metal-cooled reactor. Theoretical models of breeders with liquid sodium coolant flowing through tubes inside fuel elements ("tube-in-shell" construction) suggest breeding ratios of at least 1.8 are possible on an industrial scale. The Soviet BR-1 test reactor achieved a breeding ratio of 2.5 under non-commercial conditions.

Types of breeder reactor

Production of heavy transuranic actinides in current thermal-neutron fission reactors through neutron capture and decays. Starting at uranium-238, isotopes of plutonium, americium, and curium are all produced. In a fast neutron-breeder reactor, all these isotopes may be burned as fuel.

Many types of breeder reactor are possible:

A "breeder" is simply a reactor designed for very high neutron economy with an associated conversion rate higher than 1.0. In principle, almost any reactor design could be tweaked to become a breeder. For example, the Light Water Reactor, a very heavily moderated thermal design, evolved into the Super Fast Reactor concept, using light water in an extremely low-density supercritical form to increase the neutron economy enough to allow breeding.

Aside from water-cooled, there are many other types of breeder reactor currently envisioned as possible. These include molten-salt cooled, gas cooled, and liquid-metal cooled designs in many variations. Almost any of these basic design types may be fueled by uranium, plutonium, many minor actinides, or thorium, and they may be designed for many different goals, such as creating more fissile fuel, long-term steady-state operation, or active burning of nuclear wastes.

Extant reactor designs are sometimes divided into two broad categories based upon their neutron spectrum, which generally separates those designed to use primarily uranium and transuranics from those designed to use thorium and avoid transuranics. These designs are:

• Fast breeder reactors (FBR's) which use 'fast' (i.e. unmoderated) neutrons to breed fissile plutonium (and possibly higher transuranics) from fertile uranium-238. The fast spectrum is flexible enough that it can also breed fissile uranium-233 from thorium, if desired.
• Thermal breeder reactors which use 'thermal-spectrum' or 'slow' (i.e. moderated) neutrons to breed fissile uranium-233 from thorium (thorium fuel cycle). Due to the behavior of the various nuclear fuels, a thermal breeder is thought commercially feasible only with thorium fuel, which avoids the buildup of the heavier transuranics.

Reprocessing

Fission of the nuclear fuel in any reactor unavoidably produces neutron-absorbing fission products. One must reprocess the fertile material from a breeder reactor to remove those neutron poisons. This step is required to fully utilize the ability to breed as much or more fuel than is consumed. All reprocessing can present a proliferation concern, since it can extract weapons-usable material from spent fuel. The most common reprocessing technique, PUREX, presents a particular concern, since it was expressly designed to separate pure plutonium. Early proposals for the breeder-reactor fuel cycle posed an even greater proliferation concern because they would use PUREX to separate plutonium in a highly attractive isotopic form for use in nuclear weapons.

Several countries are developing reprocessing methods that do not separate the plutonium from the other actinides. For instance, the non-water-based pyrometallurgical electrowinning process, when used to reprocess fuel from an integral fast reactor, leaves large amounts of radioactive actinides in the reactor fuel. More conventional water-based reprocessing systems include SANEX, UNEX, DIAMEX, COEX, and TRUEX, and proposals to combine PUREX with those and other co-processes.

All these systems have moderately better proliferation resistance than PUREX, though their adoption rate is low.

In the thorium cycle, thorium-232 breeds by converting first to protactinium-233, which then decays to uranium-233. If the protactinium remains in the reactor, small amounts of uranium-232 are also produced, which has the strong gamma emitter thallium-208 in its decay chain. Similar to uranium-fueled designs, the longer the fuel and fertile material remain in the reactor, the more of these undesirable elements build up. In the envisioned commercial thorium reactors, high levels of uranium-232 would be allowed to accumulate, leading to extremely high gamma-radiation doses from any uranium derived from thorium. These gamma rays complicate the safe handling of a weapon and the design of its electronics; this explains why uranium-233 has never been pursued for weapons beyond proof-of-concept demonstrations.

While the thorium cycle may be proliferation-resistant with regard to uranium-233 extraction from fuel (because of the presence of uranium-232), it poses a proliferation risk from an alternate route of uranium-233 extraction, which involves chemically extracting protactinium-233 and allowing it to decay to pure uranium-233 outside of the reactor. This process is an obvious chemical operation which is not required for normal operation of these reactor designs, but it could feasibly happen beyond the oversight of organizations such as the International Atomic Energy Agency (IAEA), and thus must be safeguarded against.

Waste reduction

Actinides and fission products by half-life
Actinides by decay chain				Half-life range (a)	Fission products of 235U by yield
4n	4n + 1	4n + 2	4n + 3		4.5–7%	0.04–1.25%	<0.001%
228Ra№				4–6 a		155Euþ
244Cmƒ	241Puƒ	250Cf	227Ac№	10–29 a	90Sr	85Kr	113mCdþ
232Uƒ		238Puƒ	243Cmƒ	29–97 a	137Cs	151Smþ	121mSn
248Bk	249Cfƒ	242mAmƒ		141–351 a	No fission products have a half-life in the range of 100 a–210 ka ...
	241Amƒ		251Cfƒ	430–900 a
		226Ra№	247Bk	1.3–1.6 ka
240Pu	229Th	246Cmƒ	243Amƒ	4.7–7.4 ka
	245Cmƒ	250Cm		8.3–8.5 ka
			239Puƒ	24.1 ka
		230Th№	231Pa№	32–76 ka
236Npƒ	233Uƒ	234U№		150–250 ka	99Tc₡	126Sn
248Cm		242Pu		327–375 ka		79Se₡
				1.53 Ma	93Zr
	237Npƒ			2.1–6.5 Ma	135Cs₡	107Pd
236U			247Cmƒ	15–24 Ma		129I₡
244Pu				80 Ma	... nor beyond 15.7 Ma
232Th№		238U№	235Uƒ№	0.7–14.1 Ga
₡,  has thermal neutron capture cross section in the range of 8–50 barns ƒ,  fissile №,  primarily a naturally occurring radioactive material (NORM) þ,  neutron poison (thermal neutron capture cross section greater than 3k barns)

Nuclear waste became a greater concern by the 1990s. Breeding fuel cycles attracted renewed interest because of their potential to reduce actinide wastes, particularly various isotopes of plutonium and the minor actinides (neptunium, americium, curium, etc). Since breeder reactors on a closed fuel cycle would use nearly all of the isotopes of these actinides fed into them as fuel, their fuel requirements would be reduced by a factor of about 100. The volume of waste they generate would be reduced by a factor of about 100 as well. While there is a huge reduction in the volume of waste from a breeder reactor, the activity of the waste is about the same as that produced by a light-water reactor.

