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Retrieved 2 September Retrieved 24 January Berlin, Germany. Retrieved 1 November Munich, Germany. These can split more fissile material, resulting in a continued chain reaction. Examples of fissile fuels are U, U and Pu The second type of fuel is called fertile.
Examples of fertile fuel are Th mined thorium and U mined uranium. In order to become fissile these nuclides must first absorb a neutron that's been produced in the process of fission, to become Th and U respectively. After two sequential beta decays , they transmute into fissile isotopes U and Pu respectively.
This process is called breeding. All reactors breed some fuel this way, [17] but today's solid fueled thermal reactors don't breed enough new fuel from the fertile to make up for the amount of fissile they consume. This is because today's reactors use the mined uranium-plutonium cycle in a moderated neutron spectrum.
Such a fuel cycle, using slowed down neutrons, gives back less than 2 new neutrons from fissioning the bred plutonium. Since 1 neutron is required to sustain the fission reaction, this leaves a budget of less than 1 neutron per fission to breed new fuel. In addition, the materials in the core such as metals, moderators and fission products absorb some neutrons, leaving too few neutrons to breed enough fuel to continue operating the reactor.
As a consequence they must add new fissile fuel periodically and swap out some of the old fuel to make room for the new fuel. In a reactor that breeds at least as much new fuel as it consumes, it is not necessary to add new fissile fuel. Only new fertile fuel is added, which breeds to fissile inside the reactor. In addition the fission products need to be removed. This type of reactor is called a breeder reactor. If it breeds just as much new fissile from fertile to keep operating indefinitely, it is called a break-even breeder or isobreeder.
A LFTR is usually designed as a breeder reactor: thorium goes in, fission products come out. Reactors that use the uranium-plutonium fuel cycle require fast reactors to sustain breeding, because only with fast moving neutrons does the fission process provide more than 2 neutrons per fission. With thorium, it is possible to breed using a thermal reactor. This was proven to work in the Shippingport Atomic Power Station , whose final fuel load bred slightly more fissile from thorium than it consumed, despite being a fairly standard light water reactor.
Thermal reactors require less of the expensive fissile fuel to start, but are more sensitive to fission products left in the core. There are two ways to configure a breeder reactor to do the required breeding. One can place the fertile and fissile fuel together, so breeding and splitting occurs in the same place.
Alternatively, fissile and fertile can be separated. The latter is known as core-and-blanket, because a fissile core produces the heat and neutrons while a separate blanket does all the breeding. Oak Ridge investigated both ways to make a breeder for their molten salt breeder reactor. Because the fuel is liquid, they are called the "single fluid" and "two fluid" thorium thermal breeder molten salt reactors.
The one-fluid design includes a large reactor vessel filled with fluoride salt containing thorium and uranium. Graphite rods immersed in the salt function as a moderator and to guide the flow of salt. In the ORNL MSBR molten salt breeder reactor design [18] a reduced amount of graphite near the edge of the reactor core would make the outer region under-moderated, and increased the capture of neutrons there by the thorium. With this arrangement, most of the neutrons were generated at some distance from the reactor boundary, and reduced the neutron leakage to an acceptable level.
In a breeder configuration, extensive fuel processing was specified to remove fission products from the fuel salt. The MSRE was a core region only prototype reactor. According to estimates of Japanese scientists, a single fluid LFTR program could be achieved through a relatively modest investment of roughly — million dollars over 5—10 years to fund research to fill minor technical gaps and build a small reactor prototype comparable to the MSRE.
The two-fluid design is mechanically more complicated than the "single fluid" reactor design. The "two fluid" reactor has a high-neutron-density core that burns uranium from the thorium fuel cycle.
A separate blanket of thorium salt absorbs neutrons and slowly converts its thorium to protactinium Protactinium can be left in the blanket region where neutron flux is lower, so that it slowly decays to U fissile fuel, [23] rather than capture neutrons.