In addition, the waste from a breeder reactor has a different decay behavior, because it is made up of different materials. Breeder reactor waste is mostly fission products, while light-water reactor waste is mostly un-used uranium isotopes and a large quantity of transuranics. After spent nuclear fuel has been removed from a light-water reactor for longer than 100,000 years, the transuranics would be the main source of radioactivity. Eliminating them would eliminate much of the long-term radioactivity from the spent fuel.

In principle, breeder fuel cycles can recycle and consume all actinides, leaving only fission products. As the graphic in this section indicates, fission products have a peculiar 'gap' in their aggregate half-lives, such that no fission products have a half-life between 91 years and two hundred thousand years. As a result of this physical oddity, after several hundred years in storage, the activity of the radioactive waste from a Fast Breeder Reactor would quickly drop to the low level of the long-lived fission products. However, to obtain this benefit requires the highly efficient separation of transuranics from spent fuel. If the fuel reprocessing methods used leave a large fraction of the transuranics in the final waste stream, this advantage would be greatly reduced.

Both types of breeding cycles can reduce actinide wastes:

• The fast breeder reactor's fast neutrons can fission actinide nuclei with even numbers of both protons and neutrons. Such nuclei usually lack the low-speed "thermal neutron" resonances of fissile fuels used in LWRs.
• The thorium fuel cycle inherently produces lower levels of heavy actinides. The fertile material in the thorium fuel cycle has an atomic weight of 232, while the fertile material in the uranium fuel cycle has an atomic weight of 238. That mass difference means that thorium-232 requires six more neutron capture events per nucleus before the transuranic elements can be produced. In addition to this simple mass difference, the reactor gets two chances to fission the nuclei as the mass increases: First as the effective fuel nuclei U233, and as it absorbs two more neutrons, again as the fuel nuclei U235.

A reactor whose main purpose is to destroy actinides, rather than increasing fissile fuel-stocks, is sometimes known as a burner reactor. Both breeding and burning depend on good neutron economy, and many designs can do either. Breeding designs surround the core by a breeding blanket of fertile material. Waste burners surround the core with non-fertile wastes to be destroyed. Some designs add neutron reflectors or absorbers.

Breeder reactor concepts

There are several concepts for breeder reactors; the two main ones are:

• Reactors with a fast neutron spectrum are called fast breeder reactors (FBR's) – these typically utilize uranium-238 as fuel.
• Reactors with a thermal neutron spectrum are called thermal breeder reactors – these typically utilize thorium-232 as fuel.

Fast breeder reactor

Schematic diagram showing the difference between the Loop and Pool types of LMFBR.

In 2006 all large-scale fast breeder reactor (FBR) power stations were liquid metal fast breeder reactors (LMFBR) cooled by liquid sodium. These have been of one of two designs:

• Loop type, in which the primary coolant is circulated through primary heat exchangers outside the reactor tank (but inside the biological shield due to radioactive ²⁴
  Na in the primary coolant)

Experimental Breeder Reactor II, which served as the prototype for the Integral Fast Reactor

• Pool type, in which the primary heat exchangers and pumps are immersed in the reactor tank

All current fast neutron reactor designs use liquid metal as the primary coolant, to transfer heat from the core to steam used to power the electricity generating turbines. FBRs have been built cooled by liquid metals other than sodium—some early FBRs used mercury, other experimental reactors have used a sodium-potassium alloy called NaK. Both have the advantage that they are liquids at room temperature, which is convenient for experimental rigs but less important for pilot or full-scale power stations. Lead and lead-bismuth alloy have also been used.

Three of the proposed generation IV reactor types are FBRs:

• Gas-cooled fast reactor (GFR) cooled by helium.
• Sodium-cooled fast reactor (SFR) based on the existing liquid-metal FBR (LMFBR) and integral fast reactor designs.
• Lead-cooled fast reactor (LFR) based on Soviet naval propulsion units.

FBRs usually use a mixed oxide fuel core of up to 20% plutonium dioxide (PuO₂) and at least 80% uranium dioxide (UO₂). Another fuel option is metal alloys, typically a blend of uranium, plutonium, and zirconium (used because it is "transparent" to neutrons). Enriched uranium can also be used on its own.

Many designs surround the core in a blanket of tubes that contain non-fissile uranium-238, which, by capturing fast neutrons from the reaction in the core, converts to fissile plutonium-239 (as is some of the uranium in the core), which is then reprocessed and used as nuclear fuel. Other FBR designs rely on the geometry of the fuel itself (which also contains uranium-238), arranged to attain sufficient fast neutron capture. The plutonium-239 (or the fissile uranium-235) fission cross-section is much smaller in a fast spectrum than in a thermal spectrum, as is the ratio between the ²³⁹Pu/²³⁵U fission cross-section and the ²³⁸U absorption cross-section. This increases the concentration of ²³⁹Pu/²³⁵U needed to sustain a chain reaction, as well as the ratio of breeding to fission. On the other hand, a fast reactor needs no moderator to slow down the neutrons at all, taking advantage of the fast neutrons producing a greater number of neutrons per fission than slow neutrons. For this reason ordinary liquid water, being a moderator and neutron absorber, is an undesirable primary coolant for fast reactors. Because large amounts of water in the core are required to cool the reactor, the yield of neutrons and therefore breeding of ²³⁹Pu are strongly affected. Theoretical work has been done on reduced moderation water reactors, which may have a sufficiently fast spectrum to provide a breeding ratio slightly over 1. This would likely result in an unacceptable power derating and high costs in a liquid-water-cooled reactor, but the supercritical water coolant of the supercritical water reactor (SCWR) has sufficient heat capacity to allow adequate cooling with less water, making a fast-spectrum water-cooled reactor a practical possibility.

The type of coolants, temperatures and fast neutron spectrum puts the fuel cladding material (normally austenitic stainless or ferritic-martensitic steels) under extreme conditions. The understanding of the radiation damage, coolant interactions, stresses and temperatures are necessary for the safe operation of any reactor core. All materials used to date in sodium-cooled fast reactors have known limits, as explored in ONR-RRR-088 review. Oxide Dispersion Strengthened (ODS) steel is viewed as the long-term radiation resistant fuel-cladding material that overcome the shortcomings of today's material choices.