This bred fissile U can be recovered by injecting additional fluorine to create uranium hexafluoride, a gas which can be captured as it comes out of solution. Once reduced again to uranium tetrafluoride, a solid, it can be mixed into the core salt medium to fission. The core's salt is also purified, first by fluorination to remove uranium, then vacuum distillation to remove and reuse the carrier salts. The still bottoms left after the distillation are the fission products waste of a LFTR.
One weakness of the two-fluid design is the necessity of periodically replacing the core-blanket barrier due to fast neutron damage. The effect of neutron radiation on graphite is to slowly shrink and then swell it, causing an increase in porosity and a deterioration in physical properties. Another weakness of the two-fluid design is its complex plumbing. ORNL thought a complex interleaving of core and blanket tubes was necessary to achieve a high power level with acceptably low power density.
However, more recent research has questioned the need for ORNL's complex interleaving graphite tubing, suggesting a simple elongated tube-in-shell reactor that would allow high power output without complex tubing, accommodate thermal expansion, and permit tube replacement.
A two fluid reactor that has thorium in the fuel salt is sometimes called a "one and a half fluid" reactor, or 1. Like the 1 fluid reactor, it has thorium in the fuel salt, which complicates the fuel processing. And yet, like the 2 fluid reactor, it can use a highly effective separate blanket to absorb neutrons that leak from the core. The added disadvantage of keeping the fluids separate using a barrier remains, but with thorium present in the fuel salt there are fewer neutrons that must pass through this barrier into the blanket fluid.
This results in less damage to the barrier. Any leak in the barrier would also be of lower consequence, as the processing system must already deal with thorium in the core. The main design question when deciding between a one and a half or two fluid LFTR is whether a more complicated reprocessing or a more demanding structural barrier will be easier to solve.
In addition to electricity generation , concentrated thermal energy from the high-temperature LFTR can be used as high-grade industrial process heat for many uses, such as ammonia production with the Haber process or thermal Hydrogen production by water splitting, eliminating the efficiency loss of first converting to electricity.
The Rankine cycle is the most basic thermodynamic power cycle. The simplest cycle consists of a steam generator , a turbine, a condenser, and a pump. The working fluid is usually water. A Rankine power conversion system coupled to a LFTR could take advantage of increased steam temperature to improve its thermal efficiency.
The Brayton cycle generator has a much smaller footprint than the Rankine cycle, lower cost and higher thermal efficiency, but requires higher operating temperatures. It is therefore particularly suitable for use with a LFTR. The working gas can be helium, nitrogen, or carbon dioxide. The low-pressure warm gas is cooled in an ambient cooler. The low-pressure cold gas is compressed to the high-pressure of the system.
The high-pressure working gas is expanded in a turbine to produce power. Often the turbine and the compressor are mechanically connected through a single shaft. A Brayton cycle heat engine can operate at lower pressure with wider diameter piping. The LFTR needs a mechanism to remove the fission products from the fuel.
Fission products left in the reactor absorb neutrons and thus reduce neutron economy. This is especially important in the thorium fuel cycle with few spare neutrons and a thermal neutron spectrum, where absorption is strong. The minimum requirement is to recover the valuable fissile material from used fuel. Removal of fission products is similar to reprocessing of solid fuel elements; by chemical or physical means, the valuable fissile fuel is separated from the waste fission products.
Ideally the fertile fuel thorium or U and other fuel components e. However, for economic reasons they may also end up in the waste.
On site processing is planned to work continuously, cleaning a small fraction of the salt every day and sending it back to the reactor.
There is no need to make the fuel salt very clean; the purpose is to keep the concentration of fission products and other impurities e. The concentrations of some of the rare earth elements must be especially kept low, as they have a large absorption cross section.
Some other elements with a small cross section like Cs or Zr may accumulate over years of operation before they are removed. As the fuel of a LFTR is a molten salt mixture, it is attractive to use pyroprocessing , high temperature methods working directly with the hot molten salt. Pyroprocessing does not use radiation sensitive solvents and is not easily disturbed by decay heat. It can be used on highly radioactive fuel directly from the reactor. Ideally everything except new fuel thorium and waste fission products stays inside the plant.
One potential advantage of a liquid fuel is that it not only facilitates separating fission-products from the fuel, but also isolating individual fission products from one another, which is lucrative for isotopes that are scarce and in high-demand for various industrial radiation sources for testing welds via radiography , agricultural sterilizing produce via irradiation , and medical uses Molybdenum which decays into Technetiumm , a valuable radiolabel dye for marking cancerous cells in medical scans.
The more noble metals Pd , Ru , Ag , Mo , Nb , Sb , Tc do not form fluorides in the normal salt, but instead fine colloidal metallic particles. They can plate out on metal surfaces like the heat exchanger, or preferably on high surface area filters which are easier to replace. Still, there is some uncertainty where they end up, as the MSRE only provided a relatively short operating experience and independent laboratory experiments are difficult.
Gases like Xe and Kr come out easily with a sparge of helium. In addition, some of the "noble" metals are removed as an aerosol.
The quick removal of Xe is particularly important, as it is a very strong neutron poison and makes reactor control more difficult if unremoved; this also improves neutron economy. The gas mainly He, Xe and Kr is held for about 2 days until almost all Xe and other short lived isotopes have decayed.
Most of the gas can then be recycled. After an additional hold up of several months, radioactivity is low enough to separate the gas at low temperatures into helium for reuse , xenon for sale and krypton, which needs storage e. For cleaning the salt mixture several methods of chemical separation were proposed. The pyroprocesses of the LFTR salt already starts with a suitable liquid form, so it may be less expensive than using solid oxide fuels.
Although pressurized water reactors are more susceptible to nuclear meltdown in the absence of active safety measures, this is not a universal feature of civilian nuclear reactors. Much of the research in civilian nuclear reactors is for designs with passive nuclear safety features that may be less susceptible to meltdown, even if all emergency systems failed.
For example, pebble bed reactors are designed so that complete loss of coolant for an indefinite period does not result in the reactor overheating.
The ideal is to have reactors that fail-safe through physics rather than through redundant safety systems or human intervention. Certain fast breeder reactor designs may be more susceptible to meltdown than other reactor types, due to their larger quantity of fissile material and the higher neutron flux inside the reactor core. It was tested in April , just before the Chernobyl failure, to simulate loss of coolant pumping power, by switching off the power to the primary pumps.
As designed, it shut itself down, in about seconds, as soon as the temperature rose to a point designed as higher than proper operation would require. This was well below the boiling point of the unpressurised liquid metal coolant, which had entirely sufficient cooling ability to deal with the heat of fission product radioactivity, by simple convection.
The second test, deliberate shut-off of the secondary coolant loop that supplies the generators, caused the primary circuit to undergo the same safe shutdown.
This test simulated the case of a water-cooled reactor losing its steam turbine circuit, perhaps by a leak. This is a list of the major reactor failures in which damage of the reactor core played a role: [26]. The China syndrome loss-of-coolant accident is a nuclear reactor operations accident characterized by the severe meltdown of the core components of the reactor, which then burn through the containment vessel and the housing building, then figuratively through the crust and body of the Earth until reaching the opposite end, presumed to be in "China".
Moreover, any tunnel behind the material would be closed by immense lithostatic pressure. Furthermore, the antipodes of the continental US are, in fact, located in the Indian Ocean , not China. Complete melt-through can occur in several days, even through several meters of concrete; the corium then penetrates several meters into the underlying soil, spreads around, cools, and solidifies. The real scare, however, came from a quote in the film The China Syndrome , which stated, "It melts right down through the bottom of the plant—theoretically to China, but of course, as soon as it hits ground water, it blasts into the atmosphere and sends out clouds of radioactivity.