There are only two commercially operating breeder reactors as of 2017: the BN-600 reactor, at 560 MWe, and the BN-800 reactor, at 880 MWe. Both are Russian sodium-cooled reactors.

Integral fast reactor

Main article: Integral Fast Reactor

One design of fast neutron reactor, specifically conceived to address the waste disposal and plutonium issues, was the integral fast reactor (IFR, also known as an integral fast breeder reactor, although the original reactor was designed to not breed a net surplus of fissile material).

To solve the waste disposal problem, the IFR had an on-site electrowinning fuel-reprocessing unit that recycled the uranium and all the transuranics (not just plutonium) via electroplating, leaving just short half-life fission products in the waste. Some of these fission products could later be separated for industrial or medical uses and the rest sent to a waste repository. The IFR pyroprocessing system uses molten cadmium cathodes and electrorefiners to reprocess metallic fuel directly on-site at the reactor. Such systems not only co-mingle all the minor actinides with both uranium and plutonium, they are compact and self-contained, so that no plutonium-containing material needs to be transported away from the site of the breeder reactor. Breeder reactors incorporating such technology would most likely be designed with breeding ratios very close to 1.00, so that after an initial loading of enriched uranium and/or plutonium fuel, the reactor would then be refueled only with small deliveries of natural uranium metal. A quantity of natural uranium metal equivalent to a block about the size of a milk crate delivered once per month would be all the fuel such a 1 gigawatt reactor would need. Such self-contained breeders are currently envisioned as the final self-contained and self-supporting ultimate goal of nuclear reactor designers. The project was canceled in 1994 by United States Secretary of Energy Hazel O'Leary.

Other fast reactors

The graphite core of the Molten Salt Reactor Experiment

Another proposed fast reactor is a fast molten salt reactor, in which the molten salt's moderating properties are insignificant. This is typically achieved by replacing the light metal fluorides (e.g. LiF, BeF₂) in the salt carrier with heavier metal chlorides (e.g., KCl, RbCl, ZrCl₄).

Several prototype FBRs have been built, ranging in electrical output from a few light bulbs' equivalent (EBR-I, 1951) to over 1,000 MWe. As of 2006, the technology is not economically competitive to thermal reactor technology, but India, Japan, China, South Korea and Russia are all committing substantial research funds to further development of fast breeder reactors, anticipating that rising uranium prices will change this in the long term. Germany, in contrast, abandoned the technology due to safety concerns. The SNR-300 fast breeder reactor was finished after 19 years despite cost overruns summing up to a total of €3.6 billion, only to then be abandoned.

India is also developing FBR technology using both uranium and thorium feedstocks.

Thermal breeder reactor

The Shippingport Reactor, used as a prototype light water breeder for five years beginning in August 1977

The advanced heavy water reactor (AHWR) is one of the few proposed large-scale uses of thorium. India is developing this technology, motivated by substantial thorium reserves; almost a third of the world's thorium reserves are in India, which lacks significant uranium reserves.

The third and final core of the Shippingport Atomic Power Station 60 MWe reactor was a light water thorium breeder, which began operating in 1977. It used pellets made of thorium dioxide and uranium-233 oxide; initially, the U-233 content of the pellets was 5–6% in the seed region, 1.5–3% in the blanket region and none in the reflector region. It operated at 236 MWt, generating 60 MWe and ultimately produced over 2.1 billion kilowatt hours of electricity. After five years, the core was removed and found to contain nearly 1.4% more fissile material than when it was installed, demonstrating that breeding from thorium had occurred.

The liquid fluoride thorium reactor (LFTR) is also planned as a thorium thermal breeder. Liquid-fluoride reactors may have attractive features, such as inherent safety, no need to manufacture fuel rods and possibly simpler reprocessing of the liquid fuel. This concept was first investigated at the Oak Ridge National Laboratory Molten-Salt Reactor Experiment in the 1960s. From 2012 it became the subject of renewed interest worldwide. Japan, India, China, the UK, as well as private US, Czech and Australian companies have expressed intent to develop and commercialize the technology.

Discussion

Like many aspects of nuclear power, fast breeder reactors have been subject to much controversy over the years. In 2010 the International Panel on Fissile Materials said "After six decades and the expenditure of the equivalent of tens of billions of dollars, the promise of breeder reactors remains largely unfulfilled and efforts to commercialize them have been steadily cut back in most countries". In Germany, the United Kingdom, and the United States, breeder reactor development programs have been abandoned. The rationale for pursuing breeder reactors—sometimes explicit and sometimes implicit—was based on the following key assumptions:

• It was expected that uranium would be scarce and high-grade deposits would quickly become depleted if fission power were deployed on a large scale; the reality, however, is that since the end of the cold war, uranium has been much cheaper and more abundant than early designers expected.
• It was expected that breeder reactors would quickly become economically competitive with the light-water reactors that dominate nuclear power today, but the reality is that capital costs are at least 25% more than water-cooled reactors.
• It was thought that breeder reactors could be as safe and reliable as light-water reactors, but safety issues are cited as a concern with fast reactors that use a sodium coolant, where a leak could lead to a sodium fire.
• It was expected that the proliferation risks posed by breeders and their "closed" fuel cycle, in which plutonium would be recycled, could be managed. But since plutonium-breeding reactors produce plutonium from U238, and thorium reactors produce fissile U233 from thorium, all breeding cycles could theoretically pose proliferation risks. However U232, which is always present in U233 produced in breeder reactors, is a strong gamma-emitter via its daughter products, and would make weapon handling extremely hazardous and the weapon easy to detect.

There are some past anti-nuclear advocates that have become pro-nuclear power as a clean source of electricity since breeder reactors effectively recycle most of their waste. This solves one of the most-important negative issues of nuclear power. In the documentary Pandora's Promise, a case is made for breeder reactors because they provide a real high-kW alternative to fossil fuel energy. According to the movie, one pound of uranium provides as much energy as 5,000 barrels of oil.

FBRs have been built and operated in the United States, the United Kingdom, France, the former USSR, India and Japan. The experimental FBR SNR-300 was built in Germany but never operated and eventually shut down amid political controversy following the Chernobyl disaster. As of 2019, two FBRs are being operated for power generation in Russia. Several reactors are planned, many for research related to the Generation IV reactor initiative.

Development and notable breeder reactors

The Soviet Union (comprising Russia and other countries, dissolved in 1991) constructed a series of fast reactors, the first being mercury-cooled and fueled with plutonium metal, and the later plants sodium-cooled and fueled with plutonium oxide.