The number of people killed would depend on which way the wind was blowing, rendering an area the size of Pennsylvania permanently uninhabitable. Release of the fission products was less than normal background radiation, thus no radioactive related injuries or illnesses. Radioactivity and its related injuries and illnesses were tracked over a 30 year period in the surrounding area with no significant findings.
Though there was public confusion due to several miscommunications, there was no evacuation. A similar concern arose during the Chernobyl disaster: after the reactor was destroyed, a liquid corium mass from the melting core began to breach the concrete floor of the reactor vessel, which was situated above the bubbler pool a large water reservoir for emergency pumps, also designed to safely contain steam pipe ruptures.
The RBMK-type reactor had no allowance or planning for core meltdowns, and the imminent interaction of the core mass with the bubbler pool would have produced a considerable steam explosion, increasing the spread and magnitude of the radioactive plume. It was therefore necessary to drain the bubbler pool before the corium reached it. The initial explosion, however, had broken the control circuitry which allowed the pool to be emptied.
Three station workers volunteered to manually operate the valves necessary to drain this pool, and later images of the corium mass in the pipes of the bubbler pool's basement reinforced the prudence of their actions.
The system design of the nuclear power plants built in the late s raised questions of operational safety , and raised the concern that a severe reactor accident could release large quantities of radioactive materials into the atmosphere and environment. By , there were doubts about the ability of the emergency core cooling system of a nuclear reactor to cope with the effects of a loss of coolant accident and the consequent meltdown of the fuel core; the subject proved popular in the technical and the popular presses.
From Wikipedia, the free encyclopedia. Severe nuclear reactor accident that results in core damage from overheating. Inlet 2B Inlet 1A Cavity Loose core debris Crust Previously molten material Lower plenum debris Possible region depleted in uranium Ablated incore instrument guide Hole in baffle plate Coating of previously molten material on bypass region interior surfaces Upper grid damage.
This section does not cite any sources. Please help improve this section by adding citations to reliable sources. Unsourced material may be challenged and removed. April Learn how and when to remove this template message. Main article: Chernobyl disaster. For the film, see The China Syndrome. See also: Core catcher. The New York Times.
Nuclear Regulatory; Rasmussen, Norman C. Commercial Nuclear Power Plants". Hein — via Google Books. ISBN Retrieved 17 August Retrieved 3 October Introduction to nuclear power. Retrieved 5 June Archived from the original PDF on 3 January Retrieved 25 May Managing water addition to a degraded core. OSTI Beltsville, MD: U. Nuclear Regulatory Commission. Retrieved 23 November Retrieved 24 December Archived from the original on 30 October International Atomic Energy Agency. Retrieved 24 February United States Nuclear Regulatory Commission.
Retrieved 1 December Samuel Vij Books India Pvt Ltd. Howieson; H. Shapiro; J. Rogers; P. Mostert; R. Nuclear Safety. J Arias. Retrieved 9 September Archived from the original on 10 October Archived from the original PDF on 15 February Retrieved 26 January Archived from the original on 20 May Retrieved 20 May Retrieved 11 December ABC World News.
Elements of nuclear safety. And how do scientists measure its temperature? Scientific American. PM Explains ". Popular Mechanics. Nuclear safety. Exposing the Chornobyl Myths in Russian. Archived from the original on 8 November Retrieved 8 November Post Chernobyl in Russian. Archived from the original on 26 April Retrieved 3 May Look up China syndrome in Wiktionary, the free dictionary.
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A nuclear meltdown core meltdown , core melt accident , meltdown or partial core melt [2] is a severe nuclear reactor accident that results in core damage from overheating. A core meltdown accident occurs when the heat generated by a nuclear reactor exceeds the heat removed by the cooling systems to the point where at least one nuclear fuel element exceeds its melting point. This differs from a fuel element failure , which is not caused by high temperatures. A meltdown may be caused by a loss of coolant , loss of coolant pressure, or low coolant flow rate or be the result of a criticality excursion in which the reactor is operated at a power level that exceeds its design limits.
Alternatively, an external fire may endanger the core, leading to a meltdown. Once the fuel elements of a reactor begin to melt, the fuel cladding has been breached, and the nuclear fuel such as uranium , plutonium , or thorium and fission products such as caesium , krypton , or iodine within the fuel elements can leach out into the coolant.
Subsequent failures can permit these radioisotopes to breach further layers of containment. Superheated steam and hot metal inside the core can lead to fuel—coolant interactions , hydrogen explosions , or steam hammer , any of which could destroy parts of the containment. A meltdown is considered very serious because of the potential for radioactive materials to breach all containment and escape or be released into the environment , resulting in radioactive contamination and fallout , and potentially leading to radiation poisoning of people and animals nearby.
Nuclear power plants generate electricity by heating fluid via a nuclear reaction to run a generator. If the heat from that reaction is not removed adequately, the fuel assemblies in a reactor core can melt. A core damage incident can occur even after a reactor is shut down because the fuel continues to produce decay heat.
A core damage accident is caused by the loss of sufficient cooling for the nuclear fuel within the reactor core. The reason may be one of several factors, including a loss-of-pressure-control accident , a loss-of-coolant accident LOCA , an uncontrolled power excursion or, in reactors without a pressure vessel , a fire within the reactor core. Failures in control systems may cause a series of events resulting in loss of cooling.
Contemporary safety principles of defense in depth ensure that multiple layers of safety systems are always present to make such accidents unlikely.
The containment building is the last of several safeguards that prevent the release of radioactivity to the environment. Many commercial reactors are contained within a 1. Before the core of a light-water nuclear reactor can be damaged, two precursor events must have already occurred:.
The Three Mile Island accident was a compounded group of emergencies that led to core damage. What led to this was an erroneous decision by operators to shut down the ECCS during an emergency condition due to gauge readings that were either incorrect or misinterpreted; this caused another emergency condition that, several hours after the fact, led to core exposure and a core damage incident.
If the ECCS had been allowed to function, it would have prevented both exposure and core damage. During the Fukushima incident the emergency cooling system had also been manually shut down several minutes after it started. If such a limiting fault were to occur, and a complete failure of all ECCS divisions were to occur, both Kuan, et al and Haskin, et al describe six stages between the start of the limiting fault the loss of cooling and the potential escape of molten corium into the containment a so-called "full meltdown" : [8] [9].
At the point at which the corium relocates to the lower plenum, Haskin, et al relate that the possibility exists for an incident called a fuel—coolant interaction FCI to substantially stress or breach the primary pressure boundary when the corium relocates to the lower plenum of the reactor pressure vessel "RPV".
The American Nuclear Society has commented on the TMI-2 accident, that despite melting of about one-third of the fuel, the reactor vessel itself maintained its integrity and contained the damaged fuel. There are several possibilities as to how the primary pressure boundary could be breached by corium.
As previously described, FCI could lead to an overpressure event leading to RPV fail, and thus, primary pressure boundary fail. Haskin et al report that in the event of a steam explosion, failure of the lower plenum is far more likely than ejection of the upper plenum in the alpha mode.
In the event of lower plenum failure, debris at varied temperatures can be expected to be projected into the cavity below the core. The containment may be subject to overpressure, though this is not likely to fail the containment. The alpha-mode failure will lead to the consequences previously discussed. It is quite possible, especially in pressurized water reactors, that the primary loop will remain pressurized following corium relocation to the lower plenum. As such, pressure stresses on the RPV will be present in addition to the weight stress that the molten corium places on the lower plenum of the RPV; when the metal of the RPV weakens sufficiently due to the heat of the molten corium, it is likely that the liquid corium will be discharged under pressure out of the bottom of the RPV in a pressurized stream, together with entrained gases.
This mode of corium ejection may lead to direct containment heating DCH. Haskin et al identify six modes by which the containment could be credibly challenged; some of these modes are not applicable to core melt accidents. If the melted core penetrates the pressure vessel, there are theories and speculations as to what may then occur. In modern Russian plants, there is a "core catching device" in the bottom of the containment building.