BR-1 (1955) was 100W (thermal) was followed by BR-2 at 100 kW and then the 5MW BR-5.

BOR-60 (first criticality 1969) was 60 MW, with construction started in 1965.

BN-600 (1981), followed by Russia's BN-800 (2016)

Future plants

The Chinese Experimental Fast Reactor is a 65 MW (thermal), 20 MW (electric), sodium-cooled, pool-type reactor with a 30-year design lifetime and a target burnup of 100 MWd/kg.

India has been trying to develop fast breeder reactors for decades but suffered repeated delays. In 2012 an FBR called the Prototype Fast Breeder Reactor was due to be completed and commissioned. The program is intended to use fertile thorium-232 to breed fissile uranium-233. India is also pursuing thorium thermal breeder reactor technology. India's focus on thorium is due to the nation's large reserves, though known worldwide reserves of thorium are four times those of uranium. India's Department of Atomic Energy (DAE) said in 2007 that it would simultaneously construct four more breeder reactors of 500 MWe each including two at Kalpakkam.

BHAVINI, an Indian nuclear power company, was established in 2003 to construct, commission and operate all stage II fast breeder reactors outlined in India's three stage nuclear power programme. To advance these plans, the Indian FBR-600 is a pool-type sodium-cooled reactor with a rating of 600 MWe.

The China Experimental Fast Reactor (CEFR) is a 25 MW(e) prototype for the planned China Prototype Fast Reactor (CFRP). It started generating power on 21 July 2011.

China also initiated a research and development project in thorium molten-salt thermal breeder-reactor technology (liquid fluoride thorium reactor), formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011. Its ultimate target was to investigate and develop a thorium-based molten salt nuclear system over about 20 years.

Kirk Sorensen, former NASA scientist and chief nuclear technologist at Teledyne Brown Engineering, has long been a promoter of thorium fuel cycle and particularly liquid fluoride thorium reactors. In 2011, Sorensen founded Flibe Energy, a company aimed to develop 20–50 MW LFTR reactor designs to power military bases.

South Korea is developing a design for a standardized modular FBR for export, to complement the standardized PWR (pressurized water reactor) and CANDU designs they have already developed and built, but has not yet committed to building a prototype.

A cutaway model of the BN-600 reactor, superseded by the BN-800 reactor family.

 

Construction of the BN-800 reactor

Russia has a plan for increasing its fleet of fast breeder reactors significantly. A BN-800 reactor (800 MWe) at Beloyarsk was completed in 2012, succeeding a smaller BN-600. In June 2014 the BN-800 was started in the minimum power mode. Working at 35% of nominal efficiency, the reactor contributed to the energy network on 10 December 2015. It reached its full power production in August 2016.

Plans for the construction of a larger BN-1200 reactor (1,200 MWe) was scheduled for completion in 2018, with two additional BN-1200 reactors built by the end of 2030. However, in 2015 Rosenergoatom postponed construction indefinitely to allow fuel design to be improved after more experience of operating the BN-800 reactor, and among cost concerns.

An experimental lead-cooled fast reactor, BREST-300 will be built at the Siberian Chemical Combine (SCC) in Seversk. The BREST (Russian: bystry reaktor so svintsovym teplonositelem, English: fast reactor with lead coolant) design is seen as a successor to the BN series and the 300 MWe unit at the SCC could be the forerunner to a 1,200 MWe version for wide deployment as a commercial power generation unit. The development program is as part of an Advanced Nuclear Technologies Federal Program 2010–2020 that seeks to exploit fast reactors for uranium efficiency while 'burning' radioactive substances that would otherwise be disposed of as waste. Its core would measure about 2.3 metres in diameter by 1.1 metres in height and contain 16 tonnes of fuel. The unit would be refuelled every year, with each fuel element spending five years in total within the core. Lead coolant temperature would be around 540 °C, giving a high efficiency of 43%, primary heat production of 700 MWt yielding electrical power of 300 MWe. The operational lifespan of the unit could be 60 years. The design is expected to be completed by NIKIET in 2014 for construction between 2016 and 2020.

On 16 February 2006, the United States, France and Japan signed an "arrangement" to research and develop sodium-cooled fast reactors in support of the Global Nuclear Energy Partnership. In April 2007 the Japanese government selected Mitsubishi Heavy Industries (MHI) as the "core company in FBR development in Japan". Shortly thereafter, MHI started a new company, Mitsubishi FBR Systems (MFBR) to develop and eventually sell FBR technology.

The Marcoule Nuclear Site in France, location of the Phénix (on the left).

In September 2010 the French government allocated €651.6 million to the Commissariat à l'énergie atomique to finalize the design of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a 600 MW fourth-generation reactor design to be finalized in 2020. As of 2013 the UK had shown interest in the PRISM reactor and was working in concert with France to develop ASTRID. In 2019, CEA announced this design would not be built before mid-century.

In October 2010 GE Hitachi Nuclear Energy signed a memorandum of understanding with the operators of the US Department of Energy's Savannah River Site, which should allow the construction of a demonstration plant based on the company's S-PRISM fast breeder reactor prior to the design receiving full Nuclear Regulatory Commission (NRC) licensing approval. In October 2011 The Independent reported that the UK Nuclear Decommissioning Authority (NDA) and senior advisers within the Department for Energy and Climate Change (DECC) had asked for technical and financial details of PRISM, partly as a means of reducing the country's plutonium stockpile.

The traveling wave reactor (TWR) proposed in a patent by Intellectual Ventures is a fast breeder reactor designed to not need fuel reprocessing during the decades-long lifetime of the reactor. The breed-burn wave in the TWR design does not move from one end of the reactor to the other but gradually from the inside out. Moreover, as the fuel's composition changes through nuclear transmutation, fuel rods are continually reshuffled within the core to optimize the neutron flux and fuel usage at any given point in time. Thus, instead of letting the wave propagate through the fuel, the fuel itself is moved through a largely stationary burn wave. This is contrary to many media reports, which have popularized the concept as a candle-like reactor with a burn region that moves down a stick of fuel. By replacing a static core configuration with an actively managed "standing wave" or "soliton" core, TerraPower's design avoids the problem of cooling a highly variable burn region. Under this scenario, the reconfiguration of fuel rods is accomplished remotely by robotic devices; the containment vessel remains closed during the procedure, and there is no associated downtime.

Natural nuclear fission reactor

From Wikipedia, the free encyclopedia

https://en.wikipedia.org/wiki/Natural_nuclear_fission_reactor 

 

Geological situation in Gabon leading to natural nuclear fission reactors

1. Nuclear reactor zones
2. Sandstone
3. Uranium ore layer
4. Granite

A natural nuclear fission reactor is a uranium deposit where self-sustaining nuclear chain reactions occur. The conditions under which a natural nuclear reactor could exist had been predicted in 1956 by Paul Kuroda. The remnants of an extinct or fossil nuclear fission reactor, where self-sustaining nuclear reactions have occurred in the past, can be verified by analysis of isotope ratios of uranium and of the fission products (and the stable daughter nuclides of those fission products). An example of this phenomenon was discovered in 1972 in Oklo, Gabon by Francis Perrin under conditions very similar to Kuroda's predictions.

Oklo is the only location where this phenomenon is known to have occurred, and consists of 16 sites with patches of centimeter-sized ore layers. Here self-sustaining nuclear fission reactions are thought to have taken place approximately 1.7 billion years ago, during the Statherian period of the Paleoproterozoic, and continued for a few hundred thousand years, probably averaging less than 100 kW of thermal power during that time.

History

In May 1972 at the Tricastin uranium enrichment site at Pierrelatte in France, routine mass spectrometry comparing UF₆ samples from the Oklo Mine, located in Gabon, showed a discrepancy in the amount of the ²³⁵
U
isotope. Normally the concentration is 0.72% while these samples had only 0.60%, a significant difference (some 17% less U-235 was contained in the samples than expected). This discrepancy required explanation, as all civilian uranium handling facilities must meticulously account for all fissionable isotopes to ensure that none are diverted to the construction of nuclear weapons. Furthermore since fissile material is why people mine uranium, a significant amount "going missing" was also of direct economic concern.

Thus the French Commissariat à l'énergie atomique (CEA) began an investigation. A series of measurements of the relative abundances of the two most significant isotopes of the uranium mined at Oklo showed anomalous results compared to those obtained for uranium from other mines. Further investigations into this uranium deposit discovered uranium ore with a ²³⁵
U
concentration as low as 0.44% (almost 40% below the normal value). Subsequent examination of isotopes of fission products such as neodymium and ruthenium also showed anomalies, as described in more detail below. However, the trace radioisotope ²³⁴
U did not deviate significantly in its concentration from other natural samples. Both depleted uranium and reprocessed uranium will usually have ²³⁴
U concentrations significantly different from the secular equilibrium of 55 ppm ²³⁴
U relative to ²³⁸
U. This is due to ²³⁴
U being enriched together with ²³⁵
U and due to it being both consumed by neutron capture and produced from ²³⁵
U by fast neutron induced (n,2n) reactions in nuclear reactors. In Oklo any possible deviation of ²³⁴
U concentration present at the time the reactor was active would have long since decayed away. ²³⁶
U must have also been present in higher than usual ratios during the time the reactor was operating, but due to its half life of 2.348×10⁷ years being almost two orders of magnitude shorter than the time elapsed since the reactor operated, it has decayed to roughly 1.4×10⁻²² its original value and thus basically nothing and below any abilities of current equipment to detect.

This loss in ²³⁵
U
is exactly what happens in a nuclear reactor. A possible explanation was that the uranium ore had operated as a natural fission reactor in the distant geological past. Other observations led to the same conclusion, and on 25 September 1972 the CEA announced their finding that self-sustaining nuclear chain reactions had occurred on Earth about 2 billion years ago. Later, other natural nuclear fission reactors were discovered in the region.

Fission product isotope signatures

Isotope signatures of natural neodymium and fission product neodymium from ²³⁵
U
which had been subjected to thermal neutrons.

Neodymium

Neodymium and other elements were found with isotopic compositions different from what is usually found on Earth. For example, Oklo contained less than 6% of the ¹⁴²
Nd
isotope while natural neodymium contains 27%; however Oklo contained more of the ¹⁴³
Nd
isotope. Subtracting the natural isotopic Nd abundance from the Oklo-Nd, the isotopic composition matched that produced by the fission of ²³⁵
U
.

Ruthenium

Isotope signatures of natural ruthenium and fission product ruthenium from ²³⁵
U
which had been subjected to thermal neutrons. The ¹⁰⁰
Mo
(an extremely long-lived double beta emitter) has not had time to decay to ¹⁰⁰
Ru
in more than trace quantities over the time since the reactors stopped working.

Similar investigations into the isotopic ratios of ruthenium at Oklo found a much higher ⁹⁹
Ru
concentration than otherwise naturally occurring (27–30% vs. 12.7%). This anomaly could be explained by the decay of ⁹⁹
Tc
to ⁹⁹
Ru
. In the bar chart the normal natural isotope signature of ruthenium is compared with that for fission product ruthenium which is the result of the fission of ²³⁵
U
with thermal neutrons. It is clear that the fission ruthenium has a different isotope signature. The level of ¹⁰⁰
Ru
in the fission product mixture is low because fission produces neutron rich isotopes which subsequently beta decay and ¹⁰⁰
Ru
would only be produced in appreciable quantities by double beta decay of the very long-lived (half life 7.1×10¹⁸ years) molybdenum isotope ¹⁰⁰
Mo. On the timescale of when the reactors were in operation, very little (about 0.17 ppb) decay to ¹⁰⁰
Ru
will have occurred. Other pathways of ¹⁰⁰
Ru production like neutron capture in ⁹⁹
Ru or ⁹⁹
Tc (quickly followed by beta decay) can only have occurred during high neutron flux and thus ceased when the fission chain reaction stopped.

Mechanism

The natural nuclear reactor formed when a uranium-rich mineral deposit became inundated with groundwater, which could act as a moderator for the neutrons produced by nuclear fission. A chain reaction took place, producing heat that caused the groundwater to boil away; without a moderator that could slow the neutrons, however, the reaction slowed or stopped. The reactor thus had a negative void coefficient of reactivity, something employed as a safety mechanism in human-made light water reactors. After cooling of the mineral deposit, the water returned, and the reaction restarted, completing a full cycle every 3 hours. The fission reaction cycles continued for hundreds of thousands of years and ended when the ever-decreasing fissile materials, coupled with the build-up of neutron poisons, no longer could sustain a chain reaction.