The melted core is supposed to hit a thick layer of a "sacrificial metal" that would melt, dilute the core and increase the heat conductivity, and finally the diluted core can be cooled down by water circulating in the floor. There has never been any full-scale testing of this device, however. In Western plants there is an airtight containment building.
Though radiation would be at a high level within the containment, doses outside of it would be lower. Containment buildings are designed for the orderly release of pressure without releasing radionuclides, through a pressure release valve and filters. In a melting event, one spot or area on the RPV will become hotter than other areas, and will eventually melt.
When it melts, corium will pour into the cavity under the reactor. Though the cavity is designed to remain dry, several NUREG-class documents advise operators to flood the cavity in the event of a fuel melt incident. This water will become steam and pressurize the containment. Automatic water sprays will pump large quantities of water into the steamy environment to keep the pressure down. Catalytic recombiners will rapidly convert the hydrogen and oxygen back into water.
One positive effect of the corium falling into water is that it is cooled and returns to a solid state. Extensive water spray systems within the containment along with the ECCS, when it is reactivated, will allow operators to spray water within the containment to cool the core on the floor and reduce it to a low temperature.
These procedures are intended to prevent release of radioactivity. In the Three Mile Island event in , a theoretical person standing at the plant property line during the entire event would have received a dose of approximately 2 millisieverts millirem , between a chest X-ray's and a CT scan's worth of radiation.
This was due to outgassing by an uncontrolled system that, today, would have been backfitted with activated carbon and HEPA filters to prevent radionuclide release.
In the Fukushima incident, however, this design failed. Despite the efforts of the operators at the Fukushima Daiichi nuclear power plant to maintain control, the reactor cores in units 1—3 overheated, the nuclear fuel melted and the three containment vessels were breached.
Hydrogen was released from the reactor pressure vessels, leading to explosions inside the reactor buildings in units 1, 3 and 4 that damaged structures and equipment and injured personnel. Radionuclides were released from the plant to the atmosphere and were deposited on land and on the ocean.
There were also direct releases into the sea. As the natural decay heat of the corium eventually reduces to an equilibrium with convection and conduction to the containment walls, it becomes cool enough for water spray systems to be shut down and the reactor to be put into safe storage. The containment can be sealed with release of extremely limited offsite radioactivity and release of pressure.
After perhaps a decade for fission products to decay, the containment can be reopened for decontamination and demolition. Another scenario sees a buildup of potentially explosive hydrogen, but passive autocatalytic recombiners inside the containment are designed to prevent this. In Fukushima, the containments were filled with inert nitrogen, which prevented hydrogen from burning; the hydrogen leaked from the containment to the reactor building, however, where it mixed with air and exploded.
There were initial concerns that the hydrogen might ignite and damage the pressure vessel or even the containment building; but it was soon realized that lack of oxygen prevented burning or explosion. One scenario consists of the reactor pressure vessel failing all at once, with the entire mass of corium dropping into a pool of water for example, coolant or moderator and causing extremely rapid generation of steam. The pressure rise within the containment could threaten integrity if rupture disks could not relieve the stress.
Exposed flammable substances could burn, but there are few, if any, flammable substances within the containment. Another theory, called an "alpha mode" failure by the Rasmussen WASH study, asserted steam could produce enough pressure to blow the head off the reactor pressure vessel RPV. The containment could be threatened if the RPV head collided with it. The WASH report was replaced by better-based [ original research?
By , there were doubts about the ability of the emergency cooling systems of a nuclear reactor to prevent a loss-of-coolant accident and the consequent meltdown of the fuel core; the subject proved popular in the technical and the popular presses.
The hypothesis derived from a report by a group of nuclear physicists, headed by W. It has not been determined to what extent a molten mass can melt through a structure although that was tested in the loss-of-fluid-test reactor described in Test Area North 's fact sheet [20].