Fission of uranium normally produces five known isotopes of the fission-product gas xenon; all five have been found trapped in the remnants of the natural reactor, in varying concentrations. The concentrations of xenon isotopes, found trapped in mineral formations 2 billion years later, make it possible to calculate the specific time intervals of reactor operation: approximately 30 minutes of criticality followed by 2 hours and 30 minutes of cooling down (exponentially decreasing residual decay heat) to complete a 3-hour cycle. Xenon-135 is the strongest known neutron poison. However, it is not produced directly in appreciable amounts but rather as a decay product of Iodine-135 (or one of its parent nuclides). Xenon-135 itself is unstable and decays to Caesium-135 if not allowed to absorb neutrons. While Caesium-135 is relatively long lived, all Caesium-135 produced by the Oklo reactor has since decayed further to stable Barium-135. Meanwhile Xenon-136, the product of neutron capture in Xenon-135 only decays extremely slowly via double beta decay and thus scientists were able to determine the neutronics of this reactor by calculations based on those isotope ratios almost two billion years after it stopped fissioning uranium.

A key factor that made the reaction possible was that, at the time the reactor went critical 1.7 billion years ago, the fissile isotope ²³⁵
U
made up about 3.1% of the natural uranium, which is comparable to the amount used in some of today's reactors. (The remaining 96.9% was non-fissile ²³⁸
U
and roughly 55 ppm ²³⁴
U.) Because ²³⁵
U
has a shorter half-life than ²³⁸
U
, and thus decays more rapidly, the current abundance of ²³⁵
U
in natural uranium is only 0.72%. A natural nuclear reactor is therefore no longer possible on Earth without heavy water or graphite.

The Oklo uranium ore deposits are the only known sites in which natural nuclear reactors existed. Other rich uranium ore bodies would also have had sufficient uranium to support nuclear reactions at that time, but the combination of uranium, water and physical conditions needed to support the chain reaction was unique, as far as is currently known, to the Oklo ore bodies. It is also possible, that other natural nuclear fission reactors were once operating but have since been geologically disturbed so much as to be unrecognizable, possibly even "diluting" the uranium so far that the isotope ratio would no longer serve as a "fingerprint". Only a small part of the continental crust and no part of the oceanic crust reaches the age of the deposits at Oklo or an age during which isotope ratios of natural uranium would have allowed a self sustaining chain reaction with water as a moderator.

Another factor which probably contributed to the start of the Oklo natural nuclear reactor at 2 billion years, rather than earlier, was the increasing oxygen content in the Earth's atmosphere. Uranium is naturally present in the rocks of the earth, and the abundance of fissile ²³⁵
U
was at least 3% or higher at all times prior to reactor startup. Uranium is soluble in water only in the presence of oxygen. Therefore, increasing oxygen levels during the aging of the Earth may have allowed uranium to be dissolved and transported with groundwater to places where a high enough concentration could accumulate to form rich uranium ore bodies. Without the new aerobic environment available on Earth at the time, these concentrations probably could not have taken place.

It is estimated that nuclear reactions in the uranium in centimeter- to meter-sized veins consumed about five tons of ²³⁵
U
and elevated temperatures to a few hundred degrees Celsius. Most of the non-volatile fission products and actinides have only moved centimeters in the veins during the last 2 billion years. Studies have suggested this as a useful natural analogue for nuclear waste disposal. The overall mass defect from the fission of five tons of ²³⁵
U is about 4.6 kilograms (10 lb). Over its lifetime the reactor produced roughly 100 megatonnes of TNT (420 PJ) in thermal energy, including neutrinos. If one ignores fission of plutonium (which makes up roughly a third of fission events over the course of normal burnup in modern humanmade light water reactors), then fission product yields amount to roughly 129 kilograms (284 lb) of Technetium-99 (since decayed to Ruthenium-99) 108 kilograms (238 lb) of Zirconium-93 (since decayed to Niobium-93), 198 kilograms (437 lb) of Caesium-135 (since decayed to Barium-135, but the real value is probably lower as its parent nuclide, Xenon-135, is a strong neutron poison and will have absorbed neutrons before decaying to Cs-135 in some cases), 28 kilograms (62 lb) of Palladium-107 (since decayed to Silver), 86 kilograms (190 lb) of Strontium-90 (long since decayed to Zirconium) and 185 kilograms (408 lb) of Caesium-137 (long since decayed to Barium).

Relation to the atomic fine-structure constant

The natural reactor of Oklo has been used to check if the atomic fine-structure constant α might have changed over the past 2 billion years. That is because α influences the rate of various nuclear reactions. For example, ¹⁴⁹
Sm
captures a neutron to become ¹⁵⁰
Sm
, and since the rate of neutron capture depends on the value of α, the ratio of the two samarium isotopes in samples from Oklo can be used to calculate the value of α from 2 billion years ago.

Several studies have analysed the relative concentrations of radioactive isotopes left behind at Oklo, and most have concluded that nuclear reactions then were much the same as they are today, which implies α was the same too.

Photodiode

From Wikipedia, the free encyclopedia

https://en.wikipedia.org/wiki/Photodiode 

 

One Ge (top) and three Si (bottom) photodiodes
Type	Passive, diode
Working principle	Converts light into current
Pin configuration	anode and cathode
Electronic symbol

A photodiode is a light-sensitive semiconductor diode. It produces current when it absorbs photons.

The package of a photodiode allows light (or infrared or ultraviolet radiation, or X-rays) to reach the sensitive part of the device. The package may include lenses or optical filters. Devices designed for use specially as a photodiode use a PIN junction rather than a p–n junction, to increase the speed of response. Photodiodes usually have a slower response time as their surface area increases. A photodiode is designed to operate in reverse bias. A solar cell used to generate electric solar power is a large area photodiode.

Photodiodes are used in scientific and industrial instruments to measure light intensity, either for its own sake or as a measure of some other property (density of smoke, for example). A photodiode can be used as the receiver of data encoded on an infrared beam, as in household remote controls. Photodiodes can be used to form an optocoupler, allowing transmission of signals between circuits without a direct metallic connection between them, allowing isolation from high voltage differences.

Principle of operation

A photodiode is a PIN structure or p–n junction. When a photon of sufficient energy strikes the diode, it creates an electron–hole pair. This mechanism is also known as the inner photoelectric effect. If the absorption occurs in the junction's depletion region, or one diffusion length away from it, these carriers are swept from the junction by the built-in electric field of the depletion region. Thus holes move toward the anode, and electrons toward the cathode, and a photocurrent is produced. The total current through the photodiode is the sum of the dark current (current that is generated in the absence of light) and the photocurrent, so the dark current must be minimized to maximize the sensitivity of the device.

To first order, for a given spectral distribution, the photocurrent is linearly proportional to the irradiance.

Photovoltaic mode

I-V characteristic of a photodiode. The linear load lines represent the response of the external circuit: I=(Applied bias voltage-Diode voltage)/Total resistance. The points of intersection with the curves represent the actual current and voltage for a given bias, resistance and illumination.

In photovoltaic mode (zero bias), photocurrent flows into the anode through a short circuit to the cathode. If the circuit is opened or has a load impedance, restricting the photocurrent out of the device, a voltage builds up in the direction that forward biases the diode, that is, anode positive with respect to cathode. If the circuit is shorted or the impedance is low, a forward current will consume all or some of the photocurrent. This mode exploits the photovoltaic effect, which is the basis for solar cells – a traditional solar cell is just a large area photodiode. For optimum power output, the photovoltaic cell will be operated at a voltage that causes only a small forward current compared to the photocurrent.

Photoconductive mode

In photoconductive mode the diode is reverse biased, that is, with the cathode driven positive with respect to the anode. This reduces the response time because the additional reverse bias increases the width of the depletion layer, which decreases the junction's capacitance and increases the region with an electric field that will cause electrons to be quickly collected. The reverse bias also creates dark current without much change in the photocurrent.

Although this mode is faster, the photoconductive mode can exhibit more electronic noise due to dark current or avalanche effects. The leakage current of a good PIN diode is so low (<1 nA) that the Johnson–Nyquist noise of the load resistance in a typical circuit often dominates.

Related devices

Avalanche photodiodes are photodiodes with structure optimized for operating with high reverse bias, approaching the reverse breakdown voltage. This allows each photo-generated carrier to be multiplied by avalanche breakdown, resulting in internal gain within the photodiode, which increases the effective responsivity of the device.

Electronic symbol for a phototransistor

A phototransistor is a light-sensitive transistor. A common type of phototransistor, the bipolar phototransistor, is in essence a bipolar transistor encased in a transparent case so that light can reach the base–collector junction. It was invented by Dr. John N. Shive (more famous for his wave machine) at Bell Labs in 1948 but it was not announced until 1950. The electrons that are generated by photons in the base–collector junction are injected into the base, and this photodiode current is amplified by the transistor's current gain β (or h_fe). If the base and collector leads are used and the emitter is left unconnected, the phototransistor becomes a photodiode. While phototransistors have a higher responsivity for light they are not able to detect low levels of light any better than photodiodes. Phototransistors also have significantly longer response times. Another type of phototransistor, the field-effect phototransistor (also known as photoFET), is a light-sensitive field-effect transistor. Unlike photobipolar transistors, photoFETs control drain-source current by creating a gate voltage.

A solaristor is a two-terminal gate-less phototransistor. A compact class of two-terminal phototransistors or solaristors have been demonstrated in 2018 by ICN2 researchers. The novel concept is a two-in-one power source plus transistor device that runs on solar energy by exploiting a memresistive effect in the flow of photogenerated carriers.

Materials

The material used to make a photodiode is critical to defining its properties, because only photons with sufficient energy to excite electrons across the material's bandgap will produce significant photocurrents.

Materials commonly used to produce photodiodes are listed in the table below.

Material	Electromagnetic spectrum wavelength range (nm)
Silicon	190–1100
Germanium	400–1700
Indium gallium arsenide	800–2600
Lead(II) sulfide	<1000–3500
Mercury cadmium telluride	400–14000

Because of their greater bandgap, silicon-based photodiodes generate less noise than germanium-based photodiodes.

Binary materials, such as MoS₂, and graphene emerged as new materials for the production of photodiodes.

Unwanted and wanted photodiode effects

Any p–n junction, if illuminated, is potentially a photodiode. Semiconductor devices such as diodes, transistors and ICs contain p–n junctions, and will not function correctly if they are illuminated by unwanted electromagnetic radiation (light) of wavelength suitable to produce a photocurrent. This is avoided by encapsulating devices in opaque housings. If these housings are not completely opaque to high-energy radiation (ultraviolet, X-rays, gamma rays), diodes, transistors and ICs can malfunction due to induced photo-currents. Background radiation from the packaging is also significant. Radiation hardening mitigates these effects.

In some cases, the effect is actually wanted, for example to use LEDs as light-sensitive devices (see LED as light sensor) or even for energy harvesting, then sometimes called light-emitting and light-absorbing diodes (LEADs).

Features

Response of a silicon photo diode vs wavelength of the incident light

Critical performance parameters of a photodiode include spectral responsivity, dark current, response time and noise-equivalent power.

Spectral responsivity

The spectral responsivity is a ratio of the generated photocurrent to incident light power, expressed in A/W when used in photoconductive mode. The wavelength-dependence may also be expressed as a quantum efficiency or the ratio of the number of photogenerated carriers to incident photons which is a unitless quantity.

Dark current

The dark current is the current through the photodiode in the absence of light, when it is operated in photoconductive mode. The dark current includes photocurrent generated by background radiation and the saturation current of the semiconductor junction. Dark current must be accounted for by calibration if a photodiode is used to make an accurate optical power measurement, and it is also a source of noise when a photodiode is used in an optical communication system.

Response time

The response time is the time required for the detector to respond to an optical input. A photon absorbed by the semiconducting material will generate an electron–hole pair which will in turn start moving in the material under the effect of the electric field and thus generate a current. The finite duration of this current is known as the transit-time spread and can be evaluated by using Ramo's theorem. One can also show with this theorem that the total charge generated in the external circuit is e and not 2e as one might expect by the presence of the two carriers. Indeed, the integral of the current due to both electron and hole over time must be equal to e. The resistance and capacitance of the photodiode and the external circuitry give rise to another response time known as RC time constant (${\displaystyle \tau =RC}$). This combination of R and C integrates the photoresponse over time and thus lengthens the impulse response of the photodiode. When used in an optical communication system, the response time determines the bandwidth available for signal modulation and thus data transmission.

Noise-equivalent power

Noise-equivalent power (NEP) is the minimum input optical power to generate photocurrent, equal to the rms noise current in a 1 hertz bandwidth. NEP is essentially the minimum detectable power. The related characteristic detectivity (${\displaystyle D}$) is the inverse of NEP (1/NEP) and the specific detectivity (${\displaystyle D^{\star }}$) is the detectivity multiplied by the square root of the area (${\displaystyle A}$) of the photodetector (${\displaystyle D^{\star }=D\sqrt{A}}$) for a 1 Hz bandwidth. The specific detectivity allows different systems to be compared independent of sensor area and system bandwidth; a higher detectivity value indicates a low-noise device or system. Although it is traditional to give (${\displaystyle D^{\star }}$) in many catalogues as a measure of the diode's quality, in practice, it is hardly ever the key parameter.