The Three Mile Island accident provided real-life experience with an actual molten core: the corium failed to melt through the reactor pressure vessel after over six hours of exposure due to dilution of the melt by the control rods and other reactor internals, validating the emphasis on defense in depth against core damage incidents.
Other types of reactors have different capabilities and safety profiles than the LWR does. Advanced varieties of several of these reactors have the potential to be inherently safe. The first is the bulk heavy-water moderator a separate system from the coolant , and the second is the light-water-filled shield tank or calandria vault. These backup heat sinks are sufficient to prevent either the fuel meltdown in the first place using the moderator heat sink , or the breaching of the core vessel should the moderator eventually boil off using the shield tank heat sink.
One type of Western reactor, known as the advanced gas-cooled reactor or AGR , built by the United Kingdom, is not very vulnerable to loss-of-cooling accidents or to core damage except in the most extreme of circumstances. By virtue of the relatively inert coolant carbon dioxide , the large volume and high pressure of the coolant, and the relatively high heat transfer efficiency of the reactor, the time frame for core damage in the event of a limiting fault is measured in days.
Restoration of some means of coolant flow will prevent core damage from occurring. Other types of highly advanced gas cooled reactors, generally known as high-temperature gas-cooled reactors HTGRs such as the Japanese High Temperature Test Reactor and the United States' Very High Temperature Reactor , are inherently safe, meaning that meltdown or other forms of core damage are physically impossible, due to the structure of the core, which consists of hexagonal prismatic blocks of silicon carbide reinforced graphite infused with TRISO or QUADRISO pellets of uranium, thorium, or mixed oxide buried underground in a helium-filled steel pressure vessel within a concrete containment.
Though this type of reactor is not susceptible to meltdown, additional capabilities of heat removal are provided by using regular atmospheric airflow as a means of backup heat removal, by having it pass through a heat exchanger and rising into the atmosphere due to convection , achieving full residual heat removal. This reactor will use a gas as a coolant, which can then be used for process heat such as in hydrogen production or for the driving of gas turbines and the generation of electricity.
A prototype of a very similar type of reactor has been built by the Chinese , HTR , and has worked beyond researchers' expectations, leading the Chinese to announce plans to build a pair of follow-on, full-scale MWe, inherently safe, power production reactors based on the same concept.
See Nuclear power in the People's Republic of China for more information. Recently heavy liquid metal, such as lead or lead-bismuth, has been proposed as a reactor coolant. The PIUS process inherent ultimate safety designs, originally engineered by the Swedes in the late s and early s, are LWRs that by virtue of their design are resistant to core damage. No units have ever been built.
Power reactors, including the Deployable Electrical Energy Reactor , a larger-scale mobile version of the TRIGA for power generation in disaster areas and on military missions, and the TRIGA Power System, a small power plant and heat source for small and remote community use, have been put forward by interested engineers, and share the safety characteristics of the TRIGA due to the uranium zirconium hydride fuel used. The Hydrogen Moderated Self-regulating Nuclear Power Module , a reactor that uses uranium hydride as a moderator and fuel, similar in chemistry and safety to the TRIGA, also possesses these extreme safety and stability characteristics, and has attracted a good deal of interest in recent times.
The liquid fluoride thorium reactor is designed to naturally have its core in a molten state, as a eutectic mix of thorium and fluorine salts. As such, a molten core is reflective of the normal and safe state of operation of this reactor type. In the event the core overheats, a metal plug will melt, and the molten salt core will drain into tanks where it will cool in a non-critical configuration. Since the core is liquid, and already melted, it cannot be damaged.
Advanced liquid metal reactors, such as the U. Soviet-designed RBMK reactors Reaktor Bolshoy Moshchnosti Kanalnyy , found only in Russia and other post-Soviet states and now shut down everywhere except Russia, do not have containment buildings, are naturally unstable tending to dangerous power fluctuations , and have emergency cooling systems ECCS considered grossly inadequate by Western safety standards.
RBMK emergency core cooling systems only have one division and little redundancy within that division.