When a photodiode is used in an optical communication system, all these parameters contribute to the sensitivity of the optical receiver which is the minimum input power required for the receiver to achieve a specified bit error rate.

Applications

P–n photodiodes are used in similar applications to other photodetectors, such as photoconductors, charge-coupled devices (CCD), and photomultiplier tubes. They may be used to generate an output which is dependent upon the illumination (analog for measurement), or to change the state of circuitry (digital, either for control and switching or for digital signal processing).

Photodiodes are used in consumer electronics devices such as compact disc players, smoke detectors, medical devices and the receivers for infrared remote control devices used to control equipment from televisions to air conditioners. For many applications either photodiodes or photoconductors may be used. Either type of photosensor may be used for light measurement, as in camera light meters, or to respond to light levels, as in switching on street lighting after dark.

Photosensors of all types may be used to respond to incident light or to a source of light which is part of the same circuit or system. A photodiode is often combined into a single component with an emitter of light, usually a light-emitting diode (LED), either to detect the presence of a mechanical obstruction to the beam (slotted optical switch) or to couple two digital or analog circuits while maintaining extremely high electrical isolation between them, often for safety (optocoupler). The combination of LED and photodiode is also used in many sensor systems to characterize different types of products based on their optical absorbance.

Photodiodes are often used for accurate measurement of light intensity in science and industry. They generally have a more linear response than photoconductors.

They are also widely used in various medical applications, such as detectors for computed tomography (coupled with scintillators), instruments to analyze samples (immunoassay), and pulse oximeters.

PIN diodes are much faster and more sensitive than p–n junction diodes, and hence are often used for optical communications and in lighting regulation.

P–n photodiodes are not used to measure extremely low light intensities. Instead, if high sensitivity is needed, avalanche photodiodes, intensified charge-coupled devices or photomultiplier tubes are used for applications such as astronomy, spectroscopy, night vision equipment and laser rangefinding.

Comparison with photomultipliers

Advantages compared to photomultipliers:

1. Excellent linearity of output current as a function of incident light
2. Spectral response from 190 nm to 1100 nm (silicon), longer wavelengths with other semiconductor materials
3. Low noise
4. Ruggedized to mechanical stress
5. Low cost
6. Compact and light weight
7. Long lifetime
8. High quantum efficiency, typically 60–80%
9. No high voltage required

Disadvantages compared to photomultipliers:

1. Small area
2. No internal gain (except avalanche photodiodes, but their gain is typically 10²–10³ compared to 10⁵-10⁸ for the photomultiplier)
3. Much lower overall sensitivity
4. Photon counting only possible with specially designed, usually cooled photodiodes, with special electronic circuits
5. Response time for many designs is slower
6. Latent effect

Pinned photodiode

Not to be confused with PIN photodiode.

The pinned photodiode (PPD) has a shallow implant (P+ or N+) in N-type or P-type diffusion layer, respectively, over a P-type or N-type (respectively) substrate layer, such that the intermediate diffusion layer can be fully depleted of majority carriers, like the base region of a bipolar junction transistor. The PPD (usually PNP) is used in CMOS active-pixel sensors; a precursor NPNP triple junction variant with the MOS buffer capacitor and the back-light illumination scheme with complete charge transfer and no image lag was invented by Sony in 1975. This scheme was widely used in many applications of charge transfer devices.

Early charge-coupled device image sensors suffered from shutter lag. This was largely explained with the re-invention of the pinned photodiode. It was developed by Nobukazu Teranishi, Hiromitsu Shiraki and Yasuo Ishihara at NEC in 1980. Sony in 1975 recognized that lag can be eliminated if the signal carriers could be transferred from the photodiode to the CCD. This led to their invention of the pinned photodiode, a photodetector structure with low lag, low noise, high quantum efficiency and low dark current. It was first publicly reported by Teranishi and Ishihara with A. Kohono, E. Oda and K. Arai in 1982, with the addition of an anti-blooming structure. The new photodetector structure invented by Sony in 1975, developed by NEC in 1982 by Kodak in 1984 was given the name "pinned photodiode" (PPD) by B.C. Burkey at Kodak in 1984. In 1987, the PPD began to be incorporated into most CCD sensors, becoming a fixture in consumer electronic video cameras and then digital still cameras.

Advanced CMOS scaling process technology was on the way. And in 1994, Eric Fossum, while working at NASA's Jet Propulsion Laboratory (JPL), explained in public an improvement to the CMOS sensor: the integration of the pinned photodiode. A CMOS sensor with PPD technology was first fabricated in 1995 by a joint JPL and Kodak team that included Fossum along with P.P.K. Lee, R.C. Gee, R.M. Guidash and T.H. Lee. Since then, the PPD has been used in nearly all CMOS sensors. The CMOS sensor with PPD technology was further advanced and refined by R.M. Guidash in 1997, K. Yonemoto and H. Sumi in 2000, and I. Inoue in 2003. This led to CMOS sensors achieve imaging performance on par with CCD sensors, and later exceeding CCD sensors.

Photodiode array

A one-dimensional photodiode array chip with more than 200 diodes in the line across the center

 

A two-dimensional photodiode array of only 4 × 4 pixels occupies the left side of the first optical mouse sensor chip, c. 1982.

A one-dimensional array of hundreds or thousands of photodiodes can be used as a position sensor, for example as part of an angle sensor. A two-dimensional array is used in image sensors and optical mice.

In some applications, photodiode arrays allow for high-speed parallel readout, as opposed to integrating scanning electronics as in a charge-coupled device (CCD) or CMOS sensor. The optical mouse chip shown in the photo has parallel (not multiplexed) access to all 16 photodiodes in its 4 × 4 array.

Passive-pixel image sensor

The passive-pixel sensor (PPS) is a type of photodiode array. It was the precursor to the active-pixel sensor (APS). A passive-pixel sensor consists of passive pixels which are read out without amplification, with each pixel consisting of a photodiode and a MOSFET switch. In a photodiode array, pixels contain a p–n junction, integrated capacitor, and MOSFETs as selection transistors. A photodiode array was proposed by G. Weckler in 1968, predating the CCD. This was the basis for the PPS.

The noise of photodiode arrays is sometimes a limitation to performance. It was not possible to fabricate active pixel sensors with a practical pixel size in the 1970s, due to limited microlithography technology at the time.